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Sample records for s-saltcake actual waste

  1. Cold Dissolved Saltcake Waste Simulant Development, Preparation, and Analysis

    International Nuclear Information System (INIS)

    Rassat, Scot D.; Mahoney, Lenna A.; Russell, Renee L.; Bryan, Samuel A.; Sell, Rachel L.

    2003-01-01

    CH2M HILL Hanford Group, Inc. is identifying and developing supplemental process technologies to accelerate the Hanford tank waste cleanup mission. Bulk vitrification, containerized grout, and steam reforming are three technologies under consideration for treatment of the radioactive saltcake wastes in 68 single-shell tanks. To support development and testing of these technologies, Pacific Northwest National Laboratory (PNNL) was tasked with developing a cold dissolved saltcake simulant formulation to be representative of an actual saltcake waste stream, preparing 25- and 100-L batches of the simulant, and analyzing the composition of the batches to ensure conformance to formulation targets. Lacking a defined composition for dissolved actual saltcake waste, PNNL used available tank waste composition information and an equilibrium chemistry model (Environmental Simulation Program [ESP(trademark)]) to predict the concentrations of analytes in solution. Observations of insoluble solids in initial laboratory preparations for the model-predicted formulation prompted reductions in the concentration of phosphate and silicon in the final simulant formulation. The analytical results for the 25- and 100-L simulant batches, prepared by an outside vendor to PNNL specifications, agree within the expected measurement accuracy (∼10%) of the target concentrations and are highly consistent for replicate measurements, with a few minor exceptions. In parallel with the production of the 2nd simulant batch (100-L), a 1-L laboratory control sample of the same formulation was carefully prepared at PNNL to serve as an analytical standard. The instrumental analyses indicate that the vendor prepared batches of solution adequately reflect the as-formulated simulant composition. In parallel with the simulant development effort, a nominal 5-M (molar) sodium actual waste solution was prepared at the Hanford Site from a limited number of tank waste samples. Because this actual waste solution w

  2. Modeling water retention of sludge simulants and actual saltcake tank wastes

    International Nuclear Information System (INIS)

    Simmons, C.S.

    1996-07-01

    The Ferrocyanide Tanks Safety Program managed by Westinghouse hanford Company has been concerned with the potential combustion hazard of dry tank wastes containing ferrocyanide chemical in combination with nitrate salts. Pervious studies have shown that tank waste containing greater than 20 percent of weight as water could not be accidentally ignited. Moreover, a sustained combustion could not be propagated in such a wet waste even if it contained enough ferrocyanide to burn. Because moisture content is a key critical factor determining the safety of ferrocyanide-containing tank wastes, physical modeling was performed by Pacific Northwest National laboratory to evaluate the moisture-retaining behavior of typical tank wastes. The physical modeling reported here has quantified the mechanisms by which two main types of tank waste, sludge and saltcake, retain moisture in a tank profile under static conditions. Static conditions usually prevail after a tank profile has been stabilized by pumping out any excess interstitial liquid, which is not naturally retained by the waste as a result of physical forces such as capillarity

  3. Hydrological Methods can Separate Cesium from Nuclear Waste Saltcake

    International Nuclear Information System (INIS)

    Brooke, J.N.; Peters, J.F.; Staheli, K.

    1999-01-01

    Interstitial Fluid Displacement (IFD) is a new and novel method for separating cesium from saltcake waste. Hydrologic modeling of liquid flow through porous saltcake suggests that the cesium, potassium and sodium hydroxide can be separated at high recovery and low volume using IFD.'

  4. Characterization, Leaching, and Filtration Testing for Bismuth Phosphate Sludge (Group 1) and Bismuth Phosphate Saltcake (Group 2) Actual Waste Sample Composites

    International Nuclear Information System (INIS)

    Lumetta, Gregg J.; Buck, Edgar C.; Daniel, Richard C.; Draper, Kathryn; Edwards, Matthew K.; Fiskum, Sandra K.; Hallen, Richard T.; Jagoda, Lynette K.; Jenson, Evan D.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Shimskey, Rick W.; Sinkov, Sergey I.; Snow, Lanee A.

    2009-01-01

    A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan.() The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. The actual waste-testing program included homogenizing the samples by group, characterizing the solids and aqueous phases, and performing parametric leaching tests. Two of the eight defined groups - bismuth phosphate sludge (Group 1) and bismuth phosphate saltcake (Group 2) - are the subjects of this report. The Group 1 waste was anticipated to be high in phosphorus and was implicitly assumed to be present as BiPO4 (however, results presented here indicate that the phosphate in Group 1 is actually present as amorphous iron(III) phosphate). The Group 2 waste was also anticipated to be high in phosphorus, but because of the relatively low bismuth content and higher aluminum content, it was anticipated that the Group 2 waste would contain a mixture of gibbsite, sodium phosphate, and aluminum phosphate. Thus, the focus of the Group 1 testing was on determining the behavior of P removal during caustic leaching, and the focus of the Group 2 testing was on the removal of both P and Al. The waste-type definition, archived sample conditions, homogenization activities, characterization (physical, chemical, radioisotope, and crystal habit), and caustic leaching behavior as functions of time, temperature, and hydroxide concentration are discussed in this report. Testing was conducted according to TP-RPP-WTP-467

  5. Developing a model for moisture in saltcake waste tanks: Progress report

    International Nuclear Information System (INIS)

    Simmons, C.S.; Aimo, N.; Fayer, M.J.; White, M.D.

    1997-07-01

    This report describes a modeling effort to provide a computer simulation capability for estimating the distribution and movement of moisture in the saltcake-type waste contained in Hanford's single-shell radioactive waste storage tanks. This moisture model goes beyond an earlier version because it describes water vapor movement as well as the interstitial liquid held in a saltcake waste. The work was performed by Pacific Northwest National Laboratory to assist Duke Engineering and Services Hanford with the Organic Tank Safety Program. The Organic Tank Safety Program is concerned whether saltcake waste, when stabilized by jet pumping, will retain sufficient moisture near the surface to preclude any possibility of an accidental ignition and propagation of burning. The nitrate/nitrite saltcake, which might also potentially include combustible organic chemicals might not always retain enough moisture near the surface to preclude any such accident. Draining liquid from a tank by pumping, coupled with moisture evaporating into a tank's head space, may cause a dry waste surface that is not inherently safe. The moisture model was devised to help examine this safety question. The model accounts for water being continually cycled by evaporation into the head space and returned to the waste by condensation or partly lost through venting to the external atmosphere. Water evaporation occurs even in a closed tank, because it is driven by the transfer to the outside of the heat load generated by radioactivity within the waste. How dry a waste may become over time depends on the particular hydraulic properties of a saltcake, and the model uses those properties to describe the capillary flow of interstitial liquid as well as the water vapor flow caused by thermal differences within the porous waste

  6. Filtration and Leach Testing for REDOX Sludge and S-Saltcake Actual Waste Sample Composites

    Energy Technology Data Exchange (ETDEWEB)

    Shimskey, Rick W.; Billing, Justin M.; Buck, Edgar C.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Geeting, John GH; Hallen, Richard T.; Jenson, Evan D.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Snow, Lanee A.; Swoboda, Robert G.

    2009-02-20

    A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan.( ) The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Under test plan TP-RPP-WTP-467, eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. Under this test plan, a waste-testing program was implemented that included: • Homogenizing the archive samples by group as defined in the test plan • Characterizing the homogenized sample groups • Performing parametric leaching testing on each group for compounds of interest • Performing bench-top filtration/leaching tests in the hot cell for each group to simulate filtration and leaching activities if they occurred in the UFP2 vessel of the WTP Pretreatment Facility. This report focuses on filtration/leaching tests performed on two of the eight waste composite samples and follow-on parametric tests to support aluminum leaching results from those tests.

  7. Saltcake Dissolution FY 2000 Status Report

    International Nuclear Information System (INIS)

    HERTING, D.L.

    2000-01-01

    Laboratory tests were completed on the dissolution characteristics of Hanford saltcake waste from single-shell waste tanks 241-TX- 113, 241-BY-102, 241-BY-106, 241-A-101, and 241-S-102 (henceforth referred to as TX-113, BY-102, BY-106, A-101, and S-102, respectively). This work was funded by the Tanks Focus Area (EM-50) under Technical Task Plan Number RL0-8-WT-41, ''PHMC Pretreatment--Saltcake Dissolution''. The tests performed on saltcake from tank TX-113 were similar in scope to those completed in previous years on waste from tanks BY-102, BY-106, B-106, A-101, and S-102 (Herting 1998, 1999). In addition to the ''standard'' dissolution tests, new types of tests were performed this year related to feed stability and radionuclide distribution. The River Protection Project (RPP) is tasked with retrieving waste from double-shell and single-shell tanks to provide feed for vitrification. The RPP organization needs chemical and physical data to evaluate technologies for retrieving the waste. Little significant laboratory testing has been done to evaluate in-tank dissolution parameters for the various types of saltcake wastes that exist in single-shell tanks. A computer modeling program known as the Environmental Simulation Program (ESP), produced by OLI Systems, Inc of Morris Plains, New Jersey, is being used by the RPP organization to predict solubilities during dilution and retrieval of all tank waste types. Data from this task are provided to ESP users to support evaluation, refinement, and validation of the ESP model

  8. Waste Incidental to Reprocessing Evaluation for Disposing Saltcake to Saltstone

    International Nuclear Information System (INIS)

    Jones, R.T.

    2002-01-01

    This Waste Incidental to Reprocessing Evaluation is performed in accordance with Department of Energy Order 435.1, Radioactive Waste Management. This evaluation is performed in order to determine whether saltcake currently stored in the Tank Farms, when separated from supernate, meets WIR requirements and can therefore be managed as Low Level Waste and disposed in the Saltstone Production and Disposal Facility in Z-Area

  9. Bench Scale Saltcake Dissolution Test Report

    International Nuclear Information System (INIS)

    BECHTOLD, D.B.; PACQUET, E.A.

    2000-01-01

    A potential scenario for retrieving saltcake from single shell tanks is the ''Rainbird(reg s ign) sprinkler'' method. Water is distributed evenly across the surface of the saltcake and allowed to percolate by gravity through the waste. The salt dissolves in the water, forming a saturated solution. The saturated liquid is removed by a saltwell pump situated near the bottom of the tank. By this method, there is never a large inventory of liquid in the tank that could pose a threat of leakage. There are many variables or factors that can influence the hydrodynamics of this retrieval process. They include saltcake porosity; saltwell pumping rate; salt dissolution chemistry; factors that could promote flow channeling (e.g. tank walls, dry wells, inclusions or discontinuities in the saltcake); method of water distribution; plug formation due to crystal formations or accumulation of insoluble solids. A brief literature search indicates that very little experimental data exist on these aspects of saltcake dissolution (Wiersma 1996, 1997). The tests reported here were planned (Herting, 2000) to provide preliminary data and information for planning future, scaled-up tests of the sprinkler method

  10. FY 2000 Saltcake Dissolution and Feed Stability Workshop

    International Nuclear Information System (INIS)

    Hunt, R.D.; McGinnis, C.P.; Weber, C.F.; Welch, T.D.; Jewett, J.R.

    2000-01-01

    The Tanks Focus Area (TFA) continues to work closely with the Office of River Protection (ORP) to better understand the chemistry involved with the retrieval, transport, and pretreatment of nuclear wastes at Hanford. Since a private contractor is currently responsible for the pretreatment and immobilization activities in this remediation effort, the TFA has concentrated on saltcake dissolution and waste transport at the request of the ORP. Researchers at Hanford have performed a series of dissolution experiments on actual saltcake samples. Staff members at Mississippi State University (MSU) continue to model the dissolution results with the Environmental Simulation Program (ESP), which is used extensively by ORP personnel. Several ways to improve the predictive capabilities of the ESP were identified. Since several transfer lines at Hanford have become plugged, TFA tasks at AEA Technologies, Florida International University (FIU), MSU, and Oak Ridge National Laboratory (ORNL) are investigating the behavior of the supernatants and slurries during transport. A combination of experimental and theoretical techniques is used to study the transport chemistry. This effort is expected to develop process control tools for waste transfer. The results from these TFA tasks were presented to ORP personnel during the FY 2000 Saltcake Dissolution and Feed Stability Workshop, which was held on May 16-17 in Richland, Washington. The minutes from this workshop are provided in this report

  11. Tank 10H Saltcake Core Sample Analysis

    International Nuclear Information System (INIS)

    MARTINO, CHRISTOPHERJ

    2004-01-01

    In support of Low-Curie Salt (LCS) process validation at the Savannah River Site (SRS), Liquid Waste Disposition (LWD) has undertaken a program of tank characterization, including salt sampling. As part of this initiative, they sampled the surface and subsurface of Tank 10H saltcake using a series of three 12-inch long sample tubes. These tubes each contain 1-foot long segments of the saltcake from one location, representing the top three feet of saltcake. The primary objective of the characterization that will be useful to the selection and processing of the next waste tanks. Most important is the determination of the Cs-137 concentration and liquid retention properties of Tank 10H saltcake to confirm acceptability of processing. Additional chemical analyses are performed to provide information on salt elemental, ionic, and radiological composition to aid in assessment of the suitability of processing drained and dissolved material and in refining the information in the waste characterization system (WCS)

  12. Test Objectives for the Saltcake Dissolution Retrieval Demonstration

    International Nuclear Information System (INIS)

    DEFIGH PRICE, C.

    2000-01-01

    This document describes the objectives the Saltcake Dissolution Retrieval Demonstration. The near term strategy for single-shell tank waste retrieval activities has shifted from focusing on maximizing the number of tanks entered for retrieval (regardless of waste volume or content) to a focus on scheduling the retrieval of wastes from those single-shell tanks with a high volume of contaminants of concern. These contaminants are defined as mobile, long-lived radionuclides that have a potential of reaching the groundwater and the Columbia River. This strategy also focuses on the performance of key retrieval technology demonstrations, including the Saltcake Dissolution Retrieval Demonstration, in a variety of waste forms and tank farm locations to establish a technical basis for future work. The work scope will also focus on the performance of risk assessment, retrieval performance evaluations (RPE) and incorporating vadose zone characterization data on a tank-by-tank basis, and on updating tank farm closure/post closure work plans. The deployment of a retrieval technology other than Past-Practice Sluicing (PPS) allows determination of limits of technical capabilities, as well as, providing a solid planning basis for future SST retrievals. This saltcake dissolution technology deployment test will determine if saltcake dissolution is a viable retrieval option for SST retrieval. CH2M Hill Hanford Group (CHG) recognizes the SST retrieval mission is key to the success of the River Protection Project (RPP) and the overall completion of the Hanford Site cleanup. The objectives outlined in this document will be incorporated into and used to develop the test and evaluation plan for saltcake dissolution retrievals. The test and evaluation plan will be developed in fiscal year 2001

  13. A simplified model of saltcake moisture distribution. Letter report

    International Nuclear Information System (INIS)

    Simmons, C.S.

    1995-09-01

    This letter report describes the formulation of a simplified model for finding the moisture distribution in a saltcake waste profile that has been stabilized by pumping out the drainable interstitial liquid. The model is based on assuming that capillarity mainly governs the distribution of moisture in the porous saltcake waste. A stead upward flow of moisture driven by evaporation from the waste surface is conceptualized to occur for isothermal conditions. To obtain hydraulic parameters for unsaturated conditions, the model is calibrated or matched to the relative saturation distribution as measured by neutron probe scans. The model is demonstrated on Tanks 104-BY and 105-TX as examples. A value of the model is that it identifies the key physical parameters that control the surface moisture content in a waste profile. Moreover, the model can be used to estimate the brine application rate at the waste surface that would raise the moisture content there to a safe level. Thus, the model can be applied to help design a strategy for correcting the moisture conditions in a saltcake waste tank

  14. Tanks 3F and 2F Saltcake Core and Supernate Sample Analysis

    International Nuclear Information System (INIS)

    MARTINO, CHRISTOPHERJ

    2004-01-01

    In support of Low-Curie Salt (LCS) process validation at the Savannah River Site (SRS), Liquid Waste Disposition (LWD) has undertaken a program of tank waste characterization, including salt sampling. As part of this initiative, they sampled the surface of the saltcake in Tank 3F and Tank 2F using approximately 12-inch long sample tubes. A series of three saltcake samples were taken of the upper crust in Tank 3F and a single saltcake sample was taken from the bottom of a liquid-filled well in Tank 2F. In addition to analysis of the solid saltcake samples, the liquid contained in the Tank 3F samples and a separate supernate sample from Tank 2F were studied. The primary objective of the characterization is to gather information that will be useful to the selection and processing of the next waste tanks. Most important is the determination of the 137Cs concentration and liquid retention properties of Tank 3F and Tank 2F saltcake to enable projection of drained, dissolved salt composition. Additional information will aid in refining the waste characterization system (WCS) and could assist the eventual salt treatment or processing

  15. Saltcake dissolution FY 1998 status report

    International Nuclear Information System (INIS)

    HERTING, D.L.

    1999-01-01

    A laboratory scouting study was completed on the dissolution characteristics of Hanford waste from three single-shell waste tanks: 241-BY-102, 241-BY-106, and 241-B-106. Gross dissolution behavior (percent undissolved solids as a function of dilution) is explained in terms of characteristics of individual salts in the waste. The percentage of the sodium inventory retrievable from the tanks by dissolving saltcake at reasonable dilution levels is estimated at 86% of the total sodium for tank BY-102, 98% for BY-106, and 79% for B-106

  16. Process control plan for Single Shell Tank (SST) Saltcake Dissolution Proof of Concept

    International Nuclear Information System (INIS)

    ESTEY, S.D.

    2001-01-01

    This document describes the process controls for the tank 241-U-107 (U-107) saltcake dissolution proof-of-concept operations. Saltcake dissolution is defined as a method by which water-soluble salts will be retrieved from the Hanford Site radioactive waste tanks utilizing dissolution as the mobilizing mechanism. The proof-of-concept operations will monitor the retrieval process and transfer at least 100 kgal of fluid from tank U-107 to the double-shell tank (DST) system during the performance period. Tank U-107 has been identified as posing the highest long-term risk to the Columbia River of all single shell tanks (SSTs). This is because of the high content of mobile, long-lived radionuclides mostly in the saltcake waste in the tank. To meet current contractual and consent decree commitments, tank U-107 is being prepared for interim stabilization in August 2001. It is currently scheduled for saltcake retrieval in 2023, near the end of the SST retrieval campaign because of a lack of infrastructure in U-Farm. The proof-of-concept test will install a system to dissolve and retrieve a portion of the saltcake as part of, and operating in parallel with, the standard interim stabilization system to be installed on tank U-107. This proof-of-concept should provide key information on spray nozzle selection and effective spray patterns, leak detection, monitoring, and mitigation (LDMM) and in-tank saltcake solubility data that will help in the design of a full-tank retrieval demonstration system

  17. Characterization and Leach Testing for REDOX Sludge and S-Saltcake Actual Waste Sample Composites

    International Nuclear Information System (INIS)

    Fiskum, Sandra K.; Buck, Edgar C.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hubler, Timothy L.; Jagoda, Lynette K.; Jenson, Evan D.; Kozelisky, Anne E.; Lumetta, Gregg J.; MacFarlan, Paul J.; McNamara, Bruce K.; Peterson, Reid A.; Sinkov, Sergey I.; Snow, Lanee A.; Swoboda, Robert G.

    2008-01-01

    This report describes processing and analysis results of boehmite waste type (Group 5) and insoluble high Cr waste type (Group 6). The sample selection, compositing, subdivision, physical and chemical characterization are described. Extensive batch leach testing was conducted to define kinetics and leach factors of selected analytes as functions of NaOH concentration and temperature. Testing supports issue M-12 resolution for the Waste Treatment Plant

  18. Characterization and Leach Testing for REDOX Sludge and S-Saltcake Actual Waste Sample Composites

    Energy Technology Data Exchange (ETDEWEB)

    Fiskum, Sandra K.; Buck, Edgar C.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hubler, Timothy L.; Jagoda, Lynette K.; Jenson, Evan D.; Kozelisky, Anne E.; Lumetta, Gregg J.; MacFarlan, Paul J.; McNamara, Bruce K.; Peterson, Reid A.; Sinkov, Sergey I.; Snow, Lanee A.; Swoboda, Robert G.

    2008-07-10

    This report describes processing and analysis results of boehmite waste type (Group 5) and insoluble high Cr waste type (Group 6). The sample selection, compositing, subdivision, physical and chemical characterization are described. Extensive batch leach testing was conducted to define kinetics and leach factors of selected analytes as functions of NaOH concentration and temperature. Testing supports issue M-12 resolution for the Waste Treatment Plant.

  19. ANALYSIS OF TANK 28F SALTCAKE CORE SAMPLES FTF-456 - 467

    Energy Technology Data Exchange (ETDEWEB)

    Martino, C; Daniel McCabe, D; Tommy Edwards, T; Ralph Nichols, R

    2007-02-28

    Twelve LM-75 core samplers from Tank 28F sampling were received by SRNL for saltcake characterization. Of these, nine samplers contained mixtures of free liquid and saltcake, two contained only liquid, and one was empty. The saltcake contents generally appeared wet. A summary of the major tasks performed in this work are as follows: (1) Individual saltcake segments were extruded from the samplers and separated into saltcake and free liquid portions. (2) Free liquids were analyzed to estimate the amount of traced drill-string fluid contained in the samples. (3) The saltcake from each individual segment was homogenized, followed by analysis in duplicate. The analysis used more cost-effective and bounding radiochemical analyses rather than using the full Saltstone WAC suite. (4) A composite was created using an approximately equal percentage of each segment's saltcake contents. Supernatant liquid formed upon creation of the composite was decanted prior to use of the composite, but the composite was not drained. (5) A dissolution test was performed on the sample by contacting the composite with water at a 4:1 mass ratio of water to salt. The resulting soluble and insoluble fractions were analyzed. Analysis focused on a large subset of the Saltstone WAC constituents.

  20. Determination of uranium distribution in the evaporation of simulated Savannah River Site waste

    International Nuclear Information System (INIS)

    Barnes, M.J.; Chandler, G.T.

    1995-01-01

    The results of an experimental program addressing the distribution of uranium in saltcake and supernate for two Savannah River Site waste compositions are presented. Successive batch evaporations were performed on simulated H-Area Modified Purex low-heat and post-aluminum dissolution wastes spiked with depleted uranium. Waste compositions and physical data were obtained for supernate and saltcake samples. For the H-Area Modified Purex low-heat waste, the product saltcake contained 42% of the total uranium from the original evaporator feed solution. However, precipitated solids only accounted for 10% of the original uranium mass; the interstitial liquid within the saltcake matrix contained the remainder of the uranium. In the case of the simulated post-aluminum dissolution waste; the product saltcake contained 68% of the total uranium from the original evaporator feed solution. Precipitated solids accounted for 52% of the original uranium mass; again, the interstitial liquid within the saltcake matrix contained the remainder of the uranium. An understanding of the distribution of uranium between supernatant liquid, saltcake, and sludge is required to develop a material balance for waste processing operations. This information is necessary to address nuclear criticality safety concerns

  1. Removal of 137-Cs from Dissolved Hanford Tank Saltcake by Treatment with IE-911

    International Nuclear Information System (INIS)

    Rapko, Brian M.; Sinkov, Sergei I.; Levitskaia, Tatiana G.

    2003-01-01

    The U.S. Department of Energy's Richland Operations Office plans to accelerate the cleanup of the Hanford Site. Testing new technology for the accelerated cleanup will require dissolved saltcake from single-shell tanks. However, the 137Cs will need to be removed from the saltcake to alleviate radiation hazards. A saltcake composite constructed from archived samples from Hanford Site single-shell tanks 241-S-101, 241-S-109, 241-S-110, 241-S-111, 241-U-106, and 241-U-109 was dissolved in water, adjusted to 5 M Na, and transferred from the 222-S Laboratory to the Radiochemical Processing Laboratory (RPL). At the RPL, the approximately 5.5 liters of solution was passed through a 0.2-micron polyethersulfone filter, collected, and homogenized. The filtered solution then was passed through an ion exchange column containing approximately 150 mL IONSIV(reg s ign) IE-911, an engineered form of crystalline silicotitanate available from UOP, at approximately 200 mL/hour in a continuous operation until all of the feed solution had been run through the column. An analysis of the 137Cs concentrations in the initial feed solution and combined column effluent indicates that > 99.999 percent of the Cs in the feed solution was removed by this operation

  2. Test procedures and instructions for single shell tank saltcake cesium removal with crystalline silicotitanate

    Energy Technology Data Exchange (ETDEWEB)

    Duncan, J.B.

    1997-01-07

    This document provides specific test procedures and instructions to implement the test plan for the preparation and conduct of a cesium removal test, using Hanford Single Shell Tank Saltcake from tanks 24 t -BY- I 10, 24 1 -U- 108, 24 1 -U- 109, 24 1 -A- I 0 1, and 24 t - S-102, in a bench-scale column. The cesium sorbent to be tested is crystalline siticotitanate. The test plan for which this provides instructions is WHC-SD-RE-TP-024, Hanford Single Shell Tank Saltcake Cesium Removal Test Plan.

  3. Optimization of the Retrieval of Waste from Hanford Tank S-109 through Numerical Modeling

    International Nuclear Information System (INIS)

    Patel, R.; Tachiev, G.; Mulchandani, A.; Roelant, D.

    2009-01-01

    This report covers 10 different retrieval scenarios to support the U.S. Department of Energy's Office of River Protection in its mission to facilitate the retrieval and treatment of high-level radioactive waste stored in underground tanks at the Hanford site by investigating the transport properties of the salt-cake. Salt-cake consists of salts precipitated out of the brines during evaporation and storage. The main objective of this study is to gain a better understanding of the dissolution process that will occur in Tank 241-S-109 as it is retrieved to provide waste for Vitrification at the Demonstration Bulk Vitrification System Facility (DBVS). Double Shell Tank (DST) space is extremely limited and will continue to be until the Waste Treatment Plant becomes operational. Maximizing the utilization of DST space is the goal of the S-109 Partial Waste Retrieval Project that will provide waste feed to the Demonstration Bulk Vitrification System (DBVS). Florida International University, FIU has developed a 2-D axisymmetric numerical model which will assist the Department of Energy (DOE) and Savannah River Site (SRS) in evaluating the potential of selective salt-cake retrieval for schedule acceleration and significant cost savings by analyzing the performance of different retrieval scenarios with the prediction of Cs breakthrough curves in the resulting salt-cake brine and to determine the displacement patterns of Cs. This predictive information is critical for scheduling and operational purposes. Ten retrieval scenarios which include addition of flushing liquid at the entire surface of the tank or at a side peripheral channel were simulated. All retrieval scenarios were analyzed for incremental retrieval (saturation of the tank with flushing liquid followed by complete drainage at the central well) versus continuous retrieval (water is continuously added at the top and retrieved at a central well). Furthermore, the specifics of the tank hydrology were approximated

  4. Caustic-Side Solvent Extraction: Prediction of Cesium Extraction from Actual Wastes and Actual Waste Simulants

    International Nuclear Information System (INIS)

    Delmau, L.H.; Haverlock, T.J.; Sloop, F.V. Jr.; Moyer, B.A.

    2003-01-01

    This report presents the work that followed the CSSX model development completed in FY2002. The developed cesium and potassium extraction model was based on extraction data obtained from simple aqueous media. It was tested to ensure the validity of the prediction for the cesium extraction from actual waste. Compositions of the actual tank waste were obtained from the Savannah River Site personnel and were used to prepare defined simulants and to predict cesium distribution ratios using the model. It was therefore possible to compare the cesium distribution ratios obtained from the actual waste, the simulant, and the predicted values. It was determined that the predicted values agree with the measured values for the simulants. Predicted values also agreed, with three exceptions, with measured values for the tank wastes. Discrepancies were attributed in part to the uncertainty in the cation/anion balance in the actual waste composition, but likely more so to the uncertainty in the potassium concentration in the waste, given the demonstrated large competing effect of this metal on cesium extraction. It was demonstrated that the upper limit for the potassium concentration in the feed ought to not exceed 0.05 M in order to maintain suitable cesium distribution ratios

  5. Hanford Waste Physical and Rheological Properties: Data and Gaps

    Energy Technology Data Exchange (ETDEWEB)

    Wells, Beric E.; Kurath, Dean E.; Mahoney, Lenna A.; Onishi, Yasuo; Huckaby, James L.; Cooley, Scott K.; Burns, Carolyn A.; Buck, Edgar C.; Tingey, Joel M.; Daniel, Richard C.; Anderson, K. K.

    2011-08-01

    The Hanford Site in Washington State manages 177 underground storage tanks containing approximately 250,000 m3 of waste generated during past defense reprocessing and waste management operations. These tanks contain a mixture of sludge, saltcake and supernatant liquids. The insoluble sludge fraction of the waste consists of metal oxides and hydroxides and contains the bulk of many radionuclides such as the transuranic components and 90Sr. The saltcake, generated by extensive evaporation of aqueous solutions, consists primarily of dried sodium salts. The supernates consist of concentrated (5-15 M) aqueous solutions of sodium and potassium salts. The 177 storage tanks include 149 single-shell tanks (SSTs) and 28 double -hell tanks (DSTs). Ultimately the wastes need to be retrieved from the tanks for treatment and disposal. The SSTs contain minimal amounts of liquid wastes, and the Tank Operations Contractor is continuing a program of moving solid wastes from SSTs to interim storage in the DSTs. The Hanford DST system provides the staging location for waste feed delivery to the Department of Energy (DOE) Office of River Protection’s (ORP) Hanford Tank Waste Treatment and Immobilization Plant (WTP). The WTP is being designed and constructed to pretreat and then vitrify a large portion of the wastes in Hanford’s 177 underground waste storage tanks.

  6. Waste Preparation and Transport Chemistry: Results of the FY 2001 Studies

    Energy Technology Data Exchange (ETDEWEB)

    Hunt, R.D.

    2002-03-25

    of researchers from AEA Technology, Florida International University (FIU), Fluor Hanford, Mississippi State University (MSU), Oak Ridge National Laboratory (ORNL), and Savannah River Technology Center (SRTC) to evaluate various aspects of the waste preparation and transport chemistry. The majority of this effort was focused on saltcake dissolution and saltwell pumping. The results of the AEA Technology, FIU, and MSU studies of saltcake dissolution and slurry transfers for Hanford are discussed in detail in a companion report prepared by T. D. Welch in 2001 (ORNIJTM-2001097). Staff members at Fluor Hanford have continued to conduct saltcake dissolution tests on actual tank waste (documented in reports prepared by D. L. Herting in 2000 and 2001). It should be noted that full-scale saltcake dissolution at Hanford is scheduled to begin in FY 2002. While the Hanford effort is focused on the transfer of waste from one tank to another, the objective of the SRTC study is the formation of aluminosilicates at elevated temperatures, which are present in the waste evaporator.

  7. Alternatives for definse waste-salt disposal

    International Nuclear Information System (INIS)

    Benjamin, R.W.; McDonell, W.R.

    1983-01-01

    Alternatives for disposal of decontaminated high-level waste salt at Savannah River were reviewed to estimate costs and potential environmental impact for several processes. In this review, the reference process utilizing intermediate-depth burial of salt-concrete (saltcrete) monoliths was compared with alternatives including land application of the decontaminated salt as fertilizer for SRP pine stands, ocean disposal with and without containment, and terminal storage as saltcake in existing SRP waste tanks. Discounted total costs for the reference process and its modifications were in the same range as those for most of the alternative processes; uncontained ocean disposal with truck transport to Savannah River barges and storage as saltcake in SRP tanks had lower costs, but presented other difficulties. Environmental impacts could generally be maintained within acceptable limits for all processes except retention of saltcake in waste tanks, which could result in chemical contamination of surrounding areas on tank collapse. Land application would require additional salt decontamination to meet radioactive waste disposal standards, and ocean disposal without containment is not permitted in existing US practice. The reference process was judged to be the only salt disposal option studied which would meet all current requirements at an acceptable cost

  8. Baseline Design Compliance Matrix for the Type 4 In Situ Vapor Samplers and Supernate and Sludge and Soft Saltcake Grab Sampling

    International Nuclear Information System (INIS)

    BOGER, R.M.

    2000-01-01

    The DOE has identified a need to sample vapor space, exhaust ducts, supernate, sludge, and soft saltcake in waste tanks that store radioactive waste. This document provides the Design Compliance Matrix (DCM) for the Type 4 In-Situ Vapor Sampling (ISVS) system and the Grab Sampling System that are used for completing this type of sampling function. The DCM identifies the design requirements and the source of the requirements for the Type 4 ISVS system and the Grab Sampling system. The DCM is a single-source compilation design requirements for sampling and sampling support equipment and supports the configuration management of these systems

  9. FRACTIONAL CRYSTALLIZATION FLOWSHEET TESTS WITH ACTUAL TANK WASTE

    International Nuclear Information System (INIS)

    HERTING, D.L.

    2006-01-01

    Laboratory-scale flowsheet tests of the fractional crystallization process were conducted with actual tank waste samples in a hot cell at the 222-S Laboratory. The process is designed to separate medium-curie liquid waste into a low-curie stream for feeding to supplemental treatment and a high-curie stream for double-shell tank storage. Separations criteria (for Cs-137 sulfate, and sodium) were exceeded in all three of the flowsheet tests that were performed

  10. Probabilistic safety assessment for Hanford high-level waste tanks

    International Nuclear Information System (INIS)

    MacFarlane, D.R.; Stack, D.S.; Kindinger, J.P.; Deremer, R.K.

    1995-01-01

    This paper gives results from the first comprehensive level-3 probabilistic safety assessment (PSA), including consideration of external events, for the Hanford tank farm (HTF). This work was sponsored by the U.S. Department of Energy/Environmental Restoration and Waste Management Division (DOE/EM). At the HTF, there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/saltcake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is ∼60 million gal, containing ∼200 million Ci of radioactivity

  11. Axial Dispersion during Hanford Saltcake Washing

    International Nuclear Information System (INIS)

    Josephson, Gary B.; Geeting, John GH; Lessor, Delbert L.; Barton, William B.

    2006-01-01

    Clean up of Hanford salt cake wastes begins with dissolution retrieval of the sodium rich salts that make up the dominant majority of mass in the tanks. Water moving through the porous salt cake dissolves the soluble components and also displaces the soluble radionuclides (e.g. 137Cs and 99TcO4- ). The separation that occurs from this displacement, known as Selective dissolution, is an important component in Hanford?s pretreatment of low activity wastes for subsequent Supplemental treatment. This paper describes lab scale testing conducted to evaluate Selective dissolution of cesium from non-radioactive Hanford tank 241-S-112 salt cake simulant containing the primary chemicals found the actual tank. An modified axial dispersion model with increasing axial dispersion was developed to predict cesium removal. The model recognizes that water dissolves the salt cake during washing, which causes an increase in the axial dispersion during the wash. This model was subsequently compared with on-line cesium measurements from the retrieval of tank 241-S-112. The model had remarkably good agreement with both the lab scale and full scale data

  12. Cesium ion exchange using actual waste: Column size considerations

    International Nuclear Information System (INIS)

    Brooks, K.P.

    1996-04-01

    It is presently planned to remove cesium from Hanford tank waste supernates and sludge wash solutions using ion exchange. To support the development of a cesium ion exchange process, laboratory experiments produced column breakthrough curves using wastes simulants in 200 mL columns. To verify the validity of the simulant tests, column runs with actual supernatants are being planned. The purpose of these actual waste tests is two-fold. First, the tests will verify that use of the simulant accurately reflects the equilibrium and rate behavior of the resin compared to actual wastes. Batch tests and column tests will be used to compare equilibrium behaviors and rate behaviors, respectively. Second, the tests will assist in clarifying the negative interactions between the actual waste and the ion exchange resin, which cannot be effectively tested with simulant. Such interactions include organic fouling of the resin and salt precipitation in the column. These effects may affect the shape of the column breakthrough curve. The reduction in column size also may change the shape of the curve, making the individual effects even more difficult to sort out. To simplify the evaluation, the changes due to column size must be either understood or eliminated. This report describes the determination of the column size for actual waste testing that best minimizes the effect of scale-down. This evaluation will provide a theoretical basis for the dimensions of the column. Experimental testing is still required before the final decision can be made. This evaluation will be confined to the study of CS-100 and R-F resins with NCAW simulant and to a limited extent DSSF waste simulant. Only the cesium loading phase has been considered

  13. Quantity assessment of waste in the dismantlement of liquid waste treatment plant and its actual state

    International Nuclear Information System (INIS)

    Uchiyama, Takafumi; Mitsuhashi, Ishi; Matsumoto, Tetsuo; Morishima, Kayoko; Tanzawa, Tomio

    2016-01-01

    From the progress of decommissioning project work of Tokyo City University Atomic Energy Research Institute, this paper reports the comparison between the actual amount of the waste generated during dismantlement work at liquid waste treatment facilities and the assessment quantity before starting the dismantlement. The quantity assessment was made on the basis of the installation license application, design specifications, drawings, records, history of use, site investigation results, etc. Since this quantity assessment did not take into account the dismantling contents of reservoir concrete, the assessed quantity of non-radioactive waste (NR waste) did not match the sum of actual NR waste. However, if an actually generated quantity of concrete of radioactive waste was added to the quantity assessment as NR waste, the quantity of actually generated NR waste and that of assessed NR waste were nearly consistent, which verified the validity of this assessment. This method is considered to be able to be utilized in the future quantity assessment of decommissioning work and the like. On the other hand, it was found that the number of drums that were actually stored tended to increase more than the estimated number of drum conversion. In old buildings, it is necessary to take into account the generation of waste other than radioactive materials in the quantity assessment stage and dismantlement stage. (A.O.)

  14. Advanced liquid waste processing technologies: Theoretical versus actual application

    International Nuclear Information System (INIS)

    Barker, Tracy A.

    1992-01-01

    This paper provides an overview of Chem-Nuclear Systems, Inc. (CNSI) experience with turn-key chromate removal at the Maine Yankee Nuclear Plant. Theoretical and actual experiences are addressed on topics such as processing duration, laboratory testing, equipment requirements, chromate removal, waste generation, and waste processing. Chromate salts are used in industrial recirculation cooling water systems as a corrosion inhibitor. However, chromates are toxic at concentrations necessary for surface inhibition. As a result, Chem-Nuclear was contracted to perform turn-key chromate removal and waste disposal by demineralization. This project was unique in that prior to on-site mobilization, a composite sample of chromated waste was shipped to CNSI laboratories for treatment through a laboratory scale system. Removal efficiency, process media requirements, and waste processing methodology were determined from this laboratory testing. Samples of the waste resulting from this testing were processed by dewatering and solidification, respectively. TCLP tests were performed on the actual processed waste, and based on the TCLP results, pre-approval for media waste disposal was obtained. (author)

  15. The Hanford Site Tank Waste Remediation System: An update

    International Nuclear Information System (INIS)

    Alumkal, W.T.; Babad, H.; Harmon, H.D.; Wodrich, D.D.

    1994-01-01

    The U.S. Department of Energy's Hanford Site, located in southeastern Washington State, has the most diverse and largest amount of highly radioactive waste in the United States. High-level radioactive waste has been stored in large underground tanks since 1944. Approximately 230,000 m 3 (61 Mgal) of caustic liquids, slurries, saltcakes, and sludges have 137 Cs accumulated in 177 tanks. In addition, significant amounts of 90 Sr and were removed from the tank waste, converted to salts, doubly encapsulated in metal containers., and stored in water basins. A Tank Waste Remediation System Program was established by the U.S. Department of Energy in 1991 to safely manage and immobilize these wastes in anticipation of permanent disposal of the high-level waste fraction in a geologic repository. Since 1991, progress has been made resolving waste tank safety issues, upgrading Tank Farm facilities and operations, and developing a new strategy for retrieving, treating, and immobilizing the waste for disposal

  16. Organic Tank Safety Project: Effect of water partial pressure on the equilibrium water content of waste samples from Hanford Tank 241-U-105

    International Nuclear Information System (INIS)

    Scheele, R.D.; Bredt, P.R.; Sell, R.L.

    1997-09-01

    Water content plays a crucial role in the strategy developed by Webb et al. to prevent propagating or sustainable chemical reactions in the organic-bearing wastes stored in the 20 Organic Tank Watch List tanks at the U.S. Department of Energy''s Hanford Site. Because of water''s importance in ensuring that the organic-bearing wastes continue to be stored safely, Duke Engineering and Services Hanford commissioned the Pacific Northwest National Laboratory to investigate the effect of water partial pressure (P H2O ) on the water content of organic-bearing or representative wastes. Of the various interrelated controlling factors affecting the water content in wastes, P H2O is the most susceptible to being controlled by the and Hanford Site''s environmental conditions and, if necessary, could be managed to maintain the water content at an acceptable level or could be used to adjust the water content back to an acceptable level. Of the various waste types resulting from weapons production and waste-management operations at the Hanford Site, determined that saltcake wastes are the most likely to require active management to maintain the wastes in a Conditionally Safe condition. Webb et al. identified Tank U-105 as a Conditionally Safe saltcake tank. A Conditionally Safe waste is one that is currently safe based on waste classification criteria but could, if dried, be classified as open-quotes Unsafe.close quotes To provide information on the behavior of organic-bearing wastes, the Westinghouse Hanford Company provided us with four waste samples taken from Tank 241-U-105 (U-105) to determine the effect of P H2O on their equilibrium water content

  17. Characterization, Leaching, and Filtration Testing for Tributyl Phosphate (TBP, Group 7) Actual Waste Sample Composites

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, Matthew K.; Billing, Justin M.; Blanchard, David L.; Buck, Edgar C.; Casella, Amanda J.; Casella, Andrew M.; Crum, J. V.; Daniel, Richard C.; Draper, Kathryn E.; Fiskum, Sandra K.; Jagoda, Lynette K.; Jenson, Evan D.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Shimskey, Rick W.; Snow, Lanee A.; Swoboda, Robert G.

    2009-03-09

    .A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. The actual waste-testing program included homogenizing the samples by group, characterizing the solids and aqueous phases, and performing parametric leaching tests. The tributyl phosphate sludge (TBP, Group 7) is the subject of this report. The Group 7 waste was anticipated to be high in phosphorus as well as aluminum in the form of gibbsite. Both are believed to exist in sufficient quantities in the Group 7 waste to address leaching behavior. Thus, the focus of the Group 7 testing was on the removal of both P and Al. The waste-type definition, archived sample conditions, homogenization activities, characterization (physical, chemical, radioisotope, and crystal habit), and caustic leaching behavior as functions of time, temperature, and hydroxide concentration are discussed in this report. Testing was conducted according to TP-RPP-WTP-467.

  18. Actual problems of municipal cleaner’s waste waters

    Directory of Open Access Journals (Sweden)

    Konko¾ová Patrícia

    2000-03-01

    Full Text Available In paper are evaluated social and economical changes in water economy with emphasis on complex evaluation of municipal cleaner’s waste waters with respect of legislative, position of ownerskip relationskips and financial security of public experiences of water economy.

  19. Demonstration of retrieval methods for Westinghouse Hanford Corporation October 20, 1995

    International Nuclear Information System (INIS)

    1996-10-01

    Westinghouse Hanford Corporation has been pursuing strategies to break up and retrieve the radioactive waste material in single shell storage tanks at the Hanford Nuclear Reservation, by working with non-radioactive ''saltcake'' and sludge material that simulate the actual waste. It has been suggested that the use of higher volumes of water than used in the past (10 gpm nozzles at 10,000 psi) might be successful in breaking down the hard waste simulants. Additionally, the application of these higher volumes of water might successfully be applied through commercially available tooling using methods similar to those used in the deslagging of large utility boilers. NMW Industrial Services, Inc., has proposed a trial consisting of three approaches each to dislodging both the solid (saltcake) simulant and the sludge simulant

  20. Demonstration of retrieval methods for Westinghouse Hanford Corporation October 20, 1995

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-10-01

    Westinghouse Hanford Corporation has been pursuing strategies to break up and retrieve the radioactive waste material in single shell storage tanks at the Hanford Nuclear Reservation, by working with non-radioactive ``saltcake`` and sludge material that simulate the actual waste. It has been suggested that the use of higher volumes of water than used in the past (10 gpm nozzles at 10,000 psi) might be successful in breaking down the hard waste simulants. Additionally, the application of these higher volumes of water might successfully be applied through commercially available tooling using methods similar to those used in the deslagging of large utility boilers. NMW Industrial Services, Inc., has proposed a trial consisting of three approaches each to dislodging both the solid (saltcake) simulant and the sludge simulant.

  1. Organic analyses of an actual and simulated mixed waste. Hanford's organic complexant waste revisited

    International Nuclear Information System (INIS)

    Toste, A.P.; Osborn, B.C.; Polach, K.J.; Lechner-Fish, T.J.

    1995-01-01

    Reanalysis of the organics in a mixed waste, an organic complexant waste, from the U.S. Department of Energy's Hanford Site, has yielded an 80.4% accounting of the waste's total organic content. In addition to several complexing and chelating agents (citrate, EDTA, HEDTA and NTA), 38 chelator/complexor fragments have been identified, compared to only 11 in the original analysis, all presumably formed via organic degradation. Moreover, a mis identification, methanetricarboxylic acid, has been re-identified as the chelator fragment N-(methylamine)imino-diacetic acid (MAIDA). A nonradioactive simulant of the actual waste, containing the parent organics (citrate, EDTA, HEDTA and NTA), was formulated and stored in the dark at ambient temperature for 90 days. Twenty chelator and complexor fragments were identified in the simulant, along with several carboxylic acids, confirming that myriad chelator and complexor fragments are formed via degradation of the parent organics. Moreover, their abundance in the simulant (60.9% of the organics identified) argues that the harsh chemistries of mixed wastes like Hanford's organic degradation, even in the absence of radiation. (author). 26 refs., 2 tabs

  2. Organic tanks safety program waste aging studies. Final report, Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Camaioni, D.M.; Samuels, W.D.; Linehan, J.C. [and others

    1998-09-01

    Uranium and plutonium production at the Hanford Site produced large quantities of radioactive byproducts and contaminated process chemicals that are stored in underground tanks awaiting treatment and disposal. Having been made strongly alkaline and then subjected to successive water evaporation campaigns to increase storage capacity, the wastes now exist in the physical forms of saltcakes, metal oxide sludges, and aqueous brine solutions. Tanks that contain organic process chemicals mixed with nitrate/nitrite salt wastes might be at risk for fuel-nitrate combustion accidents. This project started in fiscal year 1993 to provide information on the chemical fate of stored organic wastes. While historical records had identified the organic compounds originally purchased and potentially present in wastes, aging experiments were needed to identify the probable degradation products and evaluate the current hazard. The determination of the rates and pathways of degradation have facilitated prediction of how the hazard changes with time and altered storage conditions. Also, the work with aged simulated waste contributed to the development of analytical methods for characterizing actual wastes. Finally, the results for simulants provide a baseline for comparing and interpreting tank characterization data.

  3. Organic tanks safety program waste aging studies. Final report, Revision 1

    International Nuclear Information System (INIS)

    Camaioni, D.M.; Samuels, W.D.; Linehan, J.C.

    1998-09-01

    Uranium and plutonium production at the Hanford Site produced large quantities of radioactive byproducts and contaminated process chemicals that are stored in underground tanks awaiting treatment and disposal. Having been made strongly alkaline and then subjected to successive water evaporation campaigns to increase storage capacity, the wastes now exist in the physical forms of saltcakes, metal oxide sludges, and aqueous brine solutions. Tanks that contain organic process chemicals mixed with nitrate/nitrite salt wastes might be at risk for fuel-nitrate combustion accidents. This project started in fiscal year 1993 to provide information on the chemical fate of stored organic wastes. While historical records had identified the organic compounds originally purchased and potentially present in wastes, aging experiments were needed to identify the probable degradation products and evaluate the current hazard. The determination of the rates and pathways of degradation have facilitated prediction of how the hazard changes with time and altered storage conditions. Also, the work with aged simulated waste contributed to the development of analytical methods for characterizing actual wastes. Finally, the results for simulants provide a baseline for comparing and interpreting tank characterization data

  4. Radioactive tank waste remediation focus area

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-08-01

    EM`s Office of Science and Technology has established the Tank Focus Area (TFA) to manage and carry out an integrated national program of technology development for tank waste remediation. The TFA is responsible for the development, testing, evaluation, and deployment of remediation technologies within a system architecture to characterize, retrieve, treat, concentrate, and dispose of radioactive waste stored in the underground stabilize and close the tanks. The goal is to provide safe and cost-effective solutions that are acceptable to both the public and regulators. Within the DOE complex, 335 underground storage tanks have been used to process and store radioactive and chemical mixed waste generated from weapon materials production and manufacturing. Collectively, thes tanks hold over 90 million gallons of high-level and low-level radioactive liquid waste in sludge, saltcake, and as supernate and vapor. Very little has been treated and/or disposed or in final form.

  5. Quantitative measurement of cyanide complexes in simulated and actual Hanford ferrocyanide wastes

    International Nuclear Information System (INIS)

    Bryan, S.A.; Pool, K.H.; Sell, R.L.; Bryan, S.L.

    1994-01-01

    Cyanide-containing radioactive waste from radiocesium scavenging processes conducted during the 1950's at Hanford is currently stored in 24 single shell tanks. As part of ongoing tank characterization efforts, the quantity and chemical form of cyanide in these tanks need to be determined. This report summarizes the results of studies conducted at Pacific Northwest Laboratory (PNL) under contract to Westinghouse Hanford Company (WHC) to develop methods for the quantification of total cyanide and identification of major cyanide-containing species in Ferrocyanide Tank Waste. Results from the application of FTIR, IC, and microdistillation procedures to simulated and actual Hanford waste are presented and compared where applicable

  6. Actual situation of radioactive waste management in Guatemala

    International Nuclear Information System (INIS)

    Gomez O, P.

    1996-01-01

    This paper focuses on the actual radioactive waste management situation in Guatemala, as well as on the sources and facilities that have obtained the license, and the way to disposal them, when they are considered as radioactive waste. The Direccion General de Energia Nuclear is the entity responsible for the proper and normal performance of the regulatory activity in the country. (author). 3 refs

  7. Organic tank safety project: Effect of water partial pressure on the equilibrium water contents of waste samples from Hanford Tank 241-BY-108

    International Nuclear Information System (INIS)

    Scheele, R.D.; Bredt, P.R.; Sell, R.L.

    1997-02-01

    Water content plays a crucial role in the strategy developed by Webb et al. to prevent propagating or sustainable chemical reactions in the organic-bearing wastes stored in the 20 Organic Tank Watch List tanks at the US Department of Energy's Hanford Site. Because of water's importance in ensuring that the organic-bearing wastes continue to be stored safely, Duke Engineering and Services Hanford commissioned the Pacific Northwest National Laboratory (PNNL) to investigate the effect of water partial pressure (P H2O ) on the water content of organic-bearing or representative wastes. Of the various interrelated controlling factors affecting the water content in wastes, P H2O is the most susceptible to being controlled by the and Hanford Site's environmental conditions and, if necessary, could be managed to maintain the water content at an acceptable level or could be used to adjust the water content back to an acceptable level. Of the various waste types resulting from weapons production and waste-management operations at the Hanford Site, Webb et al. determined that saltcake wastes are the most likely to require active management to maintain the wastes in a Conditionally Safe condition. A Conditionally Safe waste is one that satisfies the waste classification criteria based on water content alone or a combination of water content and either total organic carbon (TOC) content or waste energetics. To provide information on the behavior of saltcake wastes, two waste samples taken from Tank 241-BY-108 (BY-108) were selected for study, even though BY-108 is not on the Organic Tanks Watch List because of their ready availability and their similarity to some of the organic-bearing saltcakes

  8. Treatability studies of actual listed waste sludges from the Oak Ridge Reservation (ORR)

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Peeler, D.K.; Gilliam, T.M.; Bleier, A.; Spence, R.D.

    1996-01-01

    Oak Ridge National Laboratory (ORNL) and Savannah River Technology Center (SRTC) are investigating vitrification for various low-level and mixed wastes on the Oak Ridge Reservation (ORR). Treatability studies have included surrogate waste formulations at the laboratory-, pilot-, and field-scales and actual waste testing at the laboratory- and pilot-scales. The initial waste to be processing through SRTC's Transportable Vitrification System (TVS) is the K-1407-B and K-1407-C (B/C) Pond sludge waste which is a RCRA F-listed waste. The B/C ponds at the ORR K-25 site were used as holding and settling ponds for various waste water treatment streams. Laboratory-, pilot-, and field- scale ''proof-of-principle'' demonstrations are providing needed operating parameters for the planned field-scale demonstration with actual B/C Pond sludge waste at ORR. This report discusses the applied systems approach to optimize glass compositions for this particular waste stream through laboratory-, pilot-, and field-scale studies with surrogate and actual B/C waste. These glass compositions will maximize glass durability and waste loading while optimizing melt properties which affect melter operation, such as melt viscosity and melter refractory corrosion. Maximum waste loadings minimize storage volume of the final waste form translating into considerable cost savings

  9. Small-scale demonstration of high-level radioactive waste processing and solidification using actual SRP waste

    International Nuclear Information System (INIS)

    Okeson, J.K.; Galloway, R.M.; Wilhite, E.L.; Woolsey, G.B.; Ferguson, R.B.

    1980-01-01

    A small-scale demonstration of the high-level radioactive waste solidification process by vitrification in borosilicate glass is being conducted using 5-6 liter batches of actual waste. Equipment performance and processing characteristics of the various unit operations in the process are reported and, where appropriate, are compared to large-scale results obtained with synthetic waste

  10. Laboratory simulation of salt dissolution during waste removal

    International Nuclear Information System (INIS)

    Wiersma, B.J.; Parish, W.R.

    1997-01-01

    Laboratory experiments were performed to support the field demonstration of improved techniques for salt dissolution in waste tanks at the Savannah River Site. The tests were designed to investigate three density driven techniques for salt dissolution: (1) Drain-Add-Sit-Remove, (2) Modified Density Gradient, and (3) Continuous Salt Mining. Salt dissolution was observed to be a very rapid process as salt solutions with densities between 1.38-1.4 were frequently removed. Slower addition and removal rates and locating the outlet line at deeper levels below the top of the saltcake provided the best contact between the dissolution water and the saltcake. It was observed that dissolution with 1 M sodium hydroxide solution resulted in salt solutions that were within the current inhibitor requirements for the prevention of stress corrosion cracking. This result was independent of the density driven technique. However, if inhibited water (0.01 M sodium hydroxide and 0.011 M sodium nitrite) was utilized, the salt solutions were frequently outside the inhibitor requirements. Corrosion testing at conditions similar to the environments expected during waste removal was recommended

  11. Small-scale integrated demonstration of high-level radioactive waste processing and vitrification using actual SRP waste

    International Nuclear Information System (INIS)

    Ferguson, R.B.; Woolsey, G.B.; Galloway, R.M.; Baumgarten, P.M.; Eibling, R.E.

    1980-01-01

    Experiments have been made to demonstrate the feasibility of immobilizing SRP high-level waste in borosilicate glass. Results to date are encouraging. Equipment performance and processing characteristics for solidifying small batches of actual SRP waste have agreed well with previous experience with small- and large-scale tests synthetic waste, and with theoretical predictions

  12. Contaminant Release from Residual Waste in Closed Single-Shell Tanks and Other Waste Forms Associated with the Tanks

    International Nuclear Information System (INIS)

    Deutsch, William J.

    2008-01-01

    This chapter describes the release of contaminants from the various waste forms that are anticipated to be associated with closure of the single-shell tanks. These waste forms include residual sludge or saltcake that will remain in the tanks after waste retrieval. Other waste forms include engineered glass and cementitious materials as well as contaminated soil impacted by previous tank leaks. This chapter also describes laboratory testing to quantify contaminant release and how the release data are used in performance/risk assessments for the tank waste management units and the onsite waste disposal facilities. The chapter ends with a discussion of the surprises and lessons learned to date from the testing of waste materials and the development of contaminant release models

  13. Reaction chemistry of nitrogen species in hydrothermal systems: Simple reactions, waste simulants, and actual wastes

    International Nuclear Information System (INIS)

    Dell'Orco, P.; Luan, L.; Proesmans, P.; Wilmanns, E.

    1995-01-01

    Results are presented from hydrothermal reaction systems containing organic components, nitrogen components, and an oxidant. Reaction chemistry observed in simple systems and in simple waste simulants is used to develop a model which presents global nitrogen chemistry in these reactive systems. The global reaction path suggested is then compared with results obtained for the treatment of an actual waste stream containing only C-N-0-H species

  14. Laboratory stabilization/solidification of surrogate and actual mixed-waste sludge in glass and grout

    International Nuclear Information System (INIS)

    Spence, R.D.; Gilliam, T.M.; Mattus, C.H.; Mattus, A.J.

    1998-01-01

    Grouting and vitrification are currently the most likely stabilization/solidification technologies for mixed wastes. Grouting has been used to stabilize and solidify hazardous and low-level waste for decades. Vitrification has long been developed as a high-level-waste alternative and has been under development recently as an alternative treatment technology for low-level mixed waste. Laboratory testing has been performed to develop grout and vitrification formulas for mixed-waste sludges currently stored in underground tanks at Oak Ridge National Laboratory (ORNL) and to compare these waste forms. Envelopes, or operating windows, for both grout and soda-lime-silica glass formulations for a surrogate sludge were developed. One formulation within each envelope was selected for testing the sensitivity of performance to variations (±10 wt%) in the waste form composition and variations in the surrogate sludge composition over the range previously characterized in the sludges. In addition, one sludge sample of an actual mixed-waste tank was obtained, a surrogate was developed for this sludge sample, and grout and glass samples were prepared and tested in the laboratory using both surrogate and the actual sludge. The sensitivity testing of a surrogate tank sludge in selected glass and grout formulations is discussed in this paper, along with the hot-cell testing of an actual tank sludge sample

  15. Criticality considerations for salt-cake disolution in DOE waste tanks

    International Nuclear Information System (INIS)

    Trumble, E.F.; Niemer, K.A.

    1995-01-01

    A large amount of high-level waste is being stored in the form of salt cake at the Savannah River site (SRS) in large (1.3 x 106 gal) underground tanks awaiting startup of the Defense Waste Processing Facility (DWPF). This salt cake will be dissolved with water, and the solution will be fed to DWPF for immobilization in borosilicate glass. Some of the waste that was transferred to the tanks contained enriched uranium and plutonium from chemical reprocessing streams. As water is added to these tanks to dissolve the salt cake, the insoluble portion of this fissile material will be left behind in the tank as the salt solution is pumped out. Because the salt acts as a diluent to the fissile material, the process of repeated water addition, salt dissolution, and salt solution removal will act as a concentrating mechanism for the undissolved fissile material that will remain in the tank. It is estimated that tank 41 H at SRS contains 20 to 120 kg of enriched uranium, varying from 10 to 70% 235 U, distributed nonuniformly throughout the tank. This paper discusses the criticality concerns associated with the dissolution of salt cake in this tank. These concerns are also applicable to other salt cake waste tanks that contain significant quantities of enriched uranium and/or plutonium

  16. Ferrocyanide tank safety program: Cesium uptake capacity of simulated ferrocyanide tank waste. Final report

    International Nuclear Information System (INIS)

    Burgeson, I.E.; Bryan, S.A.

    1995-07-01

    The objective of this project is to determine the capacity for 137 Cs uptake by mixed metal ferrocyanides present in Hanford Site waste tanks, and to assess the potential for aggregation of these 137 Cs-exchanged materials to form ''hot-spots'' in the tanks. This research, performed at Pacific Northwest Laboratory (PNL) for Westinghouse Hanford Company, stems from concerns regarding possible localized radiolytic heating within the tanks. After ferrocyanide was added to 18 high-level waste tanks in the 1950s, some of the ferrocyanide tanks received considerable quantities of saltcake waste that was rich in 137 Cs. If radioactive cesium was exchanged and concentrated by the nickel ferrocyanide present in the tanks, the associated heating could cause tank temperatures to rise above the safety limits specified for the ferrocyanide-containing tanks, especially if the supernate in the tanks is pumped out and the waste becomes drier

  17. Tank waste remediation system: An update

    International Nuclear Information System (INIS)

    Alumkal, W.T.; Babad, H.; Dunford, G.L.; Honeyman, J.O.; Wodrich, D.D.

    1995-02-01

    The US Department of Energy's Hanford Site, located in southeastern Washington State, contains the largest amount and the most diverse collection of highly radioactive waste in the US. High-level radioactive waste has been stored at the Hanford Site in large, underground tanks since 1944. Approximately 217,000 M 3 (57 Mgal) of caustic liquids, slurries, saltcakes, and sludges have accumulated in 177 tanks. In addition, significant amounts of 90 Sr and 137 Cs were removed from the tank waste, converted to salts, doubly encapsulated in metal containers, and stored in water basins. The Tank Waste Remediation System Program was established by the US Department of Energy in 1991 to safely manage and immobilize these wastes in anticipation of permanent disposal of the high-level waste fraction in a geologic repository. Since 1991, significant progress has been made in resolving waste tank safety issues, upgrading Tank Farm facilities and operations, and developing a new strategy for retrieving, treating, and immobilizing the waste for disposal

  18. Separation, Concentration, and Immobilization of Technetium and Iodine from Alkaline Supernate Waste

    Energy Technology Data Exchange (ETDEWEB)

    James Harvey; Michael Gula

    1998-12-07

    Development of remediation technologies for the characterization, retrieval, treatment, concentration, and final disposal of radioactive and chemical tank waste stored within the Department of Energy (DOE) complex represents an enormous scientific and technological challenge. A combined total of over 90 million gallons of high-level waste (HLW) and low-level waste (LLW) are stored in 335 underground storage tanks at four different DOE sites. Roughly 98% of this waste is highly alkaline in nature and contains high concentrations of nitrate and nitrite salts along with lesser concentrations of other salts. The primary waste forms are sludge, saltcake, and liquid supernatant with the bulk of the radioactivity contained in the sludge, making it the largest source of HLW. The saltcake (liquid waste with most of the water removed) and liquid supernatant consist mainly of sodium nitrate and sodium hydroxide salts. The main radioactive constituent in the alkaline supernatant is cesium-137, but strontium-90, technetium-99, and transuranic nuclides are also present in varying concentrations. Reduction of the radioactivity below Nuclear Regulatory Commission (NRC) limits would allow the bulk of the waste to be disposed of as LLW. Because of the long half-life of technetium-99 (2.1 x 10 5 y) and the mobility of the pertechnetate ion (TcO 4 - ) in the environment, it is expected that technetium will have to be removed from the Hanford wastes prior to disposal as LLW. Also, for some of the wastes, some level of technetium removal will be required to meet LLW criteria for radioactive content. Therefore, DOE has identified a need to develop technologies for the separation and concentration of technetium-99 from LLW streams. Eichrom has responded to this DOE-identified need by demonstrating a complete flowsheet for the separation, concentration, and immobilization of technetium (and iodine) from alkaline supernatant waste.

  19. Investigation of thermolytic hydrogen generation rate of tank farm simulated and actual waste

    Energy Technology Data Exchange (ETDEWEB)

    Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Woodham, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Howe, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-15

    To support resolution of Potential Inadequacies in the Safety Analysis for the Savannah River Site (SRS) Tank Farm, Savannah River National Laboratory conducted research to determine the thermolytic hydrogen generation rate (HGR) with simulated and actual waste. Gas chromatography methods were developed and used with air-purged flow systems to quantify hydrogen generation from heated simulated and actual waste at rates applicable to the Tank Farm Documented Safety Analysis (DSA). Initial simulant tests with a simple salt solution plus sodium glycolate demonstrated the behavior of the test apparatus by replicating known HGR kinetics. Additional simulant tests with the simple salt solution excluding organics apart from contaminants provided measurement of the detection and quantification limits for the apparatus with respect to hydrogen generation. Testing included a measurement of HGR on actual SRS tank waste from Tank 38. A final series of measurements examined HGR for a simulant with the most common SRS Tank Farm organics at temperatures up to 140 °C. The following conclusions result from this testing.

  20. HANFORD MEDIUM-LOW CURIE WASTE PRETREATMENT ALTERNATIVES PROJECT FRACTIONAL CRYSTALLIZATION PILOT SCALE TESTING FINAL REPORT

    Energy Technology Data Exchange (ETDEWEB)

    HERTING DL

    2008-09-16

    The Fractional Crystallization Pilot Plant was designed and constructed to demonstrate that fractional crystallization is a viable way to separate the high-level and low-activity radioactive waste streams from retrieved Hanford single-shell tank saltcake. The focus of this report is to review the design, construction, and testing details of the fractional crystallization pilot plant not previously disseminated.

  1. HANFORD MEDIUM-LOW CURIE WASTE PRETREATMENT ALTERNATIVES PROJECT-FRACTIONAL CRYSTALLIZATION PILOT SCALE TESTING FINAL REPORT

    International Nuclear Information System (INIS)

    HERTING DL

    2008-01-01

    The Fractional Crystallization Pilot Plant was designed and constructed to demonstrate that fractional crystallization is a viable way to separate the high-level and low-activity radioactive waste streams from retrieved Hanford single-shell tank saltcake. The focus of this report is to review the design, construction, and testing details of the fractional crystallization pilot plant not previously disseminated

  2. Long-term leach rates of glasses containing actual waste

    International Nuclear Information System (INIS)

    Wiley, J.R.; LeRoy, J.H.

    1979-01-01

    Leach rates of borosilicate glasses that contained actual Savannah River Plant waste were measured. Leaching was done by water and by buffer solutions of pH 4, 7, and 9. Leach rates were then determined from the amount of 137 Cs, 90 Sr, and Pu released into the leach solutions. The cumulative fractions leached were fit to a mathematical model that included leaching by diffusion and glass dissolution

  3. Radioactive tank waste remediation focus area

    International Nuclear Information System (INIS)

    1996-08-01

    EM's Office of Science and Technology has established the Tank Focus Area (TFA) to manage and carry out an integrated national program of technology development for tank waste remediation. The TFA is responsible for the development, testing, evaluation, and deployment of remediation technologies within a system architecture to characterize, retrieve, treat, concentrate, and dispose of radioactive waste stored in the underground stabilize and close the tanks. The goal is to provide safe and cost-effective solutions that are acceptable to both the public and regulators. Within the DOE complex, 335 underground storage tanks have been used to process and store radioactive and chemical mixed waste generated from weapon materials production and manufacturing. Collectively, thes tanks hold over 90 million gallons of high-level and low-level radioactive liquid waste in sludge, saltcake, and as supernate and vapor. Very little has been treated and/or disposed or in final form

  4. Chemical Equilibrium of Aluminate in Hanford Tank Waste Originating from Tanks 241-AN-105 and 241-AP-108

    Energy Technology Data Exchange (ETDEWEB)

    McCoskey, Jacob K. [Washington River Protection Solutions LLC, Richland, WA (United States); Cooke, Gary A. [Washington River Protection Solutions LLC, Richland, WA (United States); Herting, Daniel L. [Washington River Protection Solutions LLC, Richland, WA (United States)

    2015-09-23

    The purposes of the study described in this document follow; Determine or estimate the thermodynamic equilibrium of gibbsite in contact with two real tank waste supernatant liquids through both dissolution of gibbsite (bottom-up approach) and precipitation of aluminum-bearing solids (top-down approach); determine or estimate the thermodynamic equilibrium of a mixture of gibbsite and real tank waste saltcake in contact with real tank waste supernatant liquid through both dissolution of gibbsite and precipitation of aluminum-bearing solids; and characterize the solids present after equilibrium and precipitation of aluminum-bearing solids.

  5. Long-term leach rates of glasses containing actual waste

    International Nuclear Information System (INIS)

    Wiley, J.R.; LeRoy, J.H.

    1979-01-01

    Leach rates of borosilicate glasses that contained actual Savannah River Plant waste were measured. Leaching was done by water and by buffer solutions of pH 4, 7, and 9. Leach rates were then determined from the amount of 137 Cs, 90 Sr, and plutonium released into the leach solutions. The cumulative fractions leached were fit to a mathematical model that included leaching by diffusion and glass dissolution. 5 figures, 3 tables

  6. Characterization, Leaching, and Filtrations Testing of Ferrocyanide Tank sludge (Group 8) Actual Waste Composite

    International Nuclear Information System (INIS)

    Fiskum, Sandra K.; Billing, Justin M.; Crum, J.V.; Daniel, Richard C.; Edwards, Matthew K.; Shimskey, Rick W.; Peterson, Reid A.; MacFarlan, Paul J.; Buck, Edgar C.; Draper, Kathryn E.; Kozelisky, Anne E.

    2009-01-01

    This is the final report in a series of eight reports defining characterization, leach, and filtration testing of a wide variety of Hanford tank waste sludges. The information generated from this series is intended to supplement the Waste Treatment and Immobilization Plant (WTP) project understanding of actual waste behaviors associated with tank waste sludge processing through the pretreatment portion of the WTP. The work described in this report presents information on a high-iron waste form, specifically the ferrocyanide tank waste sludge. Iron hydroxide has been shown to pose technical challenges during filtration processing; the ferrocyanide tank waste sludge represented a good source of the high-iron matrix to test the filtration processing

  7. Characterization, Leaching, and Filtrations Testing of Ferrocyanide Tank sludge (Group 8) Actual Waste Composite

    Energy Technology Data Exchange (ETDEWEB)

    Fiskum, Sandra K.; Billing, Justin M.; Crum, J. V.; Daniel, Richard C.; Edwards, Matthew K.; Shimskey, Rick W.; Peterson, Reid A.; MacFarlan, Paul J.; Buck, Edgar C.; Draper, Kathryn E.; Kozelisky, Anne E.

    2009-02-28

    This is the final report in a series of eight reports defining characterization, leach, and filtration testing of a wide variety of Hanford tank waste sludges. The information generated from this series is intended to supplement the Waste Treatment and Immobilization Plant (WTP) project understanding of actual waste behaviors associated with tank waste sludge processing through the pretreatment portion of the WTP. The work described in this report presents information on a high-iron waste form, specifically the ferrocyanide tank waste sludge. Iron hydroxide has been shown to pose technical challenges during filtration processing; the ferrocyanide tank waste sludge represented a good source of the high-iron matrix to test the filtration processing.

  8. Actual waste demonstration of the nitric-glycolic flowsheet for sludge batch 9 qualification

    Energy Technology Data Exchange (ETDEWEB)

    Newell, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Martino, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Reboul, S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Coleman, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Johnson, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-03-09

    For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs qualification testing to demonstrate that the sludge batch is processable. Based on the results of this actual-waste qualification and previous simulant studies, SRNL recommends implementation of the nitric-glycolic acid flowsheet in DWPF. Other recommendations resulting from this demonstration are reported in section 5.0.

  9. Small-scale integrated demonstration of high-level radioactive waste processing and vitrification using actual SRP waste

    International Nuclear Information System (INIS)

    Woolsey, G.B.; Baumgarten, P.K.; Eibling, R.E.; Ferguson, R.B.

    1981-01-01

    A small-scale pilot plant for chemical processing and vitrification of actual high-level waste has been constructed at the Savannah River Laboratory (SRL). This fully integrated facility has been constructed in six shielded cells and has eight major unit operations. Equipment performance and processing characteristics of the unit operations are reported

  10. Electrochemical destruction of organics and nitrates in simulated and actual radioactive Hanford tank waste

    International Nuclear Information System (INIS)

    Elmore, M.R.; Lawrence, W.E.

    1996-09-01

    Pacific Northwest National Laboratory has conducted an evaluation of electrochemical processing for use in radioactive tank waste cleanup activities. An electrochemical organic destruction (ECOD) process was evaluated, with the main focus being the destruction of organic compounds (especially organic complexants of radionuclides) in simulated and actual radioactive Hanford tank wastes. A primary reason for destroying the organic species in the complexant concentrate tank waste is to decomplex/defunctionalize species that chelate radionuclides. the separations processes required to remove the radionuclides are much less efficient when chelators are present. A second objective, the destruction of nitrates and nitrites in the wastes, was also assessed. Organic compounds, nitrates, and nitrites may affect waste management and safety considerations, not only at Hanford but at other US Department of Energy sites that maintain high- level waste storage tanks

  11. Numerical Modeling of Mixing of Chemically Reacting, Non-Newtonian Slurry for Tank Waste Retrieval

    International Nuclear Information System (INIS)

    Yuen, David A.; Onishi, Yasuo; Rustad, James R.; Michener, Thomas E.; Felmy, Andrew R.; Ten, Arkady A.; Hier, Catherine A.

    2000-01-01

    Many highly radioactive wastes will be retrieved by installing mixer pumps that inject high-speed jets to stir up the sludge, saltcake, and supernatant liquid in the tank, blending them into a slurry. This slurry will then be pumped out of the tank into a waste treatment facility. Our objectives are to investigate interactions-chemical reactions, waste rheology, and slurry mixing-occurring during the retrieval operation and to provide a scientific basis for the waste retrieval decision-making process. Specific objectives are to: (1) Evaluate numerical modeling of chemically active, non-Newtonian tank waste mixing, coupled with chemical reactions and realistic rheology; (2) Conduct numerical modeling analysis of local and global mixing of non-Newtonian and Newtonian slurries; and (3) Provide the bases to develop a scientifically justifiable, decision-making support tool for the tank waste retrieval operation

  12. Initial demonstration of DWPF process and product control strategy using actual radioactive waste

    International Nuclear Information System (INIS)

    Andrews, M.K.; Bibler, N.E.; Jantzen, C.M.; Beam, D.C.

    1991-01-01

    The Defense Waste Processing Facility at the Savannah River Site (SRS) will vitrify high-level nuclear waste into borosilicate glass. The waste will be mixed with properly formulated glass-making frit and fed to a melter at 1150 degrees C. Process control and product quality are ensured by proper control of the melter feed composition. Algorithms have been developed to predict the processability of the melt and the durability of the final glass based on this feed composition. To test these algorithms, an actual radioactive waste contained in a shielded facility at SRS was analyzed and a frit composition formulated using a simple computer spreadsheet which contained the algorithms. This frit was then mixed with the waste and the resulting slurry fed to a research scale joule-heated melter operated remotely. Approximately 24 kg of glass were successfully prepared. This paper will describe the frit formulation, the vitrification process, and the glass durability

  13. Sludge batch 9 follow-on actual-waste testing for the nitric-glycolic flowsheet

    Energy Technology Data Exchange (ETDEWEB)

    Martino, C. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, J. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-03-23

    An actual-waste Sludge Batch 9 qualification run with the nitric-glycolic flowsheet (SC-18) was performed in FY16. In order to supplement the knowledge base for the nitric-glycolic flowsheet, additional testing was performed on the product slurries, condensates, and intermediate samples from run SC-18.

  14. Final Report. LAW Glass Formulation to Support AP-101 Actual Waste Testing, VSL-03R3470-2, Rev. 0

    Energy Technology Data Exchange (ETDEWEB)

    Muller, I. S. [The Catholic University of America, Washington, DC (United States); Pegg, I. L. [The Catholic University of America, Washington, DC (United States); Rielley, Elizabeth [The Catholic University of America, Washington, DC (United States); Carranza, Isidro [The Catholic University of America, Washington, DC (United States); Hight, Kenneth [The Catholic University of America, Washington, DC (United States); Lai, Shan-Tao T. [The Catholic University of America, Washington, DC (United States); Mooers, Cavin [The Catholic University of America, Washington, DC (United States); Bazemore, Gina [The Catholic University of America, Washington, DC (United States); Cecil, Richard [The Catholic University of America, Washington, DC (United States); Kruger, Albert A. [The Catholic University of America, Washington, DC (United States)

    2015-06-22

    The main objective of the work was to develop and select a glass formulation for vitrification testing of the actual waste sample of LAW AP-101 at Battelle - Pacific Northwest Division (PNWD). Other objectives of the work included preparation and characterization of glasses to demonstrate compliance with contract and processing requirements, evaluation of the ability to achieve waste loading requirements, testing to demonstrate compatibility of the glass melts with melter materials of construction, comparison of the properties of simulant and actual waste glasses, and identification of glass formulation issues with respect to contract specifications and processing requirements.

  15. Filtration and Leach Testing for PUREX Cladding Sludge and REDOX Cladding Sludge Actual Waste Sample Composites

    Energy Technology Data Exchange (ETDEWEB)

    Shimskey, Rick W.; Billing, Justin M.; Buck, Edgar C.; Casella, Amanda J.; Crum, Jarrod V.; Daniel, Richard C.; Draper, Kathryn E.; Edwards, Matthew K.; Hallen, Richard T.; Kozelisky, Anne E.; MacFarlan, Paul J.; Peterson, Reid A.; Swoboda, Robert G.

    2009-03-02

    A testing program evaluating actual tank waste was developed in response to Task 4 from the M-12 External Flowsheet Review Team (EFRT) issue response plan (Barnes and Voke 2006). The test program was subdivided into logical increments. The bulk water-insoluble solid wastes that are anticipated to be delivered to the Hanford Waste Treatment and Immobilization Plant (WTP) were identified according to type such that the actual waste testing could be targeted to the relevant categories. Under test plan TP RPP WTP 467 (Fiskum et al. 2007), eight broad waste groupings were defined. Samples available from the 222S archive were identified and obtained for testing. Under this test plan, a waste testing program was implemented that included: • Homogenizing the archive samples by group as defined in the test plan. • Characterizing the homogenized sample groups. • Performing parametric leaching testing on each group for compounds of interest. • Performing bench-top filtration/leaching tests in the hot cell for each group to simulate filtration and leaching activities if they occurred in the UFP2 vessel of the WTP Pretreatment Facility. This report focuses on a filtration/leaching test performed using two of the eight waste composite samples. The sample groups examined in this report were the plutonium-uranium extraction (PUREX) cladding waste sludge (Group 3, or CWP) and reduction-oxidation (REDOX) cladding waste sludge (Group 4, or CWR). Both the Group 3 and 4 waste composites were anticipated to be high in gibbsite, thus requiring caustic leaching. WTP RPT 167 (Snow et al. 2008) describes the homogenization, characterization, and parametric leaching activities before benchtop filtration/leaching testing of these two waste groups. Characterization and initial parametric data in that report were used to plan a single filtration/leaching test using a blend of both wastes. The test focused on filtration testing of the waste and caustic leaching for aluminum, in the form

  16. Accelerated Best Basis Inventory Baselining Task

    International Nuclear Information System (INIS)

    SASAKI, L.M.

    2001-01-01

    The baselining effort was recently proposed to bring the Best-Basis Inventory (BBI) and Question No.8 of the Tank Interpretive Report (TIR) for all 177 tanks to the current standards and protocols and to prepare a TIR Question No.8 if one is not already available. This plan outlines the objectives and methodology of the accelerated BBI baselining task. BBI baselining meetings held during December 2000 resulted in a revised BBI methodology and an initial set of BBI creation rules to be used in the baselining effort. The objectives of the BBI baselining effort are to: (1) Provide inventories that are consistent with the revised BBI methodology and new BBI creation rules. (2) Split the total tank waste in each tank into six waste phases, as appropriate (Supernatant, saltcake solids, saltcake liquid, sludge solids, sludge liquid, and retained gas). In some tanks, the solids and liquid portions of the sludge and/or saltcake may be combined into a single sludge or saltcake phase. (3) Identify sampling events that are to be used for calculating the BBIs. (4) Update waste volumes for subsequent reconciliation with the Hanlon (2001) waste tank summary. (5) Implement new waste type templates. (6) Include any sample data that might have been unintentionally omitted in the previous BBI and remove any sample data that should not have been included. Sample data to be used in the BBI must be available on TWINS. (7) Ensure that an inventory value for each standard BBI analyte is provided for each waste component. Sample based inventories for supplemental BBI analytes will be included when available. (8) Provide new means and confidence interval reports if one is not already available and include uncertainties in reporting inventory values

  17. BENCH-SCALE STEAM REFORMING OF ACTUAL TANK 48H WASTE

    International Nuclear Information System (INIS)

    Burket, P; Gene Daniel, G; Charles Nash, C; Carol Jantzen, C; Michael Williams, M

    2008-01-01

    Fluidized Bed Steam Reforming (FBSR) has been demonstrated to be a viable technology to remove >99% of the organics from Tank 48H simulant, to remove >99% of the nitrate/nitrite from Tank 48H simulant, and to form a solid product that is primarily carbonate based. The technology was demonstrated in October of 2006 in the Engineering Scale Test Demonstration Fluidized Bed Steam Reformer1 (ESTD FBSR) at the Hazen Research Inc. (HRI) facility in Golden, CO. The purpose of the Bench-scale Steam Reformer (BSR) testing was to demonstrate that the same reactions occur and the same product is formed when steam reforming actual radioactive Tank 48H waste. The approach used in the current study was to test the BSR with the same Tank 48H simulant and same Erwin coal as was used at the ESTD FBSR under the same operating conditions. This comparison would allow verification that the same chemical reactions occur in both the BSR and ESTD FBSR. Then, actual radioactive Tank 48H material would be steam reformed in the BSR to verify that the actual tank 48H sample reacts the same way chemically as the simulant Tank 48H material. The conclusions from the BSR study and comparison to the ESTD FBSR are the following: (1) A Bench-scale Steam Reforming (BSR) unit was successfully designed and built that: (a) Emulated the chemistry of the ESTD FBSR Denitration Mineralization Reformer (DMR) and Carbon Reduction Reformer (CRR) known collectively as the dual reformer flowsheet. (b) Measured and controlled the off-gas stream. (c) Processed real (radioactive) Tank 48H waste. (d) Met the standards and specifications for radiological testing in the Savannah River National Laboratory (SRNL) Shielded Cells Facility (SCF). (2) Three runs with radioactive Tank 48H material were performed. (3) The Tetraphenylborate (TPB) was destroyed to > 99% for all radioactive Bench-scale tests. (4) The feed nitrate/nitrite was destroyed to >99% for all radioactive BSR tests the same as the ESTD FBSR. (5) The

  18. BENCH-SCALE STEAM REFORMING OF ACTUAL TANK 48H WASTE

    Energy Technology Data Exchange (ETDEWEB)

    Burket, P; Gene Daniel, G; Charles Nash, C; Carol Jantzen, C; Michael Williams, M

    2008-09-25

    Fluidized Bed Steam Reforming (FBSR) has been demonstrated to be a viable technology to remove >99% of the organics from Tank 48H simulant, to remove >99% of the nitrate/nitrite from Tank 48H simulant, and to form a solid product that is primarily carbonate based. The technology was demonstrated in October of 2006 in the Engineering Scale Test Demonstration Fluidized Bed Steam Reformer1 (ESTD FBSR) at the Hazen Research Inc. (HRI) facility in Golden, CO. The purpose of the Bench-scale Steam Reformer (BSR) testing was to demonstrate that the same reactions occur and the same product is formed when steam reforming actual radioactive Tank 48H waste. The approach used in the current study was to test the BSR with the same Tank 48H simulant and same Erwin coal as was used at the ESTD FBSR under the same operating conditions. This comparison would allow verification that the same chemical reactions occur in both the BSR and ESTD FBSR. Then, actual radioactive Tank 48H material would be steam reformed in the BSR to verify that the actual tank 48H sample reacts the same way chemically as the simulant Tank 48H material. The conclusions from the BSR study and comparison to the ESTD FBSR are the following: (1) A Bench-scale Steam Reforming (BSR) unit was successfully designed and built that: (a) Emulated the chemistry of the ESTD FBSR Denitration Mineralization Reformer (DMR) and Carbon Reduction Reformer (CRR) known collectively as the dual reformer flowsheet. (b) Measured and controlled the off-gas stream. (c) Processed real (radioactive) Tank 48H waste. (d) Met the standards and specifications for radiological testing in the Savannah River National Laboratory (SRNL) Shielded Cells Facility (SCF). (2) Three runs with radioactive Tank 48H material were performed. (3) The Tetraphenylborate (TPB) was destroyed to > 99% for all radioactive Bench-scale tests. (4) The feed nitrate/nitrite was destroyed to >99% for all radioactive BSR tests the same as the ESTD FBSR. (5) The

  19. Hanford Tank Waste - Near Source Treatment of Low Activity Waste

    International Nuclear Information System (INIS)

    Ramsey, William Gene

    2013-01-01

    Abstract only. Treatment and disposition of Hanford Site waste as currently planned consists of 100+ waste retrievals, waste delivery through up to 8+ miles of dedicated, in-ground piping, centralized mixing and blending operations- all leading to pre-treatment combination and separation processes followed by vitrification at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The sequential nature of Tank Farm and WTP operations requires nominally 15-20 years of continuous operations before all waste can be retrieved from many Single Shell Tanks (SSTs). Also, the infrastructure necessary to mobilize and deliver the waste requires significant investment beyond that required for the WTP. Treating waste as closely as possible to individual tanks or groups- as allowed by the waste characteristics- is being investigated to determine the potential to 1) defer, reduce, and/or eliminate infrastructure requirements, and 2) significantly mitigate project risk by reducing the potential and impact of single point failures. The inventory of Hanford waste slated for processing and disposition as LAW is currently managed as high-level waste (HLW), i.e., the separation of fission products and other radionuclides has not commenced. A significant inventory of this waste (over 20M gallons) is in the form of precipitated saltcake maintained in single shell tanks, many of which are identified as potential leaking tanks. Retrieval and transport (as a liquid) must be staged within the waste feed delivery capability established by site infrastructure and WTP. Near Source treatment, if employed, would provide for the separation and stabilization processing necessary for waste located in remote farms (wherein most of the leaking tanks reside) significantly earlier than currently projected. Near Source treatment is intended to address the currently accepted site risk and also provides means to mitigate future issues likely to be faced over the coming decades. This paper

  20. Ferrocyanide Safety Project: Comparison of actual and simulated ferrocyanide waste properties

    International Nuclear Information System (INIS)

    Scheele, R.D.; Burger, L.L.; Sell, R.L.; Bredt, P.R.; Barrington, R.J.

    1994-09-01

    In the 1950s, additional high-level radioactive waste storage capacity was needed to accommodate the wastes that would result from the production of recovery of additional nuclear defense materials. To provide this additional waste storage capacity, the Hanford Site operating contractor developed a process to decontaminate aqueous wastes by precipitating radiocesium as an alkali nickel ferrocyanide; this process allowed disposal of the aqueous waste. The radiocesium scavenging process as developed was used to decontaminate (1) first-cycle bismuth phosphate (BiPO 4 ) wastes, (2) acidic wastes resulting from uranium recovery operations, and (3) the supernate from neutralized uranium recovery wastes. The radiocesium scavenging process was often coupled with other scavenging processes to remove radiostrontium and radiocobalt. Because all defense materials recovery processes used nitric acid solutions, all of the wastes contained nitrate, which is a strong oxidizer. The variety of wastes treated, and the occasional coupling of radiostrontium and radiocobalt scavenging processes with the radiocesium scavenging process, resulted in ferrocyanide-bearing wastes having many different compositions. In this report, we compare selected physical, chemical, and radiochemical properties measured for Tanks C-109 and C-112 wastes and selected physical and chemical properties of simulated ferrocyanide wastes to assess the representativeness of stimulants prepared by WHC

  1. Hanford and Savannah River Site Programmatic and Technical Integration

    International Nuclear Information System (INIS)

    Ramsey, William Gene

    2013-01-01

    Abstract only. The Hanford Site and the Savannah River Site (SRS) were the primary plutonium production facilities within the U.S. nuclear weapons complex. Radioactive wastes were generated as part of these missions and are stored in similar fashion. The majority of radioactivity maintained by the two sites is located in underground carbon steel tanks in the physical form of supernatant, saltcake, or sludge. Disposition of SRS tank waste is ongoing by converting it into glass (pathway for sludge and radionuclides separated from supernatant or dissolved saltcake) or cement (pathway for the decontaminated supernatant and dissolved saltcake). Tank closure activity has also begun at SRS and will continue for the duration of mission. The Hanford tank waste inventory is roughly 2/3rds larger than SRS's by volume- but nominally half the radioactivity. The baseline disposition path includes high-level and low-activity waste vitrification with separate disposition of contact-handled transuranic tank waste. Retrieval of tank waste from aging single shell tanks (SSTs) into double-shell tanks (DSTs) is currently ongoing. As vitrification commences later this decade, Hanford will be in a similar operations mode as SRS. Site integration is increasing as the missions align. The ongoing integration is centered on key issues that impact both sites- regardless of mission timeframe. Three recent workshop exchanges have been held to improve communication with the primary intent of improving operations and technical work organization. The topics of these workshops are as follows: DST space utilization, optimization, and closure; Waste Feed Qualification; and, Cementitious Waste Forms. Key goals for these and future exchanges include aligning research and technology, preparing for joint initiatives (to maximize budgetary value for the customer), and reviewing lessons learned. Each site has played a leading role in the development of technology and operational practices that can be used

  2. Initial ACTR retrieval technology evaluation test material recommendations

    International Nuclear Information System (INIS)

    Powell, M.R.

    1996-04-01

    Millions of gallons of radiaoctive waste are contained in underground storage tanks at Hanford (SE Washington). Techniques for retrieving much of this waste from the storage tanks have been developed. Current baseline approach is to use sluice jets for single-shell tanks and mixer pumps for double-shell tanks. The Acquire Commercial Technology for Retrieval (ACTR) effort was initiated to identify potential improvements in or alternatives to the baseline waste retrieval methods. Communications with a variety of vendors are underway to identify improved methods that can be implemented at Hanford with little or no additional development. Commercially available retrieval methods will be evaluated by a combination of testing and system-level cost estimation. Current progress toward developing waste simulants for testing ACTR candidate methods is reported; the simulants are designed to model 4 different types of tank waste. Simulant recipes are given for wet sludge, hardpan/dried sludge,hard saltcake, and soft saltcake. Comparisons of the waste and simulant properties are documented in this report

  3. TANK WASTE RETRIEVAL LESSONS LEARNED AT THE HANFORD SITE

    International Nuclear Information System (INIS)

    DODD, R.A.

    2006-01-01

    One of the environmental remediation challenges facing the nation is the retrieval and permanent disposal of approximately 90 million gallons of radioactive waste stored in underground tanks at the US Department of Energy (DOE) facilities. The Hanford Site is located in southeastern Washington State and stores roughly 60% of this waste. An estimated 53 million gallons of high-level, transuranic, and low-level radioactive waste is stored underground in 149 single-shell tanks (SSTs) and 28 newer double-shell tanks (DSTs) at the Hanford Site. These SSTs range in size from 55,000 gallons to 1,000,000 gallon capacity. Approximately 30 million gallons of this waste is stored in SSTs. The SSTs were constructed between 1943 and 1964 and all have exceeded the nominal 20-year design life. Sixty-seven SSTs are known or suspected to have leaked an estimated 1,000,000 gallons of waste. The risk of additional SST leakage has been greatly reduced by removing more than 3 million gallons of interstitial liquids and supernatant and transferring the waste to the DST system since 1997 as part of the interim stabilization program. Retrieval of SST saltcake and sludge waste is underway to further reduce risks and stage feed materials for the Hanford Site Waste Treatment Plant. This paper presents lessons learned from retrieval of tank waste at the Hanford Site and discusses how this information is used to optimize retrieval system efficiency, improve overall cost effectiveness of retrieval operations, and ensure that HFFACO requirements are met

  4. Low-activity waste envelope definitions for the TWRS Privatization Phase I Request For Proposal

    International Nuclear Information System (INIS)

    Patello, G.K.; Lauerhass, L.; Myers, R.L.; Wiemers, K.D.

    1996-11-01

    Radioactive waste has been stored in large underground storage tanks at the Hanford Site since 1944. Approximately 212 million liters of waste containing approximately 240,000 metric tons of processed chemicals and 177 mega-curies of radionuclides are now stored in 177 tanks. These caustic wastes are in the form of liquids, slurries, saltcakes, and sludge. In 1991, the Tank Waste Remediation System (TWRS) Program was established to manage, retrieve, treat, immobilize, and dispose of these wastes in a safe, environmentally sound, and cost-effective manner. The Department of Energy (DOE) has believes that it is feasible to privatize portions of the TWRS Program. Under the privatization strategy embodied in the Request for Proposal (RFP), DOE will purchase services from a contractor-owned, contractor-operated facility under a fixed-price contract. Phase I of the TWRS privatization strategy is a proof-of-concept/commercial demonstration-scale effort. The objectives of Phase I are to demonstrate the technical and business viability of using privatized facilities to treat Hanford tank waste; define and maintain required levels of radiological, nuclear, process, and occupational safety; maintain environmental protection and compliance; and substantially reduce life-cycle costs and time required to treat Hanford tank waste. Three low-activity waste (LAW) envelopes are identified for Phase I of the privatization contract and are representative of the range of Hanford double-shelled tank (DST) waste

  5. Low-activity waste envelope definitions for the TWRS Privatization Phase I Request For Proposal

    Energy Technology Data Exchange (ETDEWEB)

    Patello, G.K.; Lauerhass, L.; Myers, R.L.; Wiemers, K.D.

    1996-11-01

    Radioactive waste has been stored in large underground storage tanks at the Hanford Site since 1944. Approximately 212 million liters of waste containing approximately 240,000 metric tons of processed chemicals and 177 mega-curies of radionuclides are now stored in 177 tanks. These caustic wastes are in the form of liquids, slurries, saltcakes, and sludge. In 1991, the Tank Waste Remediation System (TWRS) Program was established to manage, retrieve, treat, immobilize, and dispose of these wastes in a safe, environmentally sound, and cost-effective manner. The Department of Energy (DOE) has believes that it is feasible to privatize portions of the TWRS Program. Under the privatization strategy embodied in the Request for Proposal (RFP), DOE will purchase services from a contractor-owned, contractor-operated facility under a fixed-price contract. Phase I of the TWRS privatization strategy is a proof-of-concept/commercial demonstration-scale effort. The objectives of Phase I are to demonstrate the technical and business viability of using privatized facilities to treat Hanford tank waste; define and maintain required levels of radiological, nuclear, process, and occupational safety; maintain environmental protection and compliance; and substantially reduce life-cycle costs and time required to treat Hanford tank waste. Three low-activity waste (LAW) envelopes are identified for Phase I of the privatization contract and are representative of the range of Hanford double-shelled tank (DST) waste.

  6. Tank waste processing analysis: Database development, tank-by-tank processing requirements, and examples of pretreatment sequences and schedules as applied to Hanford Double-Shell Tank Supernatant Waste - FY 1993

    International Nuclear Information System (INIS)

    Colton, N.G.; Orth, R.J.; Aitken, E.A.

    1994-09-01

    This report gives the results of work conducted in FY 1993 by the Tank Waste Processing Analysis Task for the Underground Storage Tank Integrated Demonstration. The main purpose of this task, led by Pacific Northwest Laboratory, is to demonstrate a methodology to identify processing sequences, i.e., the order in which a tank should be processed. In turn, these sequences may be used to assist in the development of time-phased deployment schedules. Time-phased deployment is implementation of pretreatment technologies over a period of time as technologies are required and/or developed. The work discussed here illustrates how tank-by-tank databases and processing requirements have been used to generate processing sequences and time-phased deployment schedules. The processing sequences take into account requirements such as the amount and types of data available for the tanks, tank waste form and composition, required decontamination factors, and types of compact processing units (CPUS) required and technology availability. These sequences were developed from processing requirements for the tanks, which were determined from spreadsheet analyses. The spreadsheet analysis program was generated by this task in FY 1993. Efforts conducted for this task have focused on the processing requirements for Hanford double-shell tank (DST) supernatant wastes (pumpable liquid) because this waste type is easier to retrieve than the other types (saltcake and sludge), and more tank space would become available for future processing needs. The processing requirements were based on Class A criteria set by the U.S. Nuclear Regulatory Commission and Clean Option goals provided by Pacific Northwest Laboratory

  7. Transportable Vitrification System: Operational experience gained during vitrification of simulated mixed waste

    International Nuclear Information System (INIS)

    Whitehouse, J.C.; Burket, P.R.; Crowley, D.A.; Hansen, E.K.; Jantzen, C.M.; Smith, M.E.; Singer, R.P.; Young, S.R.; Zamecnik, J.R.; Overcamp, T.J.; Pence, I.W. Jr.

    1996-01-01

    The Transportable Vitrification System (TVS) is a large-scale, fully-integrated, transportable, vitrification system for the treatment of low-level nuclear and mixed wastes in the form of sludges, soils, incinerator ash, and similar waste streams. The TVS was built to demonstrate the vitrification of actual mixed waste at U. S. Department of Energy (DOE) sites. Currently, Westinghouse Savannah River Company (WSRC) is working with Lockheed Martin Energy Systems (LMES) to apply field scale vitrification to actual mixed waste at Oak Ridge Reservation's (ORR) K-25 Site. Prior to the application of the TVS to actual mixed waste it was tested on simulated K-25 B and C Pond waste at Clemson University. This paper describes the results of that testing and preparations for the demonstration on actual mixed waste

  8. Environmental monitoring report for commercial low-level radioactive waste disposal sites (1960`s through 1990`s)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-11-01

    During the time period covered in this report (1960`s through early 1990`s), six commercial low-level radioactive waste (LLRW) disposal facilities have been operated in the US. This report provides environmental monitoring data collected at each site. The report summarizes: (1) each site`s general design, (2) each site`s inventory, (3) the environmental monitoring program for each site and the data obtained as the program has evolved, and (4) what the program has indicated about releases to off-site areas, if any, including a statement of the actual health and safety significance of any release. A summary with conclusions is provided at the end of each site`s chapter. The six commercial LLRW disposal sites discussed are located near: Sheffield, Illinois; Maxey Flats, Kentucky; Beatty, Nevada; West Valley, New York; Barnwell, South Carolina; Richland, Washington.

  9. FRACTIONAL CRYSTALLIZATION TESTING WITH INTERIM PRETREATMENT SYSTEM FEEDS

    International Nuclear Information System (INIS)

    HERTING DL

    2008-01-01

    The fractional crystallization process was developed as a pretreatment method for saltcake waste retrieved from Hanford single-shell tanks (SST). The process separates the retrieved SST waste into a high-level waste stream containing the bulk of the radionuclides and a low-activity waste stream containing the bulk of the nonradioactive sodium salts. The Interim Pretreatment System project shifted the focus on pretreatment planning from SST waste to double-shell tank waste

  10. Independent technical review of Savannah River Site Defense Waste Processing Facility technical issues

    International Nuclear Information System (INIS)

    1992-07-01

    The Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) will vitrify high-level radioactive waste that is presently stored as liquid, salt-cake, and sludge in 51 waste-storage tanks. Construction of the DWPF began in 1984, and the Westinghouse Savannah Company (WSRC) considers the plant to be 100% turned over from construction and 91% complete. Cold-chemical runs are scheduled to begin in November 1992, and hot start up is projected for June 1994. It is estimated that the plant lifetime must exceed 15 years to complete the vitrification of the current, high-level tank waste. In a memo to the Assistant Secretary for Defense Programs (DP-1), the Assistant Secretary for Environmental Restoration and Waste management (EM-1) established the need for an Independent Technical Review (ITR), or the Red Team, to ''review process technology issues preventing start up of the DWPF.'' This report documents the findings of an Independent Technical Review (ITR) conducted by the Department of Energy (DOE), Office of Environmental Restoration and Waste Management (EM), at the request of the Assistant Secretary for Environmental Restoration and Waste Management, of specified aspects of Defense Waste Process Facility (DWPF) process technology. Information for the assessment was drawn from documents provided to the ITR Team by the Westinghouse Savannah River Company (WSRC), and presentations, discussions, interviews, and tours held at the Savannah River Site (SRS) during the weeks of February and March 9, 1992

  11. Alternative Chemical Cleaning Methods for High Level Waste Tanks: Actual Waste Testing with SRS Tank 5F Sludge

    Energy Technology Data Exchange (ETDEWEB)

    King, William D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hay, Michael S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-30

    Solubility testing with actual High Level Waste tank sludge has been conducted in order to evaluate several alternative chemical cleaning technologies for the dissolution of sludge residuals remaining in the tanks after the exhaustion of mechanical cleaning and sludge sluicing efforts. Tests were conducted with archived Savannah River Site (SRS) radioactive sludge solids that had been retrieved from Tank 5F in order to determine the effectiveness of an optimized, dilute oxalic/nitric acid cleaning reagent toward dissolving the bulk non-radioactive waste components. Solubility tests were performed by direct sludge contact with the oxalic/nitric acid reagent and with sludge that had been pretreated and acidified with dilute nitric acid. For comparison purposes, separate samples were also contacted with pure, concentrated oxalic acid following current baseline tank chemical cleaning methods. One goal of testing with the optimized reagent was to compare the total amounts of oxalic acid and water required for sludge dissolution using the baseline and optimized cleaning methods. A second objective was to compare the two methods with regard to the dissolution of actinide species known to be drivers for SRS tank closure Performance Assessments (PA). Additionally, solubility tests were conducted with Tank 5 sludge using acidic and caustic permanganate-based methods focused on the “targeted” dissolution of actinide species.

  12. The European atomic politics under the treaty of Lisboa and actual questions of the export of nuclear waste

    International Nuclear Information System (INIS)

    Ehlenz, Christian; Frenz, Walter

    2011-01-01

    In the context of the actual discussion considerably influenced by safety questions activated by the reactor accident in Tschernobyl in 1986, the EAG gains increasing significance. However, a substantial safety risk proceeds not only with nuclear power stations as a possible goal for attacks of terror, but also from the uncontrolled spreading with nuclear material. The latter equally affects the question on handling radioactive waste materials being straightly discussed in Germany as well as in the European Union as proven by the European Union commissioner Oettinger. This is to be considered in connection with the problem how extensively the competences of the Euratom for the nuclear safety may be interpreted. The actual changes of the legal situation by means of the Lisboa contract for the EAG altogether turned out marginally and let its demand for a reform based on the actual conditions appear more urgent.

  13. Hazardous waste dislodging and conveyance: The confined sluicing method

    International Nuclear Information System (INIS)

    Summers, D.A.; Fossey, R.D.; Mann, M.D.; Blaine, J.G.

    1994-09-01

    This report describes an investigation of a means for dislodging and conveying waste currently stored in underground storage tanks. A series of experiments have been carried out to evaluate the potential of a medium pressure, medium flow rate cutting system as a means of dislodging the waste. It has been found that waterjets at a pressure of 10,000 psi can effectively cut the material which has been chosen to simulate the hardened saltcake within the storage tanks. Based on a parameterization test it has thus been calculated that an inlet flow volume of approximately 30 gallons per minute will be sufficient to excavate 30 gallons per minute of waste from a tank. In order to transport the resulting slurry from the tank, a modified jet pump has been developed and has demonstrated its capability of conveying fluid and waste particles, up to one inch in diameter, to a height of more than 60 feet. Experiments were conducted to examine different configurations to achieve the production levels required for waste removal and to clean the walls of residual material. It was found more effective to clean the walls using an inclined angle of impact rather than a perpendicular angle of impact in order to provide a safeguard against driving the water through any cracks in the containment. It was demonstrated that excavation can take place with almost total immediate extraction of the water and debris from the cutting process. The results have qualitatively shown the potential of a medium pressure waterjet system for achieving the required results for underground storage tank waste retrieval

  14. Waste Generator Instructions: Key to Successful Implementation of the US DOE's 435.1 for Transuranic Waste Packaging Instructions (LA-UR-12-24155) - 13218

    Energy Technology Data Exchange (ETDEWEB)

    French, David M. [LANL EES-12, Carlsbad, NM, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Hayes, Timothy A. [LANL EES-12, Carlsbad, NM, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Pope, Howard L. [Aspen Resources Ltd., Inc., P.O. Box 3038, Boulder, CO 80307 (United States); Enriquez, Alejandro E. [LANL NCO-4, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States); Carson, Peter H. [LANL NPI-7, Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2013-07-01

    In times of continuing fiscal constraints, a management and operation tool that is straightforward to implement, works as advertised, and virtually ensures compliant waste packaging should be carefully considered and employed wherever practicable. In the near future, the Department of Energy (DOE) will issue the first major update to DOE Order 435.1, Radioactive Waste Management. This update will contain a requirement for sites that do not have a Waste Isolation Pilot Plant (WIPP) waste certification program to use two newly developed technical standards: Contact-Handled Defense Transuranic Waste Packaging Instructions and Remote-Handled Defense Transuranic Waste Packaging Instructions. The technical standards are being developed from the DOE O 435.1 Notice, Contact-Handled and Remote-Handled Transuranic Waste Packaging, approved August 2011. The packaging instructions will provide detailed information and instruction for packaging almost every conceivable type of transuranic (TRU) waste for disposal at WIPP. While providing specificity, the packaging instructions leave to each site's own discretion the actual mechanics of how those Instructions will be functionally implemented at the floor level. While the Technical Standards are designed to provide precise information for compliant packaging, the density of the information in the packaging instructions necessitates a type of Rosetta Stone that translates the requirements into concise, clear, easy to use and operationally practical recipes that are waste stream and facility specific for use by both first line management and hands-on operations personnel. The Waste Generator Instructions provide the operator with step-by-step instructions that will integrate the sites' various operational requirements (e.g., health and safety limits, radiological limits or dose limits) and result in a WIPP certifiable waste and package that can be transported to and emplaced at WIPP. These little known but widely

  15. Waste analysis plan for 222-S dangerous and mixed waste storage area

    International Nuclear Information System (INIS)

    Warwick, G.J.

    1994-01-01

    The 222-S Laboratory Complex, in the southeast corner of the 200 West Area, consists of the 222-S Laboratory, the 222-SA Standards Laboratory, and several ancillary facilities. Currently, 222-S Laboratory activities are in supporting efforts to characterize the waste stored in the 200 Areas single shell and double shell tanks. Besides this work, the laboratory also provides analytical services for waste-management processing plants, Tank Farms, B Plant, 242-A Evaporator Facility, Plutonium-Uranium Extraction Plant, Plutonium Finishing Plant, Uranium-Oxide Plant, Waste Encapsulation Storage Facility, environmental monitoring and surveillance programs, and activities involving essential materials and research and development. One part of the 222-SA Laboratory prepares nonradioactive standards for the 200 Area laboratories. The other section of the laboratory is used for cold (nonradioactive) process development work and standards preparation. The 219-S Waste Handling Facility has three storage tanks in which liquid acid waste from 222-S can be received, stored temporarily, and neutralized. From this facility, neutralized waste, containing radionuclides, is transferred to the Tank Farms. A 700-gallon sodium-hydroxide supply tank is also located in this facility. This plan provides the methods used to meet the acceptance criteria required by the 204-AR Waste Receiving Facility

  16. Current high-level waste solidification technology

    International Nuclear Information System (INIS)

    Bonner, W.F.; Ross, W.A.

    1976-01-01

    Technology has been developed in the U.S. and abroad for solidification of high-level waste from nuclear power production. Several processes have been demonstrated with actual radioactive waste and are now being prepared for use in the commercial nuclear industry. Conversion of the waste to a glass form is favored because of its high degree of nondispersibility and safety

  17. Environmental monitoring report for commercial low-level radioactive waste disposal sites (1960's through 1990's)

    International Nuclear Information System (INIS)

    1996-11-01

    During the time period covered in this report (1960's through early 1990's), six commercial low-level radioactive waste (LLRW) disposal facilities have been operated in the US. This report provides environmental monitoring data collected at each site. The report summarizes: (1) each site's general design, (2) each site's inventory, (3) the environmental monitoring program for each site and the data obtained as the program has evolved, and (4) what the program has indicated about releases to off-site areas, if any, including a statement of the actual health and safety significance of any release. A summary with conclusions is provided at the end of each site's chapter. The six commercial LLRW disposal sites discussed are located near: Sheffield, Illinois; Maxey Flats, Kentucky; Beatty, Nevada; West Valley, New York; Barnwell, South Carolina; Richland, Washington

  18. SIGNIFICANT PROGRESS IN THE DEPLOYMENT OF NEW TECHNOLOGIES FOR THE RETRIEVAL OF HANFORD RADIOACTIVE WASTE STORAGE TANKS

    International Nuclear Information System (INIS)

    RAYMOND RE; DODD RA; CARPENTER KE; STURGES MH

    2008-01-01

    Significant enhancements in the development and deployment of new technologies for removing waste from storage tanks at the Hanford Site have resulted in accelerated progress and reduced costs for tank cleanup. CH2M HILL Hanford Group, Inc. is the U.S. Department of Energy, Office of River Protection's prime contractor responsible for safely storing and retrieving approximately 53 million gallons of highly-radioactive and hazardous waste stored in 177 underground tanks. The waste is stored in 149 older single-shell tanks (SST) and 28 newer double-shell tanks (DST) that are grouped in 18 so-called farms near the center of the Hanford Site, located in southeastern Washington State. Tank contents include materials from years of World-War II and post-war weapons production, which account for 60 percent by volume of the nation's high-level radioactive waste. A key strategy for improved cleanup is the development and deployment of innovative technologies, which enhance worker safety, resolve technical challenges, streamline retrieval processes, and cut project costs and durations. During the past seven years of tank cleanout projects we have encountered conditions and waste chemistry that defy conventional approaches, requiring a variety of new tools and techniques. Through the deployment of advanced technology and the creative application of resources, we are finding ways to accomplish the retrieval process safely, swiftly, and economically. To date, retrieval operations have been completed in seven tanks, including a record six tanks in a two-year period. Retrieval operations are in progress for another three tanks. This paper describes the following tank cleanup technologies deployed at Hanford in the past few years: Modified waste sluicing, High pressure water lance, Mobile retrieval tools, Saltcake dissolution, Vacuum retrieval, Sparging of wastes, Selective dissolution for waste treatment, Oxalic acid dissolution, High-pressure water mixers, Variable height pumps

  19. DESTRUCTION OF TETRAPHENYLBORATE IN TANK 48H USING WET AIR OXIDATION BATCH BENCH SCALE AUTOCLAVE TESTING WITH ACTUAL RADIOACTIVE TANK 48H WASTE

    Energy Technology Data Exchange (ETDEWEB)

    Adu-Wusu, K; Paul Burket, P

    2009-03-31

    Wet Air Oxidation (WAO) is one of the two technologies being considered for the destruction of Tetraphenylborate (TPB) in Tank 48H. Batch bench-scale autoclave testing with radioactive (actual) Tank 48H waste is among the tests required in the WAO Technology Maturation Plan. The goal of the autoclave testing is to validate that the simulant being used for extensive WAO vendor testing adequately represents the Tank 48H waste. The test objective was to demonstrate comparable test results when running simulated waste and real waste under similar test conditions. Specifically: (1) Confirm the TPB destruction efficiency and rate (same reaction times) obtained from comparable simulant tests, (2) Determine the destruction efficiency of other organics including biphenyl, (3) Identify and quantify the reaction byproducts, and (4) Determine off-gas composition. Batch bench-scale stirred autoclave tests were conducted with simulated and actual Tank 48H wastes at SRNL. Experimental conditions were chosen based on continuous-flow pilot-scale simulant testing performed at Siemens Water Technologies Corporation (SWT) in Rothschild, Wisconsin. The following items were demonstrated as a result of this testing. (1) Tetraphenylborate was destroyed to below detection limits during the 1-hour reaction time at 280 C. Destruction efficiency of TPB was > 99.997%. (2) Other organics (TPB associated compounds), except biphenyl, were destroyed to below their respective detection limits. Biphenyl was partially destroyed in the process, mainly due to its propensity to reside in the vapor phase during the WAO reaction. Biphenyl is expected to be removed in the gas phase during the actual process, which is a continuous-flow system. (3) Reaction byproducts, remnants of MST, and the PUREX sludge, were characterized in this work. Radioactive species, such as Pu, Sr-90 and Cs-137 were quantified in the filtrate and slurry samples. Notably, Cs-137, boron and potassium were shown as soluble as a

  20. Hazardous waste and environmental trade: China`s issues

    Energy Technology Data Exchange (ETDEWEB)

    Ma Jiang [National Research Center for Science and Technology for Development, Beijing (China)

    1996-12-31

    By presenting some case studies, this paper analyzes China`s situation with regard to hazardous waste: its environmental trade, treatment, and management. The paper describes China`s experiences with the environmental trade of hazardous waste in both the internal and international market. Regulations for managing the import of waste are discussed, as are China`s major approaches to the trading of hazardous waste both at home and overseas. The major reasons for setting up the Asian-Pacific Regional Training Center for Technology Transfer and Environmental Sound Management of Wastes in China and the activities involved in this effort are also described. 1 tab.

  1. DOE management of high-level waste at the Hanford Site

    International Nuclear Information System (INIS)

    1993-01-01

    Approximately 60 million gallons of high-level radioactive waste--caustic liquids, slurries, saltcakes, and sludges--are stored in underground tanks at the Department of Energy's Hanford Site. At least one-third of the tanks are known to have leaked waste into the enviroranent, and there are many unresolved tank safety issues. In order to resolve the environmental and safety concerns, the Department plans to retrieve the waste, immobilize it, and dispose of it in a permanent geologic repository. Processing all of the tank waste in this manner could cost $40 billion, including $1.2 billion to construct the Hanford Waste Vitrification Plant. The purpose of our audit was to examine the reasons for cost estimate increases and schedule delays on the Hanford vitrification program. We also wanted to report on outstanding technical, safety, and environmental issues that could make the project even more costly and further delay its completion. We found that the Department managed the Hanford remediation system as a number of separate projects not fully integrated into one major system acquisition. Total costs have, therefore, been obscured, and the Department has not yet clearly defined system requirements or developed overall cost and schedule baselines. This lack of visibility could result in additional cost growth and schedule delays. We also noted a vast array of technical uncertainties, including tank safety and inadequate information about the makeup of tank waste, that could significantly affect the program's cost and ultimate success. To increase visibility of program cost and schedule, we are recommending that all separate projects relating to tank waste be included in a single major system acquisition, and that the Department complete its ongoing baselining effort to the extent practical before making major funding commitments. Management concurred with our finding and recommendations

  2. Strategic solid waste management in cities in Japan

    International Nuclear Information System (INIS)

    Tanaka, M.

    2005-01-01

    SWM (Solid Waste Management) systems have always been compatible with the societal need at every point of time. In 1950's it was oriented towards maintaining public health standards mainly to control infectious diseases. While in 1970's energy generation was considered as the vital aspect of the system. In 1990's reduction in waste generation and recycling were officially incorporated in the waste management regulation. By enacting basic law in 2000 A.D.; the society is poised to become a recycling based society in its drive towards sustainable society. The document explain the actual solid waste strategic management, and related issues, in Japan [it

  3. National emission standards for hazardous air pollutants application for approval to construct rotary mode core-sampling truck and exhauster

    International Nuclear Information System (INIS)

    1993-05-01

    Characterization of wastes in the underground single-shell tanks and double-shell tanks on the Hanford Site is crucial in developing the final disposal options for the waste and closure strategy for the Hanford Site. Additionally, characterization of tank waste is important for the waste tank safety programs. The Hanford Federal Facility Agreement and Consent Order (also referred to as the Tri-Party Agreement) Milestone M-10-00 requires the obtaining and analyzing of at least two samples from each single-shell tank, and Milestone M-10-13 specifically requires the ability to sample hard saltcake. Existing equipment does not allow sampling of all single-shell tanks within established tank safety limits. Consequently, the US Department of Energy, Richland Operations Office has developed a rotary mode core-sampling system that uses nitrogen gas to cool and clear the drill bit. A rotary mode core-sampling truck will be used on approximately 80 single-shell tanks which contain saltcake wastes, and will provide crucial information on the contents of the tanks. This application is a request for approval to construct and operate the rotary mode core-sampling truck and exhauster in the 200 East and 200 West Area Tank Farms of the Hanford Site. This request is being made pursuant to 40 CFR 61, Subpart H

  4. Kierkegaard’s Ethical Stage In Hegel’s Logical Categories: Actual Possibility, Reality And Necessity

    Directory of Open Access Journals (Sweden)

    María J. Binetti

    2007-12-01

    Full Text Available During decades, the history of philosophy has kept Kierkegaardrsquo;s and Hegelrsquo;s thought apart, and their long-standing opposition has swept through the speculative greatness of Kierkegaardian existentialism and the existential power of Hegelian philosophy. In contrast to such unfortunate misinterpretation, this article aims at showing the deep convergence that relates interiorly the Kierkegaardian ethical stage with the most important Hegelian logic categories. Kierkegaard and Hegel conceive of the idea as the real power of subjective becoming, and the existence as the actual concretion of the ideal. To both of them, the pure emeneacute;rgeia/em of freedom, which starts in the abstract and aesthetical possibility of the subjective immediacy, realizes itself as the actual concretion of finitude, assuming time and contingency by the eternal and necessary force of duty. The Kierkegaardian repetition is nothing but this powerful idea, mediating the flux of finite differences in the eternal identity of subject. However, for Kierkegaard as well as for Hegel there is an absolute contradiction, which promotes the overcoming of ethics.

  5. Organic Tank Safety Project: development of a method to measure the equilibrium water content of Hanford organic tank wastes and demonstration of method on actual waste

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, R.D.; Bredt, P.R.; Sell, R.L.

    1996-09-01

    Some of Hanford`s underground waste storage tanks contain Organic- bearing high level wastes that are high priority safety issues because of potentially hazardous chemical reactions of organics with inorganic oxidants in these wastes such as nitrates and nitrites. To ensure continued safe storage of these wastes, Westinghouse Hanford Company has placed affected tanks on the Organic Watch List and manages them under special rules. Because water content has been identified as the most efficient agent for preventing a propagating reaction and is an integral part of the criteria developed to ensure continued safe storage of Hanford`s organic-bearing radioactive tank wastes, as part of the Organic Tank Safety Program the Pacific Northwest National Laboratory developed and demonstrated a simple and easily implemented procedure to determine the equilibrium water content of these potentially reactive wastes exposed to the range of water vapor pressures that might be experienced during the wastes` future storage. This work focused on the equilibrium water content and did not investigate the various factors such as @ ventilation, tank surface area, and waste porosity that control the rate that the waste would come into equilibrium, with either the average Hanford water partial pressure 5.5 torr or other possible water partial pressures.

  6. Packaging design criteria (onsite) project W-520 immobilized low-activity waste transportation system

    International Nuclear Information System (INIS)

    BOEHNKE, W.M.

    2001-01-01

    A plan is currently in place to process the high-level radioactive wastes that resulted from uranium and plutonium recovery operations from Spent Nuclear Fuel at the Hanford Site, Richland, Washington. Currently, millions of gallons of high-level radioactive waste in the form of liquids, sludges, and saltcake are stored in many large underground tanks onsite. This waste will be processed and separated into high-level and low-activity fractions. Both fractions will then be vitrified (i.e., blended with molten borosilicate glass) in order to encapsulate the toxic radionuclides. The immobilized low-activity waste (ILAW) glass will be poured into LAW canisters, allowed to cool and harden to solid form, sealed by welding, and then transported to a double-lined trench in the 200 East Area for permanent disposal. This document presents the packaging design criteria (PDC) for an onsite LAW transportation system, which includes the ILAW canister, ILAW package, and transport vehicle and defines normal and accident conditions. This PDC provides the basis for the ILAW onsite transportation system design and fabrication and establishes the transportation safety criteria that the design will be evaluated against in the Package Specific Safety Document (PSSD). It provides the criteria for the ILAW canister, cask and transport vehicles and defines normal and accident conditions. The LAW transportation system is designed to transport stabilized waste from the vitrification facility to the ILAW disposal facility developed by Project W-520. All ILAW transport will take place within the 200 East Area (all within the Hanford Site)

  7. How People Actually Use Thermostats

    Energy Technology Data Exchange (ETDEWEB)

    Meier, Alan; Aragon, Cecilia; Hurwitz, Becky; Mujumdar, Dhawal; Peffer, Therese; Perry, Daniel; Pritoni, Marco

    2010-08-15

    Residential thermostats have been a key element in controlling heating and cooling systems for over sixty years. However, today's modern programmable thermostats (PTs) are complicated and difficult for users to understand, leading to errors in operation and wasted energy. Four separate tests of usability were conducted in preparation for a larger study. These tests included personal interviews, an on-line survey, photographing actual thermostat settings, and measurements of ability to accomplish four tasks related to effective use of a PT. The interviews revealed that many occupants used the PT as an on-off switch and most demonstrated little knowledge of how to operate it. The on-line survey found that 89% of the respondents rarely or never used the PT to set a weekday or weekend program. The photographic survey (in low income homes) found that only 30% of the PTs were actually programmed. In the usability test, we found that we could quantify the difference in usability of two PTs as measured in time to accomplish tasks. Users accomplished the tasks in consistently shorter times with the touchscreen unit than with buttons. None of these studies are representative of the entire population of users but, together, they illustrate the importance of improving user interfaces in PTs.

  8. Ferrocyanide safety project: Comparison of actual and simulated ferrocyanide waste properties

    International Nuclear Information System (INIS)

    Scheele, R.D.

    1995-10-01

    In 1995, available subsegment samples of wastes taken from the Hanford Site underground radioactive waste storage tanks 241-C-112 (C-112) and 241-C-109 (C-109) were reanalyzed to determine the nickel concentrations in the samples and to determine whether the use of a nickel crucible in the analytical sample preparation biased the reported nickel concentrations reported by Simpson and coworkers and in the original report that this report supplements. The reanalysis strategy to determine nickel was to use a sodium peroxide flux in a zirconium crucible instead of the previously used potassium hydroxide flux in a nickel crucible. This supplemental report provides the results of the reanalyses and updates tables from the original report which reflect the new nickel analyses. Nickel is important with respect to management of the potentially reactive ferrocyanide wastes as it is one of the key defining characteristics of the solids that resulted from scavenging radiocesium using ferrocyanides. In Hanford Site wastes, few other processes introduced nickel into the wastes other than radiocobalt scavenging, which was often coupled with the ferrocyanide-scavenging process. Thus the presence of nickel in a waste provides strong evidence that the original waste was or contained ferrocyanide waste at one time. Given the potential import of nickel as a defining characteristic and marker for ferrocyanide wastes, the Pacific Northwest Laboratory's (PNL) Analytical Chemistry Laboratory (ACL) reanalyzed available samples from tanks C-112 and C-109 using inductively coupled argon plasma/atomic emission spectrometry (ICP/AES) and an alternative sample preparation method which precluded contamination of the analytical samples with nickel

  9. Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    Energy Technology Data Exchange (ETDEWEB)

    Morrey, E.V.; Tingey, J.M.; Elliott, M.L.

    1996-10-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

  10. Comparison of simulants to actual neutralized current acid waste: Process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    Energy Technology Data Exchange (ETDEWEB)

    Morrey, E.V.; Tingey, J.M.

    1996-04-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs have been established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste is being performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions.

  11. Comparison of simulants to actual neutralized current acid waste: process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    International Nuclear Information System (INIS)

    Morrey, E.V.; Tingey, J.M.; Elliott, M.L.

    1996-10-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs were established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste was performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property ,models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions

  12. Comparison of simulants to actual neutralized current acid waste: Process and product testing of three NCAW core samples from Tanks 101-AZ and 102-AZ

    International Nuclear Information System (INIS)

    Morrey, E.V.; Tingey, J.M.

    1996-04-01

    A vitrification plant is planned to process the high-level waste (HLW) solids from Hanford Site tanks into canistered glass logs for disposal in a national repository. Programs have been established within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) Project to test and model simulated waste to support design, feed processability, operations, permitting, safety, and waste-form qualification. Parallel testing with actual radioactive waste is being performed on a laboratory-scale to confirm the validity of using simulants and glass property models developed from simulants. Laboratory-scale testing has been completed on three radioactive core samples from tanks 101-AZ and 102-AZ containing neutralized current acid waste (NCAW), which is one of the first waste types to be processed in the high-level waste vitrification plant under a privatization scenario. Properties of the radioactive waste measured during process and product testing were compared to simulant properties and model predictions to confirm the validity of simulant and glass property models work. This report includes results from the three NCAW core samples, comparable results from slurry and glass simulants, and comparisons to glass property model predictions

  13. Leach tests on grouts made with actual and trace metal-spiked synthetic phosphate/sulfate waste

    International Nuclear Information System (INIS)

    Serne, R.J.; Martin, W.J.; LeGore, V.L.; Lindenmeier, C.W.; McLaurine, S.B.; Martin, P.F.C.; Lokken, R.O.

    1989-10-01

    Pacific Northwest Laboratory conducted experiments to produce empirical leach rate data for phosphate-sulfate waste (PSW) grout. Effective diffusivities were measured for various radionuclides ( 90 Sr, 99 Tc, 14 C, 129 I, 137 Cs, 60 Co, 54 Mn, and U), stable major components (NO 3 - , SO 4 2- , H 3 BO 3 , K and Na) and the trace constituents Ag, As, Cd, Hg, Pb, and Se. Two types of leach tests were used on samples of actual PSW grout and synthetic PSW grout: the American Nuclear Society (ANS) 16.1 intermittent replacement leach test and a static leach test. Grout produced from both synthetic and real PSW showed low leach rates for the trace metal constituents and most of the waste radionuclides. Many of the spiked trace metals and radionuclides were not detected in any leachates. None of the effluents contained measurable quantities of 137 Cs, 60 Co, 54 Mn, 109 Cd, 51 Cr, 210 Pb, 203 Hg, or As. For those trace species with detectable leach rates, 125 I appeared to have the greatest leach rate, followed by 99 Tc, 75 Se, and finally U, 14 C, and 110m Ag. Leach rates for nitrate are between those for I and Tc, but there is much scatter in the nitrate data because of the very low nitrate inventory. 32 refs., 6 figs., 15 tabs

  14. Hanford tank initiative vehicle/based waste retrieval demonstration report phase II, track 2

    International Nuclear Information System (INIS)

    Berglin, E.J.

    1997-01-01

    Using the versatile TracPUMpTm, Environmental Specialties Group, LLC (ES) performed a successful Phase 11 demonstration of a Vehicle- Based Waste Retrieval System (VWRS) for removal of waste material and residual liquid found in the Hanford Underground Storage Tanks (ousts). The purpose of this demonstration was to address issues pertaining to the use of a VWRS in OUSTS. The demonstration also revealed the waste removal capabilities of the TracPumpTm and the most effective techniques and equipment to safely and effectively remove waste simulants. ES successfully addressed the following primary issues: I . Dislodge and convey the waste forms present in the Hanford OUSTS; 2. Access the UST through tank openings as small as twenty-four inches in diameter; 3. Traverse a variety of terrains including slopes, sludges, rocks and hard, slippery surfaces without becoming mired; 4. Dislodge and convey waste within the confinement of the Decontamination Containment Capture Vessel (DCCV) and with minimal personnel exposure; 5. Decontaminate equipment to acceptable limits during retrieval from the UST; 6. Perform any required maintenance within the confinement of the DCCV; and 7. Maintain contaminate levels ''as low as reasonably achievable'' (ALARA) within the DCCV due to its crevice and comer-free design. The following materials were used to simulate the physical characteristics of wastes found in Hanford's OUSTS: (1) Hardpan: a clay-type material that has high shear strength; (2) Saltcake: a fertilizer-based material that has high compressive strength; and (3) Wet Sludge.- a sticky, peanut- butter- like material with low shear strength. Four test beds were constructed of plywood and filled with a different simulant to a depth of eight to ten inches. Three of the test beds were of homogenous simulant material, while the fourth bed consisted of a mixture of all three simulant types

  15. Basis for a Waste Management Public Communication Policy: Actual Situation Analysis and Implementation of Corrective Actions

    Energy Technology Data Exchange (ETDEWEB)

    Jolivet, L. A.; Maset, E. R.

    2002-02-28

    Argentina will require new sites for the location of radioactive waste final disposal systems. It is currently mandatory to have social and political consensus to obtain the corresponding agreements. The experience obtained with the cancellation of the project ''Feasibility Study and Engineering Project--Repository for High Level Radioactive Waste'', reinforces even more the necessity to count with the acceptance of the public to carry out projects of this kind. The first phase of the former was developed in the 80's: geological, geophysical and hydrogeological studies were performed in a compact granitic rock located in Sierra del Medio, Chubut province. This project had to be called off in the early 90's due to strong social rejection. This decision was closely related to the poor attention given to social communication issues. The governmental decision-makers in charge underwent a lot of pressure from social groups claiming for the cancellation of the project due to the lack of information and the fear it triggered. Thus, the lesson learnt: ''social communication activities must be carefully undertaken in order to achieve the appropriate management of the radioactive waste produced in our country.'' The same as in other countries, the specific National Law demands the formulation of a Strategic Plan which will not only include the research into radioactive waste, but the design of a Social Communication Programme as well. The latter will be in charge of informing the population clearly and objectively about the latest scientific and technological advances in the issue. A tentative perception-attitude pattern of the Argentine society about the overall nuclear issue is outlined in this paper. It is meant to contribute to the understanding of the public's adverse reaction to this kind of project. A communication programme is also presented. Its objective is to install the waste management topic in the public's

  16. Integrated solid waste management of Minneapolis, Minnesota

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The subject document reports the results of an in-depth investigation of the fiscal year 1992 cost of the City of Minneapolis, Minnesota (Hennepin County) integrated municipal solid waste management (IMSWM) system, the energy consumed to operate the system, and the environmental performance requirements for each of the system`s waste-processing and disposal facilities. Actual data from records kept by participants is reported in this document. Every effort was made to minimize the use of assumptions, and no attempt is made to interpret the data reported. Analytical approaches are documented so that interested analysts may perform manipulation or further analysis of the data. As such, the report is a reference document for municipal solid waste (MSW) management professionals who are interested in the actual costs and energy consumption for a one-year period, of an operating IMSWM system.

  17. Why are our children wasting: Determinants of wasting among under 5s in Ghana.

    Science.gov (United States)

    Darteh, Eugene Kofuor Maafo; Acquah, Evelyn; Darteh, Florie

    2017-09-01

    Wasting is one of the indicators of malnutrition known to contribute to the deaths occurring from childhood malnutrition. It is the measure of body mass in relation to body length used to explain recent nutritional status. This paper examines the determinants of wasting among under 5s in Ghana. Data were drawn from the 2014 Ghana Demographic and Health Survey children's records file to examine the determinants of wasting among children. A total of 2720 children under 5 years with valid anthropometric data were used. Data on wasting were collected by measuring the weight and height of all children under 5 years of age. Bi-variate and multi-variate statistics are used to examine the determinants of wasting. The bi-variate analysis showed significant differences ( p 5s according to age of the child, region, and wealth status. On the other hand, the multi-variate analysis revealed that the odds of wasting were lower among children aged 24-35 months (Odds ratio (OR) = 0.37; p 5s. Also, efforts should be made by the relevant government agencies and other stakeholders to strengthen the socio-economic status of mothers to enable them to provide adequate nutrition and improve access to health insurance for their children in order to reduce the incidence of wasting among these children.

  18. RCRA Permit for a Hazardous Waste Management Facility Permit Number NEV HW0101 Annual Summary/Waste Minimization Report Calendar Year 2011

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Environmental Restoration

    2012-02-16

    This report summarizes the U.S. Environmental Protection Agency (EPA) identification number of each generator from which the Permittee received a waste stream; a description and quantity of each waste stream in tons and cubic feet received at the facility; the method of treatment, storage, and/or disposal for each waste stream; a description of the waste minimization efforts undertaken; a description of the changes in volume and toxicity of waste actually received; any unusual occurrences; and the results of tank integrity assessments. This Annual Summary/Waste Minimization Report is prepared in accordance with Section 2.13.3 of Permit Number NEV HW0101.

  19. The S-curve for forecasting waste generation in construction projects.

    Science.gov (United States)

    Lu, Weisheng; Peng, Yi; Chen, Xi; Skitmore, Martin; Zhang, Xiaoling

    2016-10-01

    Forecasting construction waste generation is the yardstick of any effort by policy-makers, researchers, practitioners and the like to manage construction and demolition (C&D) waste. This paper develops and tests an S-curve model to indicate accumulative waste generation as a project progresses. Using 37,148 disposal records generated from 138 building projects in Hong Kong in four consecutive years from January 2011 to June 2015, a wide range of potential S-curve models are examined, and as a result, the formula that best fits the historical data set is found. The S-curve model is then further linked to project characteristics using artificial neural networks (ANNs) so that it can be used to forecast waste generation in future construction projects. It was found that, among the S-curve models, cumulative logistic distribution is the best formula to fit the historical data. Meanwhile, contract sum, location, public-private nature, and duration can be used to forecast construction waste generation. The study provides contractors with not only an S-curve model to forecast overall waste generation before a project commences, but also with a detailed baseline to benchmark and manage waste during the course of construction. The major contribution of this paper is to the body of knowledge in the field of construction waste generation forecasting. By examining it with an S-curve model, the study elevates construction waste management to a level equivalent to project cost management where the model has already been readily accepted as a standard tool. Copyright © 2016 Elsevier Ltd. All rights reserved.

  20. DEVELOPMENT OF AN INSOLUBLE SALT SIMULANT TO SUPPORT ENHANCED CHEMICAL CLEANING TESTS

    International Nuclear Information System (INIS)

    Eibling, R

    2008-01-01

    The closure process for high level waste tanks at the Savannah River Site will require dissolution of the crystallized salts that are currently stored in many of the tanks. The insoluble residue from salt dissolution is planned to be removed by an Enhanced Chemical Cleaning (ECC) process. Development of a chemical cleaning process requires an insoluble salt simulant to support evaluation tests of different cleaning methods. The Process Science and Engineering section of SRNL has been asked to develop an insoluble salt simulant for use in testing potential ECC processes (HLE-TTR-2007-017). An insoluble salt simulant has been developed based upon the residues from salt dissolution of saltcake core samples from Tank 28F. The simulant was developed for use in testing SRS waste tank chemical cleaning methods. Based on the results of the simulant development process, the following observations were developed: (1) A composition based on the presence of 10.35 grams oxalate and 4.68 grams carbonate per 100 grams solids produces a sufficiently insoluble solids simulant. (2) Aluminum observed in the solids remaining from actual waste salt dissolution tests is probably precipitated from sodium aluminate due to the low hydroxide content of the saltcake. (3) In-situ generation of aluminum hydroxide (by use of aluminate as the Al source) appears to trap additional salts in the simulant in a manner similar to that expected for actual waste samples. (4) Alternative compositions are possible with higher oxalate levels and lower carbonate levels. (5) The maximum oxalate level is limited by the required Na content of the insoluble solids. (6) Periodic mixing may help to limit crystal growth in this type of salt simulant. (7) Long term storage of an insoluble salt simulant is likely to produce a material that can not be easily removed from the storage container. Production of a relatively fresh simulant is best if pumping the simulant is necessary for testing purposes. The insoluble

  1. Forecast of radionuclides release from actual waste form geometries

    International Nuclear Information System (INIS)

    Suarez, A.A.; Rzyski, B.M.; Sato, I.M.

    1989-01-01

    The complete understanding of the leaching mechanism of radionuclides from solid comentitious waste forms is still far from being reached. Much effort has been devoted, however, to identifying and explaining the main components which contribute to the dispersal of radionuclides out of the waste form to the environment. This is of prime importance when short term results are extrapolated into the future. The diffusion coefficient evaluation, based on experimental leaching data obtained from samples produced from the same batch was performed using the exact diffusion formulation applied to real geometric sample shape. This paper discusses the evaluation

  2. Simulation and characterization of a Hanford high-level waste slurry

    International Nuclear Information System (INIS)

    Russell, R.L.; Smith, H.D.

    1996-09-01

    The baseline waste used for this simulant is a blend of wastes from tanks 101-AZ, 102-AZ, 106-C, and 102-AY that have been through water washing. However, the simulant used in this study represents a combination of tank waste slurries and should be viewed as an example of the slurries that might be produced by blending waste from various tanks. It does not imply that this is representative of the actual waste that will be delivered to the privatization contractor(s). This blended waste sludge simulant was analyzed for grain size distribution, theological properties both as a function of concentration and aging, and calcining characteristics. The grain size distribution allows a comparison with actual waste with respect to theological properties. Slurries with similar grain size distributions of the same phases are expected to exhibit similar theological properties. Rheological properties may also change because of changes in the slurry's particulate supernate chemistry due to aging. Low temperature calcination allows the potential for hazardous gas generation to be investigated

  3. USE OF CANDEIA’S (Eremanthus erythropappus WASTE WOOD

    Directory of Open Access Journals (Sweden)

    Rosimeire Cavalcante dos Santos

    2008-09-01

    Full Text Available The candeia (Eremanthus erythropappus is a native forest species with multiple uses and specially utilized as essential oils source. The use of the candeia´s waste wood after oil extraction for particle panels production becomes a viable alternative, avoiding environmental problems and increasing the availability of these products in the consuming market. This work verified the viability of producing wood-cement panels using waste wood generated after the extraction of candeia’s oil, in association with pinus and eucalipto woods. The experiment was installed according to a completely randomized design with three repetitions. The treatments were arranged according to a factorial 2 x 3 scheme (two wooden species and three replacement percentages of the woods by candeia’s waste. The results of the physical and mechanical property tests showed high potentiality of candeia waste wood, after oil extraction, in association with pinus and eucalipto wood for manufacturing wood-cement panels.

  4. Incorporation of Savannah River Plant radioactive waste into concrete

    International Nuclear Information System (INIS)

    Stone, J.A.

    1975-01-01

    Results are reported of a laboratory-scale experimental program at the Savannah River Laboratory to gain information on the fixation of high-level radioactive wastes in concrete. Two concrete formulations, a High-Alumina Cement and a Portland Pozzalanic cement, were selected on the bases of leachability and compressive strength for the fixation of non-radioactive simulated wastes. Therefore, these two cements were selected for current studies for the fixation of actual Savannah River Plant high-level wastes. (U.S.)

  5. El estrés en personal sanitario hospitalario: estado actual

    Directory of Open Access Journals (Sweden)

    José Ignacio Aguado Martín

    2013-06-01

    Full Text Available Introducción: Los riesgos psicosociales y en particular el estrés se han convertido en uno de los principales problemas para la salud y la seguridad laboral. Se ha realizado una revisión de la literatura científica actual sobre el estrés debido a su importancia como riesgo emergente, y la repercusión de este problema en la productividad y el deterioro de la calidad de vida en el personal sanitario que desarrolla su actividad profesional en medio hospitalario, así como intentar identificar los principales métodos de evaluación del estrés actualmente utilizados. Objetivos: Conocer el estado actual del estrés en el personal hospitalario e identificar los principales métodos de evaluación. Metodología: Se realizo una búsqueda bibliográfica exploratoria utilizando las bases de datos Medline, Embase, Scirus, Wok, Psicoinfo, OSH update, Cisdoc, IBECS y LILACS, así como en repositorios como Dialnet, SciELO y Google académico. El periodo de estudio de la bibliografía publicada fue el comprendido entre el año 2009 y 2012 realizándose la última búsqueda el 10 de diciembre de 2012. Resultados: Los niveles de estrés encontrados en los artículos revisados fueron moderados-altos, con diferencias entre las distintas áreas y especialidades. Factores individuales tanto objetivos como subjetivos parecen influir en los grados de estrés. Existe diversidad de métodos utilizados para la evaluación del estrés, la mayoría tienen un nivel de fiabilidad elevado. La producción bibliográfica encontrada es mayor en Asia y Europa. Conclusiones: Los niveles de estrés son elevados en el personal sanitario del hospital ,sobre todo en algunas areas y especialidades, existen una serie de moderadores del grado de estrés como los factores individuales y el ambiente físico y no existe un consenso internacional en el método utilizado para su evaluación.

  6. Corrosion performance of Cr3C2-NiCr+0.2%Zr coated super alloys under actual medical waste incinerator environment

    Science.gov (United States)

    Ahuja, Lalit; Mudgal, Deepa; Singh, Surendra; Prakash, Satya

    2018-03-01

    Incineration techniques are widely used to dispose of various types of waste which lead to formation of very corrosive environment. Such corrosive environment leads to the degradation of the alloys used in these areas. To obviate this problem, zirconium modified Cr3C2-(NiCr) coating powder has been deposited on three superalloys namely Superni 718, Superni 600 and Superco 605 using Detonation gun technique. Corrosion test was conducted in actual medical waste incinerator environment. The samples were hung inside the secondary chamber operated at 1050°C for 1000h under cyclic condition. Corrosion kinetics was monitored using the weight gain measurements and thickness loss. Corrosion products were characterized using scanning electron microscopy, energy dispersive spectroscopy and X-ray diffraction technique. It was observed that coating is found to be successful in impeding the corrosion problem in superalloys.

  7. Corrosion of carbon steel in saturated high-level waste salt solutions

    International Nuclear Information System (INIS)

    Wiersma, B.J.; Parish, W.R.

    1997-01-01

    High level waste stored as crystallized salts is to be removed from carbon steel tanks by water dissolution. Dissolution of the saltcake must be performed in a manner which will not impact the integrity of the tank. Corrosion testing was performed to determine the amount of corrosion inhibitor that must be added to the dissolution water in order to ensure that the salt solution formed would not induce corrosion degradation of the tank materials. The corrosion testing performed included controlled potential slow strain rate, coupon immersion, and potentiodynamic polarization tests. These tests were utilized to investigate the susceptibility of the cooling coil material to stress corrosion cracking in the anticipated environments. No evidence of SCC was observed in any of the tests. Based on these results, the recommended corrosion requirements were that the temperature of the salt solution be less than 50 degrees C and that the minimum hydroxide concentration be 0.4 molar. It was also recommended that the hydroxide concentration not stay below 0.4 molar for longer than 45 days

  8. Waste minimization -- Hanford`s strategy for sustainability

    Energy Technology Data Exchange (ETDEWEB)

    Merry, D.S.

    1998-01-30

    The Hanford Site cleanup activity is an immense and challenging undertaking, which includes characterization and decommissioning of 149 single-shell storage tanks, treating waste stored in 28 double-shell tanks, safely disposing of over 2,100 metric tons of spent nuclear fuel stored onsite, removing thousands of structures, and dealing with significant solid waste, groundwater, and land restoration issues. The Pollution Prevention/Waste Minimization (P2/WMin) Program supports the Hanford Site mission to safely clean up and manage legacy waste and to develop and deploy science and technology in many ways. Once such way is through implementing and documenting over 231 waste reduction projects during the past five years, resulting in over $93 million in cost savings/avoidances. These savings/avoidances allowed other high priority cleanup work to be performed. Another way is by exceeding the Secretary of Energy`s waste reduction goals over two years ahead of schedule, thus reducing the amount of waste to be stored, treated and disposed. Six key elements are the foundation for these sustained P2/WMin results.

  9. “Cinematic” Gravity’s Rainbow: Indiscernibility of the Actual and the Virtual

    Directory of Open Access Journals (Sweden)

    Grmusa Lovorka Gruic

    2017-11-01

    Full Text Available Acknowledging that all media interpenetrate without any one being privileged as original, this paper focuses on the influence and omnipresence of screen technologies, cinematography in particular, asserting its capacities to act as a dominant technological and cultural force, which surveys and controls the economy, affects human consciousness, and implements its own ideologically tainted reality. More specifically, the analysis demonstrates the impact of cinematography within Thomas Pynchon’s Gravity’s Rainbow, emphasizing intermedial reflexivity from screen to paper, its unfolding and animation, representations and mutations, but also the notion that virtuality dominates actuality (Virilio, The Vision 63, or, in Deleuze’s terms, the two coalesce within the crystal-image, for the real and the imaginary, the actual and the virtual, although distinct, are indiscernible (Deleuze, Cinema 2 68-69. The study is grounded in Paul Virilio’s theoretical framework and his criticism of the colonization of the real by the virtual, and in Gilles Deleuze’s account on cinematic taxonomy and how cinematography partakes in the emergence of a new notion of reality and history. By appropriating, repurposing, and reshaping the techniques, forms, and contents of cinematography, and at the same time being critical of its effects, Pynchon uses paper surface for transition of his ideas, validating the repercussions of intermediation as a cultural force and unveiling literature’s performativity, its ability to shelter/entertain the “cinematic,” which in turn revolutionizes and animates fiction.

  10. Comparison of organic and inorganic ion exchangers for removal of cesium and strontium from simulated and actual Hanford 241-AW-101 DSSF tank waste

    International Nuclear Information System (INIS)

    Brown, G.N.; Bray, L.A.; Carlson, C.D.

    1996-04-01

    A number of organic and inorganic exchangers are being developed and evaluated for cesium removal from Hanford tank wastes. The exchangers of interest that are investigated in this work include powdered (IONSIV reg-sign IE-910; referred to as IE-910) and engineered (IONSIV reg-sign IE-911; referred to as IE-911) forms of the crystalline silico-titanate (CST) inorganic sorbent developed by Sandia National Laboratories (SNL)/Texas A and M and prepared by UOP; a phenol-formaldehyde (CS-100) resin developed by Rohm and Haas; a resorcinol-formaldehyde (R-F) polymer developed at the Westinghouse Savannah River Company (WSRC) and produced by Boulder Scientific; an inorganic zeolite exchanger produced by UOP (IONSIV reg-sign TIE-96; referred to as TIE-96); an inorganic sodium titanate produced by Allied Signal/Texas A and M (NaTi); and a macrocyclic organic resin developed and produced by IBC Advanced Technologies (SuperLig reg-sign 644; referred to as SL-644). Several of these materials are still under development and may not be in the optimal form. The work described in this report involves the direct comparison of the ion exchange materials for the pretreatment of actual and simulated Hanford tank waste. Data on the performance of all of the exchangers with simulated and actual double shell slurry feed (DSSF) is included. The DSSF waste is a mixture of the supernate from tanks 101-AW (70%), 106-AP (20%) and 102-AP (10%). The comparative parameters include radionuclide removal efficiency under a variety of conditions and material properties (e.g., bed density and percent removable water). Cesium and strontium distribution (K d ), lambda (λ = K d x ρ b ), and decontamination factors (DF) are compared as a function of exchanger contact duration, solution composition (Na and Cs concentration), exchanger/waste phase ratio, and multiple sequential contacts

  11. Let’s limit our waste production and let’s’ sort it!

    CERN Multimedia

    HSE Unit

    2013-01-01

    Let’s limit our waste production! – Why ? Preventing the production of waste is the best solution to avoid environmental issues, economic impacts and technical constraints. So, whenever you are involved in the design, manufacturing, distribution, use or dismantling of a product or an activity in general, always remember that the best waste is that which is not produced. The limitation of waste production being an HSE objective declared in 2013 by the CERN Director-General, we encourage everyone to help limit the amount of waste produced through CERN activities. Let’s sort it! – Why ? Since the 90s, CERN has implemented a policy to promote recovery of the waste* generated by its activities. Nowadays, CERN is committed to continuously improving its sorting and recovery and therefore various initiatives have been started by GS-IS to improve the recovery of waste (e.g. recovery of organic waste from restaurants; implementation of solar trash compactors - see Bulletin 27-...

  12. Basis for a Waste Management Public Communication Policy: Actual Situation Analysis and Implementation of Corrective Actions

    International Nuclear Information System (INIS)

    Jolivet, L. A.; Maset, E. R.

    2002-01-01

    Argentina will require new sites for the location of radioactive waste final disposal systems. It is currently mandatory to have social and political consensus to obtain the corresponding agreements. The experience obtained with the cancellation of the project ''Feasibility Study and Engineering Project--Repository for High Level Radioactive Waste'', reinforces even more the necessity to count with the acceptance of the public to carry out projects of this kind. The first phase of the former was developed in the 80's: geological, geophysical and hydrogeological studies were performed in a compact granitic rock located in Sierra del Medio, Chubut province. This project had to be called off in the early 90's due to strong social rejection. This decision was closely related to the poor attention given to social communication issues. The governmental decision-makers in charge underwent a lot of pressure from social groups claiming for the cancellation of the project due to the lack of information and the fear it triggered. Thus, the lesson learnt: ''social communication activities must be carefully undertaken in order to achieve the appropriate management of the radioactive waste produced in our country.'' The same as in other countries, the specific National Law demands the formulation of a Strategic Plan which will not only include the research into radioactive waste, but the design of a Social Communication Programme as well. The latter will be in charge of informing the population clearly and objectively about the latest scientific and technological advances in the issue. A tentative perception-attitude pattern of the Argentine society about the overall nuclear issue is outlined in this paper. It is meant to contribute to the understanding of the public's adverse reaction to this kind of project. A communication programme is also presented. Its objective is to install the waste management topic in the public's opinion with a positive real outlook

  13. Wow Technology’s innovative radioactive liquid waste treatment

    Energy Technology Data Exchange (ETDEWEB)

    Marin, A.

    2015-07-01

    WOW presents its revolutionary technology and equipment for liquid radioactive waste treatment: outperforming ultimate water decontamination and purification process, enhanced sludge concentration, no secondary waste nor consumables, fully automated, remote controlled and self-decontaminating device. The WOW’s technology is based upon a never before observed discovery of fluid dynamics science: the possibility of performing a molecular separation between solute and suspended elements and the solvent. The combination of such a molecular separation process with a standard vacuum evaporation improves the abatement performances by thousands of times, with respect to those of the state of the art vacuum evaporators. In addition to this, no secondary waste is produced during the process, as no filters, membranes, resins or additives are used. WOW equipment, automated and remote controlled, self decontaminates after use and can be designed and constructed either tailored to the application needs or with a modular approach for enhanced transportability and application flexibility. After the preliminary verification by CNR, the Italian National Research Center, Wow Technology decontamination device was tested c/o LENA, the Laboratory of Applied Nuclear Energy of the University of Pavia, Italy with a simulated solution 6000 times more contaminated than the nuclear reactor’s cooling water of Fukushima-Daiichi NPP. In addition to that, WOW Technology was also used in a real case at the Radiochemistry laboratory of the Pavia’s University Chemistry department. Both the above mentioned contaminated fluids have been successfully decontaminated without production of additional or secondary waste WOW Technology has already performed on industrial scale c/o the Nuclear Repository of S.S.M. in Saluggia, Italy: 45000 liters of acid radioactive solution have been successfully decontaminated to a Decontamination Factor (DF) of 335000 for Cs-137 by one single evaporation step and

  14. ELECTROCHEMICAL STUDIES OF CARBON STEEL CORROSION IN HANFORD DOUBLE SHELL TANK (DST) WASTE

    Energy Technology Data Exchange (ETDEWEB)

    DUNCAN, J.B.; WINDISCH, C.F.

    2006-10-13

    This paper reports on the electrochemical scans for the supernatant of Hanford double-shell tank (DST) 241-SY-102 and the electrochemical scans for the bottom saltcake layer for Hanford DST 241-AZ-102. It further reports on the development of electrochemical test cells adapted to both sample volume and hot cell constraints.

  15. Analysis of organic carbon and moisture in Hanford single-shell tank waste

    Energy Technology Data Exchange (ETDEWEB)

    Toth, J.J.; Heasler, P.G.; Lerchen, M.E.; Hill, J.G.; Whitney, P.D.

    1995-05-01

    This report documents a revised analysis performed by Pacific Northwest Laboratory involving the organic carbon laboratory measurement data for Hanford single-shell tanks (SSTs) obtained from a review of the laboratory analytical data. This activity has as its objective to provide a best-estimate, including confidence levels, of total organic carbon (TOC) and moisture in each of the 149 SSTs at Hanford. The TOC and moisture information presented in this report is useful as part of the criteria to identify SSTs for additional measurements, or monitoring for the Organic Safety Program. In April 1994, an initial study of the organic carbon in Hanford single-shell tanks was completed at PNL. That study reflected the estimates of TOC based on tank characterizations datasets that were available at the time. Also in that study, estimation of dry basis TOC was based on generalized assumptions pertaining to the moisture of the tank wastes. The new information pertaining to tank moisture and TOC data that has become available from the current study influences the best estimates of TOC in each of the SSTs. This investigation of tank TOC and moisture has resulted in improved estimates based on waste phase: saltcake, sludge, or liquid. This report details the assumptions and methodologies used to develop the estimates of TOC and moisture in each of the 149 SSTs at Hanford.

  16. Analysis of organic carbon and moisture in Hanford single-shell tank waste

    International Nuclear Information System (INIS)

    Toth, J.J.; Heasler, P.G.; Lerchen, M.E.; Hill, J.G.; Whitney, P.D.

    1995-05-01

    This report documents a revised analysis performed by Pacific Northwest Laboratory involving the organic carbon laboratory measurement data for Hanford single-shell tanks (SSTs) obtained from a review of the laboratory analytical data. This activity has as its objective to provide a best-estimate, including confidence levels, of total organic carbon (TOC) and moisture in each of the 149 SSTs at Hanford. The TOC and moisture information presented in this report is useful as part of the criteria to identify SSTs for additional measurements, or monitoring for the Organic Safety Program. In April 1994, an initial study of the organic carbon in Hanford single-shell tanks was completed at PNL. That study reflected the estimates of TOC based on tank characterizations datasets that were available at the time. Also in that study, estimation of dry basis TOC was based on generalized assumptions pertaining to the moisture of the tank wastes. The new information pertaining to tank moisture and TOC data that has become available from the current study influences the best estimates of TOC in each of the SSTs. This investigation of tank TOC and moisture has resulted in improved estimates based on waste phase: saltcake, sludge, or liquid. This report details the assumptions and methodologies used to develop the estimates of TOC and moisture in each of the 149 SSTs at Hanford

  17. Waste Isolation Pilot Plant RH TRU waste preoperational checkout: Final report

    International Nuclear Information System (INIS)

    1988-06-01

    This report documents the results of the Waste Isolation Pilot Plant (WIPP) Remote-Handled Transuranic (RH TRU) Waste Preoperational Checkout. The primary objective of this checkout was to demonstrate the process of handling RH TRU waste packages, from receipt through emplacement underground, using equipment, personnel, procedures, and methods to be used with actual waste packages. A further objective was to measure operational time lines to provide bases for confirming the WIPP design through put capability and for projecting operator radiation doses. Successful completion of this checkout is a prerequisite to the receipt of actual RH TRU waste. This checkout was witnessed in part by members of the Environmental Evaluation Group (EEG) of the state of New Mexico. Further, this report satisfies a key milestone contained in the Agreement for Consultation and Cooperation with the state of New Mexico. 4 refs., 26 figs., 4 tabs

  18. Nova Scotia's solid waste-resource management strategy

    Energy Technology Data Exchange (ETDEWEB)

    Friesen, B. [Nova Scotia Dept. of the Environment, Halifax, NS (Canada)

    2000-07-01

    The efforts that the province of Nova Scotia has made to establish a sustainable economy and environment were discussed. In 1989, the province of 930,000 people generated about 630,000 tonnes of municipal waste annually. At the time, there were no recycling or recovery facilities. The province's 1995 Solid Waste-Resource Management Strategy has changed this around with the following key measures: a commitment to 50 per cent diversion of solid waste by 2000, regionalisation to optimize costs, an expanded deposit-refund system for all beverage containers, increased disposal standards for landfilling/incineration, industry example for waste products, support for economic opportunities in resource recovery, and bans on the disposal of recyclable wastes. These measures have produced dramatic results in trying to balance resource supply with demand. The materials sectors which are examples of the strategy's success include fibre, used tires, plastics and composting.

  19. Integrated solid waste management of Seattle, Washington

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The subject document reports the results of an in-depth investigation of the fiscal year 1992 cost of the City of Seattle, Washington, integrated municipal solid waste management (IMSWM) system, the energy consumed to operate the system, and the environmental performance requirements for each of the system`s waste-processing and disposal facilities. Actual data from records kept by participants is reported in this document. Every effort was made to minimize the use of assumptions, and no attempt is made to interpret the data reported. Analytical approaches are documented so that interested analysts may perform manipulation or further analysis of the data. As such, the report is a reference document for MSW management professionals who are interested in the actual costs and energy consumption for a one-year period, of an operating IMSWM systems.

  20. Integrated solid waste management of Sevierville, Tennessee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The subject document reports the results of an in-depth investigation of the fiscal year 1992 cost of the City of Sevierville, Tennessee integrated municipal solid waste management (IMSWM) system, the energy consumed to operate the system, and the environmental performance requirements for each of the system`s waste-processing and disposal facilities. Actual data from records kept by participants is reported in this document. Every effort was made to minimize the use of assumptions, and no attempt is made to interpret the data reported. Analytical approaches are documented so that interested analysts may perform manipulation or further analysis of the data. As such, the report is a reference document for MSW management professionals who are interested in the actual costs and energy consumption for a one-year period, of an operating IMSWM systems.

  1. RCRA Permit for a Hazardous Waste Management Facility Permit Number NEV HW0101 Annual Summary/Waste Minimization Report Calendar Year 2012, Nevada National Security Site, Nevada

    Energy Technology Data Exchange (ETDEWEB)

    Arnold, P. M.

    2013-02-21

    This report summarizes the U.S. Environmental Protection Agency (EPA) identification number of each generator from which the Permittee received a waste stream, a description and quantity of each waste stream in tons and cubic feet received at the facility, the method of treatment, storage, and/or disposal for each waste stream, a description of the waste minimization efforts undertaken, a description of the changes in volume and toxicity of waste actually received, any unusual occurrences, and the results of tank integrity assessments. This Annual Summary/Waste Minimization Report is prepared in accordance with Section 2.13.3 of Permit Number NEV HW0101, issued 10/17/10.

  2. Glass forms for immobilization of Hanford wastes

    International Nuclear Information System (INIS)

    Schulz, W.W.; Dressen, A.L.; Hobbick, C.W.; Babad, H.

    1975-03-01

    Approximately 140 million liters of solid salt cake (mainly NaNO 3 ), produced by evaporation of aged alkaline high-level liquid wastes, will be stored in underground tanks when the present Hanford Waste Management Program is completed in the early 1980's. At this time also, large volumes of various other solid radioactive wastes (sludges, excavated Pu-contaminated soil, and doubly encapsulated 137 CsCl and 90 SrF 2 ) will be stored on the Hanford Reservation. All these solid wastes can be converted to immobile silicate and aluminosilicate glasses of low water leachability by melting them at 1100 0 to 1400 0 C with appropriate amounts of basalt (or sand) and other glass-formers such as B 2 O 3 or CaO. Reviewed in this paper are formulations and other melt conditions used successfully in batch tests to make glasses from actual and synthetic wastes; leachability and other properties of these glasses show them to be satisfactory vehicles for immobilization of the Hanford wastes. (U.S.)

  3. Program planning for future improvement in managing ORNL's radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    1982-01-01

    This report is intended to serve as a reference document and guide in developing the long-term improvements section of ORNL's radioactive waste management plan. The report reviews ORNL's operations and future program needs in terms of currently applicable DOE regulations and also in terms of regulations and accepted practices of the commerical sector of the nuclear power industry so that the impact of potential future adoption of these regulations and standards on ORNL's operations can be fully evaluated. The principal conclusion reached after reviewing ORNL's waste management operations is that these operations are currently being conducted in a manner that does not endanger the health or safety of workers or the general public and that does not have an adverse effect on the environment. Although nineteen specific problem areas have been identified all of these problems can be attributed to one of the following: a) the legacy of past practices; b) gradual deterioration of systems which have reached (or are near to reaching) the end of their reasonable design lives; and c) potential changes in regulations applicable to ORNL. All of the programs designed to improve or correct these problem areas could be accomplished within a four year period. However, given current limitations on manpower and capital, these programs would more likely be spread out over a five to ten year period of time if they were all to be undertaken. The cost of undertaking all of these projects concurrently is estimated to be between 60 and 100 million dollars. Due to the many unknowns and uncertainties associated with the problem areas, actual total costs for specific projects could vary from those presented in this report by as much as 300 percent. (DMC)

  4. 222-S radioactive liquid waste line replacement and 219-S secondary containment upgrade, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1995-01-01

    The U.S. Department of Energy (DOE) is proposing to: (1) replace the 222-S Laboratory (222-S) radioactive liquid waste drain lines to the 219-S Waste Handling Facility (219-S); (2) upgrade 219-S by replacing or upgrading the waste storage tanks and providing secondary containment and seismic restraints to the concrete cells which house the tanks; and (3) replace the transfer lines from 219-S to the 241-SY Tank Farm. This environmental assessment (EA) has been prepared in compliance with the National Environmental Policy Act (NEPA) of 1969, as amended, the Council on Environmental Quality Regulations for Implementing the Procedural Provisions of NEPA (40 Code of Federal Regulations [CFR] 1500-1508), and the DOE Implementing Procedures for NEPA (10 CFR 1021). 222-S is used to perform analytical services on radioactive samples in support of the Tank Waste Remediation System and Hanford Site environmental restoration programs. Activities conducted at 222-S include decontamination of analytical processing and support equipment and disposal of nonarchived radioactive samples. These activities generate low-level liquid mixed waste. The liquid mixed waste is drained through pipelines in the 222-S service tunnels and underground concrete encasements, to two of three tanks in 219-S, where it is accumulated. 219-S is a treatment, storage, and/or disposal (TSD) unit, and is therefore required to meet Washington Administrative Code (WAC) 173-303, Dangerous Waste Regulations, and the associated requirements for secondary containment and leak detection. The service tunnels are periodically inspected by workers and decontaminated as necessary to maintain as low as reasonably achievable (ALARA) radiation levels. Although no contamination is reaching the environment from the service tunnels, the risk of worker exposure is present and could increase. 222-S is expected to remain in use for at least the next 30 years to serve the Hanford Site environmental cleanup mission

  5. Waste management issues, a set of technologies

    International Nuclear Information System (INIS)

    Gautrot, J.J.

    2000-01-01

    As any other industry, nuclear fuel cycle back-end raises the major issue of waste management. In France, spent fuel is considered as valuable materials and only the ultimate waste are considered as actual waste. Accordingly, waste issue is as follows: a sorting out has to be done, in order to separate valuable materials from actual wastes, put any outlet flow under a stable form and condition them appropriately to their respective recycling or disposal routes. This implies the implementation of a comprehensive set of technologies. Actually, it is an industrial reality, as the COGEMA Group has for a long time set up a reprocessing and conditioning strategy in its plants. Waste management issues are common to many activities. European as well as French regulators already introduced the twofold necessity to reduce waste volumes, and to dispose of only ''ultimate waste'' as concerns industrial and household waste mainly. In this objective, French nuclear reprocessing and recycling industry may be seen as a breeding ground of well-proven technologies and management options. Actually, processes used can also give an answer to such different issues as excess plutonium immobilization, sites cleaning up (including for instance treatment of the liquid HLW legacy), dismantling wastes management. There are a number of operations to be dealt with worldwide that will find a solution in any of the technologies implemented and optimized in COGEMA facilities. Based on the COGEMA Group know-how, the present paper will describe those technologies and explain how they can solve the other stringent waste management issues worldwide. (author)

  6. Rethinking the Hanford Tank Waste Program

    International Nuclear Information System (INIS)

    Parker, F. L.; Clark, D. E.; Morcos, N.

    2002-01-01

    The program to treat and dispose of the highly radioactive wastes stored in underground tanks at the U.S. Department of Energy's Hanford site has been studied. A strategy/management approach to achieve an acceptable (technically sound) end state for these wastes has been developed in this study. This approach is based on assessment of the actual risks and costs to the public, workers, and the environment associated with the wastes and storage tanks. Close attention should be given to the technical merits of available waste treatment and stabilization methodologies, and application of realistic risk reduction goals and methodologies to establish appropriate tank farm cleanup milestones. Increased research and development to reduce the mass of non-radioactive materials in the tanks requiring sophisticated treatment is highly desirable. The actual cleanup activities and milestones, while maintaining acceptable safety standards, could be more focused on a risk-to-benefit cost effectiveness, as agreed to by the involved stakeholders and in accordance with existing regulatory requirements. If existing safety standards can be maintained at significant cost savings under alternative plans but with a change in the Tri-Party Agreement (a regulatory requirement), those plans should be carried out. The proposed strategy would also take advantage of the lessons learned from the activities and efforts in the first phase of the two-phased cleanup of the Hanford waste tank farms

  7. Cesium Removal From Tanks 241-AN-103 and 241-SX-105 and 241-AZ-101 and 241-AZ-102 Composite For Testing In Bench Scale Steam Reformer

    International Nuclear Information System (INIS)

    Duncan, J.B.; Huber, H.J.

    2011-01-01

    This report documents the preparation of three actual Hanford tank waste samples for shipment to the Savannah River National Laboratory (SRNL). Two of the samples were dissolved saltcakes from tank 241-AN-103 (hereafter AN-103) and tank 241-SX-105 (hereafter SX-105); one sample was a supernate composite from tanks 241-AZ-101 and 241-AZ-102 (hereafter AZ-101/102). The preparation of the samples was executed following the test plans LAB-PLAN-10-00006, Test Plan for the Preparation of Samples from Hanford Tanks 241-SX-105, 241-AN-103, 241-AN-107, and LAB-PLN-l0-00014, Test Plan for the Preparation of a Composite Sample from Hanford Tanks 241-AZ-101 and 241-AZ-102 for Steam Reformer Testing at the Savannah River National Laboratory. All procedural steps were recorded in laboratory notebook HNF-N-274 3. Sample breakdown diagrams for AN-103 and SX-105 are presented in Appendix A. The tank samples were prepared in support of a series of treatability studies of the Fluidized Bed Steam Reforming (FBSR) process using a Bench-Scale Reformer (BSR) at SRNL. Tests with simulants have shown that the FBSR mineralized waste form is comparable to low-activity waste glass with respect to environmental durability (WSRC-STI-2008-00268, Mineralization of Radioactive Wastes by Fluidized Bed Steam Reforming (FBSR): Comparisons to Vitreous Waste Forms and Pertinent Durability Testing). However, a rigorous assessment requires long-term performance data from FBSR product formed from actual Hanford tank waste. Washington River Protection Solutions, LLC (WRPS) has initiated a Waste Form Qualification Program (WP-5.2.1-2010-001, Fluidized Bed Steam Reformer Low-level Waste Form Qualification) to gather the data required to demonstrate that an adequate FBSR mineralized waste form can be produced. The documentation of the selection process of the three tank samples has been separately reported in RPP-48824, Sample Selection Process for Bench-Scale Steam Reforming Treatability Studies Using

  8. CESIUM REMOVAL FROM TANKS 241-AN-103 & 241-SX-105 & 241-AZ-101 & 241AZ-102 COMPOSITE FOR TESTING IN BENCH SCALE STEAM REFORMER

    Energy Technology Data Exchange (ETDEWEB)

    DUNCAN JB; HUBER HJ

    2011-04-21

    This report documents the preparation of three actual Hanford tank waste samples for shipment to the Savannah River National Laboratory (SRNL). Two of the samples were dissolved saltcakes from tank 241-AN-103 (hereafter AN-103) and tank 241-SX-105 (hereafter SX-105); one sample was a supernate composite from tanks 241-AZ-101 and 241-AZ-102 (hereafter AZ-101/102). The preparation of the samples was executed following the test plans LAB-PLAN-10-00006, Test Plan for the Preparation of Samples from Hanford Tanks 241-SX-105, 241-AN-103, 241-AN-107, and LAB-PLN-l0-00014, Test Plan for the Preparation of a Composite Sample from Hanford Tanks 241-AZ-101 and 241-AZ-102 for Steam Reformer Testing at the Savannah River National Laboratory. All procedural steps were recorded in laboratory notebook HNF-N-274 3. Sample breakdown diagrams for AN-103 and SX-105 are presented in Appendix A. The tank samples were prepared in support of a series of treatability studies of the Fluidized Bed Steam Reforming (FBSR) process using a Bench-Scale Reformer (BSR) at SRNL. Tests with simulants have shown that the FBSR mineralized waste form is comparable to low-activity waste glass with respect to environmental durability (WSRC-STI-2008-00268, Mineralization of Radioactive Wastes by Fluidized Bed Steam Reforming (FBSR): Comparisons to Vitreous Waste Forms and Pertinent Durability Testing). However, a rigorous assessment requires long-term performance data from FBSR product formed from actual Hanford tank waste. Washington River Protection Solutions, LLC (WRPS) has initiated a Waste Form Qualification Program (WP-5.2.1-2010-001, Fluidized Bed Steam Reformer Low-level Waste Form Qualification) to gather the data required to demonstrate that an adequate FBSR mineralized waste form can be produced. The documentation of the selection process of the three tank samples has been separately reported in RPP-48824, Sample Selection Process for Bench-Scale Steam Reforming Treatability Studies Using

  9. Radiotoxic hazard measure for buried solid radioactive waste

    International Nuclear Information System (INIS)

    Hamstra, J.

    1975-01-01

    The radiotoxic hazards resulting from the disposal of highlevel reprocessing wastes into a deep geological formation are reviewed. The term radiotoxic hazard measure (RHM), used to measure the hazard from buried radioactive wastes, is based on the maximum radionuclide concentration permissible in water. Calculations are made of the RHM levels for the high-level reprocessing wastes of both light-water-reactor and fast breeder reactor fuels. In comparing these RHM levels with that for the natural activity of an equivalent amount of uranium ore and its mill tailings, it is concluded that an actual additional radiotoxic hazard for buried high-level reprocessing waste only exists for the first 300 to 500 years after burial. (U.S.)

  10. Process chemistry for the pretreatment of Hanford tank wastes

    International Nuclear Information System (INIS)

    Lumetta, G.J.; Swanson, J.L.; Barker, S.A.

    1992-08-01

    Current guidelines for disposing radioactive wastes stored in underground tanks at the US Department of Energy's Hanford Site call for the vitrification of high-level waste in borosilicate glass and disposal of the glass canisters in a deep geologic repository. Low-level waste is to be cast in grout and disposed of on site in shallow burial vaults. Because of the high cost of vitrification and geologic disposal, methods are currently being developed to minimize the volume of high-level waste requiring disposal. Two approaches are being considered for pretreating radioactive tank sludges: (1) leaching of selected components from the sludge and (2) acid dissolution of the sludge followed by separation of key radionuclides. The leaching approach offers the advantage of simplicity, but the acid dissolution/radionuclide extraction approach has the potential to produce the least number of glass canisters. Four critical components (Cr, P, S, and Al) were leached from an actual Hanford tank waste-Plutonium Finishing Plant sludge. The Al, P, and S were removed from the sludge by digestion of the sludge with 0.1 M NaOH at 100 degrees C. The Cr was leached by treating the sludge with alkaline KMnO 4 at 100 degrees C. Removing these four components from the sludge will dramatically lower the number of glass canisters required to dispose of this waste. The transuranic extraction (TRUEX) solvent extraction process has been demonstrated at a bench scale using an actual Hanford tank waste. The process, which involves extraction of the transuranic elements with octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO), separated 99.9% of the transuranic elements from the bulk components of the waste. Several problems associated with the TRUEX processing of this waste have been addressed and solved

  11. Potential for waste reduction

    International Nuclear Information System (INIS)

    Warren, J.L.

    1990-01-01

    The author focuses on wastes considered hazardous under the Resource Conservation and Recovery Act. This chapter discusses wastes that are of interest as well as the factors affecting the quantity of waste considered available for waste reduction. Estimates are provided of the quantities of wastes generated. Estimates of the potential for waste reduction are meaningful only to the extent that one can understand the amount of waste actually being generated. Estimates of waste reduction potential are summarized from a variety of government and nongovernment sources

  12. Nuclear waste management

    International Nuclear Information System (INIS)

    Rodger, W.A.

    1985-01-01

    Most of our activities have always produced waste products of one sort or another. Huxley gives a humorous account of wastes throughout antiquity. So it should come as no surprise that some radioactive materials end up as waste products requiring management and disposal. Public perception of nuclear waste hazards places them much higher on the ''worry scale'' than is justified by the actual hazard involved. While the public perception of these hazards appears to revolve mostly around high-level wastes, there are several other categories of wastes that must also be controlled and managed. The major sources of radioactive wastes are discussed

  13. Status of U.S. nuclear waste disposal programs

    International Nuclear Information System (INIS)

    Eastering, J.B.

    1987-01-01

    Since the days of the Manhattan Project in the early 1940's, the United States has recognized the need for a system of effective nuclear waste management. The evolution of that system has been directed by a series of legislative actions to determine responsibilities and develop procedures. Under these laws, the U.S. Department of Energy (DOE), as the lead agency, is required to plan and implement the programs that will ensure safe and efficient management of nuclear wastes from both civilian and defense activities. This overview paper describes the organization of these programs

  14. The actual practice of air cleaning in Belgian nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Goossens, W.R. [PEGO, Mol (Belgium)

    1995-02-01

    With 60% of its power generation from nuclear stations Belgium has 7 nuclear power stations in operation with a total capacity of 5.4 MWe. Enriched uranium is imported and converted to fuel assemblies. The actinides of reprocessed fuel are recycled as MOX fuel. A main waste conditioning operation has been performed in the PAMELA vitrifier. The actual practice of nuclear air cleaning in the Belgian PWR station DOEL-4 and in the PAMELA -vitrification plant for high level liquid waste is reviewed.

  15. Radioactive waste management policy

    International Nuclear Information System (INIS)

    Morrison, R.W.

    1983-06-01

    The speaker discusses the development of government policy regarding radioactive waste disposal in Canada, indicates overall policy objectives, and surveys the actual situation with respect to radioactive wastes in Canada. He also looks at the public perceptions of the waste management situation and how they relate to the views of governmental decision makers

  16. Waste treatability guidance program. User`s guide. Revision 0

    Energy Technology Data Exchange (ETDEWEB)

    Toth, C.

    1995-12-21

    DOE sites across the country generate and manage radioactive, hazardous, mixed, and sanitary wastes. It is necessary for each site to find the technologies and associated capacities required to manage its waste. One role of DOE HQ Office of Environmental Restoration and Waste Management is to facilitate the integration of the site- specific plans into coherent national plans. DOE has developed a standard methodology for defining and categorizing waste streams into treatability groups based on characteristic parameters that influence waste management technology needs. This Waste Treatability Guidance Program automates the Guidance Document for the categorization of waste information into treatability groups; this application provides a consistent implementation of the methodology across the National TRU Program. This User`s Guide provides instructions on how to use the program, including installations instructions and program operation. This document satisfies the requirements of the Software Quality Assurance Plan.

  17. Waste Heat Recovery. Technology and Opportunities in U.S. Industry

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Ilona [BCS, Inc., Laurel, MD (United States); Choate, William T. [BCS, Inc., Laurel, MD (United States); Davidson, Amber [BCS, Inc., Laurel, MD (United States)

    2008-03-01

    This study was initiated in order to evaluate RD&D needs for improving waste heat recovery technologies. A bottomup approach is used to evaluate waste heat quantity, quality, recovery practices, and technology barriers in some of the largest energyconsuming units in U.S. manufacturing. The results from this investigation serve as a basis for understanding the state of waste heat recovery and providing recommendations for RD&D to advance waste heat recovery technologies.

  18. MANAGING HANFORD'S LEGACY NO-PATH-FORWARD WASTES TO DISPOSITION

    Energy Technology Data Exchange (ETDEWEB)

    WEST LD

    2011-01-13

    The U.S. Department of Energy (DOE) Richland Operations Office (RL) has adopted the 2015 Vision for Cleanup of the Hanford Site. This vision will protect the Columbia River, reduce the Site footprint, and reduce Site mortgage costs. The CH2M HILL Plateau Remediation Company's (CHPRC) Waste and Fuels Management Project (W&FMP) and their partners support this mission by providing centralized waste management services for the Hanford Site waste generating organizations. At the time of the CHPRC contract award (August 2008) slightly more than 9,000 m{sup 3} of waste was defined as 'no-path-forward waste.' The majority of these wastes are suspect transuranic mixed (TRUM) wastes which are currently stored in the low-level Burial Grounds (LLBG), or stored above ground in the Central Waste Complex (CWC). A portion of the waste will be generated during ongoing and future site cleanup activities. The DOE-RL and CHPRC have collaborated to identify and deliver safe, cost-effective disposition paths for 90% ({approx}8,000 m{sup 3}) of these problematic wastes. These paths include accelerated disposition through expanded use of offsite treatment capabilities. Disposal paths were selected that minimize the need to develop new technologies, minimize the need for new, on-site capabilities, and accelerate shipments of transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico.

  19. Characterization of underground storage tank sludge using fourier transform infrared photoacoustic spectroscopy

    International Nuclear Information System (INIS)

    Luo, S.; Bajic, S.J.; Jones, R.W.

    1994-01-01

    Analysis of underground storage tank (UST) contents is critical for the determination of proper disposal protocols and storage procedures of nuclear waste materials. Tank volume reduction processes during the 1940's and 50's have produced a waste form that compositionally varies widely and has a consistency that ranges from paste like sludge to saltcake. The heterogeneity and chemical reactivity of the waste form makes analysis difficult by most conventional methods which require extensive sample preparation. In this paper, a method is presented to characterize nuclear waste from UST's at the Westinghouse Hanford Site in Washington State, using Fourier transform infrared-photoacoustic spectroscopy (FTIR-PAS). FTIR-PAS measurements on milligram amounts of surrogate sludge samples have been used to accurately identify phosphate, sulfate, nitrite, nitrate and ferrocyanide components. A simple sample preparation method was followed to provide a reproducible homogeneous sample for quantitative analysis. The sample preparation method involved freeze drying the sludge sample prior to analysis to prevent the migration of soluble species. Conventional drying (e.g., air or, oven) leads to the formation of crystals near the surface where evaporation occurs. Sample preparation as well as the analytical utility of this method will be discussed

  20. THE ACTUALITY OF NIETZSCHE’S CRITICISM OF EDUCATIONAL SYSTEM REGARDING MODERN KNOWLEDGE SOCIETY

    Directory of Open Access Journals (Sweden)

    Andreas SCHREIBER

    2016-06-01

    Full Text Available „Thus, my friends do not confound this education, this delicately footed, spoiled, ethereal goddess, with that usable maid which occasionally also is called ‘education’, but which is merely the intellectual maid and counselor for distress of life, for earnings, for neediness.”1 In his five lectures on the “future of educational system” Nietzsche thoroughly criticized the German educational system of his time and the decay of education in general. Although he was very sharp in accusation and pessimistic about his contemporary situation, he nevertheless cherished some hopes regarding some profound and revolutionary changes in the near future that might take place and rescue the classical concept of education. Looking back from today to Nietzsche’s assertions – like the quote above – we can recognize that his critical sketch of former deficiencies of educational system has lost none of its actuality and truth. We rather notice that some of his points just have been actualized only in modern knowledge society than in his time – or at least they have aggravated. Nowadays, knowledge is considered as an economical factor of production, and education is the fundamental precondition for employability. On the way to neoliberal, capitalistic system we have started with exploitation of the knower and have come to reach a status of voluntary self-exploitation of the educated one. Knowledge and education, and the notion of lifelong learning as well as self-actualization and personality growth have become efficient and perfect instruments of neoliberal systems, as they use the freedom and free volition of the individual to enforce their self-exploitation in order to become high-potential consumers. This contribution points out the still existing actuality of Nietzsche’s critical arguments regarding the decay of the concept of education. Moreover, it is to be shown that in the age of the so called knowledge society in neoliberal systems we

  1. Trends in U.S.A. on environmental assessment of the common industrial wastes

    International Nuclear Information System (INIS)

    Yamauchi, Kiyoshi

    1994-01-01

    In Japan, the waste materials are divided to two parts of radioactive wastes and common wastes according to the law on treatment and cleaning of the wastes, and the latter is divided further to two parts of life type wastes and business type wastes. Among them, a part of wastes called the industrial wastes in the business wastes causes some troubles. In Japan, maintenance of approval level of the wastes treatment was conducted by selection of treating method due to feature of the wastes, selection of treating matters due to the treating site and method, and environmental assessment of each treatment project program. However, in U.S.A., regulation on environmental impact due to various human actions has been executed since 1960, and environmental assessment is controlled by a law. And, regulations on the wastes, the treating facilities and the formed pollution were also determined by laws. Furthermore, technical field of the environmental assessment is already executed in U.S.A. but some parts are not in Japan. On the evaluation method, there are some differences between in U.S.A. and in Japan, but are some common points in technical informations, estimating methods and so on, where future technical cooperations and movements between them in environmental assessment on both common industrial wastes and high level radioactive wastes. (G.K.)

  2. Cairo’s Contested Garbage: Sustainable Solid Waste Management and the Zabaleen’s Right to the City

    Directory of Open Access Journals (Sweden)

    Wael Fahmi

    2010-06-01

    Full Text Available Over the decades, the Zabaleen, the traditional waste (garbage collectors of Cairo, have created what is arguably one of the world’s most efficient and sustainable resource-recovery and waste-recycling systems. Yet the continuation of this intricate relationship between community, environment and livelihood is jeopardized by the official privatization of municipal solid waste (MSW services through contracts with technology-intensive multinational corporations which threatens the sustainability of the garbage collectors’ communities by removing access to their chief economic asset, waste or garbage. The situation is exacerbated by an official policy of moving the Zabaleen and their MSW sorting, recovery, trading and recycling activities further out of the city, on the grounds that this will turn their neighbourhoods into cleaner and healthier living environments. The consumption of Cairo’s sites of MSW collection and sorting open new socio-political spaces for conflict between multi-national companies and the Zabaleen’s traditional system. This is further indicated in the way Cairo’s waste materials have been subjected to new claims and conflict, as they are seen as a ‘commodity’ by global capital entrepreneurs and multi-national corporations, and as a source of ‘livelihood’ by the disadvantaged and marginalised Zabaleen population.

  3. The very low level radioactive military waste in France

    International Nuclear Information System (INIS)

    2005-01-01

    The recommendations of the Observatory of French nuclear weapons call attention to the fact the I.N.B.S. depend on the services of the High Commissioner of Atomic Energy;call attention to the fact that the reactors of the nuclear propulsion ( only military ones) are out of regulation: independent studies have to be realised on three harbors (Brest, Toulon, Cherbourg) for an evaluation about the environmental and sanitary consequences of the presence of nuclear facilities near urban areas that is not the case for any civil nuclear installation; call attention on the dismantling operations actually in some military nuclear installations, who is in charge of the control of the different steps of these dismantling operations, call attention on the setting up of radioactive waste storage ( very low activity at Saint-Priest, Rhone, very high activity with americium at Valduc, long lived elements at Cadarache, tritium waste at Valduc, alpha waste at Valduc waste not under the control of an independent organism; call attention on the final destination of depleted uranium; call attention on the bad management of tritium by the installation of nuclear weapons fabrication, about the actual stocks in plutonium and uranium. (N.C.)

  4. Full Focus Needed on Finishing Hanford's Waste Treatment Plant - 12196

    Energy Technology Data Exchange (ETDEWEB)

    Dahl, Suzanne; Biyani, Rabindra; Holmes, Erika [Washington State Department of Ecology, Richland, WA 99354 (United States)

    2012-07-01

    The United States Department of Energy's (US DOE's) Hanford Nuclear Site has 177 underground waste storage tanks located 19 to 24 km (12 to 15 miles) from the Columbia River in south-central Washington State. Hanford's tanks now hold about 212,000 cu m (56 million gallons) of highly radioactive and chemically hazardous waste. Sixty-seven tanks have leaked an estimated 3,785 cu m (1 million gallons) of this waste into the surrounding soil. Further releases to soil, groundwater, and the Columbia River are the inevitable result of the tanks continuing to age. The risk from this waste is recognized as a threat to the Northwest by both State and Federal governments. US DOE and Bechtel National, Inc., are building the Waste Treatment and Immobilization Plant (WTP) to treat and vitrify (immobilize in glass) the waste from Hanford's tanks. As is usual for any groundbreaking project, problems have arisen that must be resolved as they occur if treatment is to take place as specified in the court-enforceable Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) and the Consent Decree, entered into by US DOE, the U.S. Environmental Protection Agency, and the Washington State Department of Ecology (Ecology). At times, US DOE's approach to solving these critical issues seems to have caused undue wastes of time, energy, and, ultimately, public funds. Upon reviewing the history of Hanford's tank waste treatment project, Ecology hopes that constructive criticism of past failures and praise of successes will inspire US DOE to consider changing practices, be more transparent with regulatory agencies and the public, and take a 'lean production' approach to successfully completing this project. All three Tri-Party Agreement agencies share the goal of completing WTP on time, ensuring it is operational and in compliance with safety standards. To do this, Ecology believes US DOE should: - Maintain focus on the primary goal of

  5. Pu speciation in actual and simulated aged wastes

    Energy Technology Data Exchange (ETDEWEB)

    Lezama-pacheco, Juan S [Los Alamos National Laboratory; Conradson, Steven D [Los Alamos National Laboratory

    2008-01-01

    X-ray Absorption Fine Structure Spectroscopy (XAFS) at the Pu L{sub II/III} edge was used to determine the speciation of this element in (1) Hanford Z-9 Pu crib samples, (2) deteriorated waste resins from a chloride process ion-exchange purification line, and (3) the sediments from two Waste Isolation Pilot Plant Liter Scale simulant brine systems. The Pu speciation in all of these samples except one is within the range previously displayed by PuO{sub 2+x-2y}(OH){sub y}{center_dot}zH{sub 2}O compounds, which is expected based on the putative thermodynamic stability of this system for Pu equilibrated with excess H{sub 2}O and O{sub 2} under environmental conditions. The primary exception was a near neutral brine experiment that displayed evidence for partial substitution of the normal O-based ligands with Cl{sup -} and a concomitant expansion of the Pu-Pu distance relative to the much more highly ordered Pu near neighbor shell in PuO{sub 2}. However, although the Pu speciation was not necessarily unusual, the Pu chemistry identified via the history of these samples did exhibit unexpected patterns, the most significant of which may be that the presence of the Pu(V)-oxo species may decrease rather than increase the overall solubility of these compounds. Several additional aspects of the Pu speciation have also not been previously observed in laboratory-based samples. The molecular environmental chemistry of Pu is therefore likely to be more complicated than would be predicted based solely on the behavior of PuO{sub 2} under laboratory conditions.

  6. Treating landfill gas hydrogen sulphide with mineral wool waste (MWW) and rod mill waste (RMW).

    Science.gov (United States)

    Bergersen, Ove; Haarstad, Ketil

    2014-01-01

    Hydrogen sulphide (H2S) gas is a major odorant at municipal landfills. The gas can be generated from different waste fractions, for example demolition waste containing gypsum based plaster board. The removal of H2S from landfill gas was investigated by filtering it through mineral wool waste products. The flow of gas varied from 0.3 l/min to 3.0 l/min. The gas was typical for landfill gas with a mean H2S concentration of ca. 4500 ppm. The results show that the sulphide gas can effectively be removed by mineral wool waste products. The ratios of the estimated potential for sulphide precipitation were 19:1 for rod mill waste (RMW) and mineral wool waste (MWW). A filter consisting of a mixture of MWW and RMW, with a vertical perforated gas tube through the center of filter material and with a downward gas flow, removed 98% of the sulfide gas over a period of 80 days. A downward gas flow was more efficient in contacting the filter materials. Mineral wool waste products are effective in removing hydrogen sulphide from landfill gas given an adequate contact time and water content in the filter material. Based on the estimated sulphide removal potential of mineral wool and rod mill waste of 14 g/kg and 261 g/kg, and assuming an average sulphide gas concentration of 4500 ppm, the removal capacity in the filter materials has been estimated to last between 11 and 308 days. At the studied location the experimental gas flow was 100 times less than the actual gas flow. We believe that the system described here can be upscaled in order to treat this gas flow. Copyright © 2013 Elsevier Ltd. All rights reserved.

  7. Solidification of radioactive wastes by bituminization technology. II

    International Nuclear Information System (INIS)

    Stuchlik, S.; Brzobohaty, J.

    1985-01-01

    The bituminization line consists of a rotor film evaporator, a condenser, reserve tanks of bitumen emulsion and model concentrates, a roller train, a vapour concentrate tank, a control console, a bitumen emulsion pump, a model concentrates pump, and a tank for model concentrate preparation. Anion-active emulsion Silembit S-60 produced by Paramo Pardubice was chosen as the bitumen emulsion. A block diagram is given of the experimental bituminization line and its processing is described of model nonactive concentrates whose composition corresponds to that of actual wastes from the V-1 nuclear power plant. Successful tests of the line showed that it could be used for the disposal of radioactive wastes. (E.S.)

  8. Low and intermediate level waste in SFR-1. Reference Waste Inventory 2007

    International Nuclear Information System (INIS)

    Almkvist, Lisa; Gordon, Ann

    2007-11-01

    The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR 1 at the time of closure. The report will form the basis for the release calculation in the safety analysis for SFR 1. Three different scenarios are explored in this report; the waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 50 and 60 years and that closure of the SFR 1 repository will take place in 2040 or 2050 respectively. The third scenario is where the repository is full (one part where the activity adds up to 1016 Bq and one part where the repository is considered full regarding volume). In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemotoxic material has been identified in the waste. The inventory is estimated using the Prosit-interface which extracts information from the Triumf database. The inventory is based on so called 'waste types' and the waste types' 'reference waste package'. The reference waste package combined with a prognosis of the number of waste packages to be delivered to SFR 1 gives the final waste inventory for SFR 1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60 Co and 137 Cs in waste packages and on measurements of 239 Pu and 240 Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors

  9. Leach rate studies on glass containing actual radioactive waste

    International Nuclear Information System (INIS)

    Walker, D.D.; Wiley, J.R.; Dukes, M.D.; LeRoy, J.H.

    1980-01-01

    Borosilicate glass containing radioactive wastes from the Savannah River Plant have been leached for 900 days. The International Standards Organization's (ISO) static leach test procedure was used on glass buttons in various leachants. Leach rates based on 90 Sr and 137 Cs analyses were similar: 2 x 10 -8 to 3 x 10 -8 g/(cm 2 )(d) in distilled water, 1 x 10 -8 to 3 x 10 -7 g/(cm 2 )(d) in pH 7 buffer, 3 x 10 -7 to 7 x 10 -7 g/(cm 2 )(d) in pH 9 buffer, and 7 x 10 -6 to 8 x 10 -5 g/(cm 2 )(d) in pH 4 buffer. Rates based on Pu analyses were the same as above in distilled water and pH 9 buffer, but were lower by an order of magnitude in pH 4 and pH 7 buffers. Almost all leach rates remained constant between 200 and 900 days of leaching. Increasing the concentration of the buffering agents had no effect on the leach rates at pH 7 (phosphate) and pH 9 (carbonate), but dramatically increased the rates at pH 4 (acetate). Leach rates did not differ significantly between high aluminum and high iron waste glasses

  10. Chemodynamics of EDTA in a simulated mixed waste: the Hanford Site's complex concentrate waste

    International Nuclear Information System (INIS)

    Toste, A.P.; Ohnuki, Toshihiko

    1999-01-01

    Enormous stockpiles of mixed wastes at the USDOE's Hanford Site, the original US plutonium production facility, await permanent disposal. One mixed waste derived from reprocessing spent fuel was found to contain numerous nuclear related organics including chelating agents like EDTA and complexing agents, which have been used as decontamination agents, etc. Their presence in actual mixed wastes indicates that the organic content of nuclear wastes is dynamic and complicate waste management efforts. The subjects of this report is the chemo-degradation of EDTA degradation in a simulant Hanford's complex concentrate waste. The simulant was prepared by adding EDTA to an inorganic matrix, which was formulated based on past analyses of the actual waste. Aliquots of the EDTA simulant were withdrawn at different time points, derivatized via methylation and analyzed by gas chromatography and Gc/MS to monitor the disappearance of EDTA and the appearance of its' degradation products. This report also compares the results of EDTA's chemo-degradation to the g-radiolysis of EDTA in the simulant, the subject of a recently published article. Finally based on the results of these two studies, an assesment of the potential impact of EDTA degradation on the management of mixed wastes is offered. (J.P.N.)

  11. 2727-S Nonradioactive Dangerous Waste Storage Facility Closure Plan

    International Nuclear Information System (INIS)

    Wilczek, T.A.; Laws, J.R.; Izatt, R.D.

    1992-01-01

    This closure plan describes the activities for final closure of the 2727-S Nonradioactive Dangerous Waste Storage (NRDWS) Facility at the Hanford Site. The 2727-S NRDWS Facility provided container storage for nonradioactive dangerous and extremely hazardous wastes generated in the research and development laboratories, process operations, and maintenance and transportation functions throughout the Hanford Site. Storage operations began at the 2727-S NRDWS Facility March 14, 1983, and continued until December 30, 1986, when the last shipment of materials from the facility took place. These storage operations have been moved to the new 616 NRDWS Facility, which is an interim status unit located between the 200 East and 200 West Areas of the Hanford Site

  12. Estimate of Hanford Waste Rheology and Settling Behavior

    International Nuclear Information System (INIS)

    Poloski, Adam P.; Wells, Beric E.; Tingey, Joel M.; Mahoney, Lenna A.; Hall, Mark N.; Thomson, Scott L.; Smith, Gary Lynn; Johnson, Michael E.; Meacham, Joseph E.; Knight, Mark A.; Thien, Michael G.; Davis, Jim J.; Onishi, Yasuo

    2007-01-01

    The U.S. Department of Energy (DOE) Office of River Protection's Waste Treatment and Immobilization Plant (WTP) will process and treat radioactive waste that is stored in tanks at the Hanford Site. Piping, pumps, and mixing vessels have been selected to transport, store, and mix the high-level waste slurries in the WTP. This report addresses the analyses performed by the Rheology Working Group (RWG) and Risk Assessment Working Group composed of Pacific Northwest National Laboratory PNNL, Bechtel National Inc. (BNI), CH2M HILL, DOE Office of River Protection (ORP) and Yasuo Onishi Consulting, LLC staff on data obtained from documented Hanford waste analyses to determine a best-estimate of the rheology of the Hanford tank wastes and their settling behavior. The actual testing activities were performed and reported separately in referenced documentation. Because of this, many of the required topics below do not apply and are so noted

  13. Boundaries matter: Greenhouse gas emission reductions from alternative waste treatment strategies for California’s municipal solid waste

    DEFF Research Database (Denmark)

    Vergara, Sintana E.; Damgaard, Anders; Horvathc, Arpad

    2011-01-01

    How waste is managed – whether as a nuisance to be disposed of, or as a resource to be reused – directly affects local and global environmental quality. This analysis explores the GHG benefits of five treatment options for residual municipal solid waste (MSW) in California: Business As Usual...... landfills. Using two different waste LCA models, EASEWASTE (a Danish model) and WARM (a U.S. model), we find that improved biogenic waste management through anaerobic digestion and waste reduction can lead to life-cycle GHG savings when compared to Business As Usual. The magnitude of the benefits depends...... strongly on a number of model assumptions: the type of electricity displaced by waste-derived energy, how biogenic carbon is counted as a contributor to atmospheric carbon stocks, and the landfill gas collection rate. Assuming that natural gas is displaced by waste-derived energy, that 64% of landfill gas...

  14. Ten questions on nuclear wastes

    International Nuclear Information System (INIS)

    Guillaumont, R.; Bacher, P.

    2004-01-01

    The authors give explanations and answers to ten issues related to nuclear wastes: when a radioactive material becomes a waste, how radioactive wastes are classified and particularly nuclear wastes in France, what are the risks associated with radioactive wastes, whether the present management of radioactive wastes is well controlled in France, which wastes are raising actual problems and what are the solutions, whether amounts and radio-toxicity of wastes can be reduced, whether all long life radionuclides or part of them can be transmuted, whether geologic storage of final wastes is inescapable, whether radioactive material can be warehoused over long durations, and how the information on radioactive waste management is organised

  15. Modeling of Sulfate Double-Salt in Nuclear Wastes

    International Nuclear Information System (INIS)

    Toghiani, B.; Lindner, J.S.; Weber, C.F.; Hunt, R.D.

    2000-01-01

    The Environmental Simulation Program (ESP) continues to adequately predict the solubility of most key chemical systems in the Hanford tank waste. For example, the ESP predictions were in fair agreement with the solubility experiments for the fluoride-phosphate system, although ESP probably underestimates the aqueous amounts. Due to the importance of this system in the formation of pipeline plugs, additional experiments have been made at elevated temperatures, and improvements to the ESP database will be made. ESP encountered problems with sulfate systems because the Public database for ESP does not include anhydrous sodium sulfate in mixed solutions below 32.4 C. This limitation leads to convergence problems and to spurious predictions of solubility near the transition point with sodium sulfate decahydrate when other salts such as sodium nitrate are present. However, ESP was able to make reasonable solubility predictions with a corrected database, demonstrating the need to validate and document the various databases that can be used by ESP. Even though ESP does not include the sulfate-nitrate double salt, this omission does not appear to be a major problem. The solubility predictions with and without the sulfate-nitrate double salt are comparable. In sharp contrast, the sulfate-fluoride double salt is included, but ESP still underestimates solubility in some cases. This problem can misrepresent the ionic strength of the solution, which is an important factor in the formation of pipeline plugs. Solubility tests on the sulfate-fluoride system are planned to provide additional data at higher temperatures and in caustic solutions. These results will be used to improve the range and accuracy of ESP predictions. ESP will continue to provide important predictions for waste processing operations while being evaluated and improved. For example, ESP will be used to determine the amount of water for the saltcake dissolution efforts at Hanford. When ESP underestimates the

  16. Actualism and Fictional Characters

    Directory of Open Access Journals (Sweden)

    André Leclerc

    2016-04-01

    Full Text Available http://dx.doi.org/10.5007/1808-1711.2016v20n1p61 In what follows, I present only part of a program that consists in developing a version of actualism as an adequate framework for the metaphysics of intentionality. I will try to accommodate in that framework suggestions found in Kripke’s works and some positions developed by Amie Thomasson. What should we change if we accept “fictional entities” in the domain of the actual world? Actualism is the thesis that everything that exists belongs to the domain of the actual world and that there are no possibilia. I shall defend that there are abstract artefacts, like fictional characters, and institutions. My argument could be seen as a version of Moore’s paradox: it is paradoxical to say: “I made (created it, but I do not believe it exists”. Moreover, there are true sentences about them. I will examine what it means to include abstract artefacts in the domain of the actual world. I favour a use of “exist” that includes beings with no concrete occupation of tri-dimensional space; to exist, it is enough to have been introduced at some moment in history. Abstract artefacts, like fictional characters, exist in that sense. I argue that it is important to distinguish two perspectives (internal and external in order to clarify the kind of knowledge we have of fictional characters. However, their existence presupposes a relation of dependence to a material basis and the mental activities of many people.

  17. Stabilization and disposal of Argonne-West low-level mixed wastes in ceramicrete waste forms

    International Nuclear Information System (INIS)

    Barber, D. B.; Singh, D.; Strain, R. V.; Tlustochowicz, M.; Wagh, A. S.

    1998-01-01

    The technology of room-temperature-setting phosphate ceramics or Ceramicretetrademark technology, developed at Argonne National Laboratory (ANL)-East is being used to treat and dispose of low-level mixed wastes through the Department of Energy complex. During the past year, Ceramicretetrademark technology was implemented for field application at ANL-West. Debris wastes were treated and stabilized: (a) Hg-contaminated low-level radioactive crushed light bulbs and (b) low-level radioactive Pb-lined gloves (part of the MWIR number s ign AW-W002 waste stream). In addition to hazardous metals, these wastes are contaminated with low-level fission products. Initially, bench-scale waste forms with simulated and actual waste streams were fabricated by acid-base reactions between mixtures of magnesium oxide powders and an acid phosphate solution, and the wastes. Size reduction of Pb-lined plastic glove waste was accomplished by cryofractionation. The Ceramicretetrademark process produces dense, hard ceramic waste forms. Toxicity Characteristic Leaching Procedure (TCLP) results showed excellent stabilization of both Hg and Pb in the waste forms. The principal advantage of this technology is that immobilization of contaminants is the result of both chemical stabilization and subsequent microencapsulation of the reaction products. Based on bench-scale studies, Ceramicretetrademark technology has been implemented in the fabrication of 5-gal waste forms at ANL-West. Approximately 35 kg of real waste has been treated. The TCLP is being conducted on the samples from the 5-gal waste forms. It is expected that because the waste forms pass the limits set by the EPAs Universal Treatment Standard, they will be sent to a radioactive-waste disposal facility

  18. Waste-to-Energy Evaluation: U.S. Virgin Islands

    Energy Technology Data Exchange (ETDEWEB)

    Davis, J.; Hasse, S.; Warren, A.

    2011-08-01

    This NREL technical report evaluates the environmental impact and fundamental economics of waste-to-energy (WTE) technology based on available data from commercially operating WTE facilities in the United States. In particular, it considers life-cycle impacts of WTE as compared to landfill disposal and various forms of electrical generation, as well as WTE impacts on source reduction or recycling programs. In addition, it evaluates the economics and potential environmental impact of WTE in the U.S. Virgin Islands (USVI) based on existing USVI waste stream characterization data, recycling challenges unique to the USVI, and the results of cost and environmental modeling of four municipal solid waste (MSW) management options, including landfill, refuse-derived fuel (RDF) production, recycling, and gassification plus RDF.

  19. Energy recovery from municipal solid wastes in Italy: Actual study and perspective for future

    International Nuclear Information System (INIS)

    Brunetti, N.; Ciampa, F.; De Cecco, C.

    1992-01-01

    Materials and energy recovery from municipal solid wastes (MSW) and assimilable waste, and their re-use is one of strong points of current regulations and tendencies, both at the national and at community level in Europe. In Italy, the interest in energy recovery from renewable sources has been encouraged by energy-savings law which included financial incentives for thermal plant building if low grade fuels such as MSW were employed. New electric power prices imposed by Italian Electric Power Authority, ENEL, encourage energy recovery from waste burners. This paper aims to point out the present state of energy recovery from wastes in Italy, trends and prospects to satisfy, with new plants, the need for waste thermal destruction and part of the demand for energy in the different Italian regions: only about 10% of MSW are burned and just a small percentage of the estimated amount of recoverable energy (2 MTOE/y) is recuperated. Different technological cycles are discussed: incineration of untreated wastes and energy recovery; incineration (or gasification) of RDF (refuse derived fuels) and heat-electricity co-generation; burning of RDF in industrial plants, in addition to other fuels

  20. Radioactive wastes: a comparison of U.S. military and civilian inventories

    International Nuclear Information System (INIS)

    Krugmann, H.; von Hippel, F.

    1977-01-01

    Contrary to widespread belief, the accumulated inventory of fission products generated by the still small U.S. civilian nuclear power industry may already be comparable to that generated in the past by U.S. military nuclear programs. Although the volumes of the military wastes are very large, they are on the average almost 100 times more dilute than projected commercial high-level wastes

  1. Complementary installation for very low radioactive wastes in El Cabril

    International Nuclear Information System (INIS)

    Gregorio, S. de; Garcia Sierra, J.; Navarro, M.

    2006-01-01

    The Industry and Energy Commission of the Spanish Parliament in 1998, and the Economy and Treasury Commission in 2001 and 2002, approved both resolutions and demanded the govern to take the initiatives in the way to Spain will have capacity to store very low radioactive and low risk wastes, that can not be deals like conventional waste, in such a way that will not suppose a loss of the important strategic value that means the capacity of the actual vaults at El Cabril, Designed to disposed of radioactive waste in with high specific activity. The very low activity wastes are the part of waste with less activity in the whole low and intermediate level wastes. The complementary installation will form part, of the actual facility of El Cabril. The total activity to disposed of will be hold in the Reference Inventory allowed to the C. A. El Cabril, not being needed an extension of the actual inventory. (Author)

  2. Low and intermediate level waste in SFR-1. Reference Waste Inventory 2007

    Energy Technology Data Exchange (ETDEWEB)

    Almkvist, Lisa (Vattenfall Power Consultant AB, Stockholm (SE)); Gordon, Anna (Swedish Nuclear Fuel and Waste Management Co., Stockholm (SE))

    2007-11-15

    The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR 1 at the time of closure. The report will form the basis for the release calculation in the safety analysis for SFR 1. Three different scenarios are explored in this report; the waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 50 and 60 years and that closure of the SFR 1 repository will take place in 2040 or 2050 respectively. The third scenario is where the repository is full (one part where the activity adds up to 1016 Bq and one part where the repository is considered full regarding volume). In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemotoxic material has been identified in the waste. The inventory is estimated using the Prosit-interface which extracts information from the Triumf database. The inventory is based on so called 'waste types' and the waste types' 'reference waste package'. The reference waste package combined with a prognosis of the number of waste packages to be delivered to SFR 1 gives the final waste inventory for SFR 1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60Co and 137Cs in waste packages and on measurements of 239Pu and 240Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors

  3. Low and intermediate level waste in SFR-1. Reference waste inventory

    International Nuclear Information System (INIS)

    Riggare, P.; Johansson, Claes

    2001-06-01

    The objective with this report is to describe all the waste and the waste package that is expected to be deposited in SFR-1 at the time of closure. This report is a part of the SAFE project (Safety Assessment of Final Repository for Radioactive Operational Waste), i.e. the renewed safety assessment of SFR-1. The accounted waste inventory has been used as input to the release calculation that has been performed in the SAFE project. The waste inventory is based on an estimated operational lifetime of the Swedish nuclear power plants of 40 years and that closure of the SFR repository will happen in 2030. In the report, data about geometries, weights, materials, chemicals and radionuclide are given. No chemo toxic material has been identified in the waste. The inventory is based on so called waste types and the waste types reference waste package. The reference waste package combined with a prognosis of the number of waste packages to the year 2030 gives the final waste inventory for SFR-1. All reference waste packages are thoroughly described in the appendices of this report. The reference waste packages are as far as possible based on actual experiences and measurements. The radionuclide inventory is also based on actual measurements. The inventory is based on measurements of 60 Co and 137 Cs in waste packages and on measurements 239 Pu and 240 Pu in reactor water. Other nuclides in the inventory are calculated with correlation factors. In the SAFE project's prerequisites it was said that one realistic and one conservative (pessimistic) inventory should be produced. The conservative one should then be used for the release calculations. In this report one realistic and one conservative radionuclide inventory is presented. The conservative one adds up to 10 16 Bq. Regarding materials there is only one inventory given since it is not certain what is a conservative assumption

  4. Thirteenth annual U.S. DOE low-level radioactive waste management conference: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1991-12-31

    The 40 papers in this document comprise the proceedings of the Department of Energy`s Thirteenth Annual Low-Level Radioactive Waste Management Conference that was held in Atlanta, Georgia, on November 19--21, 1991. General subjects addressed during the conference included: disposal facility design; greater-than-class C low-level waste; public acceptance considerations; waste certification; site characterization; performance assessment; licensing and documentation; emerging low-level waste technologies; waste minimization; mixed waste; tracking and transportation; storage; and regulatory changes. Papers have been processed separately for inclusion on the data base.

  5. Waste Handling in SVAFO's Hot Cell

    International Nuclear Information System (INIS)

    Moeller, Jennifer; Ekenborg, Fredrik; Hellsten, Erik

    2016-01-01

    The decommissioning and dismantling of nuclear installations entails the generation of significant quantities of radioactive waste that must be accepted for disposal. In order to optimise the use of the final repositories for radioactive waste it is important that the waste be sent to the correct repository; that is, that waste containing short-lived radionuclides not be designated as long-lived due to conservative characterisation procedures. The disposal of short-lived waste in a future Swedish repository for long-lived waste will result in increased costs, due to the higher volumetric cost of the disposal as well as costs associated with decades of interim storage before disposal can occur. SVAFO is a non-profit entity that is responsible for the decommissioning of nuclear facilities from historical research and development projects in Sweden. They provide interim storage for radioactive waste arising from research activities until the final repository for long-lived waste is available. SVAFO's offices and facilities are located on the Studsvik site on the east coast of Sweden near the town of Nykoeping. Some of the retired facilities that SVAFO is in the process of decommissioning are located elsewhere in Sweden. The HM facility is a small waste treatment plant owned and operated by SVAFO. The plant processes both liquid and solid radioactive wastes. The facility includes a hot cell equipped with a compactor, a saw and other tools as well as manipulators for the handling and packaging of waste with high dose rates. The cell is fitted with special systems for transporting waste in and passing it out in drums. As with most hot cells there has been an accumulation of surface contamination on the walls, floor and other surfaces during decades of operation. Until recently there has been no attempt to quantify or characterize this contamination. Current practices dictate that after waste is handled in the hot cell it is conservatively designated as long

  6. Granite disposal of U.S. high-level radioactive waste.

    Energy Technology Data Exchange (ETDEWEB)

    Freeze, Geoffrey A.; Mariner, Paul E.; Lee, Joon H.; Hardin, Ernest L.; Goldstein, Barry; Hansen, Francis D.; Price, Ronald H.; Lord, Anna Snider

    2011-08-01

    This report evaluates the feasibility of disposing U.S. high-level radioactive waste in granite several hundred meters below the surface of the earth. The U.S. has many granite formations with positive attributes for permanent disposal. Similar crystalline formations have been extensively studied by international programs, two of which, in Sweden and Finland, are the host rocks of submitted or imminent repository license applications. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in granite media. In this report we develop scoping performance analyses, based on the applicable features, events, and processes (FEPs) identified by international investigators, to support generic conclusions regarding post-closure safety. Unlike the safety analyses for disposal in salt, shale/clay, or deep boreholes, the safety analysis for a mined granite repository depends largely on waste package preservation. In crystalline rock, waste packages are preserved by the high mechanical stability of the excavations, the diffusive barrier of the buffer, and favorable chemical conditions. The buffer is preserved by low groundwater fluxes, favorable chemical conditions, backfill, and the rigid confines of the host rock. An added advantage of a mined granite repository is that waste packages would be fairly easy to retrieve, should retrievability be an important objective. The results of the safety analyses performed in this study are consistent with the results of comprehensive safety assessments performed for sites in Sweden, Finland, and Canada. They indicate that a granite repository would satisfy established safety criteria and suggest that a small number of FEPs would largely control the release and transport of radionuclides. In the event the U.S. decides to pursue a potential repository in granite, a detailed evaluation of these FEPs would be needed to inform site

  7. Exile and Self-Actualization in Pauline Kaldas’s “He Had Dreamed of Returning” and “Airport”

    Directory of Open Access Journals (Sweden)

    Nisreen M. Sawwa

    2017-01-01

    Full Text Available Against common pessimistic readings of exile in postcolonial fiction, this article employs the notion of “self-actualization” that argues for people’s desire to accomplish everything they are capable of and their need to realize their potential. Within a comparative context and using identity theory and diaspora studies, the article illustrates how self-actualization keeps the immigrants from experiencing exile in two Arab American short stories by Pauline Kaldas: “Airport” (2009a and “He Had Dreamed of Returning” (2009b. This article shows how the main characters of “Airport” and “He Had Dreamed of Returning,” Samir and Hani respectively, fulfill the American Dream and how Hoda, Samir’s wife, pictures America as the place where she can realize her ambitions. However, Nancy, Hani’s wife, achieves her potential in Egypt rather than America, where she feels needed as a teacher. Thus, Samir and Hani do not get dislocated in America, and Nancy has a sense of belonging in Egypt. Hence, the article utilizes the American Dream and a reverse side of it, and it shows how Samir’s, Hani’s, and Nancy’s self-actualization is a counter to feelings of exile. In other words, the three characters do not experience loss of identity and displacement in the countries they emigrate to. Rather, they fulfill their dreams there and find/create new identities which have been suppressed in their hometowns, which enhances a view of identity as fluid rather than fixed. Briefly put, this article presents the self-actualization of immigrants in new locales as a counter to different levels of dislocation and exile.

  8. Waste Tank Organic Safety Program: Analytical methods development. Progress report, FY 1994

    International Nuclear Information System (INIS)

    Campbell, J.A.; Clauss, S.A.; Grant, K.E.

    1994-09-01

    The objectives of this task are to develop and document extraction and analysis methods for organics in waste tanks, and to extend these methods to the analysis of actual core samples to support the Waste Tank organic Safety Program. This report documents progress at Pacific Northwest Laboratory (a) during FY 1994 on methods development, the analysis of waste from Tank 241-C-103 (Tank C-103) and T-111, and the transfer of documented, developed analytical methods to personnel in the Analytical Chemistry Laboratory (ACL) and 222-S laboratory. This report is intended as an annual report, not a completed work

  9. [Investigation of actual condition of management and disposal of medical radioactive waste in Korea].

    Science.gov (United States)

    Watanabe, Hiroshi; Nagaoka, Hiroaki; Yamaguchi, Ichiro; Horiuchi, Shoji; Imoto, Atsushi

    2009-07-20

    In order to realize the rational management and disposal of radioactive waste like DIS or its clearance as performed in Europe, North America, and Japan, we investigated the situation of medical radioactive waste in Korea and its enforcement. We visited three major Korean facilities in May 2008 and confirmed details of the procedure being used by administering a questionnaire after our visit. From the results, we were able to verify that the governmental agency had established regulations for the clearance of radioactive waste as self-disposal based on the clearance level of IAEA in Korea and that the medical facilities performed suitable management and disposal of radioactive waste based on the regulations and superintendence of a radiation safety officer. The type of nuclear medicine was almost the same as that in Japan, and the half-life of all radiopharmaceuticals was 60 days or less. While performing regulatory adjustment concerning the rational management and disposal of radioactive waste in Korea for reference also in this country, it is important to provide an enforcement procedure with quality assurance in the regulations.

  10. Emissions from US waste collection vehicles

    International Nuclear Information System (INIS)

    Maimoun, Mousa A.; Reinhart, Debra R.; Gammoh, Fatina T.; McCauley Bush, Pamela

    2013-01-01

    Highlights: ► Life-cycle emissions for alternative fuel technologies. ► Fuel consumption of alternative fuels for waste collection vehicles. ► Actual driving cycle of waste collection vehicles. ► Diesel-fueled waste collection vehicle emissions. - Abstract: This research is an in-depth environmental analysis of potential alternative fuel technologies for waste collection vehicles. Life-cycle emissions, cost, fuel and energy consumption were evaluated for a wide range of fossil and bio-fuel technologies. Emission factors were calculated for a typical waste collection driving cycle as well as constant speed. In brief, natural gas waste collection vehicles (compressed and liquid) fueled with North-American natural gas had 6–10% higher well-to-wheel (WTW) greenhouse gas (GHG) emissions relative to diesel-fueled vehicles; however the pump-to-wheel (PTW) GHG emissions of natural gas waste collection vehicles averaged 6% less than diesel-fueled vehicles. Landfill gas had about 80% lower WTW GHG emissions relative to diesel. Biodiesel waste collection vehicles had between 12% and 75% lower WTW GHG emissions relative to diesel depending on the fuel source and the blend. In 2011, natural gas waste collection vehicles had the lowest fuel cost per collection vehicle kilometer travel. Finally, the actual driving cycle of waste collection vehicles consists of repetitive stops and starts during waste collection; this generates more emissions than constant speed driving

  11. MIxed Waste Integrated Program (MWIP): Technology summary

    International Nuclear Information System (INIS)

    1994-02-01

    The mission of the Mixed Waste Integrated Program (MWIP) is to develop and demonstrate innovative and emerging technologies for the treatment and management of DOE's mixed low-level wastes (MLLW) for use by its customers, the Office of Waste Operations (EM-30) and the Office of Environmental Restoration (EM-40). The primary goal of MWIP is to develop and demonstrate the treatment and disposal of actual mixed waste (MMLW and MTRU). The vitrification process and the plasma hearth process are scheduled for demonstration on actual radioactive waste in FY95 and FY96, respectively. This will be accomplished by sequential studies of lab-scale non-radioactive testing followed by bench-scale radioactive testing, followed by field-scale radioactive testing. Both processes create a highly durable final waste form that passes leachability requirements while destroying organics. Material handling technology, and off-gas requirements and capabilities for the plasma hearth process and the vitrification process will be established in parallel

  12. 42 CFR 71.45 - Food, potable water, and waste: U.S. seaports and airports.

    Science.gov (United States)

    2010-10-01

    ... 42 Public Health 1 2010-10-01 2010-10-01 false Food, potable water, and waste: U.S. seaports and... Inspection § 71.45 Food, potable water, and waste: U.S. seaports and airports. (a) Every seaport and airport..., or waste water or other polluting materials. Arriving aircraft shall discharge such matter only at...

  13. Sources and management of hazardous waste in Papua New Guinea

    Energy Technology Data Exchange (ETDEWEB)

    Singh, K. [Univ. of Papua New Guinea (Papua New Guinea)

    1996-12-31

    Papua New Guinea (PNG) has considerable mineral wealth, especially in gold and copper. Large-scale mining takes place, and these activities are the source of most of PNG`s hazardous waste. Most people live in small farming communities throughout the region. Those living adjacent to mining areas have experienced some negative impacts from river ecosystem damage and erosion of their lands. Industry is centered mainly in urban areas and Generates waste composed of various products. Agricultural products, pesticide residues, and chemicals used for preserving timber and other forestry products also produce hazardous waste. Most municipal waste comes from domestic and commercial premises; it consists mainly of combustibles, noncombustibles, and other wastes. Hospitals generate pathogenic organisms, radioactive materials, and chemical and pharmaceutical laboratory waste. Little is known about the actual treatment of waste before disposal in PNG. Traditional low-cost waste disposal methods are usually practiced, such as use of landfills; storage in surface impoundments; and disposal in public sewers, rivers, and the sea. Indiscriminate burning of domestic waste in backyards is also commonly practiced in urban and rural areas. 10 refs., 4 tabs.

  14. FTIR fiber optic methods for the analysis of Hanford Site waste

    International Nuclear Information System (INIS)

    Rebagay, T.V.; Cash, R.J.; Dodd, D.A.

    1995-06-01

    Sampling and chemical characterization of mixed high-level waste stored in underground tanks at the Hanford Site is currently in progress. Waste tank safety concerns have provided impetus to analyze this waste. A major safety issue is the possibility of significant concentrations of fuel (ferrocyanide and/or organic compounds) in contact with oxidizers (nitrates and nitrites). It is postulated that under dry conditions and elevated temperatures, ferrocyanide- and/or organic-bearing wastes could undergo rapid exothermic reactions. To maintain the tanks in a safe condition, data are needed on the moisture and fuel concentrations in the waste. Because of the highly radioactive nature of the waste, non-radioactive waste simulants mimicking actual waste are used to provide an initial basis for identifying realistic waste tank safety concerns. Emphasis has been placed on the use of new or existing Fourier transform infrared (FTIR)-based systems with potential for field or tank deployment to perform in situ remote waste characterization. Near-infrared diffuse reflectance and mid-infrared attenuated total reflectance fiber optic probes coupled to a Bio-Rad FTS 60A spectrometry system have been evaluated. The near-infrared diffuse reflectance fiber probe system has also been used for preliminary screening of the moisture content and chemical composition of actual Hanford Site waste tank waste core samples. The attributes of this method for analyzing actual radioactive waste are discussed

  15. IAEA's radioactive waste management programme a change in direction for the 1990's

    International Nuclear Information System (INIS)

    Saire, Donald E.

    1991-01-01

    The IAEA has been assisting Member States in the management and disposal of radioactive wastes for almost three decades. This assistance has largely been in the form of providing an international form for the exchange of information on subjects of mutual interest and in offering technical assistance from developed Member States to developing Member States. In the 1980's, the Agency's programme concentrated on the collection, review and publication of state-of-the-art technology covering such subject fields as handling and processing of wastes from nuclear power plants and fuel cycle facilities, gaseous waste management, near-surface disposal of low-and intermediate-level radioactive wastes, deep geological disposal of high-level and transuranic wastes, radiological and environmental effects of waste disposal and the decontamination and decommissioning of nuclear installations. While the Agency still plans to serve as the main forum for the exchange of information on radioactive waste management two new areas of activity will receive emphasis in the 1990's

  16. Nuclear criticality safety analysis summary report: The S-area defense waste processing facility

    International Nuclear Information System (INIS)

    Ha, B.C.

    1994-01-01

    The S-Area Defense Waste Processing Facility (DWPF) can process all of the high level radioactive wastes currently stored at the Savannah River Site with negligible risk of nuclear criticality. The characteristics which make the DWPF critically safe are: (1) abundance of neutron absorbers in the waste feeds; (2) and low concentration of fissionable material. This report documents the criticality safety arguments for the S-Area DWPF process as required by DOE orders to characterize and to justify the low potential for criticality. It documents that the nature of the waste feeds and the nature of the DWPF process chemistry preclude criticality

  17. Polymer solidification of mixed wastes at the Rocky Flats Plant

    International Nuclear Information System (INIS)

    Faucette, A.M.; Logsdon, B.W.; Lucerna, J.J.; Yudnich, R.J.

    1994-01-01

    The Rocky Flats Plant is pursuing polymer solidification as a viable treatment option for several mixed waste streams that are subject to land disposal restrictions within the Resource Conservation and Recovery Act provisions. Tests completed to date using both surrogate and actual wastes indicate that polyethylene microencapsulation is a viable treatment option for several mixed wastes at the Rocky Flats Plant, including nitrate salts, sludges, and secondary wastes such as ash. Treatability studies conducted on actual salt waste demonstrated that the process is capable of producing waste forms that comply with all applicable regulatory criteria, including the Toxicity Characteristic Leaching Procedure. Tests have also been conducted to evaluate the feasibility of macroencapsulating certain debris wastes in polymers. Several methods and plastics have been tested for macroencapsulation, including post-consumer recycle and regrind polyethylene

  18. Outcome of 7-S, TQM technique for healthcare waste management.

    Science.gov (United States)

    Ullah, Junaid Habib; Ahmed, Rashid; Malik, Javed Iqbal; Khan, M Amanullah

    2011-12-01

    To assess the present waste management system of healthcare facilities (HCFs) attached with Shalamar Hospital, Lahore by applying the 7-S technique of Total Quality Management (TQM) and to find out the outcome after imparting training. Interventional quasi-experimental study. The Shalamar Hospital, Lahore, Punjab, Pakistan, November, 2009 to November, 2010. Mckinsey's 7-S, technique of TQM was applied to assess the 220 HCFs from Lahore, Gujranwala and Sheikhupura districts for segregation, collection, transportation and disposal (SCTD) of hospital waste. Direct interview method was applied. Trainings were provided in each institution. After one year action period, the status of four areas of concern was compared before and after training. The parameters studied were segregation, collection, transportation and disposal systems in the 220 HCFs. Each of these were further elaborated by strategy, structure, system, staff, skill, style and stakeholder/shared value factors. Standard error of difference of proportion was applied to assess significance using 95% confidence level. There was marked improvement in all these areas ranging from 20% to 77% following a training program of 3 months. In case of disposal of the waste strategy, structure and system an increase of 60%, 65% and 75% was observed after training. The 7-S technique played a vital role in assessing the hospital waste management system. Training for the healthcare workers played a significant role in healthcare facilities.

  19. Waste. The political significance of the city’s suppressed other

    OpenAIRE

    Chrysanthe Constantinou

    2015-01-01

    MBeneath the city’s achievement as a grand machine of accumulation of matter, in invisibility lies the vast and vague sphere of waste accumulation. In invisibility, a huge network of waste flows is possibly rewriting economies as we know them, reproducing life conditions driven by the politics of wealth. In this parallel territory, recycling is a big business for the few.

  20. Trends in pan evaporation and actual evapotranspiration across the conterminous U.S.: paradoxical or complementary?

    Science.gov (United States)

    Michael T. Hobbins; Jorge A. Ramirez; Thomas C. Brown

    2004-01-01

    Pan evaporation (ETpan) has decreased at 64% of pans in the conterminous U.S. over the past half-century. Comparing trends in ETpan and water budget-derived actual evapotranspiration (ET*a), we observe the so-called ‘‘Pan Evaporation Paradox,’’ which we confirm is no more than a...

  1. Transportable vitrification system demonstration on mixed waste. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J.R.; Whitehouse, J.C. [Westinghouse Savannah River Co., Aiken, SC (United States); Wilson, C.N. [Lockheed Martin Hanford Corp., Richland, WA (United States); Van Ryn, F.R. [Bechtel Jacobs Co., Oak Ridge, TN (United States)

    1998-04-22

    The Transportable Vitrification System (TVS) is a large scale, fully integrated, vitrification system for the treatment of low-level and mixed wastes in the form of sludges, soils, incinerator ash, and many other waste streams. It was demonstrated on surrogate waste at Clemson University and at the Oak Ridge Reservation (ORR) prior to treating actual mixed waste. Treatment of a combination of dried B and C Pond sludge and CNF sludge was successfully demonstrated at ORR in 1997. The demonstration produced 7,616 kg of glass from 7,328 kg of mixed wastes with a 60% reduction in volume. Glass formulations for the wastes treated were developed using a combination of laboratory crucible studies with the actual wastes and small melter studies at Clemson with both surrogate and actual wastes. Initial characterization of the B and C Pond sludge had not shown the presence of carbon or fluoride, which required a modified glass formulation be developed to maintain proper glass redox and viscosity. The CNF sludge challenges the glass formulations due to high levels of phosphate and iron. The demonstration was delayed several times by permitting problems, a glass leak, and electrical problems. The demonstration showed that the two wastes could be successfully vitrified, although the design glass production rate was not achieved. The glass produced met the Universal Treatment Standards and the emissions from the TVS were well within the allowable permit limits.

  2. Transportable vitrification system demonstration on mixed waste. Revision 1

    International Nuclear Information System (INIS)

    Zamecnik, J.R.; Whitehouse, J.C.; Wilson, C.N.; Van Ryn, F.R.

    1998-01-01

    The Transportable Vitrification System (TVS) is a large scale, fully integrated, vitrification system for the treatment of low-level and mixed wastes in the form of sludges, soils, incinerator ash, and many other waste streams. It was demonstrated on surrogate waste at Clemson University and at the Oak Ridge Reservation (ORR) prior to treating actual mixed waste. Treatment of a combination of dried B and C Pond sludge and CNF sludge was successfully demonstrated at ORR in 1997. The demonstration produced 7,616 kg of glass from 7,328 kg of mixed wastes with a 60% reduction in volume. Glass formulations for the wastes treated were developed using a combination of laboratory crucible studies with the actual wastes and small melter studies at Clemson with both surrogate and actual wastes. Initial characterization of the B and C Pond sludge had not shown the presence of carbon or fluoride, which required a modified glass formulation be developed to maintain proper glass redox and viscosity. The CNF sludge challenges the glass formulations due to high levels of phosphate and iron. The demonstration was delayed several times by permitting problems, a glass leak, and electrical problems. The demonstration showed that the two wastes could be successfully vitrified, although the design glass production rate was not achieved. The glass produced met the Universal Treatment Standards and the emissions from the TVS were well within the allowable permit limits

  3. Progress and Lessons Learned in Transuranic Waste Disposition at The Department of Energy's Advanced Mixed Waste Treatment Project

    Energy Technology Data Exchange (ETDEWEB)

    J.D. Mousseau; S.C. Raish; F.M. Russo

    2006-05-18

    This paper provides an overview of the Department of Energy's (DOE) Advanced Mixed Waste Treatment Project (AMWTP) located at the Idaho National Laboratory (INL) and operated by Bechtel BWXT Idaho, LLC(BBWI) It describes the results to date in meeting the 6,000-cubic-meter Idaho Settlement Agreement milestone that was due December 31, 2005. The paper further describes lessons that have been learned from the project in the area of transuranic (TRU) waste processing and waste certification. Information contained within this paper would be beneficial to others who manage TRU waste for disposal at the Waste Isolation Pilot Plant (WIPP).

  4. Waste management, decommissioning and environmental restoration for Canada's nuclear activities. Proceedings

    International Nuclear Information System (INIS)

    2011-01-01

    The Canadian Nuclear Society conference on Waste Management, Decommissioning and Environmental Restoration for Canada's Nuclear Activities was held in Toronto, Ontario, Canada on September 11-14, 2011. The conference provided a forum for discussion of the status and proposed future directions of technical, regularly, environmental, social and economic aspects of radioactive waste management, nuclear facility decommissioning, and environmental restoration activities for Canadian nuclear facilities. The conference included both plenary sessions and sessions devoted to more detailed technical issues. The plenary sessions were focussed on three broad themes: the overall Canadian program; low and intermediate waste; and, international perspectives. Topics of the technical sessions included: OPG's deep geologic repository for low and intermediate level waste; stakeholder interactions; decommissioning projects; uranium mine waste management; used fuel repository - design and safety assessment; federal policies, programs and oversight; regulatory considerations; aboriginal traditional knowledge; geological disposal - CRL site classification; geological disposal - modelling and engineered barriers; Port Hope Area Initiative; waste characterization; LILWM - treatment and processing; decommissioning projects and information management; international experience; environmental remediation; fuel cycles and waste processing.

  5. Radioactivity assessment of the waste deposited in Candonga's lake after the FundÃO Dam's collapse

    Energy Technology Data Exchange (ETDEWEB)

    Freire, Carolina Braccini; Cuccia, Valeria; Lameiras, Fernando Soares; Rocha, Zildete, E-mail: cbf@cdtn.br, E-mail: vc@cdtn.br, E-mail: fsl@cdtn.br, E-mail: rochaz@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Santos, Talita Oliveira, E-mail: talitaolsantos@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Anatomia e Imagem; Sales, Gabriel Lopes, E-mail: gabriellopessales@hotmail.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte (Brazil). Dept. de Engenharia Química

    2017-07-01

    The collapse of the Fundão dam on November 5{sup th} 2015 caused impacts along the hydrographic basin of Rio Doce. Renova Foundation has the mission to repair, restore and reconstruct the regions affected by this disaster. After the collapse, a huge amount of waste was deposited in Risoleta Neves Hydroelectric Power Plant’s lake (Candonga). This waste has been dredged out from the lake and deposited in a nearby pile. The use of the waste from this pile is being considered for civil construction. Samples of this waste pile were collected and they were characterized by X-rays diffraction and EDS analyses, which showed that it consists of 80% quartz, 11% hematite, 8% goethite and 1% clays, mainly kaolinite. There is also organic matter consisting of branches, roots and leaves of trees and plants. Previous radiological analyses made before the collapse showed that the radionuclide content of the waste deposited in Fundão dam was very low. However, the waste deposited in Candonga's lake has a new composition and it is important to evaluate its radioactivity assessment. The analyses results of the Candonda's waste samples will be presented and compared to the results obtained in the Fundão dam samples, before its collapse. (author)

  6. Radioactivity assessment of the waste deposited in Candonga's lake after the FundÃO Dam's collapse

    International Nuclear Information System (INIS)

    Freire, Carolina Braccini; Cuccia, Valeria; Lameiras, Fernando Soares; Rocha, Zildete; Santos, Talita Oliveira; Sales, Gabriel Lopes

    2017-01-01

    The collapse of the Fundão dam on November 5 th 2015 caused impacts along the hydrographic basin of Rio Doce. Renova Foundation has the mission to repair, restore and reconstruct the regions affected by this disaster. After the collapse, a huge amount of waste was deposited in Risoleta Neves Hydroelectric Power Plant’s lake (Candonga). This waste has been dredged out from the lake and deposited in a nearby pile. The use of the waste from this pile is being considered for civil construction. Samples of this waste pile were collected and they were characterized by X-rays diffraction and EDS analyses, which showed that it consists of 80% quartz, 11% hematite, 8% goethite and 1% clays, mainly kaolinite. There is also organic matter consisting of branches, roots and leaves of trees and plants. Previous radiological analyses made before the collapse showed that the radionuclide content of the waste deposited in Fundão dam was very low. However, the waste deposited in Candonga's lake has a new composition and it is important to evaluate its radioactivity assessment. The analyses results of the Candonda's waste samples will be presented and compared to the results obtained in the Fundão dam samples, before its collapse. (author)

  7. Waste. The political significance of the city’s suppressed other

    Directory of Open Access Journals (Sweden)

    Chrysanthe Constantinou

    2015-03-01

    Full Text Available MBeneath the city’s achievement as a grand machine of accumulation of matter, in invisibility lies the vast and vague sphere of waste accumulation. In invisibility, a huge network of waste flows is possibly rewriting economies as we know them, reproducing life conditions driven by the politics of wealth. In this parallel territory, recycling is a big business for the few.

  8. U.S. Nuclear Regulatory Commission Role and Activities Related to U.S. Department of Energy Incidental Waste Determinations

    International Nuclear Information System (INIS)

    Bradford, A.H.; Esh, D.W.; Ridge, A.C.

    2006-01-01

    Section 3116 of the Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005 (NDAA) requires the U.S. Department of Energy (DOE) to consult with the U.S. Nuclear Regulatory Commission (NRC) for certain non-high level waste (HLW) determinations. Under the NDAA, NRC performs consultative technical reviews of DOE's waste determinations and monitors DOE's disposal actions for such waste, but the NRC does not have regulatory authority over DOE's waste disposal activities. The NDAA provides the criteria that must be met to determine that waste is not HLW. The criteria require that the waste does not need to be disposed of in a geologic repository, that highly radioactive radionuclides be removed to the maximum extent practical, and that the performance objectives of 10 CFR 61, Subpart C, be met. The performance objectives contain criteria for protection of the public, protection of inadvertent intruders, protection of workers, and stability of the disposal site after closure. This paper describes NRC's approach to implementing its responsibilities under the NDAA, as well as similar activities being performed for sites not covered by the NDAA. (authors)

  9. Shale disposal of U.S. high-level radioactive waste.

    Energy Technology Data Exchange (ETDEWEB)

    Sassani, David Carl; Stone, Charles Michael; Hansen, Francis D.; Hardin, Ernest L.; Dewers, Thomas A.; Martinez, Mario J.; Rechard, Robert Paul; Sobolik, Steven Ronald; Freeze, Geoffrey A.; Cygan, Randall Timothy; Gaither, Katherine N.; Holland, John Francis; Brady, Patrick Vane

    2010-05-01

    This report evaluates the feasibility of high-level radioactive waste disposal in shale within the United States. The U.S. has many possible clay/shale/argillite basins with positive attributes for permanent disposal. Similar geologic formations have been extensively studied by international programs with largely positive results, over significant ranges of the most important material characteristics including permeability, rheology, and sorptive potential. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in shale media. We develop scoping performance analyses, based on the applicable features, events, and processes identified by international investigators, to support a generic conclusion regarding post-closure safety. Requisite assumptions for these analyses include waste characteristics, disposal concepts, and important properties of the geologic formation. We then apply lessons learned from Sandia experience on the Waste Isolation Pilot Project and the Yucca Mountain Project to develop a disposal strategy should a shale repository be considered as an alternative disposal pathway in the U.S. Disposal of high-level radioactive waste in suitable shale formations is attractive because the material is essentially impermeable and self-sealing, conditions are chemically reducing, and sorption tends to prevent radionuclide transport. Vertically and laterally extensive shale and clay formations exist in multiple locations in the contiguous 48 states. Thermal-hydrologic-mechanical calculations indicate that temperatures near emplaced waste packages can be maintained below boiling and will decay to within a few degrees of the ambient temperature within a few decades (or longer depending on the waste form). Construction effects, ventilation, and the thermal pulse will lead to clay dehydration and deformation, confined to an excavation disturbed zone within

  10. Predicting transport requirements for radioactive-waste slurries

    International Nuclear Information System (INIS)

    Motyka, T.; Randall, C.T.

    1983-01-01

    A method for predicting the transport requirements of radioactive waste slurries was developed. This method involved preparing nonradioactive sludge slurries chemically similar to the actual high-level waste. The rheological and settling characteristics of these synthetic waste slurries were measured and found to compare favorably with data on actual defense waste slurries. Pressure drop versus flow rate data obtained fom a 2-in. slurry test loop confirmed the Bingham plastic behavior of the slurry observed during viscometry measurements. The pipeline tests, however, yielded friction factors 30 percent lower than those predicted from viscometry data. Differences between the sets of data were attributed to inherent problems in interpreting accurate yield-stress values of slurry suspensions with Couette-type viscometers. Equivalent lengths of fittings were also determined and found to be less than that of water at a specified flow rate

  11. Development of Simulants to Support Mixing Tests for High Level Waste and Low Activity Waste

    International Nuclear Information System (INIS)

    EIBLING, RUSSELLE.

    2004-01-01

    The objectives of this study were to develop two different types of simulants to support vendor agitator design studies and mixing studies. The initial simulant development task was to develop rheologically-bounding physical simulants and the final portion was to develop a nominal chemical simulant which is designed to match, as closely as possible, the actual sludge from a tank. The physical simulants to be developed included a lower and upper rheologically bounded: pretreated low activity waste (LAW) physical simulant; LAW melter feed physical simulant; pretreated high level waste (HLW) physical simulant; HLW melter feed physical simulant. The nominal chemical simulant, hereafter referred to as the HLW Precipitated Hydroxide simulant, is designed to represent the chemical/physical composition of the actual washed and leached sludge sample. The objective was to produce a simulant which matches not only the chemical composition but also the physical properties of the actual waste sample. The HLW Precipitated Hydroxide simulant could then be used for mixing tests to validate mixing, homogeneity and representative sampling and transferring issues. The HLW Precipitated Hydroxide simulant may also be used for integrated nonradioactive testing of the WTP prior to radioactive operation

  12. Quality control concept for radioactive waste packages

    International Nuclear Information System (INIS)

    Warnecke, E.; Martens, B.R.; Odoj, R.

    1990-01-01

    In the Federal Republic of Germany a contract with the BfS for the performance of quality control measures is necessary. It is principally possible to apply two alternative methods: random checks on waste packages or qualification of conditioning processes with subsequent inspections. Priority is given to the control by the process qualification. Both methods have successfully been developed in the Federal Republic of Germany and can be applied. In the course of the qualification of conditioning processes it must be demonstrated by inactive and/or active runs that waste packages are produced which fulfil the waste acceptance requirements. The qualification results in the fixation of a handbook for the operation of the respective conditioning process including the process instrumentation and the operational margins. The qualified process will be inspected to assure the compliance of the actual operation with the conditions fixed in the handbook. (orig./DG)

  13. Nuclear waste - a fresh perspective

    International Nuclear Information System (INIS)

    Tammemagi, H.Y.

    1996-01-01

    Rather than looking at the nuclear waste problem in isolation, it should be viewed in the broader context of how society disposes of all of its wastes. A comparison of radioactive and non-radioactive wastes shows, contrary to popular perception, that the properties of these two waste types are actually very similar. However, the methods of regulation and management of the two waste types are very different. It is time that these differences were reconciled - both the nuclear and the non-nuclear waste industries have a lot to gain. There are three main categories of (non-nuclear) waste: municipal wastes, hazardous wastes, and industrial wastes. Rather than treating each of these waste types in separate, isolated compartments, there should be an integration of the principles and regulations involved in their management. The non-nuclear waste industry has much to learn from the nuclear approach

  14. DOE`s integrated low-level waste management program and strategic planning

    Energy Technology Data Exchange (ETDEWEB)

    Duggan, G. [Dept. of Energy, Washington, DC (United States). Office of Environmental Restoration and Waste Management; Hwang, J. [Science Applications International Corp., Germantown, MD (United States)

    1993-03-01

    To meet the DOE`s commitment to operate its facilities in a safe, economic, and environmentally sound manner, and to comply with all applicable federal, state, and local rules, regulations, and agreements, DOE created the Office of Environmental Restoration and Waste Management (EM) in 1989 to focus efforts on controlling waste management and cleaning up contaminated sites. In the first few years of its existence, the Office of Waste Management (EM-30) has concentrated on operational and corrective activities at the sites. In 1992, the Office of Waste Management began to apply an integrated approach to managing its various waste types. Consequently, DOE established the Low-Level Waste Management Program (LLWMP) to properly manage its complex-wide LLW in a consistent manner. The objective of the LLWMP is to build and operate an integrated, safe, and cost-effective program to meet the needs of waste generators. The program will be based on acceptable risk and sound planning, resulting in public confidence and support. Strategic planning of the program is under way and is expected to take two to three years before implementation of the integrated waste management approach.

  15. Study of physical properties, gas generation and gas retention in simulated Hanford waste

    International Nuclear Information System (INIS)

    Bryan, S.A.; Pederson, L.R.; Scheele, R.D.

    1993-04-01

    The purpose of this study was to establish the chemical and physical processes responsible for the generation and retention of gases within high-level waste from Tank 101-SY on the Hanford Site. This research, conducted using simulated waste on a laboratory scale, supports the development of mitigation/remediation strategies for Tank 101-SY. Simulated waste formulations are based on actual waste compositions. Selected physical properties of the simulated waste are compared to properties of actual Tank 101-SY waste samples. Laboratory studies using aged simulated waste show that significant gas generation occurs thermally at current tank temperatures (∼60 degrees C). Gas compositions include the same gases produced in actual tank waste, primarily N 2 , N 2 O, and H 2 . Gas stoichiometries have been shown to be greatly influenced by several organic and inorganic constituents within the simulated waste. Retention of gases in the simulated waste is in the form of bubble attachment to solid particles. This attachment phenomenon is related to the presence of organic constituents (HEDTA, EDTA, and citrate) of the simulated waste. A mechanism is discussed that relates the gas bubble/particle interactions to the partially hydrophobic surface produced on the solids by the organic constituents

  16. Hydroceramics, a ''new'' cementitious waste form material for U.S. defense-type reprocessing waste

    International Nuclear Information System (INIS)

    Siemer, Darryl D.

    2002-01-01

    A ''hydroceramic'' (HC) is a concrete which possesses mineralogy similar to the zeolitized rock indigenous to the USA's current ''basis'' high level radioactive waste (HLW) repository site, Yucca Mountain (YM). It is made by curing a mixture of inorganic waste, calcined clay, vermiculite, Na 2 S, NaOH, plus water under hydrothermal conditions. The product differs from conventional Portland cement and/or slag-based concretes (''grouts'') in that it is primarily comprised of alkali aluminosilicate ''cage minerals'' (cancrinites, sodalites, and zeolites)rather than hydrated calcium silicates (C-S-H in cement-chemistry shorthand). Consequently it microencapsulates individual salt molecules thereby rendering them less leachable than they are from conventional grouts. A fundamental difference between the formulations of HCs and radwaste-type glasses is that the latter contain insufficient aluminum to form insoluble minerals with all of the alkali metals in them. This means that the imposition of worst-case ''repository failure'' (hydrothermal) conditions would cause a substantial fraction of such glasses to alter to water-soluble forms. Since the same conditions tend to reduce the solubility of HC concretes, they constitute a more rugged immobilization sub-system. This paper compares leach characteristics of HCs with those of radwaste-type glasses and points out why hydroceramic solidification makes more sense than vitrification for US defense-type reprocessing waste. (orig.)

  17. Waste container and method for containing waste

    International Nuclear Information System (INIS)

    Ono, Akira; Matsushita, Mitsuhiro; Doi, Makoto; Nakatani, Seiichi.

    1990-01-01

    In a waste container, water-proof membranes and rare earth element layers are formed on the inner surface of a steel plate concrete container in which steel plates are embedded. Further, rear earth element detectors are disposed each from the inner side of the steel plate concrete container by way of a pressure pipe to the outer side of the container. As a method for actually containing wastes, when a plurality of vessels in which wastes are fixed are collectively enhoused to the waste container, cussioning materials are attached to the inner surface of the container and wastes fixing containers are stacked successively in a plurality of rows in a bag made of elastic materials. Subsequently, fixing materials are filled and tightly sealed in the waste container. When the waste container thus constituted is buried underground, even if it should be deformed to cause intrusion of rain water to the inside of the container, the rare earth elements in the container dissolved in the rain water can be detected by the detectors, the containers are exchanged before the rain water intruding to the inner side is leached to the surrounding ground, to previously prevent the leakage of radioactive nuclides. (K.M.)

  18. Strategic Action in Nuclear Waste Disposal: Canada's Adaptive Phased Management Approach

    Energy Technology Data Exchange (ETDEWEB)

    Durant, Darrin [York Univ., Toronto (Canada). Program in Science and Technology Studies

    2006-09-15

    It has been noted that Sweden's SKB has shifted emphasis in its waste management approach, from an emphasis upon geological barriers to engineering barriers. This shift has been called strategic adaptation, though, with the appearance of responsiveness to external demands actually allowing the piloting of previously preferred options. This paper argues that a similar case of strategic adaptation has taken place in Canada, with the NWMO using dialogue with the public and an emphasis upon adaptability in order to pilot through decade-old technical preferences. Is dialogue a cure-all for participatory democracy advocates, and what implications does strategic adaptation have for discussions of the risk society? Nuclear waste disposal programs are now routinely confronting the question of whether to limit the notion of 'risk' to a quantitative discourse about biological and physical effects or allow the discussion to broaden to considerations of social perceptions of acceptability. While the effects of a limited discourse are well studied, this paper explores the other side of the coin, analyzing institutionalized discourse about radwaste management that has broadened to include social perceptions of acceptability. The case in question is Canada's Nuclear Waste Management Organization (NWMO). Despite its inclusive rhetoric, this paper suggests the NWMO's participatory discourse is in fact strategic adaptation: the construction of 'socially acceptable risk' co-opts social acceptance in the preservation of previously preferred technical and policy options. This alternative reading is consistent with trends in other nations, while posing difficult questions for risk society theorists. In both the Finnish and Swedish cases, Lidskog and Litmanen showed that relatively stable social definitions of the risk situation have been packaged differently as local circumstances demand. Strategic arguments have been used to defend and/or justify more

  19. OPERATIONS REVIEW OF THE SAVANNAH RIVER SITE INTEGRATED SALT DISPOSITION PROCESS - 11327

    Energy Technology Data Exchange (ETDEWEB)

    Peters, T.; Poirier, M.; Fondeur, F.; Fink, S.; Brown, S.; Geeting, M.

    2011-02-07

    The Savannah River Site (SRS) is removing liquid radioactive waste from its Tank Farm. To treat waste streams that are low in Cs-137, Sr-90, and actinides, SRS developed the Actinide Removal Process and implemented the Modular Caustic Side Solvent Extraction (CSSX) Unit (MCU). The Actinide Removal Process contacts salt solution with monosodium titanate to sorb strontium and select actinides. After monosodium titanate contact, the resulting slurry is filtered to remove the monosodium titanate (and sorbed strontium and actinides) and entrained sludge. The filtrate is transferred to the MCU for further treatment to remove cesium. The solid particulates removed by the filter are concentrated to {approx} 5 wt %, washed to reduce the sodium concentration, and transferred to the Defense Waste Processing Facility for vitrification. The CSSX process extracts the cesium from the radioactive waste using a customized solvent to produce a Decontaminated Salt Solution (DSS), and strips and concentrates the cesium from the solvent with dilute nitric acid. The DSS is incorporated in grout while the strip acid solution is transferred to the Defense Waste Processing Facility for vitrification. The facilities began radiological processing in April 2008 and started processing of the third campaign ('MarcoBatch 3') of waste in June 2010. Campaigns to date have processed {approx}1.2 million gallons of dissolved saltcake. Savannah River National Laboratory (SRNL) personnel performed tests using actual radioactive samples for each waste batch prior to processing. Testing included monosodium titanate sorption of strontium and actinides followed by CSSX batch contact tests to verify expected cesium mass transfer. This paper describes the tests conducted and compares results from facility operations. The results include strontium, plutonium, and cesium removal, cesium concentration, and organic entrainment and recovery data. Additionally, the poster describes lessons learned during

  20. Hanford Tank 241-S-112 Residual Waste Composition and Leach Test Data

    Energy Technology Data Exchange (ETDEWEB)

    Cantrell, Kirk J.; Krupka, Kenneth M.; Geiszler, Keith N.; Lindberg, Michael J.; Arey, Bruce W.; Schaef, Herbert T.

    2008-08-29

    This report presents the results of laboratory characterization and testing of two samples (designated 20406 and 20407) of residual waste collected from tank S-112 after final waste retrieval. These studies were completed to characterize the residual waste and assess the leachability of contami¬nants from the solids. This is the first report from this PNNL project to describe the composition and leach test data for residual waste from a salt cake tank. All previous PNNL reports (Cantrell et al. 2008; Deutsch et al. 2006, 2007a, 2007b, 2007c) describing contaminant release models, and characterization and testing results for residual waste in single-shell tanks were based on samples from sludge tanks.

  1. Preparation and leaching of radioactive INEL waste forms

    International Nuclear Information System (INIS)

    Schuman, R.P.; Welch, J.M.; Staples, B.A.

    1982-01-01

    The purpose of this study is to prepare and leach test ceramic and glass waste form specimens produced from actual transuranic waste sludges and high-level waste calcines, respectively. Description of wastes, specimen fabrication, leaching procedure, analysis of leachates and results are discussed. The conclusion is that radioactive waste stored at INEL can be readily incorporated in fused ceramic and glass forms. Initial leach testing results indicate that these forms show great promise for safe long-term containment of radioactive wastes

  2. Methods for verifying compliance with low-level radioactive waste acceptance criteria

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-09-01

    This report summarizes the methods that are currently employed and those that can be used to verify compliance with low-level radioactive waste (LLW) disposal facility waste acceptance criteria (WAC). This report presents the applicable regulations representing the Federal, State, and site-specific criteria for accepting LLW. Typical LLW generators are summarized, along with descriptions of their waste streams and final waste forms. General procedures and methods used by the LLW generators to verify compliance with the disposal facility WAC are presented. The report was written to provide an understanding of how a regulator could verify compliance with a LLW disposal facility`s WAC. A comprehensive study of the methodology used to verify waste generator compliance with the disposal facility WAC is presented in this report. The study involved compiling the relevant regulations to define the WAC, reviewing regulatory agency inspection programs, and summarizing waste verification technology and equipment. The results of the study indicate that waste generators conduct verification programs that include packaging, classification, characterization, and stabilization elements. The current LLW disposal facilities perform waste verification steps on incoming shipments. A model inspection and verification program, which includes an emphasis on the generator`s waste application documentation of their waste verification program, is recommended. The disposal facility verification procedures primarily involve the use of portable radiological survey instrumentation. The actual verification of generator compliance to the LLW disposal facility WAC is performed through a combination of incoming shipment checks and generator site audits.

  3. Analysis on one underground nuclear waste repository rock mass in USA

    International Nuclear Information System (INIS)

    Ha Qiuling; Zhang Tiantian

    2012-01-01

    When analyzing the rock mass of a underground nuclear waste repository, the current studies are all based on the loading mechanical condition, and the unloading damage of rock mass is unconsidered. According to the different mechanical condition of actual engineering rock mass of loading and unloading, this paper implements a comprehensive analysis on the rock mass deformation of underground nuclear waste repository through the combination of present loading and unloading rock mass mechanics. It is found that the results of comprehensive analysis and actual measured data on the rock mass deformation of underground nuclear waste repository are basically the same, which provide supporting data for the underground nuclear waste repository. (authors)

  4. Extraction of CdS pigment from waste polyethylene

    NARCIS (Netherlands)

    Wanrooij, P.H.P.; Agarwal, U.S.; Meuldijk, J.; Kasteren, van J.M.N.; Lemstra, P.J.

    2006-01-01

    Cadmium sulfide has often been used as a pigment in plastics such as high-density polyethylene (HDPE). Removal of CdS after the useful life of plastics is desired since it poses an environmental hazard in the waste phase of these plastics. In this study, a process is investigated to convert the

  5. Results from simulated contact-handled transuranic waste experiments at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Molecke, M.A.; Sorensen, N.R.; Krumhansl, J.L.

    1993-01-01

    We conducted in situ experiments with nonradioactive, contact-handled transuranic (CH TRU) waste drums at the Waste Isolation Pilot Plant (WIPP) facility for about four years. We performed these tests in two rooms in rock salt, at WIPP, with drums surrounded by crushed salt or 70 wt % salt/30 wt % bentonite clay backfills, or partially submerged in a NaCl brine pool. Air and brine temperatures were maintained at ∼40C. These full-scale (210-L drum) experiments provided in situ data on: backfill material moisture-sorption and physical properties in the presence of brine; waste container corrosion adequacy; and, migration of chemical tracers (nonradioactive actinide and fission product simulants) in the near-field vicinity, all as a function of time. Individual drums, backfill, and brine samples were removed periodically for laboratory evaluations. Waste container testing in the presence of brine and brine-moistened backfill materials served as a severe overtest of long-term conditions that could be anticipated in an actual salt waste repository. We also obtained relevant operational-test emplacement and retrieval experience. All test results are intended to support both the acceptance of actual TRU wastes at the WIPP and performance assessment data needs. We provide an overview and technical data summary focusing on the WIPP CH TRU envirorunental overtests involving 174 waste drums in the presence of backfill materials and the brine pool, with posttest laboratory materials analyses of backfill sorbed-moisture content, CH TRU drum corrosion, tracer migration, and associated test observations

  6. Colloidal agglomerates in tank sludge: Impact on waste processing

    International Nuclear Information System (INIS)

    Bunker, B.C.; Martin, J.E.

    1998-01-01

    of agglomeration phenomena. This project summarizes work performed after almost two years of a three year project. Significant findings include: Particles in Actual Tank Wastes - Transmission electron microscopy of actual wastes shows that most sludges consist of agglomerates of submicron (< 10 -6 m) primary particles of hydrated oxides and insoluble salts. Model colloid suspensions for this work were selected to duplicate the compositions and particle morphologies in actual waste. Agglomeration of Primary Particles - Static light scattering measurements on both model suspensions and actual wastes show that in the basic salt solutions found in most tank wastes, primary particles undergo extensive aggregation to form fractal agglomerates. The fractal nature of the agglomerates has an enormous impact on slurry properties because fractal objects occupy much more space than dense objects at the same solids loading.'

  7. Taking into account dismantling and decommissioning waste management in conception and operation phases

    International Nuclear Information System (INIS)

    Poncet, Philippe

    2016-01-01

    Managing waste during the Dismantling and Decommissioning (D and D) phase is quite specific and different from what it was during the operation phase. Indeed, waste generated during dismantling could present some analogy especially with regards to the radionuclides spectrum and contents. However waste from dismantling and cleanup could actually presents a lower level of radiologic activity but produced in much larger quantities, which requires new solutions. Moreover the characteristics and quantities of waste to be managed during D and D are highly depending on the way the facility was designed and also how it was actually operated during its life time. Taking future D and D into consideration in the early design as well as during the operation of new facilities is becoming more and more mandatory. It is now an explicit requirement set by safety authorities, to provide - in the license application for news plants - a description of design provisions and future plans for D and D as well as anticipated technical and financial impacts,. Two major aspects are driving the cost and complexity of future D and D operations: waste volumes by categories and occupational exposure while performing the work. To reduce such impacts, key approaches are to maintain areas clean, segregate the waste types and provide appropriate provisions in the design. The paper's first part describes the related design and operation concepts derived from lessons learned, and illustrations by examples are presented in a second part. (author)

  8. Hanford Facility Dangerous Waste Permit Application, 222-S Laboratory Complex

    International Nuclear Information System (INIS)

    WILLIAMS, J.F.

    2000-01-01

    The Hanford Facility Dangerous Waste Permit Application is considered to be a single application organized into a General Information Portion (document number DOE/RL-91-28) and a Unit-Specific Portion. The scope of the Unit-Specific Portion is limited to Part B permit application documentation submitted for individual, operating treatment, storage, and/or disposal units, such as the 222-S Laboratory Complex (this document, DOE/RL-91-27). Both the General Information and Unit-Specific portions of the Hanford Facility Dangerous Waste Permit Application address the content of the Part B permit application guidance prepared by the Washington State Department of Ecology (Ecology 1987 and 1996) and the U.S. Environmental Protection Agency (40 Code of Federal Regulations 270), with additional information needs defined by the Hazardous and Solid Waste Amendments and revisions of Washington Administrative Code 173-303. For ease of reference, the Washington State Department of Ecology alpha-numeric section identifiers from the permit application guidance documentation (Ecology 1996) follow, in brackets, the chapter headings and subheadings. Documentation contained in the General Information Portion is broader in nature and could be used by multiple treatment, storage, and/or disposal units (e.g., the glossary provided in the General Information Portion). Wherever appropriate, the 222-S Laboratory Complex permit application documentation makes cross-reference to the General Information Portion, rather than duplicating text. Information provided in this 222-S Laboratory Complex permit application documentation is current as of August 2000

  9. An overview of U.S. plant performance improvement through effective radioactive waste management

    International Nuclear Information System (INIS)

    Sieberling, R.; Lyons, P.W.

    1986-01-01

    Radioactive waste volumes in the United States commercial nuclear power industry have declined over the last five years after increasing steadily in the late 1970's. This decline is considered to be especially significant because major backfits/modifications and major component repair/replacement at U.S. nuclear power stations have been accomplished during this period. This paper analyzes the key reasons for this performance and outlines the positive effects that this trend has upon overall nuclear plant performance. In reviewing plant performance data, there appears to be a direct correlation between overall plant performance and radioactive waste performance. For example, plants with high capacity factors, strong industrial safety programs, low collective man-rem, and good contamination controls, generally have a history of low radioactive waste volumes. The reasons for this excellent performance is the result of direct management and supervisory involvement in plant operations, implementation of high standards of performance, and monitoring performance against these standards. Using the techniques outlined in this paper, the U.S. nuclear power industry can continue to reduce the volume of low-level solid radioactive waste being generated

  10. Overview of management programs for plutonium-contaminated solid waste in the U.S.A

    International Nuclear Information System (INIS)

    Ramsey, R.W. Jr.; Daly, G.H.

    1975-01-01

    Programs for transuranium-contaminated solid wastes (TRU) in the U.S.A. are emphasizing a reduction in waste generation and the development of appropriate treatments to reduce the volume of wastes requiring interim storage and final disposal. Research and Development is emphasizing the establishment of sufficient information on treatment, hazards and storage to adopt a standardized procedure for handling wastes during an interim retrievable period and for final disposal. Federal responsibility for TRU waste is being proposed except for minimum amounts acceptable for commercial burial

  11. Vehicle Radiation Monitoring Systems for Medical Waste Disposal - 12102

    Energy Technology Data Exchange (ETDEWEB)

    Kondrashov, Vladislav S.; Steranka, Steve A. [RadComm Systems Corp., 2931 Portland Dr., Oakville, ON L6H 5S4 (Canada)

    2012-07-01

    Hospitals often declare their waste as being 'non-radioactive'; however this material often has excessive levels of radiation caused either by an accident or lack of control. To ensure the best possible protection against the accidental receipt of radioactive materials and as a safety precaution for their employees, waste-handling companies have installed large-scale radiation portal monitors at their weigh scales or entry gates of the incinerator plant, waste transfer station, and/or landfill. Large-volume plastic scintillator-based systems can be used to monitor radiation levels at entry points to companies handling medical waste. The recent and intensive field tests together with the thousands of accumulated hours of actual real-life vehicle scanning have proven that the plastic scintillation based system is an appropriate radiation control instrument for waste management companies. The Real-Time background compensation algorithm is flexible with automatic adjustable coefficients that will response to rapidly changing environmental and weather conditions maintaining the preset alarm threshold levels. The Dose Rate correction algorithms further enhance the system's ability to meet the stringent requirements of the waste industries need for Dose Rate measurements. (authors)

  12. Outcome of 7-S, TQM technique for health care waste management

    International Nuclear Information System (INIS)

    Ullah, J.H.; Malik, J.I.

    2011-01-01

    To assess the present waste management system of health care facilities (HCFs) attached with Shalamar Hospital, Lahore by applying the 7-S technique of Total Quality Management (TQM) and to find out the outcome after imparting training. Study Design: Interventional quasi-experimental study. Place and Duration of Study: The Shalamar Hospital, Lahore, Punjab, Pakistan, November, 2009 to November, 2010. Methodology: Mckinsey's 7-S, technique of TQM was applied to assess the 220 HCFs from Lahore, Gujranwala and Sheikhupura districts for segregation, collection, transportation and disposal (SCTD) of hospital waste. Direct interview method was applied. Trainings were provided in each institution. After one year action period, the status of four areas of concern was compared before and after training. The parameters studied were segregation, collection, transportation and disposal systems in the 220 HCFs. Each of these were further elaborated by strategy, structure, system, staff, skill, style and stake holder/shared value factors. Standard error of difference of proportion was applied to assess significance using 95% confidence level. Results: There was marked improvement in all these areas ranging from 20% to 77% following a training program of 3 months. In case of disposal of the waste strategy, structure and system an increase of 60%, 65% and 75% was observed after training. Conclusion: The 7-S technique played a vital role in assessing the hospital waste management system. Training for the health care workers played a significant role in health care facilities. (author)

  13. Thermal and Radiolytic Gas Generation Tests on Material from Tanks 241-U-103, 241-AW-101, 241-S-106, and 241-S-102: Status Report

    International Nuclear Information System (INIS)

    King, C.M.; Bryan, S.A.

    1999-01-01

    This report summarizes progress in evaluating thermal and radiolytic flammable gas generation in actual Hanford single-shell tank wastes. The work described was conducted at Pacific Northwest National Laboratory (PNNL) for the Flammable Gas Safety Project, whose purpose is to develop information to support DE and S Hanford (DESH) and Project Management Hanford Contract (PHMC) subcontractors in their efforts to ensure the safe interim storage of wastes at the Hanford Site. This work is related to gas generation studies performed by Numatec Hanford Corporation (formerly Westinghouse Hanford Company). This report describes the results of laboratory tests of gas generation from actual convective layer wastes from Tank 241-U-103 under thermal and radiolytic conditions. Accurate measurements of gas generation rates from highly radioactive tank wastes are needed to assess the potential for producing and storing flammable gases within the tanks. The gas generation capacity of the waste in Tank 241-U-103 is a high priority for the Flammable Gas Safety Program due to its potential for accumulating gases above the flammability limit (Johnson et al, 1997). The objective of this work was to establish the composition of gaseous degradation products formed in actual tank wastes by thermal and radiolytic processes as a function of temperature. The gas generation tests on Tank 241-U-103 samples focused first on the effect of temperature on the composition and rate of gas generation Generation rates of nitrogen, nitrous oxide, methane, and hydrogen increased with temperature, and the composition of the product gas mixture varied with temperature

  14. Informal E-waste recycling in developing countries: review of metal(loid)s pollution, environmental impacts and transport pathways.

    Science.gov (United States)

    Ackah, Michael

    2017-11-01

    Crude or primitive recycling practices are often adopted in material resource recovery from E-waste in developing nations. Significant human health and environmental impacts may occur because of such practices. Literature on metal(loid)s pollution during E-waste processing is fragmented. Here, I review the health and environmental impacts of E-waste recycling operations and transport pathways of metal(loid)s, dispersed during operations. This paper is organised into five sections. Section 1 relates to the background of global E-waste generation and legal/illegal trade, citing specific cases from Ghana and other developing nations. Section 2 provides a brief information on sources of metal(loid)s in E-waste. Section 3 describes characteristics of informal E-waste recycling operations in developing nations. Section 4 examines the health and environmental impacts in E-waste recycling while section 5 evaluates major transport pathways of metal(loid)s contaminants.

  15. The Waste Management Plan integration into Decommissioning Plan of the WWR-S research reactor from Romania

    International Nuclear Information System (INIS)

    Barariu, Gheorghe; Oprescu, Theodor; Filip, Mihaela; Sociu, Florin

    2008-01-01

    The paper presents the progress of the Radioactive Waste Management Plan which accompanies the Decommissioning Plan for research reactor WWR-S located in Magurele, Ilfov, near Bucharest, Romania. The new variant of the Decommissioning Plan was elaborated taking into account the IAEA recommendation concerning radioactive waste management. A new feasibility study for WWR-S decommissioning was also developed. The preferred safe management strategy for radioactive wastes produced by reactor decommissioning is outlined. The strategy must account for reactor decommissioning, as well as rehabilitation of the existing Radioactive Waste Treatment Plant and the upgrade of the Radioactive Waste Disposal Facility at Baita-Bihor. Furthermore, the final rehabilitation of the laboratories and reusing of cleaned reactor building is envisaged. An inventory of each type of radioactive waste is presented. The proposed waste management strategy is selected in accordance with the IAEA assistance. Environmental concerns are part of the radioactive waste management strategy. (authors)

  16. Attenuation of heavy metal leaching from hazardous wastes by co-disposal of wastes

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Wookeun; Shin, Eung Bai [Hanyang Univ., Ansan (Korea, Republic of); Lee, Kil Chul; Kim, Jae Hyung [National Institute of Environmental Research, Seoul (Korea, Republic of)] [and others

    1996-12-31

    The potential hazard of landfill wastes was previously evaluated by examining the extraction procedures for individual waste, although various wastes were co-disposed of in actual landfills. This paper investigates the reduction of extraction-procedure toxicity by co-disposing various combinations of two wastes. When two wastes are mixed homogeneously, the extraction of heavy metals from the waste mixture is critically affected by the extract pH. Thus, co-disposal wastes will have a resultant pH between the pH values of its constituent. The lower the resultant pH, the lower the concentrations of heavy metals in the extract. When these wastes are extracted sequentially, the latter extracted waste has a stronger influence on the final concentration of heavy metals in the extract. Small-scale lysimeter experiments confirm that when heavy-metal-bearing leachates Generated from hazardous-waste lysimeters are passed through a nonhazardous-waste lysimeter filled with compost, briquette ash, or refuse-incineration ashes, the heavy-metal concentration in the final leachates decreases significantly. Thus, the heavy-metal leaching could be attenuated if a less extraction-procedure-toxic waste were placed at the bottom of a landfill. 3 refs., 4 figs., 5 tabs.

  17. CJSC ECOMET-S facility for reprocessing and utilisation of radioactive metal waste: operating experience

    International Nuclear Information System (INIS)

    Gelbutovsky, A.B.; Kishkin, S.A.; Mochenov, M.I.; Troshev, A.V.; Cheremisin, P.I.; Chernichenko, A.A.

    2006-01-01

    The principal objective of the paper is to present operating experience in management of radioactive metal waste, originating at nuclear power facilities of the Russian Federation. Issues of radioactive metal waste recycling by melting, with the purpose of unrestricted re-use in industry, or restricted re-use within the nuclear industry, have been considered. The necessity for using a method of melting at the final stage of radioactive metal waste recycling has been proved. Priority measures to be taken and results achieved in the implementation of the Governmental purpose-oriented programme 'Radioactive Metal Waste Reprocessing and Utilization' have been considered, the CJSC ECOMET-S being the main contractor on the Programme. Main specifications and results of operating a commercial melting facility, owned by CJSC 'ECOMET-S' and used to recycle low-level radioactive metal waste originated at the Leningrad Nuclear Power Plant, have been presented. (author)

  18. U.S. Department of Energy's initiatives for proliferation prevention in Russia: results of radioactive liquid waste treatment project, year 2

    International Nuclear Information System (INIS)

    Pokhitonov, Y.; Kamachev, V.; Kelley, D.

    2010-10-01

    The objective of the project is to engage weapons scientists with training and research programs at selected nuclear sites in Russia and apply high technology polymers to immobilize legacy ILW and HLW liquids that have posed environmental challenges over the years. One compelling advantage of the projects is that V.G. Khlopin Radium Institute and Pacific Nuclear Solutions have been engaged in applied research for seven years to validate the performance and effectiveness of the polymer technology for use with radioactive liquids. With conclusive results of the research work on sixty active and simulant waste streams, the project can focus on actual applications of the technology at Ozersk (Mayak), Sever sk (SCC) Zheleznogorsk (MCC) and Gatchyna rather that on pure research. The long term objective of the project is find viable waste management solutions for serious radioactive and chemical contamination that has existed in Russia and the U. S. for several decades. The polymer technologies may be applied to all radioactive liquid. This paper summarizes the experimental work of the immobilization process and data definition of the most effective polymer compositions in addition to determining the optimum polymer to liquid ratios for economic considerations. (Author)

  19. Cermets for high level waste containment

    International Nuclear Information System (INIS)

    Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.

    1978-01-01

    Cermet materials are currently under investigation as an alternate for the primary containment of high level wastes. The cermet in this study is an iron--nickel base metal matrix containing uniformly dispersed, micron-size fission product oxides, aluminosilicates, and titanates. Cermets possess high thermal conductivity, and typical waste loading of 70 wt % with volume reduction factors of 2 to 200 and low processing volatility losses have been realized. Preliminary leach studies indicate a leach resistance comparable to other candidate waste forms; however, more quantitative data are required. Actual waste studies have begun on NFS Acid Thorex, SRP dried sludge and fresh, unneutralized SRP process wastes

  20. Risk perception on management of nuclear high-level and transuranic waste storage

    Energy Technology Data Exchange (ETDEWEB)

    Dees, Lawrence A. [Colorado Christian Univ., Lakewood, CO (United States)

    1994-08-15

    The Department of Energy`s program for disposing of nuclear High-Level Waste (HLW) and transuranic (TRU) waste has been impeded by overwhelming political opposition fueled by public perceptions of actual risk. Analysis of these perceptions shows them to be deeply rooted in images of fear and dread that have been present since the discovery of radioactivity. The development and use of nuclear weapons linked these images to reality and the mishandling of radioactive waste from the nations military weapons facilities has contributed toward creating a state of distrust that cannot be erased quickly or easily. In addition, the analysis indicates that even the highly educated technical community is not well informed on the latest technology involved with nuclear HLW and TRU waste disposal. It is not surprising then, that the general public feels uncomfortable with DOE`s management plans for with nuclear HLW and TRU waste disposal. Postponing the permanent geologic repository and use of Monitored Retrievable Storage (MRS) would provide the time necessary for difficult social and political issues to be resolved. It would also allow time for the public to become better educated if DOE chooses to become proactive.

  1. An overview of the hazardous waste remedial actions program: hazardous and mixed waste activities for the U.S. Departments of energy and defense

    International Nuclear Information System (INIS)

    Craig, Robert B.; Rothermich, Nancy E.

    1991-01-01

    In May 1987 all mixed waste generated at the U.S. Department of Energy (DOE) facilities became jointly regulated by the U.S. Environmental Protection Agency (EPA) and DOE. The Department of Defense (DOD) generates hazardous wastes and is also regulated by the EPA. To maintain or attain compliance, both DOE and DOD have initiated compliance activities on all hazardous and mixed waste streams. This compliance includes the development of innovative technologies and processes to avoid the generation of hazardous and mixed wastes, development of technologies to treat the process wastes that are unavoidably generated, development of technologies to restore the environment where wastes have been released to the environment, the cleanup of asbestos and the monitoring of radon in federal facilities, the completion of remedial investigation/feasibility studies, and development of the data systems that are necessary to compile this information. This paper will describe each of these activities as they relate to compliance with the Resource Conservation and Recovery Act and/or CERCLA and their implementing regulations

  2. The U.S. Nuclear Regulatory Commission's role in the management of radioactive wastes

    International Nuclear Information System (INIS)

    Bernero, R.M.; Kalman, K.L.

    1991-01-01

    This paper discusses the U.S. Nuclear Regulatory Commission's (NRC's) role in the management of radioactive waste. It provides an overview of the NRC's programs for the management of High-Level and Low-Level radioctive wastes and uranium mill tailings. (author) 7 refs

  3. High-level waste program. Progress report, October 1, 1978--March 31, 1979

    International Nuclear Information System (INIS)

    Blanco, R.E.; Lotts, A.L.

    1979-05-01

    Development of a low-level waste management analysis computer code has been initiated. As part of an ongoing update of the ORIGEN code, models of PWRs fueled with denatured 233 U and denatured 235 U in ThO 2 are being developed. Information on waste management for 18 alternative fuel cycles has been assembled for use by NASAP. Documentation of experimental and calculational work done for the assessment of partitioning and transmutation is progressng, while the conceptual design, cost estimation, and risk analysis of partitioning and waste disposal facilities are scheduled for completion in September 1979. Assuming diffusion control, leach studies with FUETAP concretes indicate effective diffusivities in the order of less than or equal to 10 -10 cm 2 /s for plutonium, cesium, and strontium, respectively. Hydrogen and oxygen combine rapidly at ambient temperature in the presence of FUETAP concrete. A spray calciner is currently being fabricated and installed in a trailer laboratory. New cermet sintering techniques have been developed which drastically reduce the sintering temperatures and times required to achieve satisfactory densification. The radioisotope volatility losses observed during the processing of actual wastes to form cermets were very low. Conversions of actual SRP acid and dried sludge wastes to cermets have been performed. A series of cermet leach tests conducted by PNL have been completed, and the samples are currently being examined at ORNL. Results indicate that cermets exhibit promising leach resistance and durability

  4. Progress in evaluating the hazard of ferrocyanide waste storage tanks

    International Nuclear Information System (INIS)

    Babad, Harry; Cash, Robert J.; Postma, Arlin

    1992-01-01

    There are 177 high-level waste tanks on the Hanford site. Twenty-four single-shell tanks are identified as potential safety issues. These tanks contain quantities of ferrocyanide, nitrate, and nitrite salts that potentially could explode under certain conditions. Efforts were initiated in September 1990 to determine the reactive properties of the ferrocyanide waste and to define the criteria necessary to ensure tank safety until mitigation or remediation actions, if required, could be implemented. This paper describes the results of recent chemical and physical studies on synthetic ferrocyanide waste mixtures. Data obtained from monitoring, tank behavior modeling, and research studies on waste have provided sufficient understanding of the tank behavior. The Waste Tank Safety Program is exploring whether the waste in many of the ferrocyanide tanks actually represents an unreviewed safety question. The General Accounting Office (GAO) in October 1990 suggested that ferrocyanide tank accident scenarios exceed the bounds of the Hanford Environmental Impact Statement. Using the same assumptions Westinghouse Hanford Company (WHC) staff confirmed the consistency of the GAO report calculations. The hypothetical accident scenario in the GAO report, and in the EIS, are based on several assumptions that may, or may not reflect actual tank conditions. The Ferrocyanide Stabilization Program at Westinghouse Hanford (summarized in this paper) will provide updated and new data using scientific research with synthetic wastes and characterization of actual tank samples. This new information will replace the assumptions on tank waste chemical and physical properties allowing an improved recalculation of current safety and future risk associated with these tanks. (author)

  5. Progress in evaluating the hazards of ferrocyanide waste storage tanks

    International Nuclear Information System (INIS)

    Babad, H.; Cash, R.; Postma, A.

    1992-03-01

    There are 177 high-level waste tanks on the Hanford site. Twenty-four single-shell tanks are identified as potential safety issues. These tanks contain quantities of ferrocyanide, nitrate, and nitrite salts that potentially could explode under certain conditions. Efforts were initiated in September 1990 to determine the reactive properties of the ferrocyanide waste and to define the criteria necessary to ensure tank safety until mitigation or remediation actions, if required, could be implemented. This paper describes the results of recent chemical and physical studies on synthetic ferrocyanide waste mixtures. Data obtained from monitoring, tank behavior modeling, and research studies on waste have provided sufficient understanding of the tank behavior. The Waste Tank Safety Program is exploring to determine whether the waste in many of the ferrocyanide tanks actually represents an unreviewed safety question. The General Accounting Office (GAO) in October 1990 (1) suggested that ferrocyanide-tanks accident scenarios exceed the bounds of the Hanford Environmental Impact Statement (2). Using the same assumptions Westinghouse Hanford Company (WHC) staff confirmed the consistency of the GAO report calculations. The hypothetical accident scenario in the GAO report, and in the EIS, are based on several assumptions that may, or may not reflect actual tank conditions. The Ferrocyanide Stabilization Program at Westinghouse Hanford (summarized in this paper) will provide updated and new data using scientific research with synthetic and actual waste tank characterization. This new information will replace the assumptions on tank waste chemical and physical properties allowing an improved recalculation of current safety and future risk associated with these tanks

  6. A pilot survey of the U.S. medical waste industry to determine training needs for safely handling highly infectious waste.

    Science.gov (United States)

    Le, Aurora B; Hoboy, Selin; Germain, Anne; Miller, Hal; Thompson, Richard; Herstein, Jocelyn J; Jelden, Katelyn C; Beam, Elizabeth L; Gibbs, Shawn G; Lowe, John J

    2018-02-01

    The recent Ebola outbreak led to the development of Ebola virus disease (EVD) best practices in clinical settings. However, after the care of EVD patients, proper medical waste management and disposal was identified as a crucial component to containing the virus. Category A waste-contaminated with EVD and other highly infectious pathogens-is strictly regulated by governmental agencies, and led to only several facilities willing to accept the waste. A pilot survey was administered to determine if U.S. medical waste facilities are prepared to handle or transport category A waste, and to determine waste workers' current extent of training to handle highly infectious waste. Sixty-eight percent of survey respondents indicated they had not determined if their facility would accept category A waste. Of those that had acquired a special permit, 67% had yet to modify their permit since the EVD outbreak. This pilot survey underscores gaps in the medical waste industry to handle and respond to category A waste. Furthermore, this study affirms reports a limited number of processing facilities are capable or willing to accept category A waste. Developing the proper management of infectious disease materials is essential to close the gaps identified so that states and governmental entities can act accordingly based on the regulations and guidance developed, and to ensure public safety. Copyright © 2018 Association for Professionals in Infection Control and Epidemiology, Inc. Published by Elsevier Inc. All rights reserved.

  7. Sampling and analysis plan for sampling of liquid waste streams generated by 222-S Laboratory Complex operations

    International Nuclear Information System (INIS)

    Benally, A.B.

    1997-01-01

    This Sampling and Analysis Plan (SAP) establishes the requirements and guidelines to be used by the Waste Management Federal Services of Hanford, Inc. personnel in characterizing liquid waste generated at the 222-S Laboratory Complex. The characterization process to verify the accuracy of process knowledge used for designation and subsequent management of wastes consists of three steps: to prepare the technical rationale and the appendix in accordance with the steps outlined in this SAP; to implement the SAP by sampling and analyzing the requested waste streams; and to compile the report and evaluate the findings to the objectives of this SAP. This SAP applies to portions of the 222-S Laboratory Complex defined as Generator under the Resource Conservation and Recovery Act (RCRA). Any portion of the 222-S Laboratory Complex that is defined or permitted under RCRA as a treatment, storage, or disposal (TSD) facility is excluded from this document. This SAP applies to the liquid waste generated in the 222-S Laboratory Complex. Because the analytical data obtained will be used to manage waste properly, including waste compatibility and waste designation, this SAP will provide directions for obtaining and maintaining the information as required by WAC173-303

  8. Development of decommissioning engineering support system for fugen. Development of support system during actual dismantlement works

    International Nuclear Information System (INIS)

    Masanori Izumi; Yukihiro Iguchi; Yoshiki Kannehira

    2005-01-01

    The Advanced Thermal Reactor, Fugen Nuclear Power Station was permanently shut down in March 2003, and is now preparing for decommissioning. We have been developing Decommissioning Engineering Support System (DEXUS) aimed at planning optimal dismantlement process and carrying out dismantlement work safely and precisely. DEXUS consists of 'decommissioning planning support system' and 'dismantling support system'. The dismantling support system is developed aiming at using during actual dismantling work. It consists of three subsystems such as 'Worksite Visualization System', 'Dismantling Data Collection System' and 'Generated Waste Management System'. 'Worksite Visualization System' is a support system designed to provide the necessary information to workers during actual dismantlement works. And this system adopts AR (Augmented Reality) technology, overlapping calculation information into real world. 'Dismantling Data Collection System' is to collect necessary data for improving accuracy of decommissioning planning by evaluating work content and worker equipage, work time for dismantlement works. 'Generated Waste Management system' is a system recording necessary information by attaching the barcode to dismantled wastes or the containers. We can get the information of generated waste by recording generation place, generated time, treatment method and the contents. These subsystems enable to carry out reasonable and safe decommissioning of Fugen. In addition, we expect that those systems will be used for decommissioning of other nuclear facilities in the future. (authors)

  9. RADIOACTIVE DEMONSTRATION OF FINAL MINERALIZED WASTE FORMS FOR HANFORD WASTE TREATMENT PLANT SECONDARY WASTE BY FLUIDIZED BED STEAM REFORMING USING THE BENCH SCALE REFORMER PLATFORM

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C.; Burket, P.; Cozzi, A.; Daniel, W.; Jantzen, C.; Missimer, D.

    2012-02-02

    ceramic (mineral) waste form. The mineral waste form that is produced by co-processing waste with kaolin clay in an FBSR process has been shown to be as durable as LAW glass. Monolithing of the granular FBSR product is being investigated to prevent dispersion during transport or burial/storage, but is not necessary for performance. A Benchscale Steam Reformer (BSR) was designed and constructed at the SRNL to treat actual radioactive wastes to confirm the findings of the non-radioactive FBSR pilot scale tests and to qualify the waste form for applications at Hanford. BSR testing with WTP SW waste surrogates and associated analytical analyses and tests of granular products (GP) and monoliths began in the Fall of 2009, and then was continued from the Fall of 2010 through the Spring of 2011. Radioactive testing commenced in 2010 with a demonstration of Hanford's WTP-SW where Savannah River Site (SRS) High Level Waste (HLW) secondary waste from the Defense Waste Processing Facility (DWPF) was shimmed with a mixture of {sup 125/129}I and {sup 99}Tc to chemically resemble WTP-SW. Prior to these radioactive feed tests, non-radioactive simulants were also processed. Ninety six grams of radioactive granular product were made for testing and comparison to the non-radioactive pilot scale tests. The same mineral phases were found in the radioactive and non-radioactive testing.

  10. Commercial treatability study capabilities for application to the US Department of Energy`s anticipated mixed waste streams

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-07-01

    The U.S. Department of Energy (DOE) has established the Mixed Waste Focus Area (MWFA), which represents a national effort to develop and coordinate treatment solutions for mixed waste among all DOE facilities. The hazardous waste component of mixed waste is regulated under the Resource Conservation and Recovery Act (RCRA), while the radioactive component is regulated under the Atomic Energy Act, as implemented by the DOE, making mixed waste one of the most complex types of waste for the DOE to manage. The MWFA has the mission to support technologies that meet the needs of the DOE`s waste management efforts to characterize, treat, and dispose of mixed waste being generated and stored throughout the DOE complex. The technologies to be supported must meet all regulatory requirements, provide cost and risk improvements over available technologies, and be acceptable to the public. The most notable features of the DOE`s mixed-waste streams are the wide diversity of waste matrices, volumes, radioactivity levels, and RCRA-regulated hazardous contaminants. Table 1-1 is constructed from data from the proposed site treatment plans developed by each DOE site and submitted to DOE Headquarters. The table shows the number of mixed-waste streams and their corresponding volumes. This table illustrates that the DOE has a relatively small number of large-volume mixed-waste streams and a large number of small-volume mixed-waste streams. There are 1,033 mixed-waste streams with volumes less than 1 cubic meter; 1,112 mixed-waste streams with volumes between 1 and 1,000 cubic meters; and only 61 mixed-waste streams with volumes exceeding 1,000 cubic meters.

  11. Waste-to-energy: Technical, economic and ecological point of views

    International Nuclear Information System (INIS)

    Cassitto, L.

    1997-01-01

    Overwhelming waste-recycling should be considered more as a psychological than as a technological method to deal with wastes. The best waste disposal systems should actually grant mass or energy recovery from technical, economic and ecological point-of-views. Highest results seem to be granted by waste-to-energy technologies since energy content is the best preserved property after using materials

  12. Waste incineration

    International Nuclear Information System (INIS)

    Rumplmayr, A.; Sammer, G.

    2001-01-01

    Waste incineration can be defined as the thermal conversion processing of solid waste by chemical oxidation. The types of wastes range from solid household waste and infectious hospital waste through to toxic solid, liquid and gaseous chemical wastes. End products include hot incineration gases, composed primarily of nitrogen, carbon dioxide, water vapor and to a smaller extend of non-combustible residue (ash) and air pollutants (e. g. NO x ). Energy can be recovered by heat exchange from the hot incineration gases, thus lowering fossil fuel consumption that in turn can reduce emissions of greenhouse gases. Burning of solid waste can fulfil up to four distinctive objectives (Pera, 2000): 1. Volume reduction: volume reduction of about 90 %, weight reduction of about 70 %; 2. Stabilization of waste: oxidation of organic input; 3. Recovery of energy from waste; 4. Sanitization of waste: destruction of pathogens. Waste incineration is not a means to make waste disappear. It does entail emissions into air as well as water and soil. The generated solid residues are the topic of this task force. Unlike other industrial processes discussed in this platform, waste incineration is not a production process, and is therefore not generating by-products, only residues. Residues that are isolated from e. g. flue gas, are concentrated in another place and form (e. g. air pollution control residues). Hence, there are generally two groups of residues that have to be taken into consideration: residues generated in the actual incineration process and others generated in the flue gas cleaning system. Should waste incineration finally gain public acceptance, it will be necessary to find consistent regulations for both sorts of residues. In some countries waste incineration is seen as the best option for the treatment of waste, whereas in other countries it is seen very negative. (author)

  13. Evaluation of S-type fiberglass composites for use in high-level radioactive waste environments

    International Nuclear Information System (INIS)

    Parra, S.A.

    1996-01-01

    Two types of S-type fiberglass materials were evaluated for use in a high-level radioactive waste environment. The S-type fiberglass composites tested were in the form of tubes and were exposed to a simulated high-level radioactive waste environment consisting of corrosive chemicals, high gamma radiation, and elevated temperatures. The physical properties of the exposed and unexposed tube samples were compared to determine the effects of the simulated environment on the S-type fiberglass composites

  14. Waste management and disposal in Czechoslovakia: Practices and proposals

    International Nuclear Information System (INIS)

    Marek, J.

    1984-01-01

    An overview is presented on the actual practices and planning for the management of radioactive wastes in Czechoslovakia. Types and specific arisings of wastes, applied immobilization processes, and the planning for disposal of reactor wastes are outlined. A comprehensive R and D programme is focussed on the management of reactor wastes, as the spent fuel is returned to the Sovjet Union after a 10 year cooling time. (orig.)

  15. Methodology for generating waste volume estimates

    International Nuclear Information System (INIS)

    Miller, J.Q.; Hale, T.; Miller, D.

    1991-09-01

    This document describes the methodology that will be used to calculate waste volume estimates for site characterization and remedial design/remedial action activities at each of the DOE Field Office, Oak Ridge (DOE-OR) facilities. This standardized methodology is designed to ensure consistency in waste estimating across the various sites and organizations that are involved in environmental restoration activities. The criteria and assumptions that are provided for generating these waste estimates will be implemented across all DOE-OR facilities and are subject to change based on comments received and actual waste volumes measured during future sampling and remediation activities. 7 figs., 8 tabs

  16. NUEVOS RETOS PARA LA ENSEÑANZA DEL INGLÉS EN LA UNIVERSIDAD ECUATORIANA ACTUAL

    Directory of Open Access Journals (Sweden)

    Adrian Abreus González

    2017-01-01

    Full Text Available El artículo presenta las consideraciones de los autores en torno a las perspectivas para la enseñanza de los idiomas extranjeros en Ecuador. En su núcleo se enfatiza en los retos que representan para las instituciones de idiomas la obtención del requisito de graduación para los estudiantes universitarios. Además, se analizan las políticas actuales que contiene el Reglamento de Régimen Académico de Ecuador sobre este particular y se fundamentan los criterios sobre la base de la enseñanza comunicativa de lenguas y los criterios de Brown (2001 y Richards (2006. Finalmente, se establecen los desafíos que en el orden de organización didáctica enfrentan las instituciones de idiomas en este contexto y bajo los preceptos actuales.PALABRAS CLAVE: enseñanza del inglés; instituciones de educación superior; universidad ecuatoriana.NEW CHALLENGES FOR THE TEACHING OF ENGLISH IN ECUADORIAN UNIVERSITIES ABSTRACTThe article presents the authors´ opinions in regards to the perspectives for the teaching of foreign languages in Ecuador. Its core explores the challenges that the graduation requisite for university students represent to language institutes. The current policies of the Ecuadorian Academic Rules around language teaching and learning are also analyzed, which main foundation is based on the ideas of Brown (2001 and Richards (2006. The article also presents the didactic challenges for languages institutions in this context and under the current rules.    KEYWORDS: English teaching; higher education institutions; ecuadorian university.

  17. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

    International Nuclear Information System (INIS)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.

    1995-04-01

    This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

  18. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 3: Appendixes C-H

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J. [and others

    1995-04-01

    This report contains the Appendices for the Analysis of Accident Sequences and Source Terms at Waste Treatment and Storage Facilities for Waste Generated by the U.S. Department of Energy Waste Management Operations. The main report documents the methodology, computational framework, and results of facility accident analyses performed as a part of the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report.

  19. The Cementitious Barriers Partnership Experimental Programs and Software Advancing DOE@@@s Waste Disposal/Tank Closure Efforts @@@ 15436

    International Nuclear Information System (INIS)

    Burns, Heather; Flach, Greg; Smith, Frank; Langton, Christine; Brown, Kevin; Kosson, David; Samson, Eric; Mallick, Pramod

    2015-01-01

    The U.S. Department of Energy Environmental Management (DOE-EM) Office of Tank Waste Management-sponsored Cementitious Barriers Partnership (CBP) is chartered with providing the technical basis for implementing cement-based waste forms and radioactive waste containment structures for long-term disposal. DOE needs in this area include the following to support progress in final treatment and disposal of legacy waste and closure of High-Level Waste (HLW) tanks in the DOE complex: long-term performance predictions, flow sheet development and flow sheet enhancements, and conceptual designs for new disposal facilities. The DOE-EM Cementitious Barriers Partnership is producing software and experimental programs resulting in new methods and data needed for end-users involved with environmental cleanup and waste disposal. Both the modeling tools and the experimental data have already benefited the DOE sites in the areas of performance assessments by increasing confidence backed up with modeling support, leaching methods, and transport properties developed for actual DOE materials. In 2014, the CBP Partnership released the CBP Software Toolbox @@ @@Version 2.0@@@ which provides concrete degradation models for 1) sulfate attack, 2) carbonation, and 3) chloride initiated rebar corrosion, and includes constituent leaching. These models are applicable and can be used by both DOE and the Nuclear Regulatory Commission (NRC) for service life and long-term performance evaluations and predictions of nuclear and radioactive waste containment structures across the DOE complex, including future SRS Saltstone and HLW tank performance assessments and special analyses, Hanford site HLW tank closure projects and other projects in which cementitious barriers are required, the Advanced Simulation Capability for Environmental Management (ASCEM) project which requires source terms from cementitious containment structures as input to their flow simulations, regulatory reviews of DOE performance

  20. Process development for treatment of fluoride containing wastes

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Mahesh; Kanvinde, V Y [Chemical Engineering Division, Bhabha Atomic Research Centre, Mumbai (India)

    1994-06-01

    Many chemical and metallurgical industries generate liquid wastes containing high values of fluorides in association of nitrates and other metals. Due to harmful effects of fluorides these type of wastes can not be disposed off in the environment without proper treatment. Bench-scale laboratory experiments were conducted to develop a process scheme to fix the fluorides as non-leachable solid waste and fluoride free treated liquid waste for their disposal. To optimize the important parameters, simulated synthetic and actual wastes were used. For this study, three waste streams were collected from Nuclear Fuel Complex, Hyderabad. (author). 6 tabs., 1 fig.

  1. Managing America`s solid waste

    Energy Technology Data Exchange (ETDEWEB)

    1998-03-02

    This report presents an historical overview of the federal role in municipal solid waste management from 1965 to approximately 1995. Attention is focuses on the federal role in safeguarding public health, protecting the environment, and wisely using material and energy resources. It is hoped that this report will provide important background for future municipal solid waste research and development initiatives.

  2. Transportable Vitrification System Demonstration on Mixed Waste

    International Nuclear Information System (INIS)

    Zamecnik, J.R.; Whitehouse, J.C.; Wilson, C.N.; Van Ryn, F.R.

    1998-01-01

    This paper describes preliminary results from the first demonstration of the Transportable Vitrification System (TVS) on actual mixed waste. The TVS is a fully integrated, transportable system for the treatment of mixed and low-level radioactive wastes. The demonstration was conducted at Oak Ridge's East Tennessee Technology Park (ETTP), formerly known as the K-25 site. The purpose of the demonstration was to show that mixed wastes could be vitrified safely on a 'field' scale using joule-heated melter technology and obtain information on system performance, waste form durability, air emissions, and costs

  3. Public information on radioactive waste management in the Netherlands

    International Nuclear Information System (INIS)

    Vrijen, J.

    1993-01-01

    In the Netherlands, the public image of radioactive waste and the management, related with it, was very negative in the early eighties. One of the reasons obviously was that the radioactive waste management system was not clearly defined up till then. The Centrale Organisatie Voor Radioacief Afval, COVRA, was founded in 1982 by Dutch government; COVRA had the task to reorganize the waste management business and also obtained the task to inform the public about the waste issue and what had to be done with it. It was possible to create a positive atmosphere for public information by bringing the public to the waste, actually showing what it was, what we do with it, what the result of this all will be and what we consider to be the actual risks involved, without exaggerating it. We tried to be as open as possible, to create access to the waste as far as possible, and we learned that the waste should be the center point of the visitors center. We tried to avoid that the public would not be able to see for itself what has be done to secure the surroundings and the environment for possibly negative effects resulting from the waste or its treatment or the interim storage of it

  4. Waste management project’s alternatives: A risk-based multi-criteria assessment (RBMCA) approach

    International Nuclear Information System (INIS)

    Karmperis, Athanasios C.; Sotirchos, Anastasios; Aravossis, Konstantinos; Tatsiopoulos, Ilias P.

    2012-01-01

    Highlights: ► We examine the evaluation of a waste management project’s alternatives. ► We present a novel risk-based multi-criteria assessment (RBMCA) approach. ► In the RBMCA the evaluation criteria are based on the quantitative risk analysis of the project’s alternatives. ► Correlation between the criteria weight values and the decision makers’ risk preferences is examined. ► Preference to the multi-criteria against the one-criterion evaluation process is discussed. - Abstract: This paper examines the evaluation of a waste management project’s alternatives through a quantitative risk analysis. Cost benefit analysis is a widely used method, in which the investments are mainly assessed through the calculation of their evaluation indicators, namely benefit/cost (B/C) ratios, as well as the quantification of their financial, technical, environmental and social risks. Herein, a novel approach in the form of risk-based multi-criteria assessment (RBMCA) is introduced, which can be used by decision makers, in order to select the optimum alternative of a waste management project. Specifically, decision makers use multiple criteria, which are based on the cumulative probability distribution functions of the alternatives’ B/C ratios. The RBMCA system is used for the evaluation of a waste incineration project’s alternatives, where the correlation between the criteria weight values and the decision makers’ risk preferences is analyzed and useful conclusions are discussed.

  5. Low-level waste packaging--a managerial perspective

    International Nuclear Information System (INIS)

    Motl, G.P.; Hebbard, L.B. Jr.

    1980-01-01

    This paper emphasizes managerial responsibility for assuring that facility waste is properly packaged. Specifically, existing packaging regulations are summarized, several actual violations are reviewed and, lastly, some recommendations are made to assist managerial personnel in fulfilling their responsibility to ensure that low-level waste is packaged safely and properly before shipment to the disposal site

  6. The actual state of nuclear fuel cycle

    International Nuclear Information System (INIS)

    Sawai, Masako

    2014-01-01

    The describing author's claims are as follows: a new mythology, semi made-in Japan energy, which 'the energy fundamental plan' creates; what is a nuclear fuel cycle?; operation processes in a reprocessing plant; the existing state against a recycle in dream; does a recycle reduce waste masses?; discharged liquid and gaseous radioactive wastes; an evaluation of exposure 'the value 22 μSv is irresponsible'; the putting off of waste problem in reprocessing; a guide in reprocessing; should a reprocessing be a duty of electric power companies? (M.H.)

  7. A comparison of costs associated with utility management options for dry active waste

    Energy Technology Data Exchange (ETDEWEB)

    Hornibrook, C. [EPRI, Palo Alto, CA (United States)

    1995-12-31

    The economics of low level waste management is receiving more attention today than ever before. This is due to four factors: (1) the increases in the cost of processing of these wastes; (2) increases in the cost of disposal; (3) the addition of storage costs for those without access to disposal; and (4) the increasing competitive nature of the electric generation industry. These pressures are forcing the industry to update it`s evaluation of the mix of processing that will afford it the best long term economics and minimize it`s risks for unforeseen costs. Whether disposal is available or not, all utilities face the same challenge of minimizing the costs associated with the management of these wastes. There are a number of variables that will impact how a utility manages their wastes but the problem is the uncertainty of what will actually happen, i.e., will disposal be available, when and at what cost. Using the EPRI-developed WASTECOST: DAW code, this paper explores a variety of LLW management options available to utilities. Along with providing the costs and benefits, other technical considerations which play an important part in the management of these wastes are also addressed.

  8. Nuclear Criticality Safety Assessment for Tank 38H Salt Dissolution

    International Nuclear Information System (INIS)

    Davis, P.L.

    1996-01-01

    This assessment report of sample results of the accumulating insoluble solids from Tank 38H demonstrates that an inherent subcritical condition for nuclear criticality safety exists during saltcake dissolution. This report also defines criteria for future sampling of Tank 38H for continued verification of the inherent subcritical condition as saltcake dissolution proceeds

  9. Integrated solid waste management of Scottsdale, Arizona

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The subject document reports the results of an in-depth investigation of the fiscal year 1992 cost of the city of Scottsdale, Arizona, integrated municipal solid waste management (IMSWM) system, the energy consumed to operate the system, and the environmental performance requirements for each of the system`s waste-processing and disposal facilities. The document reports actual data from records kept by participants. Every effort was made to minimize the use of assumptions, and no attempt is made to interpret the data reported. Analytical approaches are documented so that interested analysts may per-form manipulation or further analysis of the data. As such, the report is a reference document for municipal solid waste (MSW) management professionals who are interested in the actual costs and energy consumption, for a 1-year period, of an operating IMSWM system. The report is organized into two main parts. The first part is the executive summary and case study portion of the report. The executive summary provides a basic description of the study area and selected economic and energy information. Within the case study are detailed descriptions of each component operating during the study period; the quantities of solid waste collected, processed, and marketed within the study boundaries; the cost of MSW in Scottsdale; an energy usage analysis; a review of federal, state, and local environmental requirement compliance; a reference section; and a glossary of terms. The second part of the report focuses on a more detailed discourse on the above topics. In addition, the methodology used to determine the economic costs and energy consumption of the system components is found in the second portion of this report. The methodology created for this project will be helpful for those professionals who wish to break out the costs of their own integrated systems.

  10. Silicate Based Glass Formulations for Immobilization of U.S. Defense Wastes Using Cold Crucible Induction Melters

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Gary L.; Kim, Dong-Sang; Schweiger, Michael J.; Marra, James C.; Lang, Jesse B.; Crum, Jarrod V.; Crawford, Charles L.; Vienna, John D.

    2014-05-22

    The cold crucible induction melter (CCIM) is an alternative technology to the currently deployed liquid-fed, ceramic-lined, Joule-heated melter for immobilizing of U.S. tank waste generated from defense related reprocessing. In order to accurately evaluate the potential benefits of deploying a CCIM, glasses must be developed specifically for that melting technology. Related glass formulation efforts have been conducted since the 1990s including a recent study that is first documented in this report. The purpose of this report is to summarize the silicate base glass formulation efforts for CCIM testing of U.S. tank wastes. Summaries of phosphate based glass formulation and phosphate and silicate based CCIM demonstration tests are reported separately (Day and Ray 2013 and Marra 2013, respectively). Combined these three reports summarize the current state of knowledge related to waste form development and process testing of CCIM technology for U.S. tank wastes.

  11. Milestones for disposal of radioactive waste at the Waste Isolation Pilot Plant (WIPP) in the United States

    International Nuclear Information System (INIS)

    Rechard, R.P.

    1998-04-01

    Since its identification as a potential deep geologic repository in about 1973, the regulatory assessment process for the Waste Isolation Pilot Plant (WIPP) in New Mexico has developed over the past 25 years. National policy issues, negotiated agreements, and court settlements over the first half of the project had a strong influence on the amount and type of scientific data collected. Assessments and studies before the mid 1980s were undertaken primarily (1) to satisfy needs for environmental impact statements, (2) to develop general understanding of selected natural phenomena associated with nuclear waste disposal, or (3) to satisfy negotiated agreements with the State of New Mexico. In the last third of the project, federal compliance policy and actual regulations were sketched out, but continued to evolve until 1996. During this eight-year period, four preliminary performance assessments, one compliance performance assessment, and one verification performance assessment were performed

  12. Milestones for disposal of radioactive waste at the Waste Isolation Pilot Plant (WIPP) in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Rechard, R.P.

    1998-04-01

    Since its identification as a potential deep geologic repository in about 1973, the regulatory assessment process for the Waste Isolation Pilot Plant (WIPP) in New Mexico has developed over the past 25 years. National policy issues, negotiated agreements, and court settlements over the first half of the project had a strong influence on the amount and type of scientific data collected. Assessments and studies before the mid 1980s were undertaken primarily (1) to satisfy needs for environmental impact statements, (2) to develop general understanding of selected natural phenomena associated with nuclear waste disposal, or (3) to satisfy negotiated agreements with the State of New Mexico. In the last third of the project, federal compliance policy and actual regulations were sketched out, but continued to evolve until 1996. During this eight-year period, four preliminary performance assessments, one compliance performance assessment, and one verification performance assessment were performed.

  13. One project`s waste is another project`s resource

    Energy Technology Data Exchange (ETDEWEB)

    Short, J.

    1997-02-01

    The author describes the efforts being made toward pollution prevention within the DOE complex, as a way to reduce overall project costs, in addition to decreasing the amount of waste to be handled. Pollution prevention is a concept which is trying to be ingrained into project planning. Part of the program involves the concept that ultimately the responsibility for waste comes back to the generator. Parts of the program involve efforts to reuse materials and equipment on new projects, to recycle wastes to generate offsetting revenue, and to increase awareness, accountability and incentives so as to stimulate action on this plan. Summaries of examples are presented in tables.

  14. Organic Tank Safety Project: development of a method to measure the equilibrium water content of Hanford organic tank wastes and demonstration of method on actual waste

    International Nuclear Information System (INIS)

    Scheele, R.D.; Bredt, P.R.; Sell, R.L.

    1996-09-01

    Some of Hanford's underground waste storage tanks contain Organic- bearing high level wastes that are high priority safety issues because of potentially hazardous chemical reactions of organics with inorganic oxidants in these wastes such as nitrates and nitrites. To ensure continued safe storage of these wastes, Westinghouse Hanford Company has placed affected tanks on the Organic Watch List and manages them under special rules. Because water content has been identified as the most efficient agent for preventing a propagating reaction and is an integral part of the criteria developed to ensure continued safe storage of Hanford's organic-bearing radioactive tank wastes, as part of the Organic Tank Safety Program the Pacific Northwest National Laboratory developed and demonstrated a simple and easily implemented procedure to determine the equilibrium water content of these potentially reactive wastes exposed to the range of water vapor pressures that might be experienced during the wastes' future storage. This work focused on the equilibrium water content and did not investigate the various factors such as at sign ventilation, tank surface area, and waste porosity that control the rate that the waste would come into equilibrium, with either the average Hanford water partial pressure 5.5 torr or other possible water partial pressures

  15. Waste Processing Cost Recovery at Los Alamos National Laboratory-Analysis and Recommendations

    International Nuclear Information System (INIS)

    Booth, St. R.

    2009-01-01

    Los Alamos National Laboratory is implementing full cost recovery for waste processing in fiscal year 2009 (FY2009), after a transition year in FY2008. Waste processing cost recovery has been implemented in various forms across the nuclear weapons complex and in corporate America. The fundamental reasoning of sending accurate price signals to waste generators is economically sound, and leads to waste minimization and reduced waste expense over time. However, Los Alamos faces significant implementation challenges because of its status as a government-owned, contractor-operated national scientific institution with a diverse suite of experimental and environmental cleanup activities, and the fact that this represents a fundamental change in how waste processing is viewed by the institution. This paper describes the issues involved during the transition to cost recovery and the ultimate selection of the business model. Of the six alternative cost recovery models evaluated, the business model chosen to be implemented in FY2009 is Recharge Plus Generators Pay Distributed Direct. Under this model, all generators who produce waste must pay a distributed direct share associated with their specific waste type to use a waste processing capability. This cost share is calculated using the distributed direct method on the fixed cost only, i.e., the fixed cost share is based on each program's forecast proportion of the total Los Alamos volume forecast of each waste type. (Fixed activities are those required to establish the waste processing capability, i.e., to make the process ready, permitted, certified, and prepared to handle the first unit of waste. Therefore, the fixed cost ends at the point just before waste begins to be processed. The activities to actually process the waste are considered variable.) The volume of waste actually sent for processing is charged a unit cost based solely on the variable cost of disposing of that waste. The total cost recovered each year is the

  16. Enforcement Alert: Hazardous Waste Management Practices at Mineral Processing Facilities Under Scrutiny by U.S. EPA; EPA Clarifies 'Bevill Exclusion' Wastes and Establishes Disposal Standards

    Science.gov (United States)

    This is the enforcement alert for Hazardous Waste Management Practices at Mineral Processing Facilities Under Scrutiny by U.S. EPA; EPA Clarifies 'Bevill Exclusion' Wastes and Establishes Disposal Standards

  17. U.S. Army Corps of Engineers Waste Experiences: More Than You May Think

    International Nuclear Information System (INIS)

    Hays, D. C.; Honerlah, H. B.

    2003-01-01

    U.S. Army Corps of Engineers (USACE) works with other federal, and state agencies through several different programs on numerous Hazardous, Toxic, and Radioactive Waste (HTRW) sites. Formerly Utilized Sites Remediation Program (FUSRAP), Formerly Used Defense Sites (FUDS), EPA Superfund, Installation Restoration, Army Deactivated Nuclear Reactor Program, and many other programs present hazardous, radioactive, and mixed waste issues. While the USACE has a reputation of excellent dirt movers, little is discussed of our other waste management experiences. This paper discusses some of the challenges facing the Health Physics (HP) staff of the USACE. The HP staff is currently organized as one team, the Radiation Safety Support Team (RSST), comprised of 15 individuals at 6 locations across the country. With typical RSST missions including HP consultation to USACE activities world wide, many waste challenges arise. These challenges have involved radioactive wastes of all classifications and stability. Sealed and unsealed sources; instruments and dials; contaminated earth and debris; liquids; lab, reactor, and medical wastes are all successfully managed by the USACE. USACE also develops, evaluates, and utilizes waste treatment Types of radioactive waste at HTRW sites include: Low Level Radioactive Wastes (LLRW) (class A, B, C, and greater than C), 11e.(2), Transuranic (TRU), Mixed, and Naturally Occurring (NORM/TENORM)

  18. Low-level radioactive waste disposal: radiation protection laws

    International Nuclear Information System (INIS)

    Chapuis, A.M.; Guetat, P.; Garbay, H.

    1991-01-01

    The politics of radioactive waste management is a part of waste management and activity levels are one of the components of potential waste pollutions in order to assume man and environment safety. French regulations about personnel and public' radiation protection defines clearly the conditions of radioactive waste processing, storage, transport and disposal. But below some activity levels definite by radiation protection laws, any administrative procedures or processes can be applied for lack of legal regulations. So regulations context is not actually ready to allow a rational low-level radioactive waste management. 15 refs.; 4 tabs.; 3 figs

  19. Waste generated by the future decommissioning of the Magurele VVR-S Research Reactor

    International Nuclear Information System (INIS)

    Dragolici, F.; Turcanu, C.N.; Dragolici, A.C.

    2001-01-01

    Nuclear Research Reactor WWR-S from the National Institute of Research and Development for Physics and Nuclear Engineering 'Horia Hulubei', Bucharest-Magurele, was commissioned in July 1957 and it was shut down in December 1997. At the moment the reactor is in conservation state. During its operation this reactor worked at an average power of 2MW, almost 3216 h/year, producing a total thermal power of 230 x 10 3 MWh. No major modifications or improvements were made during the 40 years of operation to the essential parts of the reactor, respective to the primary cooling system, reactor vessel, active core and electronic devices. So, all components of the measure, control and protection systems are old, generally at the technical level of the 1950s, therefore a reason why in December 1997 the operation was ceased. At present, the reactor can be considered, by IAEA definition in the first stage (reactor shut down, but the vital functions are maintained and monitored). The survey is related to the second stage - restrictive use of the area. To develop a real decommissioning project, it was first necessary to evaluate the volume and the characteristics of the radioactive waste which will be generated. Radioactive waste generated during the decommissioning of Magurele WR-S research reactor may be classified as: Activated wastes (internal structures, horizontal channels and thermal column, biological shield); Contaminated wastes (primary circuit non-activated components, hot cells, some technological rooms as main hall, pumps room, radioactive material transfer areas, ventilation building and stack); Possibly contaminated materials from any area of reactor building and ventilation building. After 40 years of nuclear research activities, all such areas are suspected of contamination. The volume of wastes that will result from WWR-S Research Reactor decommissioning is summarized

  20. Comparison or organic and inorganic ion exchange materials for removal of cesium and strontium from Hanford waste

    Energy Technology Data Exchange (ETDEWEB)

    Brown, G.N.; Carson, K.J.; DesChane, J.R.; Elovich, R.J. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-10-01

    This work is part of an ESP-CP task to develop and evaluate high-capacity, selective, solid extractants for the uptake of cesium, strontium, and technetium (Cs, Sr, and Tc) from nuclear wastes. Pacific Northwest National Laboratory (PNNL) staff, in collaboration with researchers from industry, academia, and national laboratories are investigating these and other novel and commercial ion exchangers for use in nuclear waste remediation of groundwater, HLW, and LLW. Since FY 1995, experimental work at PNNL has focused on small-scale batch distribution (K{sub d}) testing of numerous solid sorbents with actual and simulated Hanford wastes, chemical and radiolytic stability of various organic ion exchanger resins, bench-scale column ion exchange testing in actual and simulated Complexant Concentrate (CC) and Neutralized Current Acid Waste (NCAW), and Tc and Sr removal from groundwater and LLW. In addition, PNNL has continued to support various site demonstrations at the Idaho National Engineering Laboratory, Savannah River Site, West Valley Nuclear Services, Hanford N-Springs, and Hanford N-Basin using technologies developed by their industrial partners. This summary will focus on batch distribution results from the actual waste tests. The data collected in these development and testing tasks provide a rational basis for the selection and direct comparison of various ion exchange materials in simulated and actual HLW, LLW, and groundwater. In addition, prediction of large-scale column loading performance for the materials tested is possible using smaller volumes of actual waste solution. The method maximizes information while minimizing experimental expense, time, and laboratory and process wastes.

  1. The Cementitious Barriers Partnership Experimental Programs and Software Advancing DOE’s Waste Disposal/Tank Closure Efforts – 15436

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Heather [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Flach, Greg [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith, Frank [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Langton, Christine [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Brown, Kevin [Vanderbilt Univ./CRESP, Nashville, TN (United States); Kosson, David [Vanderbilt Univ./CRESP, Nashville, TN (United States); Samson, Eric [SIMCO Technologies, Inc. (United States); Mallick, Pramod [US DOE, Washington, DC (United States)

    2015-01-27

    The U.S. Department of Energy Environmental Management (DOE-EM) Office of Tank Waste Management-sponsored Cementitious Barriers Partnership (CBP) is chartered with providing the technical basis for implementing cement-based waste forms and radioactive waste containment structures for long-term disposal. DOE needs in this area include the following to support progress in final treatment and disposal of legacy waste and closure of High-Level Waste (HLW) tanks in the DOE complex: long-term performance predictions, flow sheet development and flow sheet enhancements, and conceptual designs for new disposal facilities. The DOE-EM Cementitious Barriers Partnership is producing software and experimental programs resulting in new methods and data needed for end-users involved with environmental cleanup and waste disposal. Both the modeling tools and the experimental data have already benefited the DOE sites in the areas of performance assessments by increasing confidence backed up with modeling support, leaching methods, and transport properties developed for actual DOE materials. In 2014, the CBP Partnership released the CBP Software Toolbox –“Version 2.0” which provides concrete degradation models for 1) sulfate attack, 2) carbonation, and 3) chloride initiated rebar corrosion, and includes constituent leaching. These models are applicable and can be used by both DOE and the Nuclear Regulatory Commission (NRC) for service life and long-term performance evaluations and predictions of nuclear and radioactive waste containment structures across the DOE complex, including future SRS Saltstone and HLW tank performance assessments and special analyses, Hanford site HLW tank closure projects and other projects in which cementitious barriers are required, the Advanced Simulation Capability for Environmental Management (ASCEM) project which requires source terms from cementitious containment structures as input to their flow simulations, regulatory reviews of DOE performance

  2. Waste management research abstracts. Information on radioactive waste management research in progress or planned. Vol. 28

    International Nuclear Information System (INIS)

    2003-11-01

    This issue contains 184 abstracts that describe research in progress in the field of radioactive waste management. The research abstracts contained in the Waste Management Research Abstracts Volume 28 (WMRA 28) were collected between October 1, 2002 and September 30, 2003. The abstracts reflect research in progress, or planned, in the field of radioactive waste management. They present ongoing work in various countries and international organizations. Although the abstracts are indexed by country, some programmes are actually the result of cooperation among several countries. Indeed, a primary reason for providing this compilation of programmes, institutions and scientists engaged in research into radioactive waste management is to increase international co-operation and facilitate communications

  3. Release to the gas phase of metals, S and Cl during combustion of dedicated waste fractions

    DEFF Research Database (Denmark)

    Pedersen, Anne Juul; van Lith, Simone Cornelia; Frandsen, Flemming

    2010-01-01

    The release to the gas phase of inorganic elements such as alkali metals. Cl, S, and heavy metals in Waste-to-Energy (WtE) boilers is a challenge. Besides the risk of harmful emissions to the environment, inorganic elements released from the grate may cause severe ash deposition and corrosion...... and the link to the formation of fly ash and aerosols in full-scale waste incinerators. The release of metals, S and Cl from four dedicated waste fractions was quantified as a function of temperature in a lab-scale fixed-bed reactor. The waste fractions comprised chromated copper arsenate (CCA) impregnated....... The lab-scale release results were then compared with results from a related, full-scale partitioning study, in which test runs with the addition of similar, dedicated waste fractions to a base-load waste had been performed in a grate-fired WtE boiler. In general, the elements Al, Ca, Cr, Cu, Fe, Mg, Si...

  4. NEXT GENERATION MELTER(S) FOR VITRIFICATION OF HANFORD WASTE: STATUS AND DIRECTION

    International Nuclear Information System (INIS)

    Ramsey, W.G.; Gray, M.F.; Calmus, R.B.; Edge, J.A.; Garrett, B.G.

    2011-01-01

    Vitrification technology has been selected to treat high-level waste (HLW) at the Hanford Site, the West Valley Demonstration Project and the Savannah River Site (SRS), and low activity waste (LAW) at Hanford. In addition, it may potentially be applied to other defense waste streams such as sodium bearing tank waste or calcine. Joule-heated melters (already in service at SRS) will initially be used at the Hanford Site's Waste Treatment and Immobilization Plant (WTP) to vitrify tank waste fractions. The glass waste content and melt/production rates at WTP are limited by the current melter technology. Significant reductions in glass volumes and mission life are only possible with advancements in melter technology coupled with new glass formulations. The Next Generation Melter (NGM) program has been established by the U.S. Department of Energy's (DOE's), Environmental Management Office of Waste Processing (EM-31) to develop melters with greater production capacity (absolute glass throughput rate) and the ability to process melts with higher waste fractions. Advanced systems based on Joule-Heated Ceramic Melter (JHCM) and Cold Crucible Induction Melter (CCIM) technologies will be evaluated for HLW and LAW processing. Washington River Protection Solutions (WRPS), DOE's tank waste contractor, is developing and evaluating these systems in cooperation with EM-31, national and university laboratories, and corporate partners. A primary NGM program goal is to develop the systems (and associated flowsheets) to Technology Readiness Level 6 by 2016. Design and testing are being performed to optimize waste glass process envelopes with melter and balance of plant requirements. A structured decision analysis program will be utilized to assess the performance of the competing melter technologies. Criteria selected for the decision analysis program will include physical process operations, melter performance, system compatibility and other parameters.

  5. Ion exchange removal of strontium from simulated and actual N-Springs well water at the Hanford 100-N Area

    International Nuclear Information System (INIS)

    Brown, G.N.; Carson, K.J.; DesChane, J.R.; Elovich, R.J.; Kafka, T.M.; White, L.R.

    1996-06-01

    Experimental ion exchange studies are being conducted by the Pacific Northwest national Laboratory (PNNL) under the Efficient Separations and Processing (ESP) Crosscutting Program to evaluate newly emerging materials and technologies for removing cesium, strontium, technetium, and transuranic elements from simulated and actual wastes at Hanford. Previous work focused on applications to treat high-level alkaline tank wastes, but many of the technologies can also be applied in process and ground-water remediation. Ultimately, each process must be evaluated in terms of life-cycle costs, removal efficiency, process chemical consumption and recycle, stability of materials exposed to chemicals and radiation, compatibility with other process streams, secondary waste generation, process and maintenance costs, and final material disposal. This report assesses the performance of the 3M-designed Process Absorber Development Unit (PADU) and the AlliedSignal-produced sodium nonatitanate (NaTi) material in trace quantities of strontium from simulated and actual Hanford N-Springs ground water. The experimental objective was to determine the strontium-loading breakthrough profile of a proprietary 3M-engineered material in either disk or cartridge forms

  6. Thermal cycling and vibration response for PREPP concrete waste forms

    International Nuclear Information System (INIS)

    Nielson, R.M.; Welch, J.M.

    1983-06-01

    The Process Experimental Pilot Plant (PREPP) will process those transuranic wastes which do not satisfy the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria. Since these wastes will contain considerable quantities of combustible materials, incineration will be an integral part of the treatment process. Four basic types of PREPP ash wastes have been identified. The four types are designated high metal box waste, combustible waste, average waste, and inorganic sludge. In this process, the output of the incinerator is a mixture of ash and shredded noncombustible material (principally metals) which is separated into two sizes, -1/4 inch (under-size waste) and reverse arrow 1/4 inch (oversize waste). These wastes are solidified with hydraulic cement in 55-gallon drums. Simulated PREPP waste forms prepared by Colorado School of Mines Research Institute were subjected to thermal cycling and vibration testing to demonstrate compliance with the WIPP immobilization criterion. Although actual storage and transport conditions are expected to vary somewhat from those utilized in the testing protocol, the generation of only very small amounts of particulate suggests that the immobilization criterion should be routinely met for similar waste form formulations and production procedures. However, the behavior of waste forms containing significant quantities of off-gas scrubber sludge or considerably higher waste loadings may differ. Limited thermal cycling and vibration testing of prototype waste forms should be conducted if the final formulations or production methods used for actual waste forms differ appreciably from those tested in this study. If such testing is conducted, consideration should be given to designing the experiment to accommodate a larger number of thermal cycles more representative of the duration of storage expected

  7. New strategic solid waste management in Sicily

    International Nuclear Information System (INIS)

    Messineo, A.; Panno, D.; Ticali, D.

    2005-01-01

    The solid waste management is, today, a very critical issue. In spite of all the attempts in order to recovery and to recycle waste, the dump still remains the more followed solution, while only a small part of solid waste is going to be burnt down. But the rubbish dump isn't, actually, an environmentally sustainable solution. In the last years the waste incineration systems with energy recovery are spreading more over the territory, and if on one hand they allow to recover energy, on the other they also generate waste. So the emergency remains and it has to be faced. Today, the waste incineration system with energy recovery seems to be the best solution for this problem. the following article examinates the main strategic aspects of the solid waste management in Sicily after the General Plan of Waste Management application [it

  8. Associations among Elementary School Children’s Actual Motor Competence, Perceived Motor Competence, Physical Activity and BMI: A Cross-Sectional Study

    Science.gov (United States)

    Stodden, David; Brian, Ali; True, Larissa; Cardon, Greet; Tallir, Isabel; Haerens, Leen

    2016-01-01

    Background Positive associations between motor competence and physical activity have been identified by means of variable-centered analyses. To expand the understanding of these associations, this study used a person-centered approach to investigate whether different combinations (i.e., profiles) of actual and perceived motor competence exist (aim 1); and to examine differences in physical activity levels (aim 2) and weight status (aim 3) among children with different motor competence-based profiles. Materials and Methods Children’s (N = 361; 180 boys = 50%; Mage = 9.50±1.24yrs) actual motor competence was measured with the Test of Gross Motor Development-2 and their perceived motor competence via the Self Perception Profile for Children. We assessed physical activity via accelerometers; height through stadiometers, and weight through scales. Cluster analyses (aim 1) and MANCOVAs (aim 2 & 3) were used to analyze the data. Results The analysis generated two predictable groups: one group displaying relatively high levels of both actual (M TGMD-2 percentile = 42.54, SD = 2.33) and perceived motor competence (M = 3.42, SD = .37; high-high), and one group with relatively low levels of both (M percentile = 9.71, SD = 3.21; M PMC = 2.52, SD = .35; low-low). One additional group was also identified as having relatively low levels of actual motor competence (M percentile = 4.22, SD = 2.85) but relatively high levels of perceived motor competence (M = 3.52, SD = .30; low-high). The high-high group demonstrated higher daily physical activity (M = 48.39±2.03) and lower BMI (M = 18.13±.43) than the low-low group (MMVPA = 37.93±2.01; MBMI = 20.22±.42). The low-high group had similar physical activity-levels as the low-low group (M = 36.21±2.18) and did not significantly differ in BMI (M = 19.49±.46) from the other two groups. Conclusions A combination of high actual and perceived motor competence is related to higher physical activity and lower weight status. It is thus

  9. Test plan for Fauske and Associates to perform tube propagation experiments with simulated Hanford tank wastes

    International Nuclear Information System (INIS)

    Carlson, C.D.; Babad, H.

    1996-05-01

    This test plan, prepared at Pacific Northwest National Laboratory for Westinghouse Hanford Company, provides guidance for performing tube propagation experiments on simulated Hanford tank wastes and on actual tank waste samples. Simulant compositions are defined and an experimental logic tree is provided for Fauske and Associates (FAI) to perform the experiments. From this guidance, methods and equipment for small-scale tube propagation experiments to be performed at the Hanford Site on actual tank samples will be developed. Propagation behavior of wastes will directly support the safety analysis (SARR) for the organic tanks. Tube propagation may be the definitive tool for determining the relative reactivity of the wastes contained in the Hanford tanks. FAI have performed tube propagation studies previously on simple two- and three-component surrogate mixtures. The simulant defined in this test plan more closely represents actual tank composition. Data will be used to support preparation of criteria for determining the relative safety of the organic bearing wastes

  10. Estimating the magnitude of food waste generated in South Africa

    CSIR Research Space (South Africa)

    Oelofse, Suzanna HH

    2012-08-01

    Full Text Available Throughout the developed world, food is treated as a disposable commodity. Between one third and half of all food produced for human consumption globally is estimated to be wasted. However, attempts to quantify the actual magnitude of food wasted...

  11. Ecological risk of reprocessing of uranium wastes of the Gafurov's tailing pit

    International Nuclear Information System (INIS)

    Khakimov, N.; Nazarov, Kh.M.; Kamalov, D.D.; Mirsaidov, U.M.

    2010-01-01

    Present article is devoted to assessment ecological risk of reprocessing of uranium wastes of the Gafurov's tailing pit. Authors set as their purpose the investigation of ecological risk of reprocessing of uranium wastes of the Gafurov-city's tailing pit. Having carried out their investigations, the authors came to the conclusion that the effect of radon gas during reprocessing to workers and specialists is minimal if they follow all the required safety rules.

  12. Waste management research abstracts volume 27. Information on radioactive waste management research in progress or planned

    International Nuclear Information System (INIS)

    2002-10-01

    This issue of the Waste Management Research Abstracts (WMRA) contains 148 abstracts that describe research in progress in the field of radioactive waste management. The research abstracts contained in Volume 27 (WMRA 27) were collected between July 1, 2001 and September 30, 2002. The abstracts present ongoing work in various countries and international organizations. Although the abstracts are indexed by country, many programmes are actually the result of co-operation among several countries. Indeed, a primary reason for providing this compilation of programmes, institutions and scientists engaged in research into radioactive waste management is to increase international co-operation and facilitate communications

  13. Disposal of radioactive waste in Romania. Present and future strategy

    International Nuclear Information System (INIS)

    Rodna, A.; Garlea, C.

    2002-01-01

    The paper begins with the presentation of the actual situation of radioactive waste management in Romania. The organizations responsible for radioactive waste management and their capabilities are described, including radioactive waste disposal. The main provisions of the 'Draft law regarding the management of nuclear spent fuel and radioactive waste, in view of their final disposal' are also presented, with accent on the responsibilities of the National Radioactive Waste Agency (ANDRAD) and on the fund for radioactive waste and spent fuel management and for decommissioning. The paper ends with the presentation of the future radioactive waste and spent fuel management strategy. (author)

  14. Potential benefits of waste transmutation to the U.S. high-level waste respository

    Energy Technology Data Exchange (ETDEWEB)

    Michaels, G.E. [Oak Ridge National Laboratory, TN (United States)

    1995-10-01

    This paper reexamines the potential benefits of waste transmutation to the proposed U.S. geologic repository at the Yucca Mountain site based on recent progress in the performance assessment for the Yucca Mountain base case of spent fuel emplacement. It is observed that actinides are assumed to have higher solubility than in previous studies and that Np and other actinides now dominate the projected aqueous releases from a Yucca Mountain repository. Actinides are also indentified as the dominant source of decay heat in the repository, and the effect of decay heat in perturbing the hydrology, geochemistry, and thermal characteristics of Yucca Mountain are reviewed. It is concluded that the potential for thermally-driven, buoyant, gas-phase flow at Yucca Mountain introduces data and modeling requirements that will increase the costs of licensing the site and may cause the site to be unattractive for geologic disposal of wastes. A transmutation-enabled cold repository is proposed that might allow licensing of a repository to be based upon currently observable characteristics of the Yucca Mountain site.

  15. Risk assessment for the on-site transportation of radioactive wastes for the U.S. Department of Energy Waste Management programmatic environmental impact statement

    International Nuclear Information System (INIS)

    Biwer, B.M.; Monette, F.A.; Chen, S.Y.

    1995-04-01

    This report documents the risk assessment performed for the on-site transportation of radioactive wastes in the U.S. Department of Energy (DOE) Waste Management (WM) Programmatic Environmental Impact Statement (PEIS). Risks for the routine shipment of wastes and the impacts from potential accidental releases are analyzed for operations at the Hanford Site (Hanford) near Richland, Washington. Like other large DOE sites, Hanford conducts waste management operations for all wastes types; consequently, the impacts calculated for Hanford are expected to be greater than those for smaller sites. The risk assessment conducted for on-site transportation is intended to provide an estimate of the magnitude of the potential risk for comparison with off-site transportation risks assessed for the WM PEIS

  16. Materials in the U.S. Municipal Waste Stream, 1960 to 2012 (in tons)

    Data.gov (United States)

    U.S. Environmental Protection Agency — The U.S. Environmental Protection Agency (EPA) has collected and reported data on the generation and disposal of waste in the United States for more than 30 years....

  17. Milestones for disposal of radioactive waste at the Waste Isolation Pilot Plant (WIPP) in the United States

    International Nuclear Information System (INIS)

    RECHARD, ROBERT P.

    2000-01-01

    The opening of the Waste Isolation Pilot Plant on March 26, 1999, was the culmination of a regulatory assessment process that had taken 25 years. National policy issues, negotiated agreements, and court settlements during the first 15 years of the project had a strong influence on the amount and type of scientific data collected up to this point. Assessment activities before the mid 1980s were undertaken primarily (1) to satisfy needs for environmental impact statements, (2) to satisfy negotiated agreements with the State of New Mexico, or (3) to develop general understanding of selected natural phenomena associated with nuclear waste disposal. In the last 10 years, federal compliance policy and actual regulations were sketched out, and continued to evolve until 1996. During this period, stochastic simulations were introduced as a tool for the assessment of the WIPP's performance, and four preliminary performance assessments, one compliance performance assessment, and one verification performance assessment were performed

  18. New Equipment and Techniques for Remote Sampling of Stored Radioactive Waste

    International Nuclear Information System (INIS)

    Nance, T.A.

    2001-01-01

    Radioactive waste is stored at the Savannah River Site (SRS), part of the Department of Energy (DOE) complex. This radioactive waste is stored in buried tanks and management of the waste requires several processes, including material addition, heating, cooling, mixing, and transfer from tank to tank. During waste processing, it is necessary to know the chemical components and their characteristics to determine the steps necessary to maintain the waste form or to manipulate the waste into the form desired. Waste characterization begins by obtaining a sample for analysis. High level radioactive waste sampling is routinely done with simple, standard samplers such as a dip sample. Other sampling is non-routine or specialized, with unique, special requirements, such as sampling remote areas that are difficult to reach. Other specialized sampling includes sampling materials with unknown characteristics or material that must be gathered to obtain an adequate sample or materials that must be broken up to sample or forcibly separated from the tank. The samplers described in this paper are specialized samplers. These samplers include the Dip Filter Sampler, Soft Core Sampler, Hard Core Sampler, Circle Scrape Sampler, Small Scrape Sampler, Suction and Strain Sampler, and Vial Snapper Sampler. The Dip Filter Sampler is used to sample floating particulate matter or floating organic matter. The Soft Core Sampler and Hard Core Sampler are used to obtain samples of solids from the tank floor. The Soft Core Sampler is used on soft solids such as sludge and saltcake and the Hard Core Sampler on hardened solid deposits. The Circle Scrape Sampler is used to obtain solid samples through a small entry riser and out from under the riser. The Small Scrape Sampler enters a small entry riser and is used to scrape a sample from the tank wall. The Suction and Strain Sampler is used to gather a remote submerged sample or filter a solid sample from supernate. The Vial Snapper Grab Sampler is

  19. Defense waste solidification studies, 200-S area. Savannah River Plant work request 860504, Project S-1780

    International Nuclear Information System (INIS)

    1977-05-01

    A scope of work and a venture guidance appraisal were prepared for a conceptual process and plant facilities for the solidification and long-term storage of radioactive wastes removed from underground storage tanks in the 241 F and H Areas at the Savannah River Plant. Conceptual design was based on incorporating the highly radioactive waste components in a borosilicate type glass. The scope of work describes facilities for: reclaiming liquid and sludge wastes from F and H area tank farms; separating the sludge from the liquid salt solution by physical processes; removing radioactive cesium from the salt solution by ion exchange techniques; incorporating the dried sludge and cesium in a borosilicate glass in stainless steel containers; evaporating the liquid salt solution and encapsulating the resulting salt cake in a stainless steel container; and storing two years' worth of glass and salt containing cyclinders in separate retrievable surface storage facilities. Operations are to be located in a new area, designated the 200-S area. A full complement of power, general, and service facilities are provided. The venture guidance appraisal based on FY 82 authorization and FY 87 turnover is $2,900,000,000. The figure is suitable for planning purposes only. The Glass-form Waste Case is a variation of the concrete-form waste case (or the Reference Plant Case) reported in DPE--3410. The new venture guidance appraisal for the concrete-form case (updated to a consistent time basis with the glass-form case) is $2,900,000,000, indicating no apparent cost advantage between the two waste product forms

  20. Waste management in Greenland: current situation and challenges

    DEFF Research Database (Denmark)

    Eisted, Rasmus; Christensen, Thomas Højlund

    2011-01-01

    Waste management in Greenland (56 000 inhabitants) is characterized by landfilling, incineration and export to Denmark of small quantities of metals and hazardous waste. The annual amount of waste is estimated to about 50 000 tons but actual data are scarce. Data on the waste composition is basic...... are small and equipped with only moderate flue gas cleaning technology. This report summarizes the current waste management situation in Greenland and identifies important challenges in improving the waste management....... is basically lacking. The scattered small towns and settlements, the climate and the long transport distances between towns and also to recycling industries abroad constitute a complex situation with respect to waste management. The landfills have no collection of gas and leachate and the incinerators...

  1. Biogas plants: Utilization of microorganisms for waste management

    Energy Technology Data Exchange (ETDEWEB)

    Stadlbauer, E A; Trieu, C

    1984-05-01

    The continuous realisation of the basic principles of environmental conservation and compatability demands concern about the problems of eliminating wastes in an industrialised society. Biogas- and therefore technical plants for methane generation by anaerobic fermentation have promoted to effective means of environmental protection and sources of alternate energy from organic wastes. Methane fermentation has been employed for the stabilisation of sludges at municipal waste treatment plants for decades. However, the anaerobic digestion process shows promising applications both for industrial effluents and agricultural wastes as well as municipal solid wastes and covered waste deposits. In view of the advances achieved interdisciplinary cooperation the actual potential and perspective of methane fermentation technology with respect to the solution of the increasing problems of waste management and energy supply is discussed.

  2. ANSTO's waste forms for the 31. century

    Energy Technology Data Exchange (ETDEWEB)

    Vance, E.R.; Begg, B. D.; Day, R. A.; Moricca, S.; Perera, D. S.; Stewart, M. W. A.; Carter, M. L.; McGlinn, P. J.; Smith, K. L.; Walls, P. A.; Robina, M. La

    2004-07-01

    ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and {sup 99}Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)

  3. U.S. policy and current practices for blending low-level radioactive waste for disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kessel, David S.; Kim, Chang Lak [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-09-15

    In the near future, many countries, including the Republic of Korea, will face a significant increase in low level radioactive waste (LLW) from nuclear power plant decommissioning. The purpose of this paper is to look at blending as a method for enhancing disposal options for low-level radioactive waste from the decommissioning of nuclear reactors. The 2007 U.S. Nuclear Regulatory Commission strategic assessment of the status of the U.S. LLW program identified the need to move to a risk-informed and performance-based regulatory approach for managing LLW. The strategic assessment identified blending waste of varying radionuclide concentrations as a potential means of enhancing options for LLW disposal. The NRC's position is that concentration averaging or blending can be performed in a way that does not diminish the overall safety of LLW disposal. The revised regulatory requirements for blending LLW are presented in the revised NRC Branch Technical Position for Concentration Averaging and Encapsulation (CA BTP 2015). The changes to the CA BTP that are the most significant for NPP operation, maintenance and decommissioning are reviewed in this paper and a potential application is identified for decommissioning waste in Korea. By far the largest volume of LLW from NPPs will come from decommissioning rather than operation. The large volumes in decommissioning present an opportunity for significant gains in disposal efficiency from blending and concentration averaging. The application of concentration averaging waste from a reactor bio-shield is also presented.

  4. Innovative cross-flow membrane system for volume reduction of mixed waste

    Energy Technology Data Exchange (ETDEWEB)

    Greene, W. [SpinTek Membrane Systems, Huntington Beach, CA (United States)

    1997-10-01

    In this task, SpinTek Membrane Systems, Inc., and the Institute of Gas Technology are completing engineering development leading to a full-scale demonstration of the SpinTek ST-II High Shear Rotary Membrane Filtration System (ST-II) under a Program Research and Development Agreement (PRDA) with the Federal Energy Technology Center-Morgantown. The SpinTek ST-II technology will be scaled-up, and a two-stage ST-II system will be designed, constructed, and operated on both surrogate and actual feed at the Los Alamos National Laboratory (LANL) Liquid Radioactive Waste Treatment Facility (LRWTF). Results from these studies on both surrogate and actual wastewater streams will also be used by LANL personnel to produce a model for determining the applicability and economics of the SpinTek ST-II system to other DOE waste and process streams. The ST-II is a unique, compact cross-flow membrane system having several advantages in performance and cost compared to currently available systems. Staff at LANL have performed pilot-scale testing with the SpinTek technology to evaluate its feasibility for enhanced radionuclide removal from wastewater at its 5- to 8-million-gallon-per-year LRWTF. Recent data have shown the system`s capabilities to remove radionuclides from the waste stream at concentration factors greater than 2000:1, and performance has exceeded both conventional and all other advanced technologies examined.

  5. The reality of waste management in primary health care units in Brazil.

    Science.gov (United States)

    Alves, Sergiane B; e Souza, Adenícia C S; Tipple, Anaclara F V; Rezende, Keyti C A D; de Resende, Fabiana R; Rodrigues, Érika G; Pereira, Milca S

    2014-09-01

    A large number of users are serviced in primary health care units in Brazil, both in health facilities and in households. These services generate waste that must be managed safely, but there is no legislation that regulates this type of waste management in Brazil. The objective of this study was to analyse the production and management of waste in primary health care. A direct observation was performed of the stages in the handling and weighing of waste generated in primary health care units in the municipality of Goiânia (Brazil). The units generated infectious, chemical, and common waste, as well as sharp objects. The generation of waste ranged between 0.027 and 0.075 kg user-day. The generated waste was classified mostly as common and recyclable. Flaws were observed in the management of all types of waste. The critical point is segregation. Only 34.1% of the waste disposed of as infectious actually belonged to this group, the rest was ordinary waste. Flaws at this stage increase the volume of infectious waste, the occupational and environmental risks, and associated costs. Intervention to change this reality is needed and it requires the careful preparation of a waste management plan, corroborating structural changes to the implementation of this plan, and professional training and public policies to guide waste management in primary care, especially those generated in households. © The Author(s) 2014.

  6. Site and facility waste transportation services planning documents

    International Nuclear Information System (INIS)

    Ratledge, J.E.; Schmid, S.; Danese, L.

    1991-01-01

    The Office of Civilian Radioactive Waste Management (OCRWM) will eventually ship Purchasers' (10 CFR 961.3) spent nuclear fuel from approximately 122 commercial nuclear facilities. The preparation and maintenance of Site- and Facility-Specific Transportation Services Planning Documents (SPDs) and Site-Specific Servicing Plans (SSSPs) provides a focus for advanced planning and the actual shipping of waste, as well as the overall development of transportation requirements for the waste transportation system. SPDs will be prepared for each of the affected nuclear waste facilities, with initial emphasis on facilities likely to be served during the earliest years of the Federal Waste Management System (FWMS) operations

  7. Heat supply from municipal solid waste incineration plants in Japan: Current situation and future challenges.

    Science.gov (United States)

    Tabata, Tomohiro; Tsai, Peii

    2016-02-01

    The use of waste-to-energy technology as part of a municipal solid waste management strategy could reduce the use of fossil fuels and contribute to prevention of global warming. In this study, we examined current heat and electricity production by incineration plants in Japan for external use. Herein, we discuss specific challenges to the promotion of heat utilisation and future municipal solid waste management strategies. We conducted a questionnaire survey to determine the actual conditions of heat production by incineration plants. From the survey results, information of about 498 incineration plants was extracted. When we investigated the relationship between heat production for external use and population density where incineration plants were located, we found that regions with a population density situation. © The Author(s) 2015.

  8. Examination plan for the soils and low-level radioactive waste forms of the NRC field testing lysimeters

    International Nuclear Information System (INIS)

    McConnell, J.W. Jr.; Rogers, R.D.; Jastrow, J.D.

    1996-01-01

    The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste forms. These experiments were recently shut down and have been examined in accordance with a detailed waste form and soil sampling plan. Ion-exchange resins from a commercial nuclear power station were solidified into waste forms using portland cement and vinyl ester-styrene. These waste forms were tested to (a) obtain information on performance of waste forms in typical disposal environments, (b) compare field results with bench leach studies, (c) develop a low-level waste data base for use in performance assessment source term calculations, and (d) apply the DUST computer code to compare predicted cumulative release to actual field data. The program, funded by the Nuclear Regulatory Commission (NRC), includes observed radionuclide releases from waste forms at two test sites over 10 years of successful operation. Lysimeters are ideal systems for obtaining actual field test data because, when properly designed and operated, they can be used to isolate soil and waste systems under actual environmental conditions. The purpose of this paper is to present the experimental plan for the examination of the waste forms and soils of the two lysimeter arrays, which have now been shut down. Vertical soil cores have been taken from the soil columns and will be analyzed with radiochemistry to define movement of radionuclides after release from the waste forms. A comparison is made of the DUST-predicted releases to those previously determined and reported from the lysimeter leachate analyses. That comparison uses new partition coefficients (Kd) recently obtained from laboratory analysis of the lysimeter soils and sand. Those DUST code results also will be compared to actual radionuclide movements through the soil columns as determined from soil core analysis

  9. Ettringite and C-S-H Portland cement phases for waste ion immobilization: A review

    International Nuclear Information System (INIS)

    Gougar, M.L.D.; Scheetz, B.E.; Roy, D.M.

    1996-01-01

    The formation, structure and chemistry of the ettringite and C-S-H phases of Portland cement have been reviewed as they relate to waste ion immobilization. The purpose of this review was to investigate the use of Portland cement as a host for priority metallic pollutants as identified by the Environmental Protection Agency and as a host for radioactive waste ions as identified in 40 CFR 191. Ettringite acts as host to a number of these ions in both the columnar and channel sections of the crystal structure. Substitutions have been made at the calcium, aluminum, hydroxide and sulfate sites. C-S-H also hosts a number of the waste species in both ionic and salt form. Immobilization mechanisms for C-S-H include sorption, phase mixing and substitution. The following ions have not apparently been reported as specifically immobilized by one of these phases: Ag, Am, Np, Pu, Ra, Tc, Th and Sn; however, some of these ions are immobilized by Portland cement

  10. Germany: Management of decommissioning waste in Germany

    International Nuclear Information System (INIS)

    Borrmann, F.; Brennecke, P.; Koch, W.; Kugel, K.; Steyer, S.

    2007-01-01

    Over the past two decades, Germany has gained a substantial amount of experience in the decommissioning of nuclear facilities of different types and sizes. Many research reactors and all prototype nuclear power plants, as well as a few larger nuclear power plants and fuel cycle facilities, are currently at varying stages of decommissioning. Several facilities have been fully dismantled and the sites have been cleared for reuse. The decommissioning projects comprise 18 power and prototype reactors, 33 research reactors and 11 fuel cycle facilities which are being or have been decommissioned. In the future, further nuclear power plants will be shut down and decommissioned in accordance with Germany?s energy policy to phase out the use of nuclear power for commercial electricity generation as given in the April 2002 amendment of the Atomic Energy Act. Radioactive waste, from operations as well as from decommissioning activities, is to be conditioned in such a way as to comply with the waste acceptance requirements of a repository. In Germany, all types of radioactive waste (i.e., short-lived and long-lived) are to be disposed of in deep geological formations. A distinction is being made for heat generating waste (i.e., high level waste) and waste with negligible heat generation (i.e., low level and intermediate level waste). Radioactive decommissioning waste is waste with negligible heat generation. Waste acceptance requirements of a repository are of particular importance for the conditioning of radioactive waste, including decommissioning waste. The waste acceptance requirements, as they resulted from the Konrad licensing procedure, are being applied by the waste generators for the conditioning of decommissioning waste. Compliance with these requirements must be demonstrated through the waste package quality control, even if the waste will be disposed of in the future. In 2002 the Konrad repository was licensed for the disposal of all types of waste with negligible

  11. Ion exchange removal of cesium from simulated and actual supernate from Hanford tanks 241-SY-101 and 241-SY-103

    International Nuclear Information System (INIS)

    Brown, G.N.; Bontha, J.R.; Carlson, C.D.

    1995-09-01

    Pacific Northwest Laboratory (PNL), in conjunction with the Process Chemistry and Statistics Section of Westinghouse Hanford Company (WHC), conducted this study as part of the Supernatant Treatment Development Task for the Initial Pretreatment Module (IPM) Applied Engineering Project. The study assesses the performance of the CS-100 ion exchange material for removing cesium from simulated and actual alkaline supernate from Hanford tanks 241-SY-101 and 241-SY-103. The objective of these experiments is to compare the cesium ion exchange loading and elution profiles of actual and simulated wastes. Specific experimental objectives include (1) demonstration of decontamination factors (DF) for cesium removal, 92) verification of simulant performance, (3) investigation of waste/exchanger chemistry, and (4) determination of the radionuclide content of the regenerated CS-100 resin prior to disposal

  12. The dark face of the present-day society: the waste - waste generation / El lado oscuro de la sociedad actual: los residuos

    Directory of Open Access Journals (Sweden)

    Manuel Díaz Cano

    2009-10-01

    Full Text Available We aim to bring to discussion the waste. We try to speak about the rubbish and not only at the end of the life of the product, even before it being, from the source, including product design, use of these, the consumption, how to get rid of them, the citizens as well as the specialist companies, recycling and reuse, technician, designer and politician. We could not miss the legislation from the different governments; it is one of the most significant parts of this quiz because here come together the different social actors and their proposals but also because it is the way to point it out so the human being, as the at the moment triumphant and the environment as the patient, can survive with great difficulty, and the most important it does not collapse in the short term because nobody want to have in their back yard the waste of the others other because in the long run this waste has a so “long life” that our legacy will be unbearable for the next generations. The society has to intervene without delaying tactics; otherwise this starts soon to smell horrible.

  13. RE-ACTUALIZATION OF BALE BANJAR FUNCTION IN DENPASAR CITY

    Directory of Open Access Journals (Sweden)

    Suryawati P.

    2018-03-01

    Full Text Available The aim of this study was to know and understand the re-actualization of bale banjar function in Denpasar City. This study was conducted due to the imbalance between the assumptions and current reality. The data sources of this study were some bale banjar-s in Denpasar City, related informants, and previous research findings. All data were collected using observation, interviews, and literature studies techniques and were analyzed using sociocultural evolution, community rationalization, and commodification theories. The research findings showed that re-actualization of bale banjar function in Denpasar City happened due to the high demand of economic space of capitalist economy and the changes of local society’s way of thinking; re-actualization of bale banjar’s form in Denpasar City could be seen from the changes in style and appearance of the spaces that are now widely designed into two floors. The society used the upper floors as a gathering place and the lower floors are rented as a center of economic activity; re-actualization of bale banjar function in Denpasar City had the implications on the way of thinking, the conception of bale banjar form and function and material culture such as the weakening of social relations among krama banjar.

  14. Full-scale leaching study of commercial reactor waste forms

    International Nuclear Information System (INIS)

    Kalb, P.D.; Colombo, P.

    1984-01-01

    This paper describes a full-scale leaching experiment which has been conducted at Brookhaven National Laboratory (BNL) to study the release of radionuclides from actual commercial reactor waste forms. While many studies characterizing the leaching behavior of simulated laboratory-scale waste forms have been performed, this program represents one of the first attempts in the United States to quantify activity releases for real, full-scale waste forms. 5 references, 5 figures, 1 table

  15. Final report on cermet high-level waste forms

    International Nuclear Information System (INIS)

    Kobisk, E.H.; Quinby, T.C.; Aaron, W.S.

    1981-08-01

    Cermets are being developed as an alternate method for the fixation of defense and commercial high level radioactive waste in a terminal disposal form. Following initial feasibility assessments of this waste form, consisting of ceramic particles dispersed in an iron-nickel base alloy, significantly improved processing methods were developed. The characterization of cermets has continued through property determinations on samples prepared by various methods from a variety of simulated and actual high-level wastes. This report describes the status of development of the cermet waste form as it has evolved since 1977. 6 tables, 18 figures

  16. Nuclear Waste Vitrification in the U.S.: Recent Developments and Future Options

    International Nuclear Information System (INIS)

    Vienna, John D.

    2010-01-01

    Nuclear power plays a key role in maintaining current world wide energy growth while minimizing the greenhouse gas emissions. A disposition path for used nuclear fuel (UNF) must be found for this technology to achieve its promise. One likely option is the recycling of UNF and immobilization of the high-level waste (HLW) by vitrification. Vitrification is the technology of choice for immobilizing HLW from defense and commercial fuel reprocessing around the world. Recent advances in both recycling technology and vitrification show great promise in closing the nuclear fuel cycle in an efficient and economical fashion. This article summarizes the recent trends developments and future options in waste vitrification for both defense waste cleanup and closing the nuclear fuel cycle in the U.S.

  17. Review of the Scientific Understanding of Radioactive Waste at the U.S. DOE Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Reid A.; Buck, Edgar C.; Chun, Jaehun; Daniel, Richard C.; Herting, Daniel L. [Washington River Protection Solutions, Richland, Washington 99354, United States; Ilton, Eugene S.; Lumetta, Gregg J.; Clark, Sue B. [Chemistry

    2018-01-02

    This paper reviews the origin and chemical and rheological complexity of radioactive waste at the U.S. Department of Energy’s Hanford Site. The waste, stored in underground tanks, was generated via three distinct processes over decades of plutonium extraction operations. Although close records were kept of original waste disposition, tank-to-tank transfers and conditions that impede equilibrium complicate our understanding of the chemistry, phase composition, and rheology of the waste. Tank waste slurries comprise particles and aggregates from nano to micron scales, with varying densities, morphologies, heterogeneous compositions, and complicated responses to flow regimes and process conditions. Further, remnant or changing radiation fields may affect the stability and rheology of the waste. These conditions pose challenges for transport through conduits or pipes to treatment plants for vitrification. Additionally, recalcitrant boehmite degrades glass quality and must be reduced prior to vitrification, but dissolves much more slowly than predicted given surface normalized rates. Existing empirical models based on ex situ experiments and observations lack true predictive capabilities. Recent advances in in situ microscopy, aberration corrected TEM, theoretical modeling across scales, and experimental methods for probing the physics and chemistry at mineral-fluid and mineral-mineral interfaces are being implemented to build robustly predictive physics-based models.

  18. Combustion of Waste Wood. Second phase of the collaboration project on waste wood combustion

    International Nuclear Information System (INIS)

    Andersson, Annika; Andersson, Christer; Eriksson, Jan; Hemstroem, Bengt; Jungstedt, Jenny; Kling, Aasa; Bahr, Bo von; Ekvall, Annika; Eskilsson, David; Tullin, Claes; Harnevie, Henrik; Sieurin, Jan; Keihaes, Juha; Mueller, Christian; Berg, Magnus; Wikman, Karin

    2003-08-01

    Combustion of waste wood has during the last decade increased dramatically and this has resulted in a number of Swedish plants using this fuel, e.g. Handeloe P11 (Norrkoeping) and ldbaecken P3 (Nykoeping), and yet other plants that are under construction (e.g. Nynaeshamn). The experience from these plants are that waste wood combustion results in a number of operational problems. To some extent these problems are different compared with the problems related to combustion of other biofuels but the situation is not directly comparable to waste incinerators. The problems are mainly related to slagging and fouling of heat exchanger surfaces and accelerated corrosion at relatively low temperature compared to the situation for ordinary biofuels. In some cases an increase in the emissions of specific substances can also result in difficulties to fulfil the EC-directive on waste combustion. Within previous projects the main problems related to combustion of waste wood have been identified and to some extent the cause of these problems has been clarified. One result of this reported investigation is a deeper understanding of the actual causes of these problems. However, the most important result is a number of recommendations for different measures on how to achieve disturbance-free combustion of waste wood. These recommendations actually summarises the most important possible solutions on how to achieve a disturbance-free operation and a lower maintenance cost for boilers combusting waste wood and can thereby be regarded as a short summery of the whole project: 1) Improving fuel quality by Improved sorting at the source and Sieving of the fuel -> Reducing the amount of metals and chlorine and Separation of fines and thereby reducing the amount of metals. 2) Combustion modifications by Avoiding reducing conditions at the heat exchanger surfaces -> Minimising slagging, fouling and corrosion. 3) Additives or co-combustion by Addition of sulphur with the fuel; Injection of

  19. Development of inorganic ion exchangers for nuclear waste remediation. 1997 annual progress report

    International Nuclear Information System (INIS)

    Clearfield, A.; Collins, J.L.; Egan, B.Z.

    1997-01-01

    'In this research program, Oak Ridge National Laboratory (ORNL) is collaborating with Texas A and M University in the development of highly selective inorganic ion exchangers for the removal of cesium and strontium from nuclear tank-waste and from groundwater. Inorganic ion exchangers are developed and characterized at Texas A and M University; ORNL is involved in preparing the powders in engineered forms and testing the performance of the sorbents in actual nuclear waste solutions. The Texas A and M studies are divided into two main categories: (1) exchangers for tank wastes and (2) exchangers for groundwater remediation. These are subdivided into exchangers for use in acid and alkaline solutions for tank wastes and those that can be recycled for use in groundwater remediation. The exchangers will also be considered for in situ immobilization of radionuclides. The approach will involve a combination of exchanger synthesis, structural characterization, and ion exchange behavior. ORNL has developed a technique for preparing inorganic ion exchangers in the form of spherules by a gel-sphere internal gelation process. This technology, which was developed and used for making nuclear fuels, has the potential of greatly enhancing the usability of many other special inorganic materials because of the improved flow dynamics of the spherules. Also, pure inorganic spherules can be made without the use of binders. ORNL also has access to actual nuclear waste in the form of waste tank supernatant solutions for testing the capabilities of the sorbents for removing the cesium and strontium radionuclides from actual waste solutions. The ORNL collaboration will involve the preparation of the powdered ion exchangers, developed and synthesized at Texas A and M, in the form of spherules, and evaluating the performance of the exchangers in real nuclear waste solutions. Selected sorbents will be provided by Texas A and M for potential incorporation into microspheres, and the performance

  20. Fourteenth annual U.S. Department of Energy low-level radioactive waste management conference: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1993-03-01

    This document contains 69 papers that were presented at the Fourteenth Annual U.S. Department of Energy Low-Level Radioactive Waste Management Conference, November 18--20, 1992, in Phoenix, Arizona. The papers address the following general topics: (a) Performance Management Track: risk assessment; waste characterization; site characterization; facility design; groundwater modeling; monitoring and modeling; and regulatory requirements; (b) Technical Track: waste minimization; new technologies; international perspectives; licensing issues; hot topics; commercial storage; DOE storage; treatment technologies; and mixed waste; and (c) Institutional Track: status report; changes in orders, regulations, and guidance; regulatory compliance issues; communicating risk; hot topics; and storage impacts. Papers have been processed separately for inclusion on the data base.

  1. ELECTROCHEMICAL CORROSION TESTING OF TANKS 241-AN-102 & 241-AP-107 & 241-AP-108 IN SUPPORT OF ULTRASONIC TESTING

    Energy Technology Data Exchange (ETDEWEB)

    WYRWAS RB; DUNCAN JB

    2008-11-20

    This report presents the results of the corrosion rates that were measured using electrochemical methods for tanks 241-AN-102 (AN-102), 241-AP-107 (AP 107), and 241-AP-108 (AP-108) performed under test plant RPP-PLAN-38215. The steel used as materials of construction for AN and AP tank farms was A537 Class 1. Test coupons of A537 Class 1 carbon steel were used for corrosion testing in the AN-107, AP-107, and AP-108 tank waste. Supernate will be tested from AN-102, AP-107, and Ap-108. Saltcake testing was performed on AP-108 only.

  2. Managing America's solid waste

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, J. A.

    1998-09-15

    This report presents an historical overview of the federal role in municipal solid waste management from 1965 to approximately 1995. Attention is focuses on the federal role in safeguarding public health, protecting the environment, and wisely using material and energy resources. It is hoped that this report will provide important background for future municipal solid waste research and development initiatives.

  3. Waste encapsulation storage facility (WESF) standards/requirements identification document (S/RIDS)

    Energy Technology Data Exchange (ETDEWEB)

    Maddox, B.S., Westinghouse Hanford

    1996-07-29

    This Standards/Requirements Identification Document (S/RID) sets forth the Environmental Safety and Health (ES{ampersand}H) standards/requirements for the Waste Encapsulation Storage Facility (WESF). This S/RID is applicable to the appropriate life cycle phases of design, construction, operation, and preparation for decommissioning. These standards/requirements are adequate to ensure the protection of the health and safety of workers, the public, and the environment.

  4. Strategies for high-level radioactive waste management: the U.S. experience

    International Nuclear Information System (INIS)

    Cotton, T.A.

    1984-01-01

    Technology exists or is under development for the safe, retrievable storage of spent fuel from commercial nuclear reactors and high-level waste from reprocessing that fuel, for many decades, and no insuperable scientific obstacles to permanent disposal of spent fuel or high-level waste in geologic repositories have been identified. However, there are significant institutional obstacles to developing such repositories: strong local opposition to siting and the requirement for a sustained commitment of money and skilled manpower over a period of decades. These create strong incentives to defer the political and economic costs of developing disposal facilities by using less expensive interim storage; yet continued deferral may affect the acceptability of nuclear power. Thus the principal strategic policy issue in high-level waste management is how rapidly to develop disposal facilities. Some countries plan decades of storage before choosing a repository site or a disposal technology, while the United States has enacted a law requiring operation of a geologic repository by 1998. This paper discusses waste management strategic issues and major provisions of the U.S. law, emphasizing those measures dealing with the institutional obstacles to developing geologic repositories. (author)

  5. 18th U.S. Department of Energy Low-Level Radioactive Waste Management Conference. Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    1997-05-20

    This conference explored the latest developments in low-level radioactive waste management through presentations from professionals in both the public and the private sectors and special guests. The conference included two continuing education seminars, a workshop, exhibits, and a tour of Envirocare of Utah, Inc., one of America's three commercial low-level radioactive waste depositories.

  6. Chemical health risk assessment for hazardous and mixed waste management units at Lawrence Livermore National Laboratory

    International Nuclear Information System (INIS)

    1992-09-01

    The Lawrence Livermore National Laboratory (LLNL) operates three Hazardous Waste Management Facilities with 24 associated waste management units for the treatment and storage of hazardous and mixed wastes. These wastes are generated by research programs and support operations. The storage and treatment units are presently operated under interim status in accordance with the requirements of the US Envirorunental Protection Agency (US EPA) and the Department of Toxic Substances Control (DTSC), a division of the California Envirorunental Protection Agency (Cal/EPA). As required by the California Hazardous Waste Control Act and the Resource Conservation and Recovery Act (RCRA), LLNL ha s applied for a Part B permit to continue operating the storage and waste treatment facilities. As part of this permitting process, LLNL is required to conduct a health risk assessment (HRA) to examine the potential health impacts to the surrounding community from continued storage and treatment of hazardous and mixed wastes. analysis document presents the results of this risk assessment. An analysis of maximum credible chemical accidents is also included in Section 7.0. This HRA was prepared in accordance with procedures set forth by the California Air Pollution Control Officers Association (CAPCOA) ''Air Toxics Assessment Manual,'' CAPCOA guidelines for preparing risk assessments under the Air Toxic ''Hot Spots'' Act (AB 2588) and requirements of the US EPA. By following these procedures, this risk assessment presents a conservative analysis of a hypothetical Maximally Exposed Individual (MEI) using many worst-case assumptions that will not apply to an actual individual. As such, the risk estimates presented should be regarded as a worst-case estimate of any actual risk that may be present

  7. Polyethylene solidification of low-level wastes

    International Nuclear Information System (INIS)

    Kalb, P.D.; Colombo, P.

    1985-02-01

    This topical report describes the results of an investigation on the solidification of low-level radioactive waste in polyethylene. Waste streams selected for this study included those which result from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Four types of commercially available low-density polyethylenes were employed which encompass a range of processing and property characteristics. Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste and polyethylene type. Property evaluation testing was performed on laboratory-scale specimens to assess the potential behavior of actual waste forms in a disposal environment. Waste form property tests included water immersion, deformation under compressive load, thermal cycling and radionuclide leaching. Recommended waste loadings of 70 wt % sodium sulfate, 50 wt % boric acid, 40 wt % incinerator ash, and 30 wt % ion exchange resins, which are based on process control and waste form performance considerations are reported. 37 refs., 33 figs., 22 tabs

  8. STS-55 crewmembers repair waste water tank under OV-102's middeck subfloor

    Science.gov (United States)

    1993-01-01

    STS-55 Pilot Terence T. Henricks uses a spotlight and pen to point out a possible problem area on a waste water tank in the bilge area below Columbia's, Orbiter Vehicle (OV) 102's, middeck. Mission Specialist 1 (MS1) and Payload Commander (PLC) Jerry L. Ross records the activity with a video camcorder. The crewmembers are participating in an inflight maintenance (IFM) exercise to counter problems experienced with the waste water tank.

  9. ANSTO`s radioactive waste management policy. Preliminary environmental review

    Energy Technology Data Exchange (ETDEWEB)

    Levins, D.M.; Airey, P.; Breadner, B.; Bull, P.; Camilleri, A.; Dimitrovski, L.; Gorman, T.; Harries, J.; Innes, R.; Jarquin, E.; Jay, G.; Ridal, A.; Smith, A.

    1996-05-01

    For over forty years, radioactive wastes have been generated by ANSTO (and its predecessor, the AAEC) from the operation of nuclear facilities, the production of radioisotopes for medical and industrial use, and from various research activities. the quantities and activities of radioactive waste currently at Lucas Heights are very small compared to many other nuclear facilities overseas, especially those in countries with nuclear power program. Nevertheless, in the absence of a repository for nuclear wastes in Australia and guidelines for waste conditioning, the waste inventory has been growing steadily. This report reviews the status of radioactive waste management at ANSTO, including spent fuel management, treatment of effluents and environmental monitoring. It gives details of: relevant legislative, regulatory and related requirements; sources and types of radioactive waste generated at ANSTO; waste quantities and activities (both cumulative and annual arisings); existing practices and procedures for waste management and environmental monitoring; recommended broad strategies for dealing with radioactive waste management issues. Detailed proposals on how the recommendations should be implemented is the subject of a companion internal document, the Radioactive Waste Management Action Plan 1996-2000 which provides details of the tasks to be undertaken, milestones and resource requirements. 44 refs., 2 tabs., 18 figs.

  10. SKB`s planning of the EIA in connection with the final disposal of nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Thegerstroem, C.; Forsstroem, H. [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden)

    1995-12-01

    The plans for the final disposal of Swedish nuclear waste are summarized. The legal requirements on Environmental Impact Statements (EIS) and their role in the program for the final disposal of nuclear waste are described. SKB`s view of the purpose of the Environmental Impact Assessment is described in the light of the experience which now exists from the work on an encapsulation facility and a deep repository. In order to obtain an adequate basis for decision-making, the EIS is of central importance. In SKB`s view, with regard to the final disposal of nuclear waste in Sweden, there is a very good possibility of fulfilling the requirements on the EIS which should be made within modern environmental protection work. 8 refs, 5 figs.

  11. Emissions from small-scale burns of simulated deployed U.S. military waste.

    Science.gov (United States)

    Woodall, Brian D; Yamamoto, Dirk P; Gullett, Brian K; Touati, Abderrahmane

    2012-10-16

    U.S. military forces have historically relied on open burning as an expedient method of volume reduction and treatment of solid waste during the conflicts in Afghanistan and Iraq. This study is the first effort to characterize a broad range of pollutants and their emission factors during the burning of military waste and the effects that recycling efforts, namely removing plastics, might have on emissions. Piles of simulated military waste were constructed, burned, and emissions sampled at the U.S. Environmental Protection Agency (EPA) Open Burn Testing Facility (OBTF), Research Triangle Park, NC. Three tests contained polyethylene terephthalate (PET #1 or PET) plastic water bottles and four did not. Emission factors for polycyclic aromatic hydrocarbons (PAHs), volatile organic compounds (VOCs), particulate matter (PM(10), PM(2.5)), polychlorinated and polybrominated dioxins/furans (PCDD/F and PBDD/F), and criteria pollutants were determined and are contained within. The average PCDD/F emission factors were 270 ng-toxic equivalency (TEQ) per kg carbon burned (ng-TEQ/kg Cb), ranging from 35 to 780 ng-TEQ/kg Cb. Limited testing suggests that targeted removal of plastic water bottles has no apparent effect on reducing pollutants and may even promote increased emissions.

  12. Materials Discarded in the U.S. Municipal Waste Stream, 1960 to 2009 (in tons)

    Data.gov (United States)

    U.S. Environmental Protection Agency — The U.S. Environmental Protection Agency (EPA) has collected and reported data on the generation and disposal of waste in the United States for more than 30 years....

  13. 76 FR 17970 - Board Meeting: April 27, 2011-Amherst, New York; the U.S. Nuclear Waste Technical Review Board...

    Science.gov (United States)

    2011-03-31

    ... the 2008-9 study on Quantitative Risk Assessment of the State Licensed Radioactive Waste Disposal Area... of vitrified high-level radioactive waste (HLW); determination of waste classification of the melter... NUCLEAR WASTE TECHNICAL REVIEW BOARD Board Meeting: April 27, 2011--Amherst, New York; the U.S...

  14. Retrieval process development and enhancements FY96 pulsed-air mixer testing and deployment study

    International Nuclear Information System (INIS)

    Powell, M.R.; Hymas, C.R.

    1996-08-01

    Millions of gallons of radioactive wastes resides in underground tanks at US Department of Energy sites. The waste was generated primarily by the processing of nuclear fuel elements to remove fissile radionuclides for use in atomic weapons. Plans call for the waste to be removed from the tanks and processed to create immobile waste forms, which will be stored to prevent release to the environment. The consistency of the waste ranges from liquid, to slurry, to sticky sludge, to hard saltcake. a variety of waste- retrieval and processing methods are being evaluated and implemented. One such method is pulsed-air mixing, which is the subject of this report. Pulsed-air mixing equipment has been successfully applied to a number of difficult mixing applications in various chemical-process industries. Most previous applications involved the mixing of particle-free viscous fluids. The study described in this report was preformed to improve the understanding of how pulsed-air mixing applies to slurries. This document describes work conducted to evaluate the potential application of pulsed-air mixers to the slurry- mixing needs of the US Department of Energy's waste-retrieval programs

  15. The regulation on commercial reactors and the management of high-level radioactive wastes in U.S

    International Nuclear Information System (INIS)

    Shimomura, Hidetsugu

    2013-01-01

    This article shows U.S. NRC's substantial and procedural regulations regarding commercial reactors and radioactive wastes. The commercial reactor's regulations are analyzed from an ensuring safety, and the radioactive waste' management is done from a locating a disposal site. (author)

  16. Engineering-scale vitrification of commercial high-level waste

    International Nuclear Information System (INIS)

    Bonner, W.F.; Bjorklund, W.J.; Hanson, M.S.; Knowlton, D.E.

    1980-04-01

    To date, technology for immobilizing commercial high-level waste (HLW) has been extensively developed, and two major demonstration projects have been completed, the Waste Solidification Engineering Prototypes (WSEP) Program and the Nuclear Waste Vitrification Project (NWVP). The feasibility of radioactive waste solidification was demonstrated in the WSEP program between 1966 and 1970 (McElroy et al. 1972) using simulated power-reactor waste composed of nonradioactive chemicals and HLW from spent, Hanford reactor fuel. Thirty-three engineering-scale canisters of solidified HLW were produced during the operations. In early 79, the NWVP demonstrated the vitrification of HLW from the processing of actual commercial nuclear fuel. This program consisted of two parts, (1) waste preparation and (2) vitrification by spray calcination and in-can melting. This report presents results from the NWVP

  17. Networks of recyclable material waste-picker’s cooperatives: An alternative for the solid waste management in the city of Rio de Janeiro

    International Nuclear Information System (INIS)

    Tirado-Soto, Magda Martina; Zamberlan, Fabio Luiz

    2013-01-01

    Highlights: ► In the marketing of recyclable materials, the waste-pickers are the least wins. ► It is proposed creating a network of recycling cooperatives to achieve viability. ► The waste-pickers contribute to waste management to the city. - Abstract: The objective of this study is to discuss the role of networks formed of waste-picker cooperatives in ameliorating problems of final disposal of solid waste in the city of Rio de Janeiro, since the city’s main landfill will soon have to close because of exhausted capacity. However, it is estimated that in the city of Rio de Janeiro there are around five thousand waste-pickers working in poor conditions, with lack of physical infrastructure and training, but contributing significantly by diverting solid waste from landfills. According to the Sustainable Development Indicators (IBGE, 2010a,b) in Brazil, recycling rates hover between 45% and 55%. In the municipality of Rio de Janeiro, only 1% of the waste produced is collected selectively by the government (COMLURB, 2010), demonstrating that recycling is mainly performed by waste-pickers. Furthermore, since the recycling market is an oligopsony that requires economies of scale to negotiate directly with industries, the idea of working in networks of cooperatives meets the demands for joint marketing of recyclable materials. Thus, this work presents a method for creating and structuring a network of recycling cooperatives, with prior training for working in networks, so that the expected synergies and joint efforts can lead to concrete results. We intend to demonstrate that it is first essential to strengthen the waste-pickers’ cooperatives in terms of infrastructure, governance and training so that solid waste management can be environmentally, socially and economically sustainable in the city of Rio de Janeiro

  18. Networks of recyclable material waste-picker’s cooperatives: An alternative for the solid waste management in the city of Rio de Janeiro

    Energy Technology Data Exchange (ETDEWEB)

    Tirado-Soto, Magda Martina, E-mail: magda@pep.ufrj.br [Program of Production Engineering, School and Research in Engineering, Federal University of Rio de Janeiro (Brazil); Zamberlan, Fabio Luiz, E-mail: fabio@pep.ufrj.br [Program of Production Engineering, School and Research in Engineering, Federal University of Rio de Janeiro (Brazil)

    2013-04-15

    Highlights: ► In the marketing of recyclable materials, the waste-pickers are the least wins. ► It is proposed creating a network of recycling cooperatives to achieve viability. ► The waste-pickers contribute to waste management to the city. - Abstract: The objective of this study is to discuss the role of networks formed of waste-picker cooperatives in ameliorating problems of final disposal of solid waste in the city of Rio de Janeiro, since the city’s main landfill will soon have to close because of exhausted capacity. However, it is estimated that in the city of Rio de Janeiro there are around five thousand waste-pickers working in poor conditions, with lack of physical infrastructure and training, but contributing significantly by diverting solid waste from landfills. According to the Sustainable Development Indicators (IBGE, 2010a,b) in Brazil, recycling rates hover between 45% and 55%. In the municipality of Rio de Janeiro, only 1% of the waste produced is collected selectively by the government (COMLURB, 2010), demonstrating that recycling is mainly performed by waste-pickers. Furthermore, since the recycling market is an oligopsony that requires economies of scale to negotiate directly with industries, the idea of working in networks of cooperatives meets the demands for joint marketing of recyclable materials. Thus, this work presents a method for creating and structuring a network of recycling cooperatives, with prior training for working in networks, so that the expected synergies and joint efforts can lead to concrete results. We intend to demonstrate that it is first essential to strengthen the waste-pickers’ cooperatives in terms of infrastructure, governance and training so that solid waste management can be environmentally, socially and economically sustainable in the city of Rio de Janeiro.

  19. The Role of Perceived and Actual Motor Competency on Children's Physical Activity and Cardiorespiratory Fitness during Middle Childhood

    Science.gov (United States)

    Gu, Xiangli; Thomas, Katherine Thomas; Chen, Yu-Lin

    2017-01-01

    Purpose: Guided by Stodden et al.'s (2008) conceptual model, the purpose of this study was to examine the associations among perceived competence, actual motor competence (MC), physical activity (PA), and cardiorespiratory fitness in elementary children. The group differences were also investigated as a function of MC levels. Methods: A…

  20. A New and Sound Technology for Biogas from Solid Waste and Biomass

    International Nuclear Information System (INIS)

    Busch, G.; Grossmann, J.; Sieber, M.; Burkhardt, M.

    2009-01-01

    Organic waste, as a main constituent of municipal solid waste, has as well as solid biomass a high potential for biogas generation. Despite the importance of biogas generation from these materials, the availability of large-scale biogas processes lacks behind the demand. A newly developed double-stage solid-liquid biogas process, consisting of an open hydrolysis stage and a fixed-bed methane reactor, allows the biogas production from almost all biodegradable solid waste and renewable resources like maize, grass, sugar cane, etc. Furthermore, residues from industrial processes, like the glycerine waste water from biodiesel production, can also be converted into biogas successfully. Due to the strong separation of hydrolysis and methanation, the process is extremely stable. No malfunction has been detected so far. The open hydrolysis releases CO 2 and allows oxidation of sulfur. Consequently, the biogas has a high methane (>72%) and low H 2 S concentration (<100 ppm). Stirrers or other agitation equipment are not necessary; only liquids are pumped. The biogas generation becomes controllable for the first time; thus, the actual generation can be easily adapted to the consumption

  1. Polyethylene encapsulation of mixed wastes: Scale-up feasibility

    International Nuclear Information System (INIS)

    Kalb, P.D.; Heiser, J.H.; Colombo, P.

    1991-01-01

    A polyethylene process for the improved encapsulation of radioactive, hazardous, and mixed wastes have been developed at Brookhaven National Laboratory (BNL). Improvements in waste loading and waste form performance have been demonstrated through bench-scale development and testing. Maximum waste loadings of up to 70 dry wt % mixed waste nitrate salt were achieved, compared with 13--20 dry wt % using conventional cement processes. Stability under anticipated storage and disposal conditions and compliance with applicable hazardous waste regulations were demonstrated through a series of lab-scale waste form performance tests. Full-scale demonstration of this process using actual or surrogate waste is currently planned. A scale-up feasibility test was successfully conducted, demonstrating the ability to process nitrate salts at production rates (up to 450 kg/hr) and the close agreement between bench- and full-scale process parameters. Cored samples from the resulting pilot-scale (114 liter) waste form were used to verify homogeneity and to provide additional specimens for confirmatory performance testing

  2. Low and medium radioactive waste management in Belgium

    International Nuclear Information System (INIS)

    Dejonghe, P.; Bonne, A.; Van de Voorde, N.; Detilleux, E.

    1986-01-01

    This paper summarizes some information on the producers and production of radioactive wastes in Belgium, the evolution of the technical thinking in this area, how the management of radioactive wastes has been and is being organized in the country, and some practical achievements. This paper does not contain much information on new methods and processes and is based primarily on actual experience. Successively the following subjects are discussed: the waste producers; the main objectives in treatment/conditioning and disposal; organization; treatment and conditioning technology; and storage and disposal

  3. Activity monitoring of alpha-bearing wastes

    International Nuclear Information System (INIS)

    Birkhoff, G.; Bondar, L.

    1980-01-01

    The paper aims at the survey on the actual situation in activity monitoring of alpha-bearing wastes. Homogeneous materials such as liquid-, gaseous- and homogeneous solid wastes are amenable to destructive analyses of representative samples. Available destructive analyses methods are sensitive and precise enough to cope with all requirements in alpha-waste monitoring. The more difficult problems are encountered with alpha-contaminated solids, when representative sampling is not practicable. Non-destructive analysis techniques are applied for monitoring this category of solid wastes. The techniques for nondestructive analysis of alpha-bearing wastes are based on the detection of gamma and/or neutron-emission of actinides. Principles and a theory of non-destructive radiometric assay of plutonium contaminated solid waste streams are explained. Guidelines for the calibration of instruments and interpretation of experimental data are given. Current theoretical and experimental development work in this problem area is reviewed. Evaluations concerning capabilities and limitations of monitoring systems for alpha-bearing solid wastes are very complex and out of the scope of this paper

  4. Waste management research abstracts. Information on radioactive waste management research in progress or planned. Vol. 29

    International Nuclear Information System (INIS)

    2004-11-01

    The twenty-ninth issue of the Waste Management Research Abstracts (WMRA) contains 96 abstracts that describe research in progress in the field of radioactive waste management. These abstracts were collected between May 1 and October 15, 2004 and present ongoing work in Brazil(1), Finland (1), Germany (7), India (11), Mauritius (1), republic of Korea (1), Russian Federation (1) and the United States of America (70). Although the abstracts are indexed by country, some programmes are actually the result of co-operation among several countries. Indeed, a primary reason for providing this compilation of programmes, institutions and scientists engaged in research into radioactive waste management is to increase international co-operation and facilitate communications

  5. New Innovations in Highly Ion Specific Media for Recalcitrant Waste stream Radioisotopes

    International Nuclear Information System (INIS)

    Denton, M. S.; Wilson, J.; Ahrendt, M.; Bostick, W. D.; DeSilva, F.; Meyers, P.

    2006-01-01

    Specialty ion specific media were examined and developed for, not only pre- and post-outage waste streams, but also for very difficult outage waste streams. This work was carried out on first surrogate waste streams, then laboratory samples of actual waste streams, and, finally, actual on-site waste streams. This study was particularly focused on PWR wastewaters such as Floor Drain Tank (FDT), Boron Waste Storage Tank (BWST), and Waste Treatment Tank (WTT, or discharge tank). Over the last half decade, or so, treatment technologies have so greatly improved and discharge levels have become so low, that certain particularly problematic isotopes, recalcitrant to current treatment skids, are all that remain prior to discharge. In reality, they have always been present, but overshadowed by the more prevalent and higher activity isotopes. Such recalcitrants include cobalt, especially Co 58 [both ionic/soluble (total dissolved solids, TDS) and colloidal (total suspended solids, TSS)] and antimony (Sb). The former is present in most FDT and BWST wastewaters, while the Sb is primarily present in BWST waste streams. The reasons Co 58 can be elusive to granulated activated carbon (GAC), ultrafiltration (UF) and ion exchange (IX) demineralizers is that it forms submicron colloids as well as has a tendency to form metal complexes with chelating agents (e.g., ethylene diamine tetraacetic acid, or EDTA). Such colloids and non-charged complexes will pass through the entire treatment skid. Antimony (Sb) on the other hand, has little or no ionic charge, and will, likewise, pass through both the filtration and de-min skids into the discharge tanks. While the latter will sometimes (the anionic vs. the cationic or neutral species) be removed on the anion bed(s), it will slough off (snow-plow effect) when a higher affinity anion (iodine slugs, etc.) comes along; thus causing effluents not meeting discharge criteria. The answer to these problems found in this study, during an actual

  6. 7th Expert meeting radiation protection. International developments, waste management, challenges for the radiation protection in aging nuclear installations

    International Nuclear Information System (INIS)

    2010-01-01

    The proceedings of the 7th Expert meeting on radiation protection include contributions to the following topics: nuclear power and public opinion, IAEA safety standards, ISOE - information system on occupational exposure, European harmonization of the radiation protection education, WANO - challenges and results, CTBTO's global radiation measurement network, state of final radioactive waste disposal in Germany and worldwide, radioactive waste management and disposal in French NPPs, preparedness for final waste disposal in Schacht Konrad, actualization of the transport study Konrad, transport of NPPs' operational radioactive waste and waste from decommissioned reactor demolition to the final repository Konrad, qualification of radioactive waste casks for the final repository Konrad, radioactive waste disposal management concept in Switzerland, aging management and radiation protection, decontamination as effective measure for dose rate reduction - long-term and sustainable dose rate reduction by primary circuit decontamination, system and component decontamination for individual and collective dose reduction - practical examples, radiation protection map - electronic assistance for work planning, EPR dismantling already today? radiation protection register 2002-2010 - knowledge based on a decade of radiation monitoring, actual information on radiation protection in medicine, mobile telecommunication - actual research results.

  7. Risk assessment for the transportation of hazardous waste and hazardous waste components of low-level mixed waste and transuranic waste for the U.S. Department of Energy waste management programmatic environmental impact statement

    International Nuclear Information System (INIS)

    Lazaro, M.A.; Policastro, A.J.; Hartmann, H.M.

    1995-04-01

    This report, a supplement to Appendix E (Transportation Risk) of the U.S. Department of Energy Waste Management Programmatic Environmental Impact Statement (WM PEIS), provides additional information supporting the accident data for chemical risk assessment and health risk methodology described in that appendix (Part II), as well as providing the uncertainty analysis and on-site risk calculations. This report focuses on hazardous material truck accident rates, release probabilities, and release quantities; provides the toxicological values derived for each hazardous chemical assessed in the WM PEIS and further details on the derivation of health criteria; describes the method used in the transportation risk assessments to address potential additivity of health effects from simultaneous exposure to several chemicals and the method used to address transportation risks for maximally exposed individuals; presents an expanded discussion of the uncertainty associated with transportation risk calculations; and includes the results of the on-site transportation risk analysis. In addition, two addenda are provided to detail the risk assessments conducted for the hazardous components of low-level mixed waste (Addendum I) and transuranic waste (Addendum II)

  8. 10 CFR 51.52 - Environmental effects of transportation of fuel and waste-Table S-4.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Environmental effects of transportation of fuel and waste... Environmental effects of transportation of fuel and waste—Table S-4. Under § 51.50, every environmental report... detailed analysis of the environmental effects of transportation of fuel and wastes to and from the reactor...

  9. PWR Users Group 10 CFR 61 Waste Form Requirements Compliance Test Program

    International Nuclear Information System (INIS)

    Rosenlof, R.C.

    1985-01-01

    In January of 1984, a PWR Users Group was formed to initiate a 10 CFR 61 Waste Form Requirements Compliance Test Program on a shared cost basis. The original Radwaste Solidification Systems sold by ATCOR ENGINEERED SYSTEMS, INC. to the utilities were required to produce a free-standing monolith with no free water. None of the other requirements of 10 CFR 61 had to be met. Current regulations, however, have substantially expanded the scope of the waste form acceptance criteria. These new criteria required that generators of radioactive waste demonstrate the ability to produce waste forms which meet certain chemical and physical requirements. This paper will present the test program used and the results obtained to insure 10 CFR 61 compliance of the three (3) typical waste streams generated by the ATCOR PWR Users Group's plants. The primary objective of the PWR Users Group was not to maximize waste loading within the masonry cement solidification media, but to insure that the users Radwaste Solidification System is capable of producing waste forms which meet the waste form criteria of 10 CFR 61. A description of the laboratory small sample certification program and the actual full scale pilot plant verification approach used is included in this paper. Also included is a discussion of the development of a Process Control Program to ensure the reproducibility of the test results with actual waste

  10. 2016 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site’s Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond

    Energy Technology Data Exchange (ETDEWEB)

    Cafferty, Kara Grace [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-02-01

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (WRU-I-0160-01, Modification 1, formerly LA 000160 01), for the wastewater reuse site at the Idaho National Laboratory Site’s Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond from November 1, 2015, through October 31, 2016.

  11. Processing and solidification of Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Kelley, J.A.

    1981-01-01

    The entire flowsheet for processing and solidification of Savannah River Plant (SRP) high-level wastes has been demonstrated. A new small-scale integrated pilot plant is operating with actual radioactive wastes, and large-scale equipment is being demonstrated with nonradioactive simulated wastes. Design of a full-scale waste solidification plant is in progress. Plant construction is expected to begin in 1983, and startup is anticipated in 1988. The plant will poduce about 500 cans of glass per year with each can containing about 1.5 tons of glass

  12. Answers to questions posed by the Michigan Governor's Nuclear Waste Disposal Task Force

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-30

    A general presentation of the National Waste Terminal Storage (NWTS) Program was given on July 26, 1976, to the Michigan Environmental Review Board and the Michigan Governor's Nuclear Waste Disposal Task Force. Following the presentation, Dr. William G. Taylor, Chairman of the Task Force, provided ERDA with a listing of questions which pertained to the NWTS program and ERDA/OWI's interest in northeast Michigan. This document contains copies of the information which was provided to Dr. Taylor in response to his inquiry.

  13. Separating and stabilizing phosphate from high-level radioactive waste: process development and spectroscopic monitoring.

    Science.gov (United States)

    Lumetta, Gregg J; Braley, Jenifer C; Peterson, James M; Bryan, Samuel A; Levitskaia, Tatiana G

    2012-06-05

    Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

  14. Nuclear waste: The problem that won't go away

    International Nuclear Information System (INIS)

    Lenssen, N.

    1991-01-01

    This book presents an overview of the problems of permanent and safe disposal of nuclear waste. The introduction has a brief history of the politics of nuclear waste. Major sections of the book include the following: permanent hazards of nuclear waste, including examples and the politics; health and radiation (history of recommended dosages, health risks, and problems of environmental transport are included); They call it disposal talks about technical options for dealing with nuclear waste, the actual number of sites in different countries, and the inadequacies of scientific knowledge in this area; Technical Fixes? Includes a discussion of other suggested ways of handling nuclear waste; The politics of nuclear waste and beyond illusion conclude the book. 105 refs., 5 tabs

  15. A NEW WASTE CLASSIFYING MODEL: HOW WASTE CLASSIFICATION CAN BECOME MORE OBJECTIVE?

    Directory of Open Access Journals (Sweden)

    Burcea Stefan Gabriel

    2015-07-01

    documents available in the virtual space, on the websites of certain international organizations involved in the wide and complex issue of waste management. The second part of the paper contains a proposal classification model with four main criteria in order to make waste classification a more objective process. The new classification model has the main role of transforming the traditional patterns of waste classification into an objective waste classification system and a second role of eliminating the strong contextuality of the actual waste classification models.

  16. Should the U.S. proceed to consider licensing deep geological disposal of high-level nuclear waste

    International Nuclear Information System (INIS)

    Curtiss, J.R.

    1993-01-01

    The United States, as well as other countries facing the question of how to handle high-level nuclear waste, has decided that the most appropriate means of disposal is in a deep geologic repository. In recent years, the Radioactive Waste Management Committee of the Nuclear Energy Agency has developed several position papers on the technical achievability of deep geologic disposal, thus demonstrating the serious consideration of deep geologic disposal in the international community. The Committee has not, as yet, formally endorsed disposal in a deep geologic repository as the preferred method of handling high-level nuclear waste. The United States, on the other hand, has studied the various methods of disposing of high-level nuclear waste, and has determined that deep geologic disposal is the method that should be developed. The purpose of this paper is to present a review of the United States' decision on selecting deep geologic disposal as the preferred method of addressing the high-level waste problem. It presents a short history of the steps taken by the U.S. in determining what method to use, discusses the NRC's waste Confidence Decision, and provides information on other issues in the U.S. program such as reconsideration of the final disposal standard and the growing inventory of spent fuel in storage

  17. Human factors, system safety, and systems engineering in the transportation of U.S. high-level waste

    International Nuclear Information System (INIS)

    Price, D.L.; Chu, S.C.

    1993-01-01

    The U.S. Nuclear Waste Technical Review Board is an independent agency charged with evaluating the technical and scientific validity of the U.S. Department of Energy's program to manage the disposal of spent fuel and defense high-level waste. The Board has continued to emphasize the importance of using a true system approach in designing the waste management system. The Board has recommended the application of basic design disciplines such as human factors, system safety, and systems engineering. A top-level system study needs to be undertaken that focuses on minimizing handling. The analysis must be well done, in a timely manner, and without the inclusion in the analysis of arbitrary and artificial constraints. (author)

  18. The systematic roles of SKI and SSI in the Swedish nuclear waste management system. Syncho`s report for project RISCOM

    Energy Technology Data Exchange (ETDEWEB)

    Espejo, R. [Syncho, Solihull (United Kingdom); Gill, A. [Syncho, Oxon (United Kingdom)

    1998-01-01

    The purpose of this report is to share and summarize our findings about the regulatory roles of SKI/SSI in the context of the Swedish Nuclear System (SNS), with an emphasis on nuclear waste management. The driving force in this review is to make decision processes more transparent. What is reported is based on interviews conducted with employees at SKI/SSI/SKB during early December 1996, the presentation to SKI/SSI in January 1997, discussions during the Shap Wells meeting in Cumbria during March 1997 and RISCOM internal discussions. We offer two hypotheses about the way the Nuclear Waste Management System (NWMS) appears to work. We choose one and derive from it a view about structural issues in SNS and NWMS. The conclusion is a set of systemic roles for the regulators. It is the comparison between these systemic roles and the actual situation that may trigger some adjustments in the system. Our hope is that these findings will make apparent feasible and desirable changes in the system in order to increase the chances for transparent decisions in the Nuclear Waste Management System. In summary, Section 2 includes a general background of the NWMS based on interviews and general information. Section 3 makes a more focused attempt to work out the issues expressed by people in the interviews. Section 4 discusses at a more conceptual level systemic ideas such as the unfolding of complexity. Section 5 is an attempt to organize viewpoints about the NWMS and offers hypotheses to support a preliminary diagnosis of the system in Section 6. We call this section `A problem of identity`. It is only in Section 7 that basic systemic arguments are unfolded with the intention of supporting an appreciation of SKI/SSI`s regulatory roles in the nuclear industry as a whole and nuclear waste management in particular. Section 8 offers a summary of conclusions.

  19. RCRA Assessment Plan for Single-Shell Tank Waste Management Area S-SX at the Hanford Site

    International Nuclear Information System (INIS)

    Chou, C.J.; Johnson, V.G.

    1999-01-01

    A groundwater quality assessment plan was prepared for waste management area S-SX at the Hanford Site. Groundwater monitoring is conducted at this facility in accordance with Title 40, Code of Federal Regulation (CFR) Part 265, Subpart F [and by reference of Washington Administrative Code (WAC) 173-303-400(3)]. The facility was placed in assessment groundwater monitoring program status after elevated waste constituents and indicator parameter measurements (i.e., chromium, technetium-99 and specific conductance) in downgradient monitoring wells were observed and confirmed. A first determination, as allowed under 40 CFR 265.93(d), provides the owner/operator of a facility an opportunity to demonstrate that the regulated unit is not the source of groundwater contamination. Based on results of the first determination it was concluded that multiple source locations in the waste management area could account for observed spatial and temporal groundwater contamination patterns. Consequently, a continued investigation is required. This plan, developed using the data quality objectives process, is intended to comply with the continued investigation requirement. Accordingly, the primary purpose of the present plan is to determine the rate and extent of dangerous waste (hexavalent chromium and nitrate) and radioactive constituents (e.g., technetium-99) in groundwater and to determine their concentrations in groundwater beneath waste management area S-SX. Comments and concerns expressed by the Washington State Department of Ecology on the initial waste management area S-SX assessment report were addressed in the descriptive narrative of this plan as well as in the planned activities. Comment disposition is documented in a separate addendum to this plan

  20. Complex-wide review of DOE`s management of low-level radioactive waste - progress to date

    Energy Technology Data Exchange (ETDEWEB)

    Letourneau, M.J.

    1995-12-31

    The Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 94-2 includes a recommendation that the Department of Energy (DOE) conduct a comprehensive, complex-wide review of the low-level waste issue to establish the dimensions of the low-level waste problem and to identify necessary corrective actions to address the safe disposition of past, present, and future volumes. DOE`s Implementation Plan calls for the conduct of a complex-wide review of low-level radioactive waste treatment, storage, and disposal sites to identify environmental, safety, and health vulnerabilities. The complex-wide review focuses on low-level waste disposal facilities through a site evaluation survey, reviews of existing documentation, and onsite observations. Low-level waste treatment and storage facilities will be assessed for their ability to meet waste acceptance criteria for disposal. Results from the complex-wide review will be used to form the basis for an integrated and planned set of actions to correct the identified vulnerabilities and to prompt development of new requirements for managing low-level waste.

  1. Progress report on safety research of high-level waste management for the period April 1987 to March 1988

    International Nuclear Information System (INIS)

    Nakamura, Haruto; Tashiro, Shingo

    1988-10-01

    Researches on high-level waste management at the High Level Waste Management Laboratory and the Waste Safety Testing Facility Operation Division of the Japan Atomic Energy Research Institute in the fiscal year of 1987 are reviewed in the three sections of the report. The topics are as follows: 1) On performance and durability of waste forms and engineered barrier materials, accelerated alpha radiation stability of glass form and Synroc has been investigated and stress corrosion cracking of canister materials was examined under simulated conditions. 2) Sorption of 237 Np on granite samples and behavior of iron during weathering of granites were studied with respect to safety evaluation for geological disposal. 3) Actual waste was transported from the Tokai Reprocessing Plant and hot operation using the actual waste was initiated at WASTEF. (author)

  2. Demonstration of the TRUEX process for the treatment of actual high activity tank waste at the INEEL using centrifugal contactors

    International Nuclear Information System (INIS)

    Law, J.D.; Brewer, K.N.; Todd, T.A.; Olson, L.G.

    1997-01-01

    The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering and Environmental Laboratory (INEEL), formerly reprocessed spent nuclear fuel to recover fissionable uranium. The radioactive raffinates from the solvent extraction uranium recovery processes were converted to granular solids (calcine) in a high temperature fluidized bed. A secondary liquid waste stream was generated during the course of reprocessing, primarily from equipment decontamination between campaigns and solvent wash activities. This acidic tank waste cannot be directly calcined due to the high sodium content and has historically been blended with reprocessing raffinates or non-radioactive aluminum nitrate prior to calcination. Fuel reprocessing activities are no longer being performed at the ICPP, thereby eliminating the option of waste blending to deplete the waste inventory. Currently, approximately 5.7 million liters of high-activity waste are temporarily stored at the ICPP in large underground stainless-steel tanks. The United States Environmental Protection Agency and the Idaho Department of Health and Welfare filed a Notice of Noncompliance in 1992 contending some of the underground waste storage tanks do not meet secondary containment. As part of a 1995 agreement between the State of Idaho, the Department of Energy, and the Department of Navy, the waste must be removed from the tanks by 2012. Treatment of the tank waste inventories by partitioning the radionuclides and immobilizing the resulting high-activity and low-activity waste streams is currently under evaluation. A recent peer review identified the most promising radionuclide separation technologies for evaluation. The Transuranic Extraction-(TRUEX) process was identified as a primary candidate for separation of the actinides from ICPP tank waste

  3. Alternatives to High-Level Waste Vitrification: The Need for Common Sense

    International Nuclear Information System (INIS)

    Bell, Jimmy T.

    2000-01-01

    The competition for government funding for remediation of defense wastes (and for other legitimate government functions) is intensifying as the United States moves toward a balanced national budget. Determining waste remediation priorities for the use of available tax dollars will likely depend on established international agreements and on the real risks posed to human health.Remediation of the U.S. Department of Energy (DOE) high-level radioactive tank wastes has been described as the most important priority in the DOE system. The proposed tank waste remediation at three DOE sites will include retrieval of the wastes from the aging storage tanks, immobilization of the wastes, and safe disposal of the processed waste. Vitrification, the current immobilization technology chosen by DOE, is very costly. The U.S. Congress and the American people may not be aware that the present cost of preparing just 1 m 3 of processed waste product at the Savannah River Site is ∼$2 million! In a smaller waste remediation project at the West Valley Site, similar waste treatment is costing >$2 million/m 3 of waste product. Privatization efforts at the Hanford Site are now estimated to cost >$4 million/m 3 of waste product. Even at the lowest current cost of $2 million/m 3 of HLW glass product, the total estimated costs for remediating the tank wastes at the three DOE sites of Savannah River, Hanford, and Idaho Falls is $75 billion.Whether our nation can afford treatment costs of this magnitude and whether Congress will be willing to appropriate these huge sums for waste vitrification when alternative technologies can provide safe disposal at considerably lower cost are questions that need to be addressed. The hazard levels posed by the DOE tank wastes do not warrant high priority in comparison to the hazards of other defense wastes. Unless DOE selects a lower-cost technology for tank waste remediation, such efforts are likely to continue in a holding pattern, with little actually

  4. [ACTUAL PROBLEMS OF HYGIENE SCIENCE AND PRACTICE IN THE PRESERVATION OF PUBLIC HEALTH].

    Science.gov (United States)

    Onishchenko, G G

    2015-01-01

    In the article there are designated the state and actual hygiene tasks on the issue of environmental pollution and its effects on health of the population. There was emphasized the growing importance of chemical contamination of various objects of environment--air water, soil, and living environment. There is presented the analysis of data on different types of treatment of municipal waste in selected countries. There were shown the significance of the developed Guidance on risk assessmentfor public health as a toolfor making sound management decisions, prospects of using of the methodology of epidemiological mapping based on geoinformational technology (GIS technology). There was marked an important role of the younger generation of hygienists and health officers in further work on both preservation and improvement the health of the population in their countries, harmonization of scientific and practical solutions of actual problems of hygiene.

  5. Considerations of human inturison in U.S. programs for deep geologic disposal of radioactive waste.

    Energy Technology Data Exchange (ETDEWEB)

    Swift, Peter N.

    2013-01-01

    Regulations in the United States that govern the permanent disposal of spent nuclear fuel and high-level radioactive waste in deep geologic repositories require the explicit consideration of hypothetical future human intrusions that disrupt the waste. Specific regulatory requirements regarding the consideration of human intrusion differ in the two sets of regulations currently in effect in the United States; one defined by the Environmental Protection Agencys 40 Code of Federal Regulations part 197, applied only to the formerly proposed geologic repository at Yucca Mountain, Nevada, and the other defined by the Environmental Protection Agencys 40 Code of Federal Regulations part 191, applied to the Waste Isolation Pilot Plant in New Mexico and potentially applicable to any repository for spent nuclear fuel and high-level radioactive waste in the United States other than the proposed repository at Yucca Mountain. This report reviews the regulatory requirements relevant to human intrusion and the approaches taken by the Department of Energy to demonstrating compliance with those requirements.

  6. The Handling of Liquid Waste at the Research Station of Studsvik, Sweden

    International Nuclear Information System (INIS)

    Lindhe, Soeren; Linder, Per

    1965-03-01

    The following quantities of radioactive waste are allowed to be released into a strait between the islands of Stora Bergoe and Studsviksholme: Total α-activity 0,2 curie/month. Total β-activity 36 curie/month of which cerium, yttrium, rare earth 15 curie/month, strontium 2,4 curie/month. Before the release the radioactive waste has to be collected and controlled. Quantities approaching or exceeding the disposal limits have to be removed and concentrated by evaporation. The liquid waste is classified in several categories depending upon the level of activity: high active and medium active waste, low active waste, process water, sanitary water, surface water and reactor cooling water. The technical dimensioning of each category was based upon expected specific production figures (volume/man - month and activity/ man - month). These figures are based upon information obtained from Harwell. Actual production figures obtained during 1963 and the first half of 1964 are shown and compared with the expected ones. As a conclusion is stated that the actual production follows the predictions fairly well

  7. The Handling of Liquid Waste at the Research Station of Studsvik, Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Lindhe, Soeren; Linder, Per

    1965-03-15

    The following quantities of radioactive waste are allowed to be released into a strait between the islands of Stora Bergoe and Studsviksholme: Total {alpha}-activity 0,2 curie/month. Total {beta}-activity 36 curie/month of which cerium, yttrium, rare earth 15 curie/month, strontium 2,4 curie/month. Before the release the radioactive waste has to be collected and controlled. Quantities approaching or exceeding the disposal limits have to be removed and concentrated by evaporation. The liquid waste is classified in several categories depending upon the level of activity: high active and medium active waste, low active waste, process water, sanitary water, surface water and reactor cooling water. The technical dimensioning of each category was based upon expected specific production figures (volume/man - month and activity/ man - month). These figures are based upon information obtained from Harwell. Actual production figures obtained during 1963 and the first half of 1964 are shown and compared with the expected ones. As a conclusion is stated that the actual production follows the predictions fairly well.

  8. Technology for commercial radioactive waste management

    International Nuclear Information System (INIS)

    1979-05-01

    Conceptual processes and facilities for treating gaseous and various transuranium (TRU) wastes produced during the past fission portion of the light water reactor fuel cycle are described in volume 2. The goal of the treatment process for TRU wastes and for long-lived radionuclides removed from the gaseous waste streams is to convert these wastes to stable products suitable for placement in geologic isolation repositories. The treatment concepts are based on available technology. They do not necessarily represent an optimum design but are representative of what could be achieved with current technology. In actual applications it is reasonable to expect that there could be some improvement over these concepts that might be reflected in either lower costs or lower environmental impacts or both. These conceptual descriptions do provide a reasonable basis for cost analysis and for development of estimates of environmental impacts. The waste treatment technologies considered here include: high-level waste solidification, packaging of fuel residue, failed equipment and noncombustible waste treatment, general trash and combustible waste treatment, degraded solvent treatment, dilute aqueous waste pretreatment, immobilization of wet and solid wastes, off-gas particle removal systems, fuel reprocessing plant dissolver off-gas treatment, process off-gas treatment, and fuel reprocessing plant atmospheric protection system

  9. Effect of microwaves on solubilization of organic fraction of municipal solid waste

    Energy Technology Data Exchange (ETDEWEB)

    Shahriari, H.; Warith, M.; Kennedy, K.J. [Ottawa Univ., Ottawa, ON (Canada). Dept. of Civil Engineering

    2009-07-01

    Landfilling is the most common method for disposing of municipal solid waste (MSW) in North America. MSW consists of nonbiodegradable fractions as well as biodegradable fractions known as the organic fraction of municipal solid waste (OFMSW). Because of its high moisture content, OFMSW produces large amounts of leachate in landfills. If not treated properly, leachates can pollute groundwater and negatively affect health and the environment. This paper reported on a study that was conducted to determine the effects of microwave (MW) irradiation on the solubilization of organic fraction of municipal solid waste (OFMSW) at different temperatures, MW ramp times, and supplemental water addition (SWA). The objective was to enhance solubilization before anaerobic digestion (AD). MW pretreatment resulted in higher soluble chemical oxygen demand (sCOD), proteins and sugars in the supernatant phase. The highest increase in sCOD was achieved at 175 degrees C. For the same condition, the free liquid volume from bound water released from OFMSW into the supernatant was about 1.39 times higher than the control. The increase in potentially bio-available sCOD increased significantly to more than 200 per cent after microwaving at high temperature. It was concluded that microwaving of OFMSW at high temperature with SWA provides the best conditions for waste solubilisation in preparation for anaerobic digestion. The actual effect of MW pre-treatment on the anaerobic digestion process has yet to be determined. 49 refs., 5 tabs., 3 figs.

  10. U.S. Department of Energy's 'initiatives for proliferation prevention' program: solidification technologies for radioactive waste treatment in Russia - 16037

    International Nuclear Information System (INIS)

    Pokhitonov, Yuri; Kelley, Dennis

    2009-01-01

    Large amounts of liquid radioactive waste have existed in the U.S. and Russia since the 1950's as a result of the Cold War. Comprehensive action to treat and dispose of waste products has been lacking due to insufficient funding, ineffective technologies or no proven technologies, low priority by governments among others. Today the U.S. and Russian governments seek new, more reliable methods to treat liquid waste, in particular the legacy waste streams. A primary objective of waste generators and regulators is to find economical and proven technologies that can provide long-term stability for repository storage. In 2001, the V.G. Khlopin Radium Institute (Khlopin), St. Petersburg, Russia, and Pacific Nuclear Solutions (PNS), Indianapolis, Indiana, began extensive research and test programs to determine the validity of polymer technology for the absorption and immobilization of standard and complex waste streams. Over 60 liquid compositions have been tested including extensive irradiation tests to verify polymer stability and possible degradation. With conclusive scientific evidence of the polymer's effectiveness in treating liquid waste, both parties have decided to enter the Russian market and offer the solidification technology to nuclear sites for waste treatment and disposal. In conjunction with these efforts, the U.S. Department of Energy (DOE) will join Khlopin and PNS to explore opportunities for direct application of the polymers at predetermined sites and to conduct research for new product development. Under DOE's 'Initiatives for Proliferation Prevention' (IPP) program, funding will be provided to the Russian participants over a three year period to implement the program plan. This paper will present updated details of U.S. DOE's IPP program, the project structure and its objectives both short and long-term, polymer tests and applications for LLW, ILW and HLW, and new product development initiatives. (authors)

  11. Determining leach rates of monolithic waste forms

    International Nuclear Information System (INIS)

    Gilliam, T.M.; Dole, L.R.

    1986-01-01

    The ANS 16.1 Leach Procedure provides a conservative means of predicting long-term release from monolithic waste forms, offering a simple and relatively quick means of determining effective solid diffusion coefficients. As presented here, these coefficients can be used in a simple model to predict maximum release rates or be used in more complex site-specific models to predict actual site performance. For waste forms that pass the structural integrity test, this model also allows the prediction of EP-Tox leachate concentrations from these coefficients. Thus, the results of the ANS 16.1 Leach Procedure provide a powerful tool that can be used to predict the waste concentration limits in order to comply with the EP-Toxicity criteria for characteristically nonhazardous waste. 12 refs., 3 figs

  12. The construction of solid waste form test facility

    International Nuclear Information System (INIS)

    Park, Hun Hwee; Kim, Joon Hyung; Lee, Byung Jik; Koo, Jun Mo; Kim, Jeong Guk; Jung, In Ha

    1990-03-01

    The solid waste form test facility (SWFTF) to test and/or evaluate the characteristics of waste forms, such as homogeniety, mechanical properties, thermal properties, waste resistance and leachability, have been constructed, and some equipments for testing actual waste forms has been purchased; radiocative monitoring system, glove box for the manipulator repair room, and uninteruppted power supply system, et al. Classifications of radioactive wastes, basic requirements and criteria to be considered during waste management were also reviewed. Some of the described items above have been standardized for the purpose of indigenigation. Therefore, safety assurance of waste forms, as well as increase in the range of participating of domestic companies in construction of further nuclear facilities could be obtained as results through constructing this facility. In the furture this facility is going to be utilized not only for the inspection of waste forms but also for the periodic decontamination for extending the life time of some expensive radiological equipments using remote handling techniques. (author)

  13. Who is Self-Actualized?

    Science.gov (United States)

    Roweton, William E.

    1981-01-01

    In an attempt to clarify Maslow's concept of self-actualization as it relates to human motivation, a class of educational psychology students wrote essays describing a self-actualized person and then attempted to decide whether public schools contribute to the production of self-actualized persons. Two-thirds of the students decided that schools…

  14. U.S. policy and current practices for blending low-level radioactive waste for disposal

    International Nuclear Information System (INIS)

    Kessel, David S.; Kim, Chang Lak

    2016-01-01

    In the near future, many countries, including the Republic of Korea, will face a significant increase in low level radioactive waste (LLW) from nuclear power plant decommissioning. The purpose of this paper is to look at blending as a method for enhancing disposal options for low-level radioactive waste from the decommissioning of nuclear reactors. The 2007 U.S. Nuclear Regulatory Commission strategic assessment of the status of the U.S. LLW program identified the need to move to a risk-informed and performance-based regulatory approach for managing LLW. The strategic assessment identified blending waste of varying radionuclide concentrations as a potential means of enhancing options for LLW disposal. The NRC's position is that concentration averaging or blending can be performed in a way that does not diminish the overall safety of LLW disposal. The revised regulatory requirements for blending LLW are presented in the revised NRC Branch Technical Position for Concentration Averaging and Encapsulation (CA BTP 2015). The changes to the CA BTP that are the most significant for NPP operation, maintenance and decommissioning are reviewed in this paper and a potential application is identified for decommissioning waste in Korea. By far the largest volume of LLW from NPPs will come from decommissioning rather than operation. The large volumes in decommissioning present an opportunity for significant gains in disposal efficiency from blending and concentration averaging. The application of concentration averaging waste from a reactor bio-shield is also presented

  15. Spectroscopic Properties of Tc(I) Tricarbonyl Species Relevant to the Hanford Tank Waste

    Energy Technology Data Exchange (ETDEWEB)

    Levitskaia, Tatiana G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Andersen, Amity [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Chatterjee, Sayandev [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hall, Gabriel B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Walter, Eric D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Washton, Nancy M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-12-04

    Technetium-99 (Tc) exists predominately in soluble forms in the liquid supernatant and salt cake fractions of the nuclear tank waste stored at the U.S. DOE Hanford Site. In the strongly alkaline environments prevalent in the tank waste, its dominant chemical form is pertechnetate (TcO4-, oxidation state +7). However, attempts to remove Tc from the Hanford tank waste using ion-exchange processes specific to TcO4- only met with limited success, particularly processing tank waste samples containing elevated concentrations of organic complexants. This suggests that a significant fraction of the soluble Tc can be present as non-pertechnetate low-valent Tc (oxidation state < +7) (non-pertechnetate). The chemical identities of these non-pertechnetate species are poorly understood. Previous analysis of the SY-101 and SY-103 tank waste samples provided strong evidence that non-pertechnetate can be comprised of [Tc(CO)3]+ complexes containing Tc in oxidation state +1 (Lukens et al. 2004). During the last two years, our team has expanded this work and demonstrated that high-ionic-strength solutions typifying tank waste supernatants promote oxidative stability of the [Tc(CO)3]+ species (Rapko et al. 2013; Levitskaia et al. 2014). It also was observed that high-ionic-strength alkaline matrices stabilize Tc(VI) and potentially Tc(IV) oxidation states, particularly in presence organic chelators, suggesting that the relevant Tc compounds can serve as important redox intermediates facilitating the reduction of Tc(VII) to Tc(I). Designing strategies for effective Tc processing, including separation and immobilization, necessitates understanding the molecular structure of these non-pertechnetate species and their identification in the actual tank waste samples. To-date, only limited information exists regarding the nature and characterization of the Tc(I), Tc(IV), and Tc(VI) species. One objective of this project is to

  16. Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 1: Sections 1-9

    International Nuclear Information System (INIS)

    Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.

    1995-04-01

    This report documents the methodology, computational framework, and results of facility accident analyses performed for the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. The methodology is in compliance with the most recent guidance from DOE. It considers the spectrum of accident sequences that could occur in activities covered by the WM PEIS and uses a graded approach emphasizing the risk-dominant scenarios to facilitate discrimination among the various WM PEIS alternatives. Although it allows reasonable estimates of the risk impacts associated with each alternative, the main goal of the accident analysis methodology is to allow reliable estimates of the relative risks among the alternatives. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

  17. Medical and Biohazardous Waste Generator's Guide (Revision2)

    Energy Technology Data Exchange (ETDEWEB)

    Waste Management Group

    2006-11-29

    These guidelines describe procedures to comply with all Federal and State laws and regulations and Lawrence Berkeley National Laboratory (LBNL) policy applicable to State-regulated medical and unregulated, but biohazardous, waste (medical/biohazardous waste). These guidelines apply to all LBNL personnel who: (1) generate and/or store medical/biohazardous waste, (2) supervise personnel who generate medical/biohazardous waste, or (3) manage a medical/biohazardous waste pickup location. Personnel generating biohazardous waste at the Joint Genome Institute/Production Genomics Facility (JGI/PGF) are referred to the guidelines contained in Section 9. Section 9 is the only part of these guidelines that apply to JGI/PGF. Medical/biohazardous waste referred to in this Web site includes biohazardous, sharps, pathological and liquid waste. Procedures for proper storage and disposal are summarized in the Solid Medical/Biohazardous Waste Disposal Procedures Chart. Contact the Waste Management Group at 486-7663 if you have any questions regarding medical/biohazardous waste management.

  18. U.S. advanced accelerator applications program: plans to develop and test waste transmutation technologies

    International Nuclear Information System (INIS)

    Van Tuyle, G.; Bennett, D.; Arthur, E.; Cappiello, M.; Finck, P.; Hill, D.; Herczeg, J.; Goldner, F.

    2001-01-01

    The primary mission of the U.S. Advanced Accelerator Applications (AAA) Program is to establish a national nuclear technology research capability that can demonstrate accelerator-based transmutation of waste and conduct transmutation research while at the same time providing a capability for the production of tritium if required. The AAA Program was created during fiscal year 2001 from the Accelerator Transmutation of Waste (ATW) Program and the Accelerator Production of Tritium (APT) Project. This paper describes the new AAA Program, as well as its two major components: development and testing of waste transmutation technologies and construction of an integrated accelerator-driven test facility (ADTF). (author)

  19. ANALYZING CERTAIN CHRACTERISTICS OF MUNICIPAL SOLID WASTE GENERATION IN THE PROCES S OF WASTE MANAGEMENT

    Directory of Open Access Journals (Sweden)

    Gábriel Györgyi T #336;ZSÉR

    2010-01-01

    Full Text Available Based on the regulations of Act XLIII/2000 on Waste Management to implement the strategic objectives and targets in the Act for the prevalence of the basic waste management principles a National Waste Management Plan II will be worked out and then accepted by the Parliament as part of the National Environmental Protection Programme. On the basis of the national plan the administrative bodies of environmental protection in accordance with the regional settlement and d evelopment programmes make a regional waste management project with the inclusion of the regional, local authorities, and other authorities concerned as well as the non governmental organisations for environmental protection. In our research we analyze the correlation between municipal solid waste per capita and urbanisation level. We have conducted similar calculations in the filed of population density and income. The study was carried out on a micro region level. Our analysis can help determine the framework conditions and factors that influence waste generation, and therefore should be taken into consideration when designing waste policies .

  20. Slurry growth and gas retention in synthetic Hanford waste

    International Nuclear Information System (INIS)

    Bryan, S.A.; Pederson, L.R.; Scheele, R.D.

    1992-09-01

    This work seeks to establish chemical and physical processes responsible for the generation and retention of gases within waste from a particular high-level waste tank on the Hanford Site, Tank 101-SY, through the use of synthetic wastes on a laboratory scale. The goal of these activities is to support the development of mitigation/remediation strategies for Tank 101-SY. Laboratory studies of aged synthetic waste have shown that gas generation occurs thermally at a significant level at current tank temperatures. Gas compositions include the same gases produced in actual tank waste, primarily N 2 , N 2 O, and H 2 . Gas stoichiometries have been shown to be greatly influenced by several organic and inorganic constituents within the synthetic waste. Retention of gases in the synthetic waste is in the form of bubble attachment to solid particles

  1. Management of radioactive wastes at UFRJ

    International Nuclear Information System (INIS)

    Soares, J.C.A.R.; Anjos, M.J.; Mattos, J.A.; Souza, H.V.; Castro, C.R.F.; Sarquis, A.C.R.; Padilha Filho, L.G.; Borges, J.C.

    1992-01-01

    Since 1987, when the radiological Control Service of UFRJ was created, a radioprotection program has been developed, including all the installations and employees involved with ionizing radiations activities. Recently, a more systematic approach for radioactive wastes produced, monthly/annual manipulated activities, procedure and capacity of storage, and forms of release to the environment. Actual situation is not distant from that required by the norms about, but there area some critical situations that deserved emergency corrections. A plan for waste management is being prepared to be executed in the short and middle terms. (author)

  2. Nuclear power plant Isar. Actualized environmental statement 2012 on the consolidated environmental statement 2010

    International Nuclear Information System (INIS)

    2012-01-01

    The actualized environmental statement 2012 for the NPP Isar covers the following issues: Direct environmental aspects: heat dissipation to the environment, water management, radioactive matter release onto the environment, environmental monitoring, waste management, operational materials supply, radioactive materials transport, safety of the nuclear facilities, remote monitoring and information system, environmental protection, communication; indirect environmental aspects; core indicators according EMAS III, development of the environmental targets and programs, reference to the valid environmental regulations, declaration of the environmental consultant.

  3. Alternative techniques for low-level waste shallow land burial

    International Nuclear Information System (INIS)

    Levin, G.B.; Mezga, L.J.

    1983-01-01

    Experience to date relative to the shallow land burial of low-level radioactive waste (LLW) indicates that the physical stability of the disposal unit and the hydrologic isolation of the waste are the two most important factors in assuring disposal site performance. Disposal unit stability can be ensured by providing stable waste packages and waste forms, compacting backfill material, and filling the void spaces between the packages. Hydrologic isolation can be achieved though a combination of proper site selection, subsurface drainage controls, internal trench drainage systems, and immobilization of the waste. A generalized design of a LLW disposal site that would provide the desired long-term isolation of the waste is discussed. While this design will be more costly than current practices, it will provide additional confidence in predicted and reliability and actual site performance

  4. Tritium waste control: April-September 1982

    International Nuclear Information System (INIS)

    Lamberger, P.H.; Rogers, M.L.

    1983-01-01

    The pilot-scale, water feed cleanup system was used to successfully remove organic and inorganic impurities from Effluent Removal System (ERS) water. Tests with activated carbon traps removed organic impurities to as low as 2.5 ppM total carbon. Traps containing Amberlite resins for removing organic impurities were not successful and actually contaminated the water with higher levels (>2000 ppM) of organics. Gas generation rates caused by radiolysis of tritiated waste materials were determined for polymer and nonpolymer-impregnated tritiated concrete and fixated and nonfixated tritiated waste vacuum pump oil. In addition, the pressure change of hydrogen cover gas over tritiated water on cement-plaster was determined. The test program to measure and compare the release of tritium from tritiated concrete with and without styrene impregnation continued. Tritium permeation data from small test blocks are given. The drum study monitoring the release of tritium from actual burial packages continued. The maximum fractional release rate for the three types of high activity, tritiated liquid waste generated is 2.97 x 10 -5 , and the maximum total permeation is 158 mCi after 8 yr. These two values represent a 13% increase for the past 6 months. Tritium release from the polymer-impregnated, tritiated concrete (PITC) and from the control (non-PITC) remains very low

  5. Overview of DOE LLWMP waste treatment, packaging, and handling activities

    International Nuclear Information System (INIS)

    Pechin, W.H.

    1982-01-01

    The program objective is to develop the best available technology for waste treatment, packaging, and handling to meet the needs of shallow land burial disposal and for greater confinement than shallow land burial. The program has reviewed many of the hardware options for appropriate usage with low-level waste, but promising options remain to be evaluated. The testing of treatment technologies with actual radioactive process wastes has been initiated. The analysis of the interaction of treatment, solidification and disposal needs to be completed

  6. TESTING OF ENHANCED CHEMICAL CLEANING OF SRS ACTUAL WASTE TANK 5F AND TANK 12H SLUDGES

    Energy Technology Data Exchange (ETDEWEB)

    Martino, C.; King, W.

    2011-08-22

    Forty three of the High Level Waste (HLW) tanks at the Savannah River Site (SRS) have internal structures that hinder removal of the last approximately five thousand gallons of waste sludge solely by mechanical means. Chemical cleaning can be utilized to dissolve the sludge heel with oxalic acid (OA) and pump the material to a separate waste tank in preparation for final disposition. This dissolved sludge material is pH adjusted downstream of the dissolution process, precipitating the sludge components along with sodium oxalate solids. The large quantities of sodium oxalate and other metal oxalates formed impact downstream processes by requiring additional washing during sludge batch preparation and increase the amount of material that must be processed in the tank farm evaporator systems and the Saltstone Processing Facility. Enhanced Chemical Cleaning (ECC) was identified as a potential method for greatly reducing the impact of oxalate additions to the SRS Tank Farms without adding additional components to the waste that would extend processing or increase waste form volumes. In support of Savannah River Site (SRS) tank closure efforts, the Savannah River National Laboratory (SRNL) conducted Real Waste Testing (RWT) to evaluate an alternative to the baseline 8 wt. % OA chemical cleaning technology for tank sludge heel removal. The baseline OA technology results in the addition of significant volumes of oxalate salts to the SRS tank farm and there is insufficient space to accommodate the neutralized streams resulting from the treatment of the multiple remaining waste tanks requiring closure. ECC is a promising alternative to bulk OA cleaning, which utilizes a more dilute OA (nominally 2 wt. % at a pH of around 2) and an oxalate destruction technology. The technology is being adapted by AREVA from their decontamination technology for Nuclear Power Plant secondary side scale removal. This report contains results from the SRNL small scale testing of the ECC process

  7. Summary of 4th general plan of radioactive wastes

    International Nuclear Information System (INIS)

    Veganzones, A.

    1995-01-01

    The last 9th of December 1994 the Council of Ministries approved the Fourth General Plan of Radioactive Waste (PGRR). The Fourth Plan actualizes former texts taking into into account new circumstances, both technical and economical, affecting radioactive waste. Some of the steps that conform the global waste management process have been revised on the light of the Spanish experience but also considering the evolution and trends in other countries. In this work some of the most important aspects included in the Fourth general Plan are overviewed. (Author)

  8. From waste minimization to ISO 14000: Taiwan`s experience and prospects

    Energy Technology Data Exchange (ETDEWEB)

    Wen-Huei Chen; Wain-Sun Hou [China Technical Consultants, Inc., Taipei (Taiwan, Province of China)

    1996-12-31

    Taiwan has completed a very successful five-year industrial waste minimization (IWM) demonstration and promotion project sponsored by the government. From 1990 to 1995, the project successfully disseminated the IWM concept of pollution prevention (P2) to industries. It effectively reduced industrial waste while significantly benefitting the economy by assisting industries in implementing in-plant IWM programs. In July 1995, the second stage of the five-year IWM and ISO 14000 promotion project was initiated for further promoting the IWM, P2, and cleaner production and, in particular, coping with the upcoming international environmental management standards (ISO 14000). To assist industries in establishing an environmental management system (EMS) and accumulating related experience, an EMS pilot demonstration project of five model industries and an ISO 14001 EMS demonstration and promotion project for 22 factories in 13 industries were initiated in October 1995 and August 1996, respectively. These projects can assist Taiwan`s industries in changing the constitution of their enterprises, enhancing competition in the international market, and helping our nation achieve the forerunner`s profits in sustainable development. 9 refs., 3 figs.

  9. Salt removal from tanks containing high-level radioactive waste

    International Nuclear Information System (INIS)

    Kiser, D.L.

    1981-01-01

    At the Savannah River Plant (SRP), there are 23 waste storage tanks containing high-level radioactive wastes that are to be retired. These tanks contain about 23 million liters of salt and about 10 million liters of sludge, that are to be relocated to new Type III, fully stress-relieved tanks with complete secondary containment. About 19 million liters of salt cake are to be dissolved. Steam jet circulators were originally proposed for the salt dissolution program. However, use of steam jet circulators raised the temperature of the tank contents and caused operating problems. These included increased corrosion risk and required long cooldown periods prior to transfer. Alternative dissolution concepts were investigated. Examination of mechanisms affecting salt dissolution showed that the ability of fresh water to contact the cake surface was the most significant factor influencing dissolution rate. Density driven and mechanical agitation techniques were developed on a bench scale and then were demonstrated in an actual waste tank. Actual waste tank demonstrations were in good agreement with bench-scale experiments at 1/85 scale. The density driven method utilizes simple equipment, but leaves a cake heel in the tank and is hindered by the presence of sludge or Zeolite in the salt cake. Mechanical agitation overcomes the problems found with both steam jet circulators and the density driven technique and is the best method for future waste tank salt removal

  10. Laboratory simulation of high-level liquid waste evaporation and storage

    International Nuclear Information System (INIS)

    Anderson, P.A.

    1978-01-01

    The reprocessing of nuclear fuel generates high-level liquid wastes (HLLW) which require interim storage pending solidification. Interim storage facilities are most efficient if the HLLW is evaporated prior to or during the storage period. Laboratory evaporation and storage studies with simulated waste slurries have yielded data which are applicable to the efficient design and economical operation of actual process equipment

  11. Occupational and Public Exposure During Normal Operation of Radioactive Waste Disposal Facilities

    OpenAIRE

    M. V. Vedernikova; I. A. Pron; M. N. Savkin; N. S. Cebakovskaya

    2017-01-01

    This paper focuses on occupational and public exposure during operation of disposal facilities receiving liquid and solid radioactive waste of various classes and provides a comparative analysis of the relevant doses: actual and calculated at the design stage. Occupational and public exposure study presented in this paper covers normal operations of a radioactive waste disposal facility receiving waste. Results: Analysis of individual and collective occupational doses was performed based on d...

  12. A model based on feature objects aided strategy to evaluate the methane generation from food waste by anaerobic digestion.

    Science.gov (United States)

    Yu, Meijuan; Zhao, Mingxing; Huang, Zhenxing; Xi, Kezhong; Shi, Wansheng; Ruan, Wenquan

    2018-02-01

    A model based on feature objects (FOs) aided strategy was used to evaluate the methane generation from food waste by anaerobic digestion. The kinetics of feature objects was tested by the modified Gompertz model and the first-order kinetic model, and the first-order kinetic hydrolysis constants were used to estimate the reaction rate of homemade and actual food waste. The results showed that the methane yields of four feature objects were significantly different. The anaerobic digestion of homemade food waste and actual food waste had various methane yields and kinetic constants due to the different contents of FOs in food waste. Combining the kinetic equations with the multiple linear regression equation could well express the methane yield of food waste, as the R 2 of food waste was more than 0.9. The predictive methane yields of the two actual food waste were 528.22 mL g -1  TS and 545.29 mL g -1  TS with the model, while the experimental values were 527.47 mL g -1  TS and 522.1 mL g -1  TS, respectively. The relative error between the experimental cumulative methane yields and the predicted cumulative methane yields were both less than 5%. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. An assessment of the current municipal solid waste management system in Lahore, Pakistan.

    Science.gov (United States)

    Masood, Maryam; Barlow, Claire Y; Wilson, David C

    2014-09-01

    The current status of solid waste management in Lahore, a metropolitan city of Pakistan, is reviewed in this article using an existing approach, the UN-Habitat city profile. This involves a systematic quantitative and qualitative assessment of physical components and governance features of the current waste management system. A material flow diagram (MFD) is developed, which allows visualisation of the current waste management system with all related inputs and outputs. This study shows that in the current system, waste collection and transportation is the main focus, however the collection coverage is only about 68%. There is no controlled or even semi-controlled waste disposal facility in Lahore. There is no official recycling system in the city. It is estimated that currently ~27% of waste by weight is being recycled through the informal sector. Making use of the organic content of the waste, a composting facility is operative in the city, producing 47,230 tonnes year(-1) of organic compost. Lahore does not perform very well in governance features. Inclusivity of users and providers of the waste management system is low in the city, as not all stakeholders are consulted in the decision making processes. Waste management costs US$20 per tonne of waste, where the main focus is only on waste collection, and the current user fees are much lower than the actual costs. This study recommends that recycling should be promoted by increasing public awareness and integrating the informal sector to make the current system sustainable and financially viable. © The Author(s) 2014.

  14. New reactor concepts. An analysis of the actual research status; Neue Reaktorkonzepte. Eine Analyse des aktuellen Forschungsstands

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias

    2017-04-15

    The report on new reactor concepts covers the following issues: characterization and survey of new reactor concepts; evaluation criteria: safety, resources for fuel supply, waste problems, economy and proliferation; comprehensive relevant aspects: thorium as alternative resource, partitioning and transmutation; actual developments and preliminary experiences for fast breeding reactor (FBR), high-temperature reactor (HTR), molten salt reactor (MSR), small modular reactor (SMR).

  15. Development of an accelerated leach test(s) for low-level waste forms

    International Nuclear Information System (INIS)

    Dougherty, D.R.; Fuhrmann, M.; Colombo, P.

    1986-01-01

    An accelerated leach test(s) is being developed to predict long-term leaching behavior of low-level radioactive waste (LLW) forms in their disposal environments. As necessary background, a literature survey of reported leaching mechanisms, available mathematical models and factors that affect leaching of LLW forms has been compiled. Mechanisms which have been identified include diffusion, dissolution, ion exchange, corrosion and surface effects. A computerized data base of LLW leaching data and mathematical models is being developed. The data is being used for model evaluation by curve fitting and statistical analysis according to standard procedures of statistical quality control. Long-term leach tests on portland cement, bitumen and vinyl ester-styrene (VES) polymer waste forms are underway which are designed to identify and evaluate factors that accelerate leaching without changing the mechanisms. Initial results on the effect of temperature on leachability indicate that the leach rates of cement and VES waste forms increase with increasing temperature, whereas, the leach rate of bitumen is little affected

  16. THE CONDITION OF WASTE MANAGEMENT IN ROMANIA

    Directory of Open Access Journals (Sweden)

    Diana ZAGAN ZELTER

    2009-01-01

    Full Text Available The present article approaches a very important and actual theme andthat is the problem of generating waste in Romania which, on one hand,affects the environment and human health, and on the other hand itreflects the inefficient way of using the natural resources in society.Probably the majority of us have thought or hoped that the naturalresources are inexhaustible, but we can see today that the unwiseexploitation of these resources is threatening our future.Waste management is a difficult and complex problem in Romania whichis far from being solved according to the environment rules of theEuropean Union. The worsening of the waste problem, especially of thedomestic waste is generated by the significant increase of its quantity, aswell as by the inappropriate way of solving different stages of wasteprocessing.

  17. Research and development program for transuranic-contaminated waste within the U.S. Energy Research and Development Administration

    International Nuclear Information System (INIS)

    Wolfe, R.A.

    1976-01-01

    This overview examines the research and development program that has been established within the U.S. Energy Research and Development Administration (ERDA) to develop the technology to treat transuranic-contaminated waste. Also considered is the waste expected within the total nuclear fuel cycle

  18. Alternative disposal options for alpha-mixed low-level waste

    International Nuclear Information System (INIS)

    Loomis, G.G.; Sherick, M.J.

    1995-01-01

    This paper presents several disposal options for the Department of Energy alpha-mixed low-level waste. The mixed nature of the waste favors thermally treating the waste to either an iron-enriched basalt or glass waste form, at which point a multitude of reasonable disposal options, including in-state disposal, are a possibility. Most notably, these waste forms will meet the land-ban restrictions. However, the thermal treatment of this waste involves considerable waste handling and complicated/expensive offgas systems with secondary waste management problems. In the United States, public perception of offgas systems in the radioactive incinerator area is unfavorable. The alternatives presented here are nonthermal in nature and involve homogenizing the waste with cryogenic techniques followed by complete encapsulation with a variety of chemical/grouting agents into retrievable waste forms. Once encapsulated, the waste forms are suitable for transport out of the state or for actual in-state disposal. This paper investigates variances that would have to be obtained and contrasts the alternative encapsulation idea with the thermal treatment option

  19. Alternative disposal options for alpha-mixed low-level waste

    Energy Technology Data Exchange (ETDEWEB)

    Loomis, G.G.; Sherick, M.J. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-12-31

    This paper presents several disposal options for the Department of Energy alpha-mixed low-level waste. The mixed nature of the waste favors thermally treating the waste to either an iron-enriched basalt or glass waste form, at which point a multitude of reasonable disposal options, including in-state disposal, are a possibility. Most notably, these waste forms will meet the land-ban restrictions. However, the thermal treatment of this waste involves considerable waste handling and complicated/expensive offgas, systems with secondary waste management problems. In the United States, public perception of off gas systems in the radioactive incinerator area is unfavorable. The alternatives presented here are nonthermal in nature and involve homogenizing the waste with cryogenic techniques followed by complete encapsulation with a variety of chemical/grouting agents into retrievable waste forms. Once encapsulated, the waste forms are suitable for transport out of the state or for actual in-state disposal. This paper investigates variances that would have to be obtained and contrasts the alternative encapsulation idea with the thermal treatment option.

  20. Motives as predictors of the public's attitudes toward solid waste issues

    Energy Technology Data Exchange (ETDEWEB)

    Ebreo, A.; Vining, J.

    2000-02-01

    Surveys focusing on solid-waste-related issues, conducted over a period of several years, provided data from independent samples of residents of a Midwestern, USA, community. The collection of these data yielded useful information about the relationship between residents' recycling motives and their attitudes toward solid waste management in light of several changes in the solid waste infrastructure of the community over that time. The initial survey assessed baseline beliefs and attitudes, while later surveys were conducted after the implementation of a community educational program and a curbside recycling program. The findings indicated that for recyclers and nonrecyclers, different motives predicted endorsement of solid waste programs and policies. Although a similar percentage of recyclers and nonrecyclers were in support of various proposed programs and policies, concern for the environment was found to be positively related to nonrecyclers' support of proposed programs, particularly before these programs were implemented. Prior to program implementation, motives other than environmental altruism were found to be related to recyclers' support of the programs. Additional findings support the idea that educational programs and increased accessibility to recycling opportunities affect the relationship between people's attitudes toward solid waste management and their recycling motives.

  1. Super analog computer for evaluating the safety of buried radioactive waste

    International Nuclear Information System (INIS)

    Cohen, B.L.

    1980-01-01

    It is argued that the past use of digital computer programs for evaluating the safety of buried radioactive waste has been largely wasteful and dangerously delusive. It is suggested to use actual rocks as the analog of buried waste. The problem of comparable rates of leaching of radioactive waste and of natural rock is discussed. Two examples are given of the use of natural rock as an ''analog computer'': one for high-level radioactive waste, and one for low-level radioactive waste. Digital computers have not contributed anything to two crucial questions: Can shafts be securely sealed. Does the heat crack the rock or have important effects on its chemistry. 4 refs

  2. Capabilities of U.S. domestic transportation systems for the shipment of radioactive wastes

    International Nuclear Information System (INIS)

    Best, R.E.; Allen, J.H.; Aucoin, P.A.; Ball, G.D.; Hoffman, C.C.; Mason, M.E.; Propes, W.A.; Vizzini, T.A.

    1977-09-01

    This document is a compilation of data and reports that provide an overview of the capabilities of U.S. domestic transportation systems for the shipment of materials that are or may be classified as radioactive wastes

  3. Criticality safety evaluation for TRU waste in storage at the RWMC

    International Nuclear Information System (INIS)

    Shaw, M.E.; Briggs, J.B.; Atkinson, C.A.; Briscoe, G.J.

    1993-11-01

    Stored containers (drums, boxes, and bins) of transuranic waste at the Radioactive Waste Management Complex (RWMC) facility located at the Idaho National Engineering Laboratory (INEL) were evaluated based on inherent neutron absorption characteristics of the waste materials. It was demonstrated that these properties are sufficient to preclude an accidental criticality accident at the actual fissile levels present in the waste stored at the RWMC. Based on the database information available, the results reported herein confirm that the waste drums, boxes, and bins currently stored at the RWMC will remain safely subcritical if rearranged, restacked, or otherwise handled. Acceptance criteria for receiving future drum shipments were established based on fully infinite systems

  4. Projected Salt Waste Production from a Commercial Pyroprocessing Facility

    Directory of Open Access Journals (Sweden)

    Michael F. Simpson

    2013-01-01

    Full Text Available Pyroprocessing of used nuclear fuel inevitably produces salt waste from electrorefining and/or oxide reduction unit operations. Various process design characteristics can affect the actual mass of such waste produced. This paper examines both oxide and metal fuel treatment, estimates the amount of salt waste generated, and assesses potential benefit of process options to mitigate the generation of salt waste. For reference purposes, a facility is considered in which 100 MT/year of fuel is processed. Salt waste estimates range from 8 to 20 MT/year from considering numerous scenarios. It appears that some benefit may be derived from advanced processes for separating fission products from molten salt waste, but the degree of improvement is limited. Waste form production is also considered but appears to be economically unfavorable. Direct disposal of salt into a salt basin type repository is found to be the most promising with respect to minimizing the impact of waste generation on the economic feasibility and sustainability of pyroprocessing.

  5. High-level radioactive waste glass and storage canister design

    International Nuclear Information System (INIS)

    Slate, S.C.; Ross, W.A.

    1979-01-01

    Management of high-level radioactive wastes is a primary concern in nuclear operations today. The main objective in managing these wastes is to convert them into a solid, durable form which is then isolated from man. A description is given of the design and evaluation of this waste form. The waste form has two main components: the solidified waste and the storage canister. The solid waste form discussed in this study is glass. Waste glasses have been designed to be inert to water attack, physically rugged, low in volatility, and stable over time. Two glass-making processes are under development at PNL. The storage canister is being designed to provide high-integrity containment for solidified wastes from processing to terminal storage. An outline is given of the steps in canister design: material selection, stress and thermal analyses, quality verification, and postfill processing. Examples are given of results obtained from actual nonradioactive demonstration tests. 14 refs

  6. Improved low-level radioactive waste management practices for hospitals and research institutions

    International Nuclear Information System (INIS)

    1983-07-01

    This report provides a general overview and a compendium of source material on low-level radioactive waste management practices in the institutional sector. Institutional sector refers to hospitals, universities, clinics, and research facilities that use radioactive materials in scientific research and the practice of medicine, and the manufacturers of radiopharmaceuticals and radiography devices. This report provides information on effective waste management practices for institutional waste to state policymakers, regulatory agency officials, and waste generators. It is not intended to be a handbook for actual waste management, but rather a sourcebook of general information, as well as a survey of the more detailed analysis

  7. Treatment of low- and intermediate-level solid radioactive wastes

    International Nuclear Information System (INIS)

    1983-01-01

    One of the essential aims in the waste management is to reduce as much as possible the waste volumes to be stored or disposed of, and to concentrate and immobilize as much as possible the radioactivity contained in the waste. This document describes the treatment of low- and intermediate-level solid waste prior to its conditioning for storage and disposal. This report aims primarily at compiling the experience gained in treating low- and intermediate-active solid wastes, one of the major waste sources in nuclear technology. Apart from the description of existing facilities and demonstrated handling schemes, this report provides the reader with the basis for a judgement that facilitates the selection of appropriate solutions for a given solid-waste management problem. It thus aims at providing guidelines in the particular field and indicates new promising approaches that are actually under investigation and development

  8. Co-Benefits of Household Waste Recycling for Local Community’s Sustainable Waste Management in Thailand

    Directory of Open Access Journals (Sweden)

    Amornchai Challcharoenwattana

    2015-06-01

    Full Text Available The study aimed to evaluate co-benefits in term of GHG reduction, and avoided landfill costs by implementing a community-based management (CBM program for municipal solid waste (MSW. Two towns of peri-urban settlement in Thailand were investigated in case studies to compare eco-performance between the towns with and without implementation of the CBM program. MSW mass flows together with MSW utilization records were analyzed based on data in year 2013. Climate co-benefits from waste utilization activities were examined. Results from the study indicated that waste banks in the CBM program can effectively divert most of recyclables from entering landfills. The performance of “waste bank—recyclable recovery program” recycling rate from the case study with CBM is 172.20 kg per member per year, which is about 926% higher than average CBMs with MSW recycling in Thailand, and the success of CBM can be attributed to its curbside pickup service and fair-pricing of recyclables. The study also found that if the town decided to divert wastes from landfilling, carbon intensity of the MSW system would be 0.47 tons of CO2-eq per ton of collected MSW. The landfilling cost would be approximately 7.41 USD per ton of MSW as landfilling cost. With CBM programs, current MSW reutilization rate has achieved 9.68% of generated waste, and 16.80% of GHG emission has been avoided, along with a reduction in landfill costs of 11.57%. Two scenarios of waste utilization in Thailand were explored and compared, in terms of which scenarios yielded the highest co-benefits. The study demonstrates that by allowing local mechanism and community involvement programs to develop with operational waste banks, the efficiency of collecting recycling wastes increased. A similar system can be applied to other communities in other countries.

  9. Angina crónica estable. Consideraciones actuales

    Directory of Open Access Journals (Sweden)

    Yanier Coll Muñoz

    2011-03-01

    Full Text Available La angina de pecho crónica estable es un síndrome clínico caracterizado por dolor o malestar precordial secundario a isquemia miocárdica y sin características clínicas de inestabilidad. El debate actual sobre esta entidad clínica incluye tres aspectos fundamentales: diagnóstico, prevalencia e impacto socioeconómico, y tratamiento médico frente a revascularización miocárdica. En los últimos cuatro años se han dado a conocer los resultados de varios estudios multicéntricos de gran importancia que compararon tratamiento médico con revascularización miocárdica y cirugía de revascularización frente a intervención coronaria percutánea; además, han mejorado las técnicas de revascularización y existe abundante información sobre la eficacia de nuevos fármacos antiisquémicos. El presente trabajo tiene como objetivo hacer una revisión de los aspectos actualess importantes relacionados con la angina de pecho estable y su tratamiento, publicados en la literatura especializada.

  10. Substantiation of ways for improvement of questionnaire-survey method of asssessment of population’s actual nutrition

    Directory of Open Access Journals (Sweden)

    Ieltsova L.B.

    2017-10-01

    Full Text Available Actual nutrition state indicates a significant prevalence of polynutrient deficits in Ukraine. One of the key issues in the fight against alimentary-dependent diseases is monitoring of population’s nutrition status. Only reliable methods for assessing actual nutrition can provide reliable information about the nature of nutrition and establish a relationship with health state. Taking into account the abovementioned, the aim of our study was to compare the existing methods of studying nutrition and nutrition status of individuals or population groups to substantiate and develop a new universal, effective method for assessing nutrition and its subsequent approval. It has been found that existing methods for assessing human food consumption have its disadvantages and advantages and provide the opportunity to receive certain types of nutrition information. There is no single universal method that would allow to make comprehensive assessment of population’s nutrition, including student youth. To solve this issue, we have substantiated and developed a questionnaire for assessing the nutrition of different population groups, including student youth, which has been evaluated and approved by specialists from leading in­stitu­tions in the field of nutrition hygiene, gastroenterology and dietetics. Created survey questionnaire is a universal tool for collecting primary information and providing an opportunity to assess the social and welfare standards of respon­dents, working conditions and living conditions, the presence of hazard factors and bad habits, involvement in sports, anthropometric indicators, as well as it allows to assess the diet, drinking regime, main food product groups consumption, weight of portions, food handling methods, priority and benefits when choosing food, food intake places, which will be taken into account in the future to develop effective measures for the prevention of alimentary and alimentary-dependent diseases.

  11. Electrical Energy Harvesting from Cooker’s Wasted Heat with Using Conduction Cooling

    Directory of Open Access Journals (Sweden)

    Amouzard Mahdiraji Wincent Ghafour

    2018-01-01

    Full Text Available In order meet the demand of electricity in current era, the need for new sources of energy even in very minimal amount, could be done with proper research and technology advancement in order to convert as much wasted energy as possible. Collecting and analyses cooker’s wasted heat as a main wasted energy source become the main interest for this research. This application can be installed either in household usage or commercial usage. Based on majority stove in household datasheet it shown that the efficiency of the stove is approximately 50%. With half of the efficiency turn into wasted heat, this application is suitable for thermoelectric generator (TEG to harvest the heat. The objective of this research is to determine whether the thermoelectric generator (TEG would able to power the 3V LED light as a small lighting system in household. Several designs with five TEGs in series circuit are tested to the application to analyses which method generated a better result. Since this research only focus in using a conduction cooling, aluminum heat sink will be utilized either for heat absorption or heat rejection. The maximum temperature differences between hot side and cold side is 209.83 °C with average power approximately 0.1 W.

  12. Degradation of hazardous chemicals in liquid radioactive wastes from biomedical research using a mixed microbial population

    International Nuclear Information System (INIS)

    Wolfram, J.H.; Radtke, M.; Wey, J.E.; Rogers, R.D.; Rau, E.H.

    1997-10-01

    As the costs associated with treatment of mixed wastes by conventional methods increase, new technologies will be investigated as alternatives. This study examines the potential of using a selected mixed population of microorganisms to treat hazardous chemical compounds in liquid low level radioactive wastes from biomedical research procedures. Microorganisms were isolated from various waste samples and enriched against compounds known to occur in the wastes. Individual isolates were tested for their ability to degrade methanol, ethanol, phenol, toluene, phthalates, acetonitrile, chloroform, and trichloroacetic acid. Following these tests, the organisms were combined in a media with a mixture of the different compounds. Three compounds: methanol, acetonitrile, and pseudocumene, were combined at 500 microliter/liter each. Degradation of each compound was shown to occur (75% or greater) under batch conditions with the mixed population. Actual wastes were tested by adding an aliquot to the media, determining the biomass increase, and monitoring the disappearance of the compounds. The compounds in actual waste were degraded, but at different rates than the batch cultures that did not have waste added. The potential of using bioprocessing methods for treating mixed wastes from biomedical research is discussed

  13. Rock solid: the geology of nuclear waste disposal

    International Nuclear Information System (INIS)

    Reid, Elspeth.

    1990-01-01

    With a number of nuclear submarines and power stations due to be decommissioned in the next decade, stores of radioactive waste, and arguments about storage increase. Whatever the direction taken by the nuclear industry in Britain, the legacy of waste remains for the foreseeable future. Geology is at the heart of the safety argument for nuclear wastes. It is claimed that rocks should act as the main safety barrier, protecting present and future generations from radiation. Rock Solid presents a clear, accessible and up to date account of the geological problems involved in building a nuclear waste repository. The author describes the geology of some of the possible UK repository sites (Sellafield, Dounreay, Altnabreac, Billingham), explains how sites are investigated (including computer models), and finally considers the crucial question: 'would geological containment of radioactive waste actually work?'. (author)

  14. Performance benefits of telerobotics and teleoperation - enhancements for an arm-based tank waste retrieval system

    Energy Technology Data Exchange (ETDEWEB)

    Horschel, D.S. [Sandia National Labs., Albuquerque, NM (United States); Gibbons, P.W. [Westinghouse Hanford Co., Richland, WA (United States); Draper, J.V. [Oak Ridge National Lab., TN (United States)] [and others

    1995-06-01

    This report evaluates telerobotic and teleoperational arm-based retrieval systems that require advanced robotic controls. These systems will be deployed in waste retrieval activities in Hanford`s Single Shell Tanks (SSTs). The report assumes that arm-based, retrieval systems will combine a teleoperational arm and control system enhanced by a number of advanced and telerobotic controls. The report describes many possible enhancements, spanning the full range of the control spectrum with the potential for technical maturation. The enhancements considered present a variety of choices and factors including: the enhancements to be included in the actual control system, safety, detailed task analyses, human factors, cost-benefit ratios, and availability and maturity of technology. Because the actual system will be designed by an offsite vendor, the procurement specifications must have the flexibility to allow bidders to propose a broad range of ideas, yet build in enough restrictions to filter out infeasible and undesirable approaches. At the same time they must allow selection of a technically promising proposal. Based on a preliminary analysis of the waste retrieval task, and considering factors such as operator limitations and the current state of robotics technology, the authors recommend a set of enhancements that will (1) allow the system to complete its waste retrieval mission, and (2) enable future upgrades in response to changing mission needs and technological advances.

  15. Projected radionuclide inventories of DWPF glass from current waste at time of production

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1993-01-01

    The Waste Acceptance Preliminary Specifications (WAPS) require that the DWPF estimate the inventory of long-lived radionuclides present in the waste glass, and report the values in the Waste Form Qualification Report. In this report, conservative (biased high) estimates of the radionuclide inventory of glass produced from waste currently in the Tank Farm are provided. In most cases, these calculated values compare favorably with actual data. In those cases where the agreement is not good, the values reported here are conservative

  16. The U.S. DOE new production reactor/heavy water reactor facility pollution prevention/waste minimization program

    International Nuclear Information System (INIS)

    Kaczmarsky, Myron M.; Tsang, Irving; Stepien, Walter P.

    1992-01-01

    A Pollution Prevention/Waste Minimization Program was established during the early design phase of the U.S. DOE's New Production Reactor/Heavy Water Reactor Facility (NPR/HWRF) to encompass design, construction, operation and decommissioning. The primary emphasis of the program was given to waste elimination, source reduction and/or recycling to minimize the quantity and toxicity of material before it enters the waste stream for treatment or disposal. The paper discusses the regulatory and programmatic background as it applies to the NPR/HWRF and the waste assessment program developed as a phased approach to pollution prevention/waste minimization for the NPR/HWRF. Implementation of the program will be based on various factors including life cycle cost analysis, which will include costs associated with personnel, record keeping, transportation, pollution control equipment, treatment, storage, disposal, liability, compliance and oversight. (author)

  17. Energy requirements for waste water treatment.

    Science.gov (United States)

    Svardal, K; Kroiss, H

    2011-01-01

    The actual mathematical models describing global climate closely link the detected increase in global temperature to anthropogenic activity. The only energy source we can rely on in a long perspective is solar irradiation which is in the order of 10,000 kW/inhabitant. The actual primary power consumption (mainly based on fossil resources) in the developed countries is in the range of 5 to 10 kW/inhabitant. The total power contained in our nutrition is in the range of 0.11 kW/inhabitant. The organic pollution of domestic waste water corresponds to approximately 0.018 kW/inhabitant. The nutrients contained in the waste water can also be converted into energy equivalents replacing market fertiliser production. This energy equivalent is in the range of 0.009 kW/inhabitant. Hence waste water will never be a relevant source of energy as long as our primary energy consumption is in the range of several kW/inhabitant. The annual mean primary power demand of conventional municipal waste water treatment with nutrient removal is in the range of 0.003-0.015 kW/inhabitant. In principle it is already possible to reduce this value for external energy supply to zero. Such plants should be connected to an electrical grid in order to keep investment costs low. Peak energy demand will be supported from the grid and surplus electric energy from the plant can be is fed to the grid. Zero 'carbon footprint' will not be affected by this solution. Energy minimisation must never negatively affect treatment efficiency because water quality conservation is more important for sustainable development than the possible reduction in energy demand. This argument is strongly supported by economical considerations as the fixed costs for waste water infrastructure are dominant.

  18. Estimation of Regional-Scale Actual Evapotranspiration in Okayama prefecture in Japan using Complementary Relationship

    Science.gov (United States)

    Moroizumi, T.; Yamamoto, M.; Miura, T.

    2008-12-01

    It is important to estimate accurately a water balance in watershed for proposing a reuse of water resources and a proper settlement of water utilization. Evapotranspiration (ET) is an important factor of water balance. Therefore, it is needed to estimate accurately the actual ET. The objective of this study is to estimate accurately monthly actual ET in Yoshii, Asahi, and Takahashi River watersheds in Okayama prefecture from 1999 to 2000. The monthly actual ET was calculated by a Morton and a modified Brutsaert and Stricker (B&S) method, using Automated Meteorological Data Acquisition Systems (AMeDAS) in the basin. The actual ET was estimated using land covers which were classified in 11 categories. The land covers includes the effects of albedo. The actual ET was related to the elevation at each AMeDAS station. Using this relationship, the actual ET at the 1 or 5 km grid-interval mesh in the basin was calculated, and finally, the distribution of actual ET was mapped. The monthly ET estimated by the modified B&S method were smaller than that by Morton method which showed a same tendency as the Penman potential ET (PET). The annual values of Morton"fs ET, modified B&S"fs ET, and PET were estimated as 796, 645, and 800 mm, respectively. The ET by the modified B&S was larger in hilly and mountainous areas than in settlement or city. In general, it was a reasonable result because city or settlement areas were covered with concrete and asphalt and the ET was controlled.

  19. Research and development of improved type radioactive waste volume reduction system

    International Nuclear Information System (INIS)

    Okamoto, Masahiro; Watanabe, Yoshifumi; Yamaoka, Katsuaki; Masaki, Tetsuo; Akagawa, Yoshihiro; Murakami, Tadashi; Miyake, Takashi.

    1985-01-01

    Development and research had been conducted since 1978 on an improved type radioactive waste volume reduction system incorporating calcining and incinerating fluidized bed type furnaces. This system can dispose of concentrated liquid wastes, combustible solid wastes, spent ion exchange resins and so forth by calcination or incineration to turn them into reduced-volume products. Recently a pilot test facility has constructed and tests has been conducted to demonstrate actual performance. Representative results of pilot tests are reported in this paper. (author)

  20. Design features of a full-scale high-level waste vitrification system

    International Nuclear Information System (INIS)

    Siemens, D.H.; Bonner, W.F.

    1976-08-01

    A system has been designed and is currently under construction for vitrification of commercial high-level waste. The process consists of a spray calciner coupled to an in-can melter. Due to the high radiation levels expected, this equipment is designed for totally remote operation and maintenance. The in-cell arrangement of this equipment has been developed cooperatively with a nuclear fuel reprocessor. The system will be demonstrated both full scale with nonradioactive simulated waste and pilot scale with actual high-level waste

  1. Landfill mining: Resource potential of Austrian landfills--Evaluation and quality assessment of recovered municipal solid waste by chemical analyses.

    Science.gov (United States)

    Wolfsberger, Tanja; Aldrian, Alexia; Sarc, Renato; Hermann, Robert; Höllen, Daniel; Budischowsky, Andreas; Zöscher, Andreas; Ragoßnig, Arne; Pomberger, Roland

    2015-11-01

    Since the need for raw materials in countries undergoing industrialisation (like China) is rising, the availability of metal and fossil fuel energy resources (like ores or coal) has changed in recent years. Landfill sites can contain considerable amounts of recyclables and energy-recoverable materials, therefore, landfill mining is an option for exploiting dumped secondary raw materials, saving primary sources. For the purposes of this article, two sanitary landfill sites have been chosen for obtaining actual data to determine the resource potential of Austrian landfills. To evaluate how pretreating waste before disposal affects the resource potential of landfills, the first landfill site has been selected because it has received untreated waste, whereas mechanically-biologically treated waste was dumped in the second. The scope of this investigation comprised: (1) waste characterisation by sorting analyses of recovered waste; and (2) chemical analyses of specific waste fractions for quality assessment regarding potential energy recovery by using it as solid recovered fuels. The content of eight heavy metals and the net calorific values were determined for the chemical characterisation tests. © The Author(s) 2015.

  2. Electrical resistivities of glass melts containing simulated SRP waste sludges

    International Nuclear Information System (INIS)

    Wiley, J.R.

    1978-08-01

    One option for the long-term management of radioactive waste at the Savannah River Plant is to solidify the waste in borosilicate glass by using a continuous, joule-heated, ceramic melter. Electrical resistivities that are needed for melter design were measured for melts of two borosilicate, glass-forming mixtures, each of which was combined with various amounts of several simulated-waste sludges. The simulated sludge spanned the composition range of actual sludges sampled from SRP waste tanks. Resistivities ranged from 6 to 10 ohm-cm at 500 0 C. Melt composition and temperature were correlated with resistivity. Resistivity was not a simple function of viscosity. 15 figures, 4 tables

  3. Gas generation from radiolytic attack of TRU-contaminated hydrogenous waste

    International Nuclear Information System (INIS)

    Zerwekh, A.

    1979-06-01

    In 1970, the Waste Management and Transportation Division of the Atomic Energy Commission ordered a segregation of transuranic (TRU)-contaminated solid wastes. Those below a contamination level of 10 nCi/g could still be buried; those above had to be stored retrievably for 20 y. The possibility that alpha-radiolysis of hydrogenous materials might produce toxic, corrosive, and flammable gases in retrievably stored waste prompted an investigation of gas identities and generation rates in the laboratory and field. Typical waste mixtures were synthesized and contaminated for laboratory experiments, and drums of actual TRU-contaminated waste were instrumented for field testing. Several levels of contamination were studied, as well as pressure, temperature, and moisture effects. G (gas) values were determined for various waste matrices, and degradation products were examined

  4. SWEPP PAN assay system uncertainty analysis: Passive mode measurements of graphite waste

    International Nuclear Information System (INIS)

    Blackwood, L.G.; Harker, Y.D.; Meachum, T.R.; Yoon, Woo Y.

    1997-07-01

    The Idaho National Engineering and Environmental Laboratory is being used as a temporary storage facility for transuranic waste generated by the U.S. Nuclear Weapons program at the Rocky Flats Plant (RFP) in Golden, Colorado. Currently, there is a large effort in progress to prepare to ship this waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Active Neutron (PAN) radioassay system. To this end a modified statistical sampling and verification approach has been developed to determine the total uncertainty of a PAN measurement. In this approach the total performance of the PAN nondestructive assay system is simulated using computer models of the assay system and the resultant output is compared with the known input to assess the total uncertainty. This paper is one of a series of reports quantifying the results of the uncertainty analysis of the PAN system measurements for specific waste types and measurement modes. In particular this report covers passive mode measurements of weapons grade plutonium-contaminated graphite molds contained in 208 liter drums (waste code 300). The validity of the simulation approach is verified by comparing simulated output against results from measurements using known plutonium sources and a surrogate graphite waste form drum. For actual graphite waste form conditions, a set of 50 cases covering a statistical sampling of the conditions exhibited in graphite wastes was compiled using a Latin hypercube statistical sampling approach

  5. High level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 6

    Energy Technology Data Exchange (ETDEWEB)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety.

  6. High level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 6

    International Nuclear Information System (INIS)

    1994-04-01

    The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety

  7. SELF-REALIZATION IS INSEPARABLE FROM SELF-ACTUALIZATION OF A PERSONALITY

    Directory of Open Access Journals (Sweden)

    Irina Ivanovna Vitkovskaya

    2016-02-01

    Full Text Available The article considers the issues of a personality’s self-realization and self-actualization under current educational standards required in the time of impetuous development in informative technologies. The practical aspect of the article describes self-realization and self-actualization of a person as their social and useful activity, interdiscipline approach, it is characterised by interdependence of an individual and their social environment apart with the self-realization of a personality in educational process. Evidently, being a substantial ca-tegory of vital «self-assertion», it should be thought over in categories of self-actualization and self-realization. So, self-realization is considered to be a social process which continues all conscientious life. In pedagogics of our country, labour and professional activity were the spheres of self-actualization therefore, it should be taken into account external factors for self-realization and self-actualization of a person. If socio-pedagogic conditions presuppose having good teachers, what conditions should be created in an individual’s micro-groups and in particular, in the family and kind of values shoud an individual get through the process of socialization. Thus, technological project of self-realization of a personality has to assume that there is the strive for self-perfection and interrelation of ways of self-realization, self-education, and adaptation. The correlation of these ways in a specific educational field is the purpose of the article.

  8. Geologic factors in nuclear waste disposal

    International Nuclear Information System (INIS)

    Towse, D.

    1978-07-01

    The study of geosciences and their relation to nuclear waste disposal and management entails analyzing the hydrology, chemistry, and geometry of the nuclear waste migration process. Hydrologic effects are determined by analyzing the porosity and permeability (natural and induced) of rock as well as pressures and gradients, dispersion, and aquifer length of the system. Chemistry parameters include radionuclide retardation factors and waste dissolution rate. Geometric parameters (i.e., parameters with dimension) evaluated include repository layer thickness, fracture zone area, tunnel length, and aquifer length. The above parameters act as natural barriers or controls to nuclear waste migration, and are evaluated in three potential geologic media: salt, shale, and crystalline rock deposits. Parametric values are assigned that correspond to many existing situations. These values, in addition to other important inputs, are lumped as a hydrology input into a computer simulation program used to model and calculate nuclear waste migration from the repository to the biosphere, and potential individual and population dose and radiation effects. These results are preliminary and show trends only; they do not represent an actual risk analysis

  9. Implementation of a management applied program for liquid radioactive waste treatment

    International Nuclear Information System (INIS)

    Lee, Y. H.; Ann, S. J.; Jo, H. S.; Son, J. S.

    2003-01-01

    A data collection of a liquid radioactive waste treatment process of a research organization became necessary while developing the RAWMIS(Radioactive Waste Management Integration System) which it can generate personal history management for efficient management of a waste, documents, all kinds of statistics. This paper introduces an input and output application program design to do to database with data in the results and a stream process of a treatment that analyzed the waste occurrence present situation and data by treatment process. Data on the actual treatment process that is not limited experiment improve by a document, human traces, saving of material resources and improve with efficiency of tracking about a radioactive waste and a process and give help to radioactive waste material valance and inventory study

  10. Summary: special waste form lysimeters - arid program

    International Nuclear Information System (INIS)

    Skaggs, R.L.; Walter, M.B.

    1987-01-01

    The purpose of the Special Waste Form Lysimeters - Arid Program is to determine the performance of solidified commercial low-level waste forms using a field-scale lysimeter facility constructed for measuring the release and migration of radionuclides from the waste forms. The performance of these waste forms, as measured by radionuclide concentrations in lysimeter effluent, will be compared to that predicted by laboratory characterization of the waste forms. Waste forms being tested include nuclear power reactor waste streams that have been solidified in cement, Dow polymer, and bitumen. To conduct the field leaching experiments a lysimeter facility was built to measure leachate under actual environmental conditions. Field-scale samples of waste were buried in lysimeters equipped to measure water balance components, effluent radionuclide concentrations, and to a limited extent, radionuclide concentrations in lysimeter soil samples. The waste forms are being characterized by standard laboratory leach tests to obtain estimates of radionuclide release. These estimates will be compared to leach rates observed in the field. Adsorption studies are being conducted to determine the amount of contaminant available for transport after the release. Theoretical solubility calculations will also be performed to investigate whether common solid phases could be controlling radionuclide release. 4 references, 8 figures, 1 table

  11. Hazardous Waste

    Science.gov (United States)

    ... chemicals can still harm human health and the environment. When you throw these substances away, they become hazardous waste. Some hazardous wastes come from products in our homes. Our garbage can include such hazardous wastes as old batteries, bug spray cans and paint thinner. U.S. residents ...

  12. VOC reduction technology deveolpment as part of the U.S. Department of Energy, Industrial Waste Reduction Program

    International Nuclear Information System (INIS)

    Cranford, B.

    1993-01-01

    A strong industry is vital to U.S. Economic health and prosperity, but U.S. industry is facing serious challenges both domestically and internationally. One of these challenges is the reduction of volatile organic compounds emissions from industrial processes and products. To assist industry with these challenges, the U.S. Department of Energy established the Industrial Waste Reduction Program to improve energy efficiency and competitiveness to private industry through cost-effective waste material reduction. This paper describes the programs and the use of joint partnerships between the Department of Energy, industry, national laboratories, universities and others, in developing technologies which reduce VOC emissions while improving energy efficiency. This paper also describes the process and selection criteria for participation in the program, and briefly describes the following five VOC reduction technologies under development: Dual Cure Coatings, Solvent Reduction through use of a No-clean Soldering Process, Solvent Waste Minimization by Supercritical CO 2 Cleaning Process, ethanol Recovery Process, and Membrane Vapor Recovery Systems. The VOC reductions as well as the energy savings and other benefits to the U.S. are discussed

  13. Technology for commercial radioactive waste management

    International Nuclear Information System (INIS)

    1979-05-01

    Conceptual facilities for interim storage of various treated transuranic (TRU) and gaseous wastes produced during fuel reprocessing and mixed oxide fuel fabrication are described in volume 3. Alternatives for interim storage of spent fuel prior to reprocessing or geologic isolation are also described. The storage concepts are based on available technology. They do not necessarily represent optimum designs, but are representative of what could be achieved with current capabilities. In actual applications it is reasonable to expect that there could be some improvements over these concepts, reflected in lower costs, lower environmental impacts, or both. These conceptual descriptions provide a reasonable basis for cost analysis and for development of estimates of environmental impacts. Sections are devoted to: storage of high-level liquid waste in large stainless steel tanks; two interim storage concepts for fuel residue waste (fuel hulls and hardware) waste storage; storage concepts for other nonhigh-level TRU waste; two alternatives for storage of solidified high-level waste; conceptual storage for large quantities of plutonium oxide; a concept for storing krypton gas cylinders; and alternatives for both short-term and extended storage of spent fuel

  14. 2011 Annual Industrial Wastewater Reuse Report for the Idaho National Laboratory Site's Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond

    Energy Technology Data Exchange (ETDEWEB)

    David Frederick

    2012-02-01

    This report describes conditions, as required by the state of Idaho Industrial Wastewater Reuse Permit (LA-000160-01), for the wastewater reuse site at the Idaho National Laboratory Site's Materials and Fuels Complex Industrial Waste Ditch and Industrial Waste Pond from November 1, 2010 through October 31, 2011. The report contains the following information: (1) Facility and system description; (2) Permit required effluent monitoring data and loading rates; (3) Groundwater monitoring data; (4) Status of special compliance conditions; and (5) Discussion of the facility's environmental impacts. During the 2011 reporting year, an estimated 6.99 million gallons of wastewater were discharged to the Industrial Waste Ditch and Pond which is well below the permit limit of 13 million gallons per year. Using the dissolved iron data, the concentrations of all permit-required analytes in the samples from the down gradient monitoring wells were below the Ground Water Quality Rule Primary and Secondary Constituent Standards.

  15. Chemical characterization, leach, and adsorption studies of solidified low-level wastes

    International Nuclear Information System (INIS)

    Walter, M.B.; Serne, R.J.; Jones, T.L.; McLaurine, S.B.

    1986-12-01

    Laboratory and field leaching experiments are beig conducted by Pacific Northwest Laboratory (PNL) to investigate the performance of solidified low-level nuclear waste in a typical, arid, near-surface disposal site. Under PNL's Special Waste Form Lysimeters-Arid Program, a field test facility was constructed to monitor the leaching of commercial solidified waste. Laboratory experiments were conducted to investigate the leaching and adsorption characteristics of the waste forms in contact with soil. Liquid radioactive wastes solidified in cement, vinyl ester-styrene, and bitumen were obtained from commercial boiling water and pressurized water reactors, and buried in a field leaching facility on the Hanford site in southeastern Washington State. Batch leaching, soil column adsorption, and soil/waste form column experiments were conducted in the laboratory, using small-scale cement waste forms and Hanford site ground water. The purpose of these experiments is to evaluate the ability of laboratory leaching tests to predict leaching under actual field conditions and to determine which mechanisms (i.e., diffusion, solubility, adsorption) actually control the concentration of radionuclides in the soil surrounding the waste form. Chemical and radionuclide analyses performed on samples collected from the field and laboratory experiments indicate strong adsorption of /sup 134,137/Cs and 85 Sr onto the Hanford site sediment. Small amounts of 60 Co are leached from the waste forms as very mobile species. Some 60 Co migrated through the soil at the same rate as water. Chemical constituents present in the reactor waste streams also found at elevated levels in the field and laboratory leachates include sodium, sulfate, magnesium, and nitrate. Plausible solid phases that could be controlling some of the chemical and radionuclide concentrations in the leachate were identified using the MINTEQ geochemical computer code

  16. Notas para The Waste Land: T. S. Eliot e a máquina literária = Notes on The Waste Land: T. S. Eliot and the literary machine

    Directory of Open Access Journals (Sweden)

    Cechinel, André

    2014-01-01

    Full Text Available Este artigo propõe-se a investigar o suposto papel explicativo atribuído às notas que T. S. Eliot acrescentou ao poema The Waste Land quando de sua publicação. Para tanto, o texto volta a sua atenção, em particular, para três das várias observações em que o poeta expõe as referências por ele consultadas e referidas nos versos, isso com o intuito de verificar se de fato há um esclarecimento interpretativo prestado ao poema graças à inclusão das notas. Em linhas gerais, o choque entre os versos de The Waste Land e as notas autorais provoca um impasse que aciona o que aqui chamamos de uma “máquina literária” produtora de teses. Essa máquina, compreendida a partir de seu mecanismo estruturante, não só recusa a sugestão de que o entendimento do poema decorre da decodificação apropriada das fontes, como também produz um efeito plural responsável por desestabilizar a hierarquia classicista e canônica que a crítica literária por décadas viu ali instalada

  17. Minimum Additive Waste Stabilization (MAWS), Phase I: Soil washing final report

    International Nuclear Information System (INIS)

    1995-08-01

    As a result of the U.S. Department of Energy's environmental restoration and technology development activities, GTS Duratek, Inc., and its subcontractors have demonstrated an integrated thermal waste treatment system at Fernald, OH, as part the Minimum Additive Waste Stabilization (MAWS) Program. Specifically, MAWS integrates soil washing, vitrification of mixed waste streams, and ion exchange to recycle and remediate process water to achieve, through a synergistic effect, a reduction in waste volume, increased waste loading, and production of a durable, leach-resistant, stable waste form suitable for disposal. This report summarizes the results of the demonstration/testing of the soil washing component of the MAWS system installed at Fernald (Plant 9). The soil washing system was designed to (1) process contaminated soil at a rate of 0.25 cubic yards per hour; (2) reduce overall waste volume and provide consistent-quality silica sand and contaminant concentrates as raw material for vitrification; and (3) release clean soil with uranium levels below 35 pCi/g. Volume reductions expected ranged from 50-80 percent; the actual volume reduction achieved during the demonstration reached 66.5 percent. The activity level of clean soil was reduced to as low as 6 pCi/g from an initial average soil activity level of 17.6 pCi/g (the highest initial level of soil provided for testing was 41 pCi/g). Although the throughput of the soil washing system was inconsistent throughout the testing period, the system was online for sufficient periods to conclude that a rate equivalent to 0.25 cubic yards per hour was achieved

  18. U.S. Navy Shipboard-Generated Plastic Waste Pilot Recycling Program

    Science.gov (United States)

    1991-03-01

    ship’s stores or vending machines often makes it way into "plastics only" containers located in work and berthing spaces. During the Escambia MRFsort...from Ships, or MARPOL). This legislation, which bans the disposal of plastics into the ocean, resulted from international and national pressures to halt...plastic by weight from a mixed, municipal waste stream produces a good product using the Er-1 extrusion technology. The Er-1 machine is a batch

  19. TRUEX partitioning from radioactive ICPP sodium bearing waste

    International Nuclear Information System (INIS)

    Herbst, R.S.; Brewer, K.N.; Tranter, T.J.; Todd, T.A.

    1995-03-01

    The Idaho Chemical Processing Plant (ICPP) located at the Idaho National Engineering Laboratory in Southeast Idaho is currently evaluating several treatment technologies applicable to waste streams generated over several decades of-nuclear fuel reprocessing. Liquid sodium bearing waste (SBW), generated primarily during decontamination activities, is one of the waste streams of interest. The TRansUranic EXtraction (TRUEX) process developed at Argonne National Laboratory is currently being evaluated to separate the actinides from SBW. On a mass basis, the amount of the radioactive species in SBW are low relative to inert matrix components. Thus, the advantage of separations is a dramatic decrease in resulting volumes of high activity waste (HAW) which must be dispositioned. Numerous studies conducted at the ICPP indicate the applicability of the TRUEX process has been demonstrated; however, these studies relied on a simulated SBW surrogate for the real waste. Consequently, a series of batch contacts were performed on samples of radioactive ICPP SBW taken from tank WM-185 to verify that actual waste would behave similarly to the simulated waste. The test results with SBW from tank WM-185 indicate the TRUEX solvent effectively extracts the actinides from the samples of actual waste. Gross alpha radioactivity, attributed predominantly to Pu and Am, was reduced from 3.14E+04 dps/mL to 1.46 dps/mL in three successive batch contacts with fresh TRUEX solvent. This reduction corresponds to a decontamination factor of DF = 20,000 or 99.995% removal of the gross a activity in the feed. The TRUEX solvent also extracted the matrix components Zr, Fe, and Hg to an appreciable extent (D Zr > 10, D Fe ∼ 2, D Hg ∼6). Iron co-extracted with the actinides can be successfully scrubbed from the organic with 0.2 M HNO 3 . Mercury can be selectively partitioned from the actinides with either sodium carbonate or nitric acid (≥ 5 M HNO 3 ) solutions

  20. Defense Waste Processing Facility (DWPF): The vitrification of high-level nuclear waste. (Latest citations from the Bibliographic database). Published Search

    International Nuclear Information System (INIS)

    1993-09-01

    The bibliography contains citations concerning a production-scale facility and the world's largest plant for the vitrification of high-level radioactive nuclear wastes (HLW) located in the United States. Initially based on the selection of borosilicate glass as the reference waste form, the citations present the history of the development including R ampersand D projects and the actual construction of the production facility at the DOE Savannah River Plant (SRP). (Contains a minimum of 177 citations and includes a subject term index and title list.)

  1. Hydrogen venting characteristics of commercial carbon-composite filters and applications to TRU waste

    International Nuclear Information System (INIS)

    Callis, E.L.; Marshall, R.S.; Cappis, J.H.

    1997-04-01

    The generation of hydrogen (by radiolysis) and of other potentially flammable gases in radioactive wastes which are in contact with hydrogenous materials is a source of concern, both from transportation and on-site storage considerations. Because very little experimental data on the generation and accumulation of hydrogen was available in actual waste materials, work was initiated to experimentally determine factors affecting the concentration of hydrogen in the waste containers, such as the hydrogen generation rate, (G-values) and the rate of loss of hydrogen through packaging and commercial filter-vents, including a new design suitable for plastic bags. This report deals only with the venting aspect of the problem. Hydrogen venting characteristics of two types of commercial carbon-composite filter-vents, and two types of PVC bag closures (heat-sealed and twist-and-tape) were measured. Techniques and equipment were developed to permit measurement of the hydrogen concentration in various layers of actual transuranic (TRU) waste packages, both with and without filter-vents. A test barrel was assembled containing known configuration and amounts of TRU wastes. Measurements of the hydrogen in the headspace verified a hydrogen release model developed by Benchmark Environmental Corporation. These data were used to calculate revised wattage Emits for TRU waste packages incorporating the new bag filter-vent

  2. Development of an accelerated leach test(s) for low-level waste forms

    International Nuclear Information System (INIS)

    Dougherty, D.R.; Fuhrmann, M.; Colombo, P.

    1985-01-01

    An accelerated leach test(s) is being developed to predict long-term leaching behavior of low-level radioactive waste (LLW) forms in their disposal environments. As necessary background, a literature survey of reported leaching mechanisms, available mathematical models and factors that affect leaching of LLW forms has been compiled. Mechanisms which have been identified include diffusion, dissolution, ion exchange, corrosion and surface effects. A computerized data base of LLW leaching data and mathematical models is being developed. The data is being used for model evaluation by curve fitting and statistical analysis according to standard procedures of statistical quality control. Long-term leach tests on portland cement, bitumen and vinyl ester-styrene (VES) polymer waste forms are underway which are designed to identify and evaluate factors that accelerate leaching without changing the mechanisms. Initial results on the effect of temperature on leachability indicate that the leach rates of cement and VES waste forms increase with increasing temperature, whereas, the leach rate of bitumen is little affected. 10 refs., 5 figs

  3. Waste isolation in geologic formations in the USA

    International Nuclear Information System (INIS)

    Zerby, C.D.; McClain, W.C.

    1976-01-01

    The ERDA program for the establishment of terminal storage facilities for commercial radioactive wastes in deep geologic formations was recently reorganized as the National Waste Terminal Storage (NWTS) program. General plans for implementing this expanded program call for geologic investigations and feasibility confirmation studies at multiple geographic locations, leading to pilot plant construction and operation with possible future conversion into a Federal Repository. The pilot plant operations will be experimental facilities having limited capacity to store actual waste in a readily retrievable configuration. The first two pilot plants are planned to start operations simultaneously in the mid-1980's. Geologic investigations are now in progress or planned in study areas of the interior basins of the Gulf Coast Salt Dome Province, in the Salina Salt basin and in the Paradox Basin in an effort to identify acceptable locations for these initial facilities. Subsequent pilot plants will be located in other formations. Preliminary geologic evaluations have been initiated in the Paleozoic shales and limestones, Triassic shale basins along the east coast, Mesozoic shales of the Gulf Coast and northern high plains and certain crystalline igneous rocks. Most of the required engineering testing of disposal in salt formations has been completed in previous programs. However, the establishment of pilot plants in the other rock types will require a sequence of in situ testing designed to develop the information necessary to both demonstrate the feasibility of waste disposal in that particular formation and provide the data for facility designs

  4. Developing a Sustainability Assessment Model to Analyze China’s Municipal Solid Waste Management Enhancement Strategy

    Directory of Open Access Journals (Sweden)

    Hua Li

    2015-01-01

    Full Text Available This study develops a sustainability assessment model for analysis and decision-making of the impact of China’s municipal solid waste management enhancement strategy options based on three waste treatment scenarios: landfill disposal, waste-to-energy incineration, and a combination of a material recovery facility and composting. The model employs life cycle assessment, health risk assessment, and full cost accounting to evaluate the treatment scenarios regarding safeguarding public health, protecting the environment and conserving resources, and economic feasibility. The model then uses an analytic hierarchy process for an overall appraisal of sustainability. Results suggest that a combination of material recovery and composting is the most efficient option. The study results clarify sustainable attributes, suitable predications, evaluation modeling, and stakeholder involvement issues in solid waste management. The demonstration of the use of sustainability assessment model (SAM provides flexibility by allowing assessment for a municipal solid waste management (MSWM strategy on a case-by-case basis, taking into account site-specific factors, therefore it has the potential for flexible applications in different communities/regions.

  5. Liquid secondary waste. Waste form formulation and qualification

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); King, W. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-01

    The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during Site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the Integrated Disposal Facility IDF). Waste form testing to support this plan is composed of work in the near term to demonstrate the waste form will provide data as input to a performance assessment (PA) for Hanford’s IDF.

  6. Radioactive Demonstration Of Final Mineralized Waste Forms For Hanford Waste Treatment Plant Secondary Waste By Fluidized Bed Steam Reforming Using The Bench Scale Reformer Platform

    International Nuclear Information System (INIS)

    Crawford, C.; Burket, P.; Cozzi, A.; Daniel, W.; Jantzen, C.; Missimer, D.

    2012-01-01

    The U.S. Department of Energy's Office of River Protection (ORP) is responsible for the retrieval, treatment, immobilization, and disposal of Hanford's tank waste. Currently there are approximately 56 million gallons of highly radioactive mixed wastes awaiting treatment. A key aspect of the River Protection Project (RPP) cleanup mission is to construct and operate the Waste Treatment and Immobilization Plant (WTP). The WTP will separate the tank waste into high-level and low-activity waste (LAW) fractions, both of which will subsequently be vitrified. The projected throughput capacity of the WTP LAW Vitrification Facility is insufficient to complete the RPP mission in the time frame required by the Hanford Federal Facility Agreement and Consent Order, also known as the Tri-Party Agreement (TPA), i.e. December 31, 2047. Therefore, Supplemental Treatment is required both to meet the TPA treatment requirements as well as to more cost effectively complete the tank waste treatment mission. In addition, the WTP LAW vitrification facility off-gas condensate known as WTP Secondary Waste (WTP-SW) will be generated and enriched in volatile components such as 137 Cs, 129 I, 99 Tc, Cl, F, and SO 4 that volatilize at the vitrification temperature of 1150 C in the absence of a continuous cold cap (that could minimize volatilization). The current waste disposal path for the WTP-SW is to process it through the Effluent Treatment Facility (ETF). Fluidized Bed Steam Reforming (FBSR) is being considered for immobilization of the ETF concentrate that would be generated by processing the WTP-SW. The focus of this current report is the WTP-SW. FBSR offers a moderate temperature (700-750 C) continuous method by which WTP-SW wastes can be processed irrespective of whether they contain organics, nitrates, sulfates/sulfides, chlorides, fluorides, volatile radionuclides or other aqueous components. The FBSR technology can process these wastes into a crystalline ceramic (mineral) waste form

  7. Mixed Waste Focus Area mercury contamination product line: An integrated approach to mercury waste treatment and disposal

    International Nuclear Information System (INIS)

    Hulet, G.A.; Conley, T.B.; Morris, M.I.

    1998-01-01

    The US Department of Energy (DOE) Mixed Waste Focus Area (MWFA) is tasked with ensuring that solutions are available for the mixed waste treatment problems of the DOE complex. During the MWFA's initial technical baseline development process, three of the top four technology deficiencies identified were related to the need for amalgamation, stabilization, and separation/removal technologies for the treatment of mercury and mercury-contaminated mixed waste. The focus area grouped mercury-waste-treatment activities into the mercury contamination product line under which development, demonstration, and deployment efforts are coordinated to provide tested technologies to meet the site needs. The Mercury Working Group (HgWG), a selected group of representatives from DOE sites with significant mercury waste inventories, is assisting the MWFA in soliciting, identifying, initiating, and managing efforts to address these areas. Based on the scope and magnitude of the mercury mixed waste problem, as defined by HgWG, solicitations and contract awards have been made to the private sector to demonstrate amalgamation and stabilization processes using actual mixed wastes. Development efforts are currently being funded under the product line that will address DOE's needs for separation/removal processes. This paper discusses the technology selection process, development activities, and the accomplishments of the MWFA to date through these various activities

  8. Characterization and reaction behavior of ferrocyanide simulants and Hanford Site high-level ferrocyanide waste

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Simpson, B.C.

    1994-02-01

    Nonradioactive waste simulants and initial ferrocyanide tank waste samples were characterized to assess potential safety concerns associated with ferrocyanide high-level radioactive waste stored at the Hanford Site in underground single-shell tanks (SSTs). Chemical, physical, thermodynamic, and reaction properties of the waste simulants were determined and compared to properties of initial samples of actual ferrocyanide wastes presently in the tanks. The simulants were shown to not support propagating reactions when subjected to a strong ignition source. The simulant with the greatest ferrocyanide concentration was shown to not support a propagating reaction that would involve surrounding waste because of its high water content. Evaluation of dried simulants indicated a concentration limit of about 14 wt% disodium mononickel ferrocyanide, below which propagating reactions could not occur in the ambient temperature bulk tank waste. For postulated localized hot spots where dried waste is postulated to be at an initial temperature of 130 C, a concentration limit of about 13 wt% disodium mononickel ferrocyanide was determined, below which propagating reactions could not occur. Analyses of initial samples of the presently stored ferrocyanide waste indicate that the waste tank ferrocyanide concentrations are considerably lower than the limit for propagation for dry waste and that the water content is near that of the as-prepared simulants. If the initial trend continues, it will be possible to show that runaway ferrocyanide reactions are not possible under present tank conditions. The lower ferrocyanide concentrations in actual tank waste may be due to tank waste mixing and/or degradation from radiolysis and/or hydrolysis, which may have occurred over approximately 35 years of storage

  9. Applications of thermal energy storage to waste heat recovery in the food processing industry

    Science.gov (United States)

    Wojnar, F.; Lunberg, W. L.

    1980-01-01

    A study to assess the potential for waste heat recovery in the food industry and to evaluate prospective waste heat recovery system concepts employing thermal energy storage was conducted. The study found that the recovery of waste heat in canning facilities can be performed in significant quantities using systems involving thermal energy storage that are both practical and economical. A demonstration project is proposed to determine actual waste heat recovery costs and benefits and to encourage system implementation by the food industry.

  10. Implementation of a management applied program for solid radioactive waste

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y. H.; Kim, T. K.; Kang, I. S.; Cho, H. S.; Son, J. S. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    Solid radioactive wastes are generated from the Post-irradiated Fuel Examination Facility, the Irradiated Material Examination Facility, the Research Reactor, and the laboratories at KAERI. A data collection of a solid radioactive waste treatment process of a research organization became necessary while developing the RAWMIS(Radioactive Waste Management Integration System) which it can generate personal history management for efficient management of a waste, documents, all kinds of statistics. This paper introduces an input and output application program design to do to database with data in the results and a stream process of a treatment that analyzed the waste occurrence present situation and data by treatment process. Data on the actual treatment process that is not limited experiment improve by a document, human traces, saving of material resources and improve with efficiency of tracking about a radioactive waste and a process and give help to radioactive waste material balance and inventory study.

  11. Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)

    Energy Technology Data Exchange (ETDEWEB)

    F. Hua; P. Pasupathi; N. Brown; K. Mon

    2005-09-19

    The safe disposal of radioactive waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels for plants, animals, and humans. All of the countries currently studying the options for disposing of high-level nuclear waste (HLW) have selected deep geologic formations to be the primary barrier for accomplishing this isolation. In U.S.A., the Nuclear Waste Policy Act of 1982 (as amended in 1987) designated Yucca Mountain in Nevada as the potential site to be characterized for high-level nuclear waste (HLW) disposal. Long-term containment of waste and subsequent slow release of radionuclides into the geosphere will rely on a system of natural and engineered barriers including a robust waste containment design. The waste package design consists of a highly corrosion resistant Ni-based Alloy 22 cylindrical barrier surrounding a Type 316 stainless steel inner structural vessel. The waste package is covered by a mailbox-shaped drip shield composed primarily of Ti Grade 7 with Ti Grade 24 structural support members. The U.S. Yucca Mountain Project has been studying and modeling the degradation issues of the relevant materials for some 20 years. This paper reviews the state-of-the-art understanding of the degradation processes based on the past 20 years studies on Yucca Mountain Project (YMP) materials degradation issues with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the 10,000 years regulatory period. This paper provides an overview of the current understanding of the likely degradation behavior of the waste package and drip shield in the repository after the permanent closure of the facility. The degradation scenario discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking and hydrogen induced

  12. Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)

    International Nuclear Information System (INIS)

    Hua, F.; Pasupathi, P.; Brown, N.; Mon, K.

    2005-01-01

    The safe disposal of radioactive waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels for plants, animals, and humans. All of the countries currently studying the options for disposing of high-level nuclear waste (HLW) have selected deep geologic formations to be the primary barrier for accomplishing this isolation. In U.S.A., the Nuclear Waste Policy Act of 1982 (as amended in 1987) designated Yucca Mountain in Nevada as the potential site to be characterized for high-level nuclear waste (HLW) disposal. Long-term containment of waste and subsequent slow release of radionuclides into the geosphere will rely on a system of natural and engineered barriers including a robust waste containment design. The waste package design consists of a highly corrosion resistant Ni-based Alloy 22 cylindrical barrier surrounding a Type 316 stainless steel inner structural vessel. The waste package is covered by a mailbox-shaped drip shield composed primarily of Ti Grade 7 with Ti Grade 24 structural support members. The U.S. Yucca Mountain Project has been studying and modeling the degradation issues of the relevant materials for some 20 years. This paper reviews the state-of-the-art understanding of the degradation processes based on the past 20 years studies on Yucca Mountain Project (YMP) materials degradation issues with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the 10,000 years regulatory period. This paper provides an overview of the current understanding of the likely degradation behavior of the waste package and drip shield in the repository after the permanent closure of the facility. The degradation scenario discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking and hydrogen induced

  13. Characterization of cement and bitumen waste forms containing simulated low-level waste incinerator ash

    International Nuclear Information System (INIS)

    Westsik, J.H. Jr.

    1984-08-01

    Incinerator ash from the combustion of general trash and ion exchange resins was immobilized in cement and bitumen. Tests were conducted on the resulting waste forms to provide a data base for the acceptability of actual low-level waste forms. The testing was done in accordance with the US Nuclear Regulatory Commission Technical Position on Waste Form. Bitumen had a measured compressive strength of 130 psi and a leachability index of 13 as measured with the ANS 16.1 leach test procedure. Cement demonstrated a compressive strength of 1400 psi and a leachability index of 7. Both waste forms easily exceed the minimum compressive strength of 50 psi and leachability index of 6 specified in the Technical Position. Irradiation to 10 8 Rad and exposure to 31 thermal cycles ranging from +60 0 ) to -30 0 C did not significantly impact these properties. Neither waste form supported bacterial or fungal growth as measured with ASTM G21 and G22 procedures. However, there is some indication of biodegradation due to co-metabolic processes. Concentration of organic complexants in leachates of the ash, cement and bitumen were too low to significantly affect the release of radionuclides from the waste forms. Neither bitumen nor cement containing incinerator ash caused any corrosion or degradation of potential container materials including steel, polyethylene and fiberglass. However, moist ash did cause corrosion of the steel

  14. Potential role of ABC-assisted repositories in U.S. plutonium and high-level waste disposition

    Energy Technology Data Exchange (ETDEWEB)

    Berwald, D.; Favale, A.; Myers, T. [Grumman Aerospace Corporation, Bethpage, NY (United States)] [and others

    1995-10-01

    This paper characterizes the issues involving deep geologic disposal of LWR spent fuel rods, then presents results of an investigation to quantify the potential role of Accelerator-Based Conversion (ABC) in an integrated national nuclear materials and high level waste disposition strategy. The investigation used the deep geological repository envisioned for Yucca Mt., Nevada as a baseline and considered complementary roles for integrated ABC transmutation systems. The results indicate that although a U.S. geologic waste repository will continue to be required, waste partitioning and accelerator transmutation of plutonium, the minor actinides, and selected long-lived fission products can result in the following substantial benefits: plutonium burndown to near zero levels, a dramatic reduction of the long term hazard associated with geologic repositories, an ability to place several-fold more high level nuclear waste in a single repository, electricity sales to compensate for capital and operating costs.

  15. How Wastes Influence Quality Management

    Directory of Open Access Journals (Sweden)

    Gratiela Dana BOCA

    2011-06-01

    Full Text Available Companies are often surprised to learn that only a fraction of their activities actually add value for their customers. A primary cause of waste is information deficits – employees simply lack the knowledge they need to do their jobs efficiently and effectively. This leads employees to waste valuable time and motion searching, waiting, retrieving, reworking or just plain future action. Companies are able to respond to changing customer desires with high variety, high quality, low cost, and with very fast throughput times. Eliminating waste along entire value streams, instead of at isolated points, creates processes that need less human effort, less space, less capital, and less time to make products and services at far less costs and with much fewer defects, compared with traditional business systems. Companies are able to respond to changing customer desires with high variety, high quality, low cost, and with very fast throughput times.

  16. Stabilization Using Phosphate Bonded Ceramics. Salt Containing Mixed Waste Treatment. Mixed Waste Focus Area. OST Reference No. 117

    International Nuclear Information System (INIS)

    1999-01-01

    Throughout the Department of Energy (DOE) complex there are large inventories of homogeneous mixed waste solids, such as wastewater treatment residues, fly ashes, and sludges that contain relatively high concentrations (greater than 15% by weight) of salts. The inherent solubility of salts (e.g., nitrates, chlorides, and sulfates) makes traditional treatment of these waste streams difficult, expensive, and challenging. One alternative is low-temperature stabilization by chemically bonded phosphate ceramics (CBPCs). The process involves reacting magnesium oxide with monopotassium phosphate with the salt waste to produce a dense monolith. The ceramic makes a strong environmental barrier, and the metals are converted to insoluble, low-leaching phosphate salts. The process has been tested on a variety of surrogates and actual mixed waste streams, including soils, wastewater, flyashes, and crushed debris. It has also been demonstrated at scales ranging from 5 to 55 gallons. In some applications, the CBPC technology provides higher waste loadings and a more durable salt waste form than the baseline method of cementitious grouting. Waste form test specimens were subjected to a variety of performance tests. Results of waste form performance testing concluded that CBPC forms made with salt wastes meet or exceed both RCRA and recommended Nuclear Regulatory Commission (NRC) low-level waste (LLW) disposal criteria. Application of a polymer coating to the CBPC may decrease the leaching of salt anions, but continued waste form evaluations are needed to fully assess the deteriorating effects of this leaching, if any, over time.

  17. Waste Characterization Methods

    Energy Technology Data Exchange (ETDEWEB)

    Vigil-Holterman, Luciana R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Naranjo, Felicia Danielle [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-02-02

    This report discusses ways to classify waste as outlined by LANL. Waste Generators must make a waste determination and characterize regulated waste by appropriate analytical testing or use of acceptable knowledge (AK). Use of AK for characterization requires several source documents. Waste characterization documentation must be accurate, sufficient, and current (i.e., updated); relevant and traceable to the waste stream’s generation, characterization, and management; and not merely a list of information sources.

  18. Waste Characterization Methods

    International Nuclear Information System (INIS)

    Vigil-Holterman, Luciana R.; Naranjo, Felicia Danielle

    2016-01-01

    This report discusses ways to classify waste as outlined by LANL. Waste Generators must make a waste determination and characterize regulated waste by appropriate analytical testing or use of acceptable knowledge (AK). Use of AK for characterization requires several source documents. Waste characterization documentation must be accurate, sufficient, and current (i.e., updated); relevant and traceable to the waste stream's generation, characterization, and management; and not merely a list of information sources.

  19. Status of Household Solid Waste and People’s Perception on its Management at Sitakunda Upazila, Chittagong

    Directory of Open Access Journals (Sweden)

    Md. Juwel Rana

    2017-02-01

    Full Text Available Urban household solid waste (HSW is well-known, but has little knowledge about the rural areas and sub-cities. The quest of the study was to investigate current HSW management system and people’s attitude towards it’s at Sitakunda Upazila. The study was conducted through a semi-structured questionnaire survey by interviewing 150 households from 3 different socioeconomic groups. The results showed that on an average each household generated 1.26 kg/day wastes which stands at 0.24 kg/person/day in the study area. However, HSW generation is positively correlated with family size (rxy=0.991, p˂0.05, average age of family members (rxy=0.455, p˂0.01, and monthly earnings (rxy=0.999, p˂0.01 of the households. Amidst the various categories of wastes, vegetable and food wastes (VFWs were identified as the highest value 68.4%. In contrast, 24.6% of the generated wastes were recyclable and 75.4% non-recyclable; 82.1% organic and 17.9% inorganic. A large percentage of sampled households opined that present HSW management involves merely partial collection that is also irregular in urban areas but absent in rural areas. Unfair HSW collection was noticed by 58.7% and partial source segregation is also rare 14%. Besides, 42.7% people were found to dispose their daily generated wastes by open dumping and 25.3% nearby lagoon. A large number of people were dissatisfied 68% and only 6.7% were satisfied with the existing HSW management system. In addition, most of the people encountered bad odor problems by 28.7% and wastes spread on roadside by 38%. Significant quantity of recyclable wastes was noticed to collect from households by hawkers/scraps in the study area and very few household were found to use VFWs with cow-dung to generate biogas.INTERNATIONAL JOURNAL OF ENVIRONMENT Volume-6, Issue-1, Dec-Feb 2016/17, page: 58-75

  20. Colloidal agglomerates in tank sludge and their impact on waste processing

    International Nuclear Information System (INIS)

    Tingey, J.M.; Bunker, B.C.; Graff, G.L.; Keefer, K.D.; Lea, A.S.; Rector, D.R.

    1999-01-01

    Disposal of millions of gallons of existing radioactive wastes in underground storage tanks is a major remediation activity for the US Department of Energy. These wastes include a substantial volume of insoluble sludges consisting of submicron colloidal particles. Processing these sludges under the proposed processing conditions presents unique challenges in retrieval transport, separation, and solidification of these waste streams. Depending on processing conditions, these colloidal particles can form agglomerated networks having high viscosities that could clog transfer lines or produce high volumes of low-density sediments that interfere with solid-liquid separations. Under different conditions, these particles can be dispersed to form very fine suspended particles that do not settle. Given the wide range of waste chemistries present at Department of Energy sites, it is impractical to measure the properties of all treatment procedures. Under the current research activities, the underlying principles of colloid chemistry and physics are being studied to predict and eventually control the physical properties of sludge suspensions and sediment layers in tank wastes and other waste processing streams. Proposed tank processing strategies include retrieval transport, and solid-liquid separations in basic (pH 10 to 14), high ionic strength (0.1 to 1.0 M) salt solutions. The effect of salt concentration, ionic strength, and salt composition on the physical properties such as viscosity, agglomerate size, and sedimentation of model suspensions containing mixtures of one or two of the major components found in actual wastes have been measured to understand how agglomeration influences processing. Property models developed from theory and experiment on these simple suspensions are then applied to explain the results obtained on actual wastes

  1. Evaluation Of Supplemental Pre-Treatment Development Requirements To Meet TRL 6: Rotary Microfiltration

    International Nuclear Information System (INIS)

    Huber, H.J.

    2011-01-01

    In spring 2011, the Technology Maturation Plan (TMP) for the Supplemental Treatment Project (RPP-PLAN-49827, Rev. 0), Technology Maturation Plan for the Treatment Project (T4S01) was developed. This plan contains all identified actions required to reach technical maturity for a field-deployable waste feed pretreatment system. The supplemental pretreatment system has a filtration and a Cs-removal component. Subsequent to issuance of the TMP, rotary microfiltration (RMF) has been identified as the prime filtration technology for this application. The prime Cs-removal technology is small column ion exchange (ScIX) using spherical resorcinol formaldehyde (sRF) as the exchange resin. During fiscal year 2011 (FY2011) some of the tasks identified in the TMP have been completed. As of September 2011, the conceptual design package has been submitted to DOE as part of the critical decision (CD-1) process. This document describes the remaining tasks identified in the TMP to reach technical maturity and evaluates the validity of the proposed tests to fill the gaps as previously identified in the TMP. The potential vulnerabilities are presented and the completed list of criteria for the DOE guide DOE G 413.3-4 different technology readiness levels are added in an attachment. This evaluation has been conducted from a technology development perspective - all programmatic and manufacturing aspects were excluded from this exercise. Compliance with the DOE G 413.3-4 programmatic and manufacturing requirements will be addressed directly by the Treatment Project during the course of engineering design. The results of this evaluation show that completion of the proposed development tasks in the TMP are sufficient to reach TRL 6 from a technological point of view. The tasks involve actual waste tests using the current baseline configuration (2nd generation disks, 40 psi differential pressure, 30 C feed temperature) and three different simulants - the PEP, an AP-Farm and an S-saltcake

  2. Portrayals in Print: Media Depictions of the Informal Sector’s Involvement in Managing E-Waste in India

    Directory of Open Access Journals (Sweden)

    Verena Radulovic

    2018-03-01

    Full Text Available For over a decade, media stories have exposed health and environmental harm caused by informal electronics recycling in less industrialized countries. Greater awareness of these risks helped inform regulations across the globe and the development of recycling standards. Yet, media depictions also shape public perceptions of informal workers and their role in handling electronic waste, or e-waste. This paper examines how mainstream print media describes the informal sector’s involvement in handling e-waste in India, especially as policymakers and other stakeholders currently grapple with how to integrate informal workers into formal, more transparent e-waste management schemes. This study evaluates depictions of the informal sector in print articles from both non-Indian and Indian news media outlets, employing controversy mapping principles and digital research tools. Findings may help inform stakeholder agendas seeking to influence public awareness on how to integrate informal workers into viable e-waste management solutions. Subsequent research based on these results could also help stakeholders understand the actors and networks that shape such media depictions. Results from the dataset show that most news articles describe informal workers negatively or problematically due to activities causing health risks and environmental damage, but usually do not discern which activities in the value chain (e.g., collection, dismantling, metals extraction represent the greatest risks. Comparatively fewer articles portray informal workers positively or as contributing to e-waste solutions. Most articles also do not explain challenges that arise when working with informal workers. As such, media depictions today often lag behind policy debates and obscure multiple facets—good and bad—of the informal sector’s involvement in managing e-waste. Thus, an opportunity exists for policymakers, manufacturers, and advocacy groups to bridge the gap between

  3. Shifting strategies and precarious progress: Nuclear waste management in Canada

    International Nuclear Information System (INIS)

    Ramana, M.V.

    2013-01-01

    Canada has a lengthy history of trying to find a path for dealing with radioactive spent fuel and nuclear waste from its nuclear reactors. In the last decade, it has taken major strides towards this goal by evolving a process through which a site for a geological repository to sequester nuclear waste is to be selected. The Canadian Nuclear Waste Management Organization (NWMO) is in the early stages of the process of finding a community that is willing to host such a repository. Differences between the broad principles underlying siting and the processes for actually selecting the site have emerged as the NWMO proceeds with engaging local governments and specific communities. These differences and other conflicts, especially over new nuclear reactor construction, might pose hurdles in the path of successfully setting up a repository. - Highlights: • Canada has set up a process for siting a geological repository for nuclear waste. • The current challenge is to find a community willing to host such a repository. • Authorities are luring communities with the promise of jobs and local investment. • Potential new nuclear reactor construction might become a locus of conflict. • Success in actually setting up a repository is by no means guaranteed

  4. Bioprocessing of a stored mixed liquid waste

    Energy Technology Data Exchange (ETDEWEB)

    Wolfram, J.H.; Rogers, R.D. [Idaho National Engineering Lab., Idaho Falls, ID (United States); Finney, R. [Mound Applied Technologies, Miamisburg, OH (United States)] [and others

    1995-12-31

    This paper describes the development and results of a demonstration for a continuous bioprocess for mixed waste treatment. A key element of the process is an unique microbial strain which tolerates high levels of aromatic solvents and surfactants. This microorganism is the biocatalysis of the continuous flow system designed for the processing of stored liquid scintillation wastes. During the past year a process demonstration has been conducted on commercial formulation of liquid scintillation cocktails (LSC). Based on data obtained from this demonstration, the Ohio EPA granted the Mound Applied Technologies Lab a treatability permit allowing the limited processing of actual mixed waste. Since August 1994, the system has been successfully processing stored, {open_quotes}hot{close_quotes} LSC waste. The initial LSC waste fed into the system contained 11% pseudocumene and detectable quantities of plutonium. Another treated waste stream contained pseudocumene and tritium. Data from this initial work shows that the hazardous organic solvent, and pseudocumene have been removed due to processing, leaving the aqueous low level radioactive waste. Results to date have shown that living cells are not affected by the dissolved plutonium and that 95% of the plutonium was sorbed to the biomass. This paper discusses the bioprocess, rates of processing, effluent, and the implications of bioprocessing for mixed waste management.

  5. Detection of waste's layers with geoelectric methods: case of Relleno Sanitario Rio Azul, Costa Rica

    International Nuclear Information System (INIS)

    Suarez, Jorge; Zuniga, Hector

    2013-01-01

    Areas of greatest potential leachate storage are identified making 25 VES (Vertical Electrical Soundings), whose effective exploration depth was around 65 meters. The proposed geophysical model consists of four layers: the first corresponds to a coverage of compacted material (20 Ωm - 54 Ωm), the second layer contains unsaturated waste (11 Ωm - 18 Ωm), the third (1 Ωm - 5.4 Ωm) is saturated waste with leachate and the fourth (49 Ωm - 360 Ωm) the basement on which was installed the landfill. Therefore, the geoelectric allowed determining the thickness and distribution of covering materials, waste, saturated waste with leachates and depth of the basement's top. (author) [es

  6. Conversion of bakery wastes to ethanol

    Energy Technology Data Exchange (ETDEWEB)

    Martin, J.E.

    1984-10-04

    The project had the following goals: (1) determine the actual yields of ethanol from the waste products of major bakeries in the Denver area by distillation in a laboratory bench scale distillation column; (2) determine the expected yields of ethanol from the major types of wastes produced by the large bakeries in the Denver area; (3) increase alcohol yields to the maximum possible by identifying any inhibitors and taking steps to remove them; (4) purify the distilled alcohol if necessary to remove any by-products distilled off with the alcohol. The production of alcohol from bakery wastes is feasible. The average yield of the bread type products is 19.8% by weight with yields ranging up to 25%. In other words, on the average, for every 1000 pounds of waste, 198 pounds or 30 gallons of alcohol would be produced. This estimate is conservative since a larger facility would tend to get a better yield (due to the difficulties of getting all the alcohol out of the small batches run). The major variable appeared to be the yeast. Fresh yeast should always be used. No yeast nutrients were tried in these experiments, since the yeast seemed to grow well. However, this could be an area of further investigation. It is possible that the yields could be kept consistently high by providing the yeast with nutrients. Finally, contamination of the alcohol with oils can be a problem, although not necessarily a very significant one. Methods do exist to remove the oil during the actual distillation, as well as, before distillation. Careful distillation also tends to lessen the problem. 7 references, 4 figures, 2 tables.

  7. Ferrocyanide tank waste stability

    International Nuclear Information System (INIS)

    Fowler, K.D.

    1993-01-01

    Ferrocyanide wastes were generated at the Hanford Site during the mid to late 1950s as a result of efforts to create more tank space for the storage of high-level nuclear waste. The ferrocyanide process was developed to remove 137 CS from existing waste and newly generated waste that resulted from the recovery of valuable uranium in Hanford Site waste tanks. During the course of research associated with the ferrocyanide process, it was recognized that ferrocyanide materials, when mixed with sodium nitrate and/or sodium nitrite, were capable of violent exothermic reaction. This chemical reactivity became an issue in the 1980s, when safety issues associated with the storage of ferrocyanide wastes in Hanford Site tanks became prominent. These safety issues heightened in the late 1980s and led to the current scrutiny of the safety issues associated with these wastes, as well as current research and waste management programs. Testing to provide information on the nature of possible tank reactions is ongoing. This document supplements the information presented in Summary of Single-Shell Tank Waste Stability, WHC-EP-0347, March 1991 (Borsheim and Kirch 1991), which evaluated several issues. This supplement only considers information particular to ferrocyanide wastes

  8. Suspicions about the practice of radioactive waste treatment

    International Nuclear Information System (INIS)

    Merz, E.

    1988-01-01

    Two quite distinct events associated with the treatment and disposal of radioactive waste have created a turmoil in the past twelve months. In the discussions about the Hanau nuclear fuel industries, the pressure buildup in conditioned waste drums, which had been known for two years and had meanwhile been investigated in various places, was mixed up with the irregularities at Transnuklear either out of ignorance or intentionally. In the public, this was bound to create the impression as if the ballooning of a large number of waste drums as a result of gas evolution were connected with the goings-on at Transnuklear. Actually, however, these are two absolutely separate issues. (orig.) [de

  9. U.S. Nuclear Regulatory Commission Process for Risk-Informing the Nuclear Waste Arena

    International Nuclear Information System (INIS)

    Leslie, B. W.

    2003-01-01

    The U.S. Nuclear Regulatory Commission (NRC) is increasing the use of risk insights and information in its regulation of nuclear materials and waste. The objective of this risk-informed regulatory effort is to improve the effectiveness and efficiency of the agency, while maintaining or increasing its focus on safety. The agency's Office of Nuclear Material Safety and Safeguards (NMSS) proposed a five-step process to carry out a framework for increasing the use of risk information and insights in its regulation of nuclear materials and waste. The office is carrying out the five-step process to risk-inform the nuclear materials and waste arenas. NMSS's actions included forming a Risk Task Group and the use of case studies to test and complete screening criteria for identifying candidate regulatory applications amenable for risk-informing. Other actions included involving stakeholders through enhanced public participation, developing safety goals for materials and waste regulatory applications, and establishing a risk training program for staff. Through the case studies, NRC staff found the draft screening criteria to be effective in deciding regulatory areas that may be amenable to an increased use of risk insights. NRC staff also found that risk information may have the potential to reduce regulatory burden and improve staff's efficiency in making decisions, while maintaining safety. Finally, staff found that it would be possible to develop safety goals for the nuclear materials and waste arenas

  10. Characterization and durability testing of a glass-bonded ceramic waste form

    International Nuclear Information System (INIS)

    Johnson, S. G.

    1998-01-01

    Argonne National Laboratory is developing a glass bonded ceramic waste form for encapsulating the fission products and transuranics from the conditioning of metallic reactor fuel. This waste form is currently being scaled to the multi-kilogram size for encapsulation of actual high level waste. This paper will present characterization and durability testing of the ceramic waste form. An emphasis on results from application of glass durability tests such as the Product Consistency Test and characterization methods such as X-ray diffraction and scanning electron microscopy. The information presented is based on a suite of tests utilized for assessing product quality during scale-up and parametric testing

  11. Final Report for X-ray Diffraction Sample Preparation Method Development

    Energy Technology Data Exchange (ETDEWEB)

    Ely, T. M. [Hanford Site (HNF), Richland, WA (United States); Meznarich, H. K. [Hanford Site (HNF), Richland, WA (United States); Valero, T. [Hanford Site (HNF), Richland, WA (United States)

    2018-01-30

    WRPS-1500790, “X-ray Diffraction Saltcake Sample Preparation Method Development Plan/Procedure,” was originally prepared with the intent of improving the specimen preparation methodology used to generate saltcake specimens suitable for XRD-based solid phase characterization. At the time that this test plan document was originally developed, packed powder in cavity supports with collodion binder was the established XRD specimen preparation method. An alternate specimen preparation method less vulnerable, if not completely invulnerable to preferred orientation effects, was desired as a replacement for the method.

  12. The disposal of Canada`s nuclear fuel waste: a study of postclosure safety of in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock volume 1: summary

    Energy Technology Data Exchange (ETDEWEB)

    Wikjord, A G; Baumgartner, P; Johnson, L H; Stanchell, F W; Zach, R; Goodwin, B W

    1996-06-01

    The concept for disposal of Canada`s nuclear fuel waste involves isolating the waste in corrosion-resistant containers emplaced and sealed within a vault at a depth of 500 to 1000 m in plutonic rock of the Canadian Shield. The case for the acceptability of the concept as a means of safely disposing of Canada`s nuclear fuel waste is presented in an Environmental Impact Statement (EIS) The disposal concept permits a choice of methods, materials, site locations and designs. The EIS presents a case study of the long-term (i.e., postclosure) performance of a hypothetical implementation of the concept, referred to in this report as the reference disposal system. The reference disposal system is based on borehole emplacement of used CANDU fuel in Grade-2 titanium alloy containers in low-permeability, sparsely fractured plutonic rock of the Canadian Shield. We evaluate the long-term performance of another hypothetical implementation of the concept based on in-room emplacement of used CANDU fuel in copper containers in permeable plutonic rock. The geological characteristics of the geosphere assumed for this study result in short groundwater travel times from the disposal vault to the surface. In the present study, the principal barrier to the movement of contaminants is the long-lasting copper container. We show that the long-lasting container can effectively compensate for a permeable host rock which results in an unfavourable groundwater flow condition. These studies illustrate the flexibility of AECL`s disposal concept to take advantage of the retention, delay, dispersion, dilution and radioactive decay of contaminants in a system of natural barriers provided by the geosphere and hydrosphere and of engineered barriers provided by the waste form, container, buffer, backfills, other vault seals and grouts. In an actual implementation, the engineered system would be designed for the geological conditions encountered at the host site. 34 refs., 2 tabs., 11 figs.

  13. Methodology for completing Hanford 200 Area tank waste physical/chemical profile estimations

    International Nuclear Information System (INIS)

    Kruger, A.A.

    1996-01-01

    The purpose of the Methodology for Completing Hanford 200 Area Tank Waste Physical/Chemical Profile Estimations is to capture the logic inherent to completing 200 Area waste tank physical and chemical profile estimates. Since there has been good correlation between the estimate profiles and actual conditions during sampling and sub-segment analysis, it is worthwhile to document the current estimate methodology

  14. Decommissioning of evaporation technology for processing liquid radioactive waste in UJV Rez, a. s

    International Nuclear Information System (INIS)

    Tous, M.; Otcovsky, T.; Podlaha, J.

    2015-01-01

    The UJV Rez, a. s. is the main leader in processing institutional radioactive waste (RAW) in the Czech Republic and the Waste Management Department has been established since the research reactor VVR-S (now LVR-15) was put in operation. Due to the large activities in nuclear research and engineering in the past, a big capacity of waste management technologies was needed. The low pressure compactor for volume reduction of solid RAW, as well as chemical pre-treatment technology of liquid RAW were installed and later the evaporation technology for effective processing the liquid RAW with the cementation and bituminization unit for final conditioning of concentrated liquid RAW were used. During the years of research reactor operation and research activities in UJV Rez, a. s. there were two installed evaporation technologies in row. After the latest evaporator lifetime, changes in liquid RAW production and together with higher decontamination factor requirements, this technology was decided to be decommissioned. The decommissioned evaporation technology was installed and put in operation in 1991. This technology was used for processing liquid aqueous RAW produced from internal research activities and of course for external producers and institutions (e.g. universities, medicine, research institutes, industry). The approved decommissioning plan was prepared and the licence for immediate decommissioning was obtained in 2012. Then the decommissioning project started. The preparing stages as dosimetric survey, expected material balance and of course initial decontamination activities were performed. Evaporation technology dismantling and processing the arising RAW were done by the internal staff of Waste Management Department. The total volume of produced RAW was 49,5 m 3 of RAW. The secondary liquid RAW (from decontamination) of amount 1,4 m 3 , contaminated sludge of amount 0,5 m 3 , solid RAW (construction steel) of amount 39,1 m 3 , solid compressible RAW (protective

  15. Dryout modeling in support of the organic tank safety project

    International Nuclear Information System (INIS)

    Simmons, C.S.

    1998-08-01

    This work was performed for the Organic Tank Safety Project to evaluate the moisture condition of the waste surface organic-nitrate bearing tanks that are classified as being conditionally safe because sufficient water is present. This report describes the predictive modeling procedure used to predict the moisture content of waste in the future, after it has been subjected to dryout caused by water vapor loss through passive ventilation. This report describes a simplified procedure for modeling the drying out of tank waste. Dryout occurs as moisture evaporates from the waste into the headspace and then exits the tank through ventilation. The water vapor concentration within the waste of the headspace is determined by the vapor-liquid equilibrium, which depends on the waste's moisture content and temperature. This equilibrium has been measured experimentally for a variety of waste samples and is described by a curve called the water vapor partial pressure isotherm. This curve describes the lowering of the partial pressure of water vapor in equilibrium with the waste relative to pure water due to the waste's chemical composition and hygroscopic nature. Saltcake and sludge are described by two distinct calculations that emphasize the particular physical behavior or each. A simple, steady-state model is devised for each type to obtain the approximate drying behavior. The report shows the application of the model to Tanks AX-102, C-104, and U-105

  16. In situ vitrification of a mixed-waste contaminated soil site: The 116-B-6A crib at Hanford

    Energy Technology Data Exchange (ETDEWEB)

    Luey, J.; Koegler, S.S.; Kuhn, W.L.; Lowery, P.S.; Winkelman, R.G.

    1992-09-01

    The first large-scale mixed-waste test of in situ vitrification (ISV) has been completed. The large-scale test was conducted at an actual contaminated soil site, the 116-B-6A crib, on the Department of Energy's Hanford Site. The large-scale test was a demonstration of the ISV technology and not an interim action for the 116-B-6A crib. This demonstration has provided technical data to evaluate the ISV process for its potential in the final disposition of mixed-waste contaminated soil sites at Hanford. Because of the test's successful completion. technical data on the vitrified soil are available on how well the process incorporates transuranics and heavy metals into the waste form. how well the form resists leaching of transuranics and heavy metals. how well the process handles sites with high combustible loadings, and the important site parameters which may affect the achievable process depth. This report describes the 116-B-6A crib site, the objectives of the ISV demonstration, the results in terms of the objectives, and the overall process performance.

  17. Minimum Additive Waste Stabilization (MAWS)

    International Nuclear Information System (INIS)

    1994-02-01

    In the Minimum Additive Waste Stabilization(MAWS) concept, actual waste streams are utilized as additive resources for vitrification, which may contain the basic components (glass formers and fluxes) for making a suitable glass or glassy slag. If too much glass former is present, then the melt viscosity or temperature will be too high for processing; while if there is too much flux, then the durability may suffer. Therefore, there are optimum combinations of these two important classes of constituents depending on the criteria required. The challenge is to combine these resources in such a way that minimizes the use of non-waste additives yet yields a processable and durable final waste form for disposal. The benefit to this approach is that the volume of the final waste form is minimized (waste loading maximized) since little or no additives are used and vitrification itself results in volume reduction through evaporation of water, combustion of organics, and compaction of the solids into a non-porous glass. This implies a significant reduction in disposal costs due to volume reduction alone, and minimizes future risks/costs due to the long term durability and leach resistance of glass. This is accomplished by using integrated systems that are both cost-effective and produce an environmentally sound waste form for disposal. individual component technologies may include: vitrification; thermal destruction; soil washing; gas scrubbing/filtration; and, ion-exchange wastewater treatment. The particular combination of technologies will depend on the waste streams to be treated. At the heart of MAWS is vitrification technology, which incorporates all primary and secondary waste streams into a final, long-term, stabilized glass wasteform. The integrated technology approach, and view of waste streams as resources, is innovative yet practical to cost effectively treat a broad range of DOE mixed and low-level wastes

  18. Energetical and economical assessment of the waste heat problem

    International Nuclear Information System (INIS)

    Demicheli, U.; Voort, E. van der; Schneiders, A.; Zegers, P.

    1977-01-01

    Electrical power plants produce large quantities of low grade heat that remain unused. For ecological reasons this waste heat must be dispersed by means of expensive cooling devices. Waste heat could be used in acquacultural and agricultural complexes this replacing large amounts of primary energy. Energetical and economical aspects are discussed. The state of the art of these and other utilisations is outlined. A different approach to the problem is to reduce the production of waste heat. Various strategies to achieve this challenge are outlined and their actual state and possible future developments are discussed. Finally, the various most promising utilizations are examined from an energetical point of view

  19. China's current status and long-term outlook of nuclear power and radioactive waste disposal management

    International Nuclear Information System (INIS)

    Li, Zhidong

    2015-01-01

    This study identified the current status and long-term outlook of China's nuclear power development and radioactive waste disposal management after the 3.11 FUKUSHIMA accidents. China strengthened the actions for achieving nuclear power safety and cost efficiency as well as safety management of radioactive waste. It is a hard work to expand the capacity to 58 GW, the governmental target in 2020. The long-term development will strongly depend on the progress in safety management of nuclear power and radioactive waste and economic competitiveness. (author)

  20. Characterization of off-gases from a small-scale, joule-heated ceramic melter for nuclear waste vitrification

    International Nuclear Information System (INIS)

    Woolsey, G.B.; Wilhite, E.L.

    1980-01-01

    This paper confirmed with actual nuclear waste the thermodynamic predictions of the fate of some of the semivolatiles in off-gas. Ruthenium behaves erratically and it is postulated that it migrates as a finely divided solid, rather than as a volatile oxide. Provisions for handling these waste off-gasses will be incorporated in the design of facilities for vitrifying SRP waste

  1. U.S. Bureau of Mines, phase I Hanford low-level waste melter tests: Melter offgas report

    International Nuclear Information System (INIS)

    Eaton, W.C.

    1995-01-01

    A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Phase 1 of the melter demonstration tests using simulated LLW was completed during fiscal year 1995. This document is the melter offgas report on testing performed by the U.S. Department of the Interior, Bureau of Mines, Albany Research Center in Albany, Oregon. The Bureau of Mines (one of the seven vendors selected) was chosen to demonstrate carbon electrode melter technology (also called carbon arc or electric arc) under WHC subcontract number MMI-SVV-384216. The document contains the complete offgas report for the first 24-hour melter test (WHC-1) as prepared by Entropy Inc. A summary of this report is also contained in the''U.S. Bureau of Mines, Phase 1 Hanford Low-Level Waste Melter Tests: Final Report'' (WHC-SD-WM-VI-030)

  2. Comparative environmental evaluation of three different waste treating methods for bio-degradable organic waste and residual waste from the six municipalities cooperating in Biogasanlaeg I/S and I/S Nordforbraending

    International Nuclear Information System (INIS)

    Cour Jansen, J. la.

    1997-03-01

    A comparative environmental evaluation has been made for three systems for processing household refuse. The first system (1) comprises thermofilic treatment of the organic waste. The inorganic parts of the waste from both systems is combusted in a cogeneration plant. The third system (3) is a cogeneration plant combusting all the wastes. The most essential issue in comparing the three waste processing systems is the possibility to use the residual products from the bio-gasification system (1) and the composting system in the agriculture as fertilizers. Without such a use it is estimated that the systems (1) and (2) will be very unfeasible, both environmentally and economically. The contents of heavy metals, organic chemicals, and plasticizers in the residual products are at present above the Danish threshold values. This problem can only be solved by reducing the contents of these pollutants in the household wastes. Based on these findings it is therefore very doubtful whether the residual waste products from bio-gasification and probably also from composting can be used in the agriculture in the future. However, it is for the other parts of the environmental analysis assumed that the problem will eventually be solved. Differences in resource utilization in the waste processing are very small in the three systems. When comparing the net energy production system (1) and (3) show the best results. The three systems are almost equal related to emissions of HC1, NO x , and SO 2 , the composting system performing the best. The three systems emit the same amount of CO 2 and other greenhouse gases. Composting, however, results in a smaller energy susbstituion and is therefore marginally a poorer solution. For reducing the heavy metals and plasticizers in the environment combustion of the organic wastes is the best system. (LN)

  3. Development of a universal solvent for the decontamination of acidic liquid radioactive wastes

    Science.gov (United States)

    Todd, T. A.; Brewer, K. N.; Law, J. D.; Wood, D. J.; Herbest, R. S.; Romanovskiy, V. N.; Esimantovskiy, V. M.; Smirnov, I. V.; Babain, V. A.

    1999-01-01

    A teritiary solvent containing chlorinated cobalt dicarbollide, polyethylene glycol and diphenylcarbamoylmethylphosphine oxide was evaluated in different non-nitroaromatic diluents for the separation of cesium, strontium, actinides and rare earth elements from acidic liquid radioactive waste. Decontamination factors of >95% for Cs, 99.7% for Sr, and 99.99% for actinides were achieved in four successive batch contacts using actual radioactive waste. Pilot plant testing in centrifugal contactors using simulated wastes, has demonstrated removal of >99% of all targeted ions.

  4. Liquid secondary waste: Waste form formulation and qualification

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-31

    The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity

  5. 78 FR 14774 - U.S. Environmental Solutions Toolkit-Universal Waste

    Science.gov (United States)

    2013-03-07

    ...--Universal Waste AGENCY: International Trade Administration, DOC. ACTION: Notice and Request for Comment... or services relevant to management of universal waste. The Department of Commerce continues to..., Web site address, contact information, and universal waste management category of interest from the...

  6. Review of the Scientific Understanding of Radioactive Waste at the U.S. DOE Hanford Site.

    Science.gov (United States)

    Peterson, Reid A; Buck, Edgar C; Chun, Jaehun; Daniel, Richard C; Herting, Daniel L; Ilton, Eugene S; Lumetta, Gregg J; Clark, Sue B

    2018-01-16

    This Critical Review reviews the origin and chemical and rheological complexity of radioactive waste at the U.S. Department of Energy Hanford Site. The waste, stored in underground tanks, was generated via three distinct processes over decades of plutonium extraction operations. Although close records were kept of original waste disposition, tank-to-tank transfers and conditions that impede equilibrium complicate our understanding of the chemistry, phase composition, and rheology of the waste. Tank waste slurries comprise particles and aggregates from nano to micro scales, with varying densities, morphologies, heterogeneous compositions, and complicated responses to flow regimes and process conditions. Further, remnant or changing radiation fields may affect the stability and rheology of the waste. These conditions pose challenges for transport through conduits or pipes to treatment plants for vitrification. Additionally, recalcitrant boehmite degrades glass quality and the high aluminum content must be reduced prior to vitrification for the manufacture of waste glass of acceptable durability. However, caustic leaching indicates that boehmite dissolves much more slowly than predicted given surface normalized rates. Existing empirical models based on ex situ experiments and observations generally only describe material balances and have not effectively predicted process performance. Recent advances in the areas of in situ microscopy, aberration-corrected transmission electron microscopy, theoretical modeling across scales, and experimental methods for probing the physics and chemistry at mineral-fluid and mineral-mineral interfaces are being implemented to build robustly predictive physics-based models.

  7. Recent Improvements in Interface Management for Hanford's Waste Treatment and Immobilization Plant - 13263

    Energy Technology Data Exchange (ETDEWEB)

    Arm, Stuart T.; Van Meighem, Jeffery S. [Washington River Protection Solutions, P.O. Box 850, Richland, Washington, 99352 (United States); Duncan, Garth M.; Pell, Michael J. [Bechtel National Inc., 2435 Stevens Center Place, Richland, Washington, 99352 (United States); Harrington, Christopher C. [Department of Energy - Office of River Protection, 2440 Stevens Center Place, Richland, Washington, 99352 (United States)

    2013-07-01

    The U.S. Department of Energy (DOE), Office of River Protection (ORP) is responsible for management and completion of the River Protection Project (RPP) mission, which includes the Hanford Site tank farms operations and the Waste Treatment and Immobilization Plant (WTP). The RPP mission is to store, retrieve and treat Hanford's tank waste; store and dispose of treated wastes; and close the tank farm waste management areas and treatment facilities by 2047. The WTP is currently being designed and constructed by Bechtel National Inc. (BNI) for DOE-ORP. BNI relies on a number of technical services from other Hanford contractors for WTP's construction and commissioning. These same services will be required of the future WTP operations contractor. Partly in response to a DNFSB recommendation, the WTP interface management process managing these technical services has recently been improved through changes in organization and issue management. The changes are documented in an Interface Management Plan. The organizational improvement is embodied in the One System Integrated Project Team that was formed by integrating WTP and tank farms staff representing interfacing functional areas into a single organization. A number of improvements were made to the issue management process but most notable was the formal appointment of technical, regulatory and safety subject matter experts to ensure accurate identification of issues and open items. Ten of the thirteen active WTP Interface Control Documents have been revised in 2012 using the improved process with the remaining three in progress. The value of the process improvements is reflected by the ability to issue these documents on schedule and accurately identify technical, regulatory and safety issues and open items. (authors)

  8. Biological processes for advancing lignocellulosic waste biorefinery by advocating circular economy.

    Science.gov (United States)

    Liguori, Rossana; Faraco, Vincenza

    2016-09-01

    The actualization of a circular economy through the use of lignocellulosic wastes as renewable resources can lead to reduce the dependence from fossil-based resources and contribute to a sustainable waste management. The integrated biorefineries, exploiting the overall lignocellulosic waste components to generate fuels, chemicals and energy, are the pillar of the circular economy. The biological treatment is receiving great attention for the biorefinery development since it is considered an eco-friendly alternative to the physico-chemical strategies to increase the biobased product recovery from wastes and improve saccharification and fermentation yields. This paper reviews the last advances in the biological treatments aimed at upgrading lignocellulosic wastes, implementing the biorefinery concept and advocating circular economy. Copyright © 2016 Elsevier Ltd. All rights reserved.

  9. Influence of radiation on the system liquid radioactive wastes: geologic formation

    International Nuclear Information System (INIS)

    Spitsyn, V.I.; Balukova, V.D.; Kabakchi, S.A.; Medvedeva, M.L.

    1979-01-01

    Introduction of liquid radioactive wastes into deep strata-collectors results in a number of physical-chemical processes: precipitation, dissolution, complex formation, sorption, etc. The area occupied by the injected waste and changes in the nature of the liquid phase depend primarily on radiolysis processes in the heterogeneous system of liquid waste-stratal material occurring at elevated temperatures and pressures. Experiments that simulate actual conditions of temperature, pressure and high radiation levels on this system have been performed. Results are presented for radiolytic gas formation and for changes in the liquid phase and sorption capacity of stratal minerals. It is shown that the temperature increase in the stratum-collector significantly enhances waste decomposition processes, promotes sorption of radionuclides and decreases the mobility of the waste in the formation

  10. 78 FR 14773 - U.S. Environmental Solutions Toolkit-Medical Waste

    Science.gov (United States)

    2013-03-07

    ...--Medical Waste AGENCY: International Trade Administration, DOC. ACTION: Notice and Request for Comment... or services relevant to management of medical waste. The Department of Commerce continues to develop... encouraged to submit their company's name, Web site address, contact information, and medical waste...

  11. Prevention of stress corrosion cracking in nuclear waste storage tanks

    International Nuclear Information System (INIS)

    Ondrejcin, R.S.

    1983-01-01

    At the Savannah River Plant, stress corrosion of carbon steel storage tanks containing alkaline nitrate radioactive waste is prevented by stress relief and specification of limits on waste composition and temperature. Actual cases of cracking have occurred in the primary steel shell of tanks designed and built before 1960 and were attributed to a combination of high residual stresses from fabrication welding and aggressiveness of fresh wastes from the reactor fuel reprocessing plants. The fresh wastes have the highest concentration of nitrate, which has been shown to be the cracking agent. Also, as the waste solutions age and are reduced in volume by evaporation of water, nitrite and hydroxide ions become more concentrated and inhibit stress corrosion. Thus, by providing a heel of aged evaporated waste in tanks that receive fresh wastes, concentrations of the inhibitor ions are maintained within specific ranges to protect against nitrate cracking. The concentration and temperature range limits to prevent cracking were determined by a series of statistically designed experiments

  12. On-line caloric value sensor and validation of dynamic models applied to municipal solid waste

    NARCIS (Netherlands)

    Kessel, van L.B.M.; Leskens, M.; Brem, G.

    2002-01-01

    This paper deals with two aspects concerning the optimization of municipal solid waste combustion (MSWC) processes. First of all, an on-line calorific value sensor is discussed by means of which the calorific value of the waste can be estimated from actual process data. Experimental results on a

  13. Radioactive waste management and regulation

    International Nuclear Information System (INIS)

    Willrich, M.; Lester, R.K.; Greenberg, S.C.; Mitchell, H.C.; Walker, D.A.

    1977-01-01

    Purpose of this book is to assist in developing public policy and institutions for the safe management of radioactive waste, currently and long term. Both high-level waste and low-level waste containing transuranium elements are covered. The following conclusions are drawn: the safe management of post-fission radioactive waste is already a present necessity and an irreversible long-term commitment; the basic goals of U.S. radioactive waste policy are unclear; the existing organization for radioactive waste management is likely to be unworkable if left unchanged; and the existing framework for radioactive waste regulation is likely to be ineffective if left unchanged. The following recommendations are made: a national Radioactive Waste Authority should be established as a federally chartered public corporation; with NRC as the primary agency, a comprehensive regulatory framework should be established to assure the safety of all radioactive waste management operations under U.S. jurisdiction or control; ERDA should continue to have primary government responsibility for R and D and demonstration of radioactive waste technology; and the U.S. government should propose that an international Radioactive Waste Commission be established under the IAEA

  14. Greater-than-Class C low-level radioactive waste characterization. Appendix A-3: Basis for greater-than-Class C low-level radioactive waste light water reactor projections

    International Nuclear Information System (INIS)

    Mancini, A.; Tuite, P.; Tuite, K.; Woodberry, S.

    1994-09-01

    This study characterizes low-level radioactive waste types that may exceed Class C limits at light water reactors, estimates the amounts of waste generated, and estimates radionuclide content and distribution within the waste. Waste types that may exceed Class C limits include metal components that become activated during operations, process wastes such as cartridge filters and decontamination resins, and activated metals from decommissioning activities. Operating parameters and current management practices at operating plants are reviewed and used to estimate the amounts of low-level waste exceeding Class C limits that is generated per fuel cycle, including amounts of routinely generated activated metal components and process waste. Radionuclide content is calculated for specific activated metals components. Empirical data from actual low-level radioactive waste are used to estimate radionuclide content for process wastes. Volumes and activities are also estimated for decommissioning activated metals that exceed Class C limits. To estimate activation levels of decommissioning waste, six typical light water reactors are modeled and analyzed. This study does not consider concentration averaging

  15. Síndrome doloroso regional complejo: Conceptos actuales

    OpenAIRE

    Ferrer Lozano, Yovanny; Oquendo Vázquez, Pablo; Ferrer Lozano, Dunia

    2006-01-01

    Se realizó una revisión bibliográfica de los conceptos actuales que se les atribuyen al síndrome doloroso regional complejo, término propuesto en 1994 por la International Association for Study of Pain. En la aparición del síndrome intervienen varios factores como son los sistemas nerviosos periférico, central y vegetativo; el sistema endocrino, los factores psicológicos, históricos, ambientales y circunstanciales; que permiten en conjunto personalizar el dolor en cada paciente.El tratamiento...

  16. Contact-Handled Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-12-29

    The purpose of this document is to summarize the waste acceptance criteria applicable to the transportation, storage, and disposal of contact-handled transuranic (CH-TRU) waste at the Waste Isolation Pilot Plant (WIPP). These criteria serve as the U.S. Department of Energy's (DOE) primary directive for ensuring that CH-TRU waste is managed and disposed of in a manner that protects human health and safety and the environment.The authorization basis of WIPP for the disposal of CH-TRU waste includes the U.S.Department of Energy National Security and Military Applications of Nuclear EnergyAuthorization Act of 1980 (reference 1) and the WIPP Land Withdrawal Act (LWA;reference 2). Included in this document are the requirements and associated criteriaimposed by these acts and the Resource Conservation and Recovery Act (RCRA,reference 3), as amended, on the CH-TRU waste destined for disposal at WIPP.|The DOE TRU waste sites must certify CH-TRU waste payload containers to thecontact-handled waste acceptance criteria (CH-WAC) identified in this document. Asshown in figure 1.0, the flow-down of applicable requirements to the CH-WAC istraceable to several higher-tier documents, including the WIPP operational safetyrequirements derived from the WIPP CH Documented Safety Analysis (CH-DSA;reference 4), the transportation requirements for CH-TRU wastes derived from theTransuranic Package Transporter-Model II (TRUPACT-II) and HalfPACT Certificates ofCompliance (references 5 and 5a), the WIPP LWA (reference 2), the WIPP HazardousWaste Facility Permit (reference 6), and the U.S. Environmental Protection Agency(EPA) Compliance Certification Decision and approval for PCB disposal (references 7,34, 35, 36, and 37). The solid arrows shown in figure 1.0 represent the flow-down of allapplicable payload container-based requirements. The two dotted arrows shown infigure 1.0 represent the flow-down of summary level requirements only; i.e., the sitesmust reference the regulatory source

  17. Characterization of Hanford tank wastes containing ferrocyanides

    International Nuclear Information System (INIS)

    Tingey, J.M.; Matheson, J.D.; McKinley, S.G.; Jones, T.E.; Pool, K.H.

    1993-02-01

    Currently, 17 storage tanks on the Hanford site that are believed to contain > 1,000 gram moles (465 lbs) of ferrocyanide compounds have been identified. Seven other tanks are classified as ferrocyanide containing waste tanks, but contain less than 1,000 gram moles of ferrocyanide compounds. These seven tanks are still included as Hanford Watch List Tanks. These tanks have been declared an unreviewed safety question (USQ) because of potential thermal reactivity hazards associated with the ferrocyanide compounds and nitrate and nitrite. Hanford tanks with waste containing > 1,000 gram moles of ferrocyanide have been sampled. Extensive chemical, radiothermical, and physical characterization have been performed on these waste samples. The reactivity of these wastes were also studied using Differential Scanning Calorimetry (DSC) and Thermogravimetric analysis. Actual tank waste samples were retrieved from tank 241-C-112 using a specially designed and equipped core-sampling truck. Only a small portion of the data obtained from this characterization effort will be reported in this paper. This report will deal primarily with the cyanide and carbon analyses, thermal analyses, and limited physical property measurements

  18. Waste acceptance and waste loading for vitrified Oak Ridge tank waste

    International Nuclear Information System (INIS)

    Harbour, J.R.; Andrews, M.K.

    1997-01-01

    The Office of Science and Technology of the DOE has funded a joint project between the Oak Ridge National Laboratory (ORNL) and the Savannah River Technology Center (SRTC) to evaluate vitrification and grouting for the immobilization of sludge from ORNL tank farms. The radioactive waste is from the Gunite and Associated Tanks (GAAT), the Melton Valley Storage Tanks (MVST), the Bethel Valley Evaporator Service Tanks (BVEST), and the Old Hydrofractgure Tanks (OHF). Glass formulation development for sludge from these tanks is discussed in an accompanying article for this conference (Andrews and Workman). The sludges contain transuranic radionuclides at levels which will make the glass waste form (at reasonable waste loadings) TRU. Therefore, one of the objectives for this project was to ensure that the vitrified waste form could be disposed of at the Waste Isolation Pilot Plant (WIPP). In order to accomplish this, the waste form must meet the WIPP Waste Acceptance Criteria (WAC). An alternate pathway is to send the glass waste forms for disposal at the Nevada Test Site (NTS). A sludge waste loading in the feed of 6 wt percent will lead to a waste form which is non-TRU and could potentially be disposed of at NTS. The waste forms would then have to meet the requirements of the NTS WAC. This paper presents SRTC''s efforts at demonstrating that the glass waste form produced as a result of vitrification of ORNL sludge will meet all the criteria of the WIPP WAC or NTS WAC

  19. Tribal Waste Management Program

    Science.gov (United States)

    The EPA’s Tribal Waste Management Program encourages environmentally sound waste management practices that promote resource conservation through recycling, recovery, reduction, clean up, and elimination of waste.

  20. Special waste-form lysimeters-arid: Three-year monitoring report

    International Nuclear Information System (INIS)

    Jones, T.L.; Serne, R.J.; Toste, A.P.

    1988-04-01

    Regulations governing the disposal of commercial low-level waste require all liquid waste to be solidified before burial. Most waste must be solidified into a rigid matrix such as cement or plastic to prevent waste consolidation and site slumping after burial. These solidification processes affect the rate at which radionuclides and other solutes are released into the soil. In 1983, a program was initiated at Pacific Northwest Laboratory to study the release of waste from samples of low-level radioactive waste that had been commercially solidified. The primary method used by this program is to bury sample waste forms in field lysimeters and monitor leachate composition from the release and transport of solutes. The lysimeter facility consists of 10 lysimeters, each containing one sample of solidified waste. Five different waste forms are being tested, allowing duplicate samples of each one to be evaluated. The samples were obtained from operating nuclear power plants and are actual waste forms routinely generated at these facilities. All solidification was accomplished by commercial processes. Sample size is a partially filled 210-L drum. All containers were removed prior to burial leaving the bare waste form in contact with the lysimeter soil. 11 refs., 14 figs., 16 tabs