Stochastic simulation of pitting degradation of multi-barrier waste container in the potential repository at Yucca Mountain

International Nuclear Information System (INIS)

Lee, J.H.; Atkins, J.E.; Andrews, R.W.

1995-01-01

A detailed stochastic waste package degradation simulation model was developed incorporating the humid-air and aqueous general and pitting corrosion models for the carbon steel corrosion-allowance outer barrier and aqueous pitting corrosion model for the Alloy 825 corrosion-resistant inner barrier. The uncertainties in the individual corrosion models were also incorporated to capture the variability in the corrosion degradation among waste packages and among pits in the same waste package. Within the scope of assumptions employed in the simulations, the corrosion modes considered, and the near-field conditions from the drift-scale thermohydrologic model, the results of the waste package performance analyses show that the current waste package design appears to meet the 'controlled design assumption' requirement of waste package performance, which is currently defined as having less than 1% of waste packages breached at 1,000 years. It was shown that, except for the waste packages that fail early, pitting corrosion of the corrosion-resistant inner barrier has a greater control on the failure of waste packages and their subsequent degradation than the outer barrier. Further improvement and substantiation of the inner barrier pitting model (currently based on an elicitation) is necessary in future waste package performance simulation model

Biogenic carbon in combustible waste: Waste composition, variability and measurement uncertainty

DEFF Research Database (Denmark)

Larsen, Anna Warberg; Fuglsang, Karsten; Pedersen, Niels H.

2013-01-01

described in the literature. This study addressed the variability of biogenic and fossil carbon in combustible waste received at a municipal solid waste incinerator. Two approaches were compared: (1) radiocarbon dating (14C analysis) of carbon dioxide sampled from the flue gas, and (2) mass and energy......, the measurement uncertainties related to the two approaches were determined. Two flue gas sampling campaigns at a full-scale waste incinerator were included: one during normal operation and one with controlled waste input. Estimation of carbon contents in the main waste types received was included. Both the 14C...... method and the balance method represented promising methods able to provide good quality data for the ratio between biogenic and fossil carbon in waste. The relative uncertainty in the individual experiments was 7–10% (95% confidence interval) for the 14C method and slightly lower for the balance method....

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.)

Retrieval process development and enhancements waste simulant compositions and defensibility

International Nuclear Information System (INIS)

Powell, M.R.; Golcar, G.R.; Geeting, J.G.H.

1997-09-01

The purpose of this report is to document the physical waste simulant development efforts of the EM-50 Tanks Focus Area at the Hanford Site. Waste simulants are used in the testing and development of waste treatment and handling processes because performing such tests using actual tank waste is hazardous and prohibitively expensive. This document addresses the simulant development work that supports the testing of waste retrieval processes using simulants that mimic certain key physical properties of the tank waste. Development and testing of chemical simulants are described elsewhere. This work was funded through the EM-50 Tanks Focus Area as part of the Retrieval Process Development and Enhancements (RPD ampersand E) Project at the Pacific Northwest National Laboratory (PNNL). The mission of RPD ampersand E is to understand retrieval processes, including emerging and existing processes, gather performance data on those processes, and relate the data to specific tank problems to provide end users with the requisite technical bases to make retrieval and closure decisions. Physical simulants are prepared using relatively nonhazardous and inexpensive materials rather than the chemicals known to be in tank waste. Consequently, only some of the waste properties are matched by the simulant. Deciding which properties need to be matched and which do not requires a detailed knowledge of the physics of the process to be tested using the simulant. Developing this knowledge requires reviews of available literature, consultation with experts, and parametric tests. Once the relevant properties are identified, waste characterization data are reviewed to establish the target ranges for each property. Simulants are then developed that possess the desired ranges of properties

Full-scale retrieval of simulated buried transuranic waste

International Nuclear Information System (INIS)

Valentich, D.J.

1993-09-01

This report describes the results of a field test conducted to determine the effectiveness of using conventional type construction equipment for the retrieval of buried transuranic (TRU) waste. A cold (nonhazardous and nonradioactive) test pit (1,100 yd 3 volume) was constructed with boxes and drums filled with simulated waste materials, such as metal, plastic, wood, concrete, and sludge. Large objects, including truck beds, tanks, vaults, pipes, and beams, were also placed in the pit. These materials were intended to simulate the type of wastes found in TRU buried waste pits and trenches. A series of commercially available equipment items, such as excavators and tracked loaders outfitted with different end effectors, were used to remove the simulated waste. Work was performed from both the abovegrade and belowgrade positions. During the demonstration, a number of observations, measurements, and analyses were performed to determine which equipment was the most effective in removing the waste. The retrieval rates for the various excavation techniques were recorded. The inherent dust control capabilities of the excavation methods used were observed. The feasibility of teleoperating reading equipment was also addressed

Incineration of Non-radioactive Simulated Waste

International Nuclear Information System (INIS)

Ahmed, A.Z.; Abdelrazek, I.D.

1999-01-01

An advanced controlled air incinerator has been investigated, developed and put into successful operation for both non radioactive simulated and other combustible solid wastes. Engineering efforts concentrated on providing an incinerator which emitted a clean, easily treatable off-gas and which produced minimum amounts of secondary waste. Feed material is fed by gravity into the gas reactor without shredding or other pretreatment. The temperature of the waste is gradually increased in a reduced oxygen atmosphere as the resulting products are introduced into the combustion chamber. Steady burning is thus accomplished under easily controlled excess air conditions with the off-gas then passing through a simple dry cleaning-up system. Experimental studies showed that, at lower temperature, CO 2 , and CH 4 contents in gas reactor effluent increase by the increase of glowing bed temperature, while H 2 O, H 2 and CO decrease . It was proved that, a burn-out efficiency (for ash residues) and a volume reduction factor appeared to be better than 95.5% and 98% respectively. Moreover, high temperature permits increased volumes of incinerated material and results in increased gasification products. It was also found that 8% by weight of ashes are separated by flue gas cleaning system as it has chemical and size uniformity. This high incineration efficiency has been obtained through automated control and optimization of process variables like temperature of the glowing bed and the oxygen feed rate to the gas reactor

WASTES: Waste System Transportation and Economic Simulation--Version 2:

International Nuclear Information System (INIS)

Sovers, R.A.; Shay, M.R.; Ouderkirk, S.J.; McNair, G.W.; Eagle, B.G.

1988-02-01

The Waste System Transportation and Economic Simulation (WASTES) Technical Reference Manual was written to describe and document the algorithms used within the WASTES model as implemented in Version 2.23. The manual will serve as a reference for users of the WASTES system. The intended audience for this manual are knowledgeable users of WASTES who have an interest in the underlying principles and algorithms used within the WASTES model. Each algorithm is described in nonprogrammers terminology, and the source and uncertainties of the constants in use by these algorithms are described. The manual also describes the general philosophy and rules used to: 1) determine the allocation and priority of spent fuel generation sources to facility destinations, 2) calculate transportation costs, and 3) estimate the cost of at-reactor ex-pool storage. A detailed description of the implementation of many of the algorithms is also included in the WASTES Programmers Reference Manual (Shay and Buxbaum 1986a). This manual is separated into sections based on the general usage of the algorithms being discussed. 8 refs., 14 figs., 2 tabs

Hanford solid waste management system simulation

International Nuclear Information System (INIS)

Shaver, S.R.; Armacost, L.L.; Konynenbelt, H.S.; Wehrman, R.R.

1994-12-01

This paper describes systems analysis and simulation model development for a proposed solid waste management system at a U.S. Department of Energy Site. The proposed system will include a central storage facility, four treatment facilities, and three disposal sites. The material managed by this system will include radioactive, hazardous, and mixed radioactive and hazardous wastes. The objective of the modeling effort is to provide a means of evaluating throughput and capacity requirements for the proposed treatment, storage, and disposal facilities. The model is used to evaluate alternative system configurations and the effect on the alternatives of changing waste stream characteristics and receipt schedules. An iterative modeling and analysis approach is used that provides macro-level models early in the project and establishes credibility with the customer. The results from the analyses based on the macro models influence system design decisions and provide information that helps focus subsequent model development. Modeling and simulation of alternative system configurations and operating strategies yield a better understanding of the solid waste system requirements. The model effectively integrates information obtained through systems analysis and waste characterization to provide a consistent basis for system and facility planning

Pitting growth rate in carbon steel exposed to simulated radioactive waste

International Nuclear Information System (INIS)

Zapp, P.E.

1995-01-01

Dilute high-level radioactive waste slurries can induce pitting corrosion in carbon steel tanks in which such waste is stored and processed. The waste is normally maintained with closely monitored nitrite and hydroxide concentrations known to prevent the initiation of pitting. Coupon immersion are being conducted in laboratory simulants of waste to determine the probability and growth rate of pitting in steel in the event of below-limits nitrite concentrations. Sets of about 36 carbon steel coupons have been immersed in known corrosive conditions (nitrite < 5% of the established limit) at a temperature of 50 C. Three sets have been removed from testing after 64, 150, and 350 days of immersion. The long immersion times introduced variability in the exposure conditions due to the evaporation and replenishment of solution. The deepest corrosive attack was measured one each coupon by optical microscopy. The deepest pits were ranked and analyzed as a type 1 extreme value distribution to extrapolate from the coupon population to the maximum pit depths in a waste tank structure. The data were compared to a power law for pit growth, although the deepest pits did not increase monotonically with time in the limited data set

WASTES II: Waste System Transportation and Economic Simulation. Version II. User's guide

International Nuclear Information System (INIS)

Shay, M.R.; Buxbaum, M.E.

1986-02-01

The WASTES II model was developed to provide detailed analyses beyond the capabilities of other available models. WASTES uses discrete event simulation techniques to model the generation of commercial spent nuclear fuel, the buildup of spent fuel inventories within the system, and the transportation requirements for the movement of radioactive waste throughout the system. The model is written in FORTRAN 77 as an extension to the SLAM commercial simulation language package. In addition to the pool storage and dry storage located at the reactors, the WASTES model provides a choice of up to ten other storage facilities of four different types. The simulation performed by WASTES may be controlled by a combination of source- and/or destination-controlled transfers that are requested by the code user. The user supplies shipping cask characteristics for truck or rail shipment casks. As part of the facility description, the user specifies which casks the facility can use. Shipments within the system can be user specified to occur optimally, or proximally. Optimized shipping can be used when exactly two destination facilities of the same facility type are open for receipt of fuel. Optimized shipping selects source/destination pairs so that the total shipping distance or total shipping costs in a given year are minimized when both facilities are fully utilized. Proximity shipping sequentially fills the closest facility to the source according to the shipment priorities without regard for the total annual shipments. This results in sub-optimal routing of waste material but can be used to approximate an optimal shipping strategy when more than two facilities of the same type are available to receive waste. WASTES is currently able to analyze each of the commercial spent fuel logistics scenarios specified in the 1985 DOE Mission Plan

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

Variability in properties of grouted Phosphate/Sulfate N-Reactor Waste

International Nuclear Information System (INIS)

Lokken, R.O.; Martin, P.F.C.; Bowen, W.M.; Harty, H.; Treat, R.L.

1987-02-01

A Transportable Grout Facility (TGF) is being constructed at the Hanford site in Washington State to convert various low-level liquid wastes to a grout waste form for onsite disposal. The TGF Project is managed by Rockwell Hanford Operations (Rockwell). Oak Ridge National Laboratory (ORNL) has provided a grout formulation for Phosphate/Sulfate N-Reactor Waste, the first waste stream scheduled for grouting beginning in late 1987. The formulation includes a blend of portland cement, fly ash, attapulgite clay, and an illitic clay. Grout will be produced by mixing the blend with Phosphate/Sulfate N-Reactor Waste. These wastes result from decontamination and ion-exchange regeneration activities at Hanford's N-Reactor. Pacific Northwest Laboratory (PNL) is conducting studies on grouted Phosphate/Sulfate N-Reactor Waste to verify that the grout can be successfully processed and, when hardened, that it will meet all performance and regulatory requirements. As part of these studies, PNL is assessing the variability that may be encountered when processing Phosphate/Sulfate N-Reactor Waste grout. Sources of variability that may affect grout properties include the composition and concentrations of the waste and dry solids, temperature, efficiency of dry solids blending, and dry blend storage time. 13 refs., 20 figs., 9 tabs

Siting simulation for low-level waste disposal facilities

International Nuclear Information System (INIS)

Roop, R.D.; Rope, R.C.

1985-01-01

The Mock Site Licensing Demonstration Project has developed the Low-Level Radioactive Waste Siting Simulation, a role-playing exercise designed to facilitate the process of siting and licensing disposal facilities for low-level waste (LLW). This paper describes the development, content, and usefulness of the siting simulation. The simulation can be conducted at a workshop or conference, involves 14 or more participants, and requires about eight hours to complete. The simulation consists of two sessions; in the first, participants negotiate the selection of siting criteria, and in the second, a preferred disposal site is chosen from three candidate sites. The project has sponsored two workshops (in Boston, Massachusetts and Richmond, Virginia) in which the simulation has been conducted for persons concerned with LLW management issues. It is concluded that the simulation can be valuable as a tool for disseminating information about LLW management; a vehicle that can foster communication; and a step toward consensus building and conflict resolution. The DOE National Low-Level Waste Management Program is now making the siting simulation available for use by states, regional compacts, and other organizations involved in development of LLW disposal facilities

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

Eternity Variables to Prove Simulation of Specifications

NARCIS (Netherlands)

Hesselink, Wim H.

2005-01-01

Simulations of specifications are introduced as a unification and generalization of refinement mappings, history variables, forward simulations, prophecy variables, and backward simulations. A specification implements another specification if and only if there is a simulation from the first one to

WIPP waste package testing on simulated DHLW: emplacement

International Nuclear Information System (INIS)

Molecke, M.A.

1984-01-01

Several series of simulated (nonradioactive) defense high-level waste (DHLW) package tests have been emplaced in the WIPP, a research and development facility authorized to demonstrate the safe disposal of defense-related wastes. The primary purpose of these 3-to-7 year duration tests is to evaluate the in situ materials performance of waste package barriers (canisters, overpacks, backfills, and nonradioactive DHLW glass waste form) for possible future application to a licensed waste repository in salt. This paper describes all test materials, instrumentation, and emplacement and testing techniques, and discusses progress of the various tests. These tests are intended to provide information on materials behavior (i.e., corrosion, metallurgical and geochemical alterations, waste form durability, surface interactions, etc.), as well as comparison between several waste package designs, fabrications details, and actual costs. These experiments involve 18 full-size simulated DHLW packages (approximately 3.0 m x 0.6 m diameter) emplaced in vertical boreholes in the salt drift floor. Six of the test packages contain internal electrical heaters (470 W/canister), and were emplace under approximately reference DHLW repository conditions. Twelve other simulated DHLW packages were emplaced under accelerated-aging or overtest conditions, including the artificial introduction of brine, and a thermal loading approximately three to four times higher than reference. Eight of these 12 test packages contain 1500 W/canister electrical heaters; the other four are filled with DHLW glass. 9 refs., 1 fig

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

Standard test method for determining liquidus temperature of immobilized waste glasses and simulated waste glasses

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

1.1 These practices cover procedures for determining the liquidus temperature (TL) of nuclear waste, mixed nuclear waste, simulated nuclear waste, or hazardous waste glass in the temperature range from 600°C to 1600°C. This method differs from Practice C829 in that it employs additional methods to determine TL. TL is useful in waste glass plant operation, glass formulation, and melter design to determine the minimum temperature that must be maintained in a waste glass melt to make sure that crystallization does not occur or is below a particular constraint, for example, 1 volume % crystallinity or T1%. As of now, many institutions studying waste and simulated waste vitrification are not in agreement regarding this constraint (1). 1.2 Three methods are included, differing in (1) the type of equipment available to the analyst (that is, type of furnace and characterization equipment), (2) the quantity of glass available to the analyst, (3) the precision and accuracy desired for the measurement, and (4) candi...

Solubilities of gases in simulated Tank 241-SY-101 wastes

International Nuclear Information System (INIS)

Norton, J.D.; Pederson, L.R.

1995-09-01

Oxygen, nitrogen, hydrogen, methane, and nitrous oxide solubilities were evaluated as a function of temperature in SYl-SIM-93B, a homogeneous simulated waste mixture containing sodium hydroxide, sodium nitrite, sodium nitrate, sodium aluminate, and sodium carbonate, the principal inorganic constituents of the wastes in Tank 241-SY-101. Ammonia solubility data for this simulated waste was obtained as a function of temperature in an earlier study. The choice of a homogeneous waste mixture in this study has the advantage of eliminating complications associated with a changing electrolyte concentration as a function of temperature that would be encountered with a slurry simulant. Dissolution is one of the means by which gases may be retained in Hanford Site wastes. While models are available to estimate gas solubilities in electrolyte solutions, few data are in existence that pertain to highly concentrated, multicomponent electrolytes such as those stored in Hanford Site waste tanks

Chemical compatibility screening results of plastic packaging to mixed waste simulants

International Nuclear Information System (INIS)

Nigrey, P.J.; Dickens, T.G.

1995-01-01

We have developed a chemical compatibility program for evaluating transportation packaging components for transporting mixed waste forms. We have performed the first phase of this experimental program to determine the effects of simulant mixed wastes on packaging materials. This effort involved the screening of 10 plastic materials in four liquid mixed waste simulants. The testing protocol involved exposing the respective materials to ∼3 kGy of gamma radiation followed by 14 day exposures to the waste simulants of 60 C. The seal materials or rubbers were tested using VTR (vapor transport rate) measurements while the liner materials were tested using specific gravity as a metric. For these tests, a screening criteria of ∼1 g/m 2 /hr for VTR and a specific gravity change of 10% was used. It was concluded that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only VITON passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture simulant mixed waste, none of the seal materials met the screening criteria. It is anticipated that those materials with the lowest VTRs will be evaluated in the comprehensive phase of the program. For specific gravity testing of liner materials the data showed that while all materials with the exception of polypropylene passed the screening criteria, Kel-F, HDPE, and XLPE were found to offer the greatest resistance to the combination of radiation and chemicals

Uncertainty analysis of NDA waste measurements using computer simulations

International Nuclear Information System (INIS)

Blackwood, L.G.; Harker, Y.D.; Yoon, W.Y.; Meachum, T.R.

2000-01-01

Uncertainty assessments for nondestructive radioassay (NDA) systems for nuclear waste are complicated by factors extraneous to the measurement systems themselves. Most notably, characteristics of the waste matrix (e.g., homogeneity) and radioactive source material (e.g., particle size distribution) can have great effects on measured mass values. Under these circumstances, characterizing the waste population is as important as understanding the measurement system in obtaining realistic uncertainty values. When extraneous waste characteristics affect measurement results, the uncertainty results are waste-type specific. The goal becomes to assess the expected bias and precision for the measurement of a randomly selected item from the waste population of interest. Standard propagation-of-errors methods for uncertainty analysis can be very difficult to implement in the presence of significant extraneous effects on the measurement system. An alternative approach that naturally includes the extraneous effects is as follows: (1) Draw a random sample of items from the population of interest; (2) Measure the items using the NDA system of interest; (3) Establish the true quantity being measured using a gold standard technique; and (4) Estimate bias by deriving a statistical regression model comparing the measurements on the system of interest to the gold standard values; similar regression techniques for modeling the standard deviation of the difference values gives the estimated precision. Actual implementation of this method is often impractical. For example, a true gold standard confirmation measurement may not exist. A more tractable implementation is obtained by developing numerical models for both the waste material and the measurement system. A random sample of simulated waste containers generated by the waste population model serves as input to the measurement system model. This approach has been developed and successfully applied to assessing the quantity of

Pitting growth rate in carbon steel exposed to simulated radioactive waste

International Nuclear Information System (INIS)

Zapp, P.E.

1996-06-01

Dilute high-level radioactive waste slurries can induce pitting corrosion in carbon steel tanks in which such waste is stored and processed. The waste is normally maintained with closely monitored nitrite and hydroxide concentrations known to prevent the initiation of pitting. Coupon immersion tests are being conducted in laboratory simulants of waste to determine the probability and growth rate of pitting in steel in the event of out-of-limits nitrite concentrations. Sets of about 36 carbon steel coupons have been immersed in known corrosive conditions (nitrite < 5 per cent of the established limit) at a temperature of 50 degrees C. Three sets have been removed from testing after 64, 150, and 350 days of immersion. The long immersion times introduced variability in the exposure conditions due to the evaporation and replenishment of solution. The deepest corrosive attack was measured on each coupon by optical microscopy. The deepest pits were ranked and analyzed as a type 1 extreme value distribution to extrapolate from the coupon population to the maximum expected pit depths in a waste tank structure. The data were compared to a power law for pit growth, although the deepest pits did not increase monotonically with time in the limited data set

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

Repository simulation tests

International Nuclear Information System (INIS)

Wicks, G.G.; Bibler, N.E.; Jantzen, C.M.; Plodinec, M.J.

1984-01-01

The repository simulation experiments described in this paper are designed to assess the performance of SRP waste glass under the most realistic repository conditions that can be obtained in the laboratory. These tests simulate the repository environment as closely as possible and introduce systematically the variability of the geology, groundwater chemistry, and waste package components during the leaching of the waste glass. The tests evaluate waste form performance under site-specific conditions, which differ for each of the geologic repositories under consideration. Data from these experiments will aid in the development of a realistic source term that can describe the release of radionuclides from SRP waste glass as a component of proposed waste packages. Hence, this information can be useful to optimize waste package design for SRP waste glass and to provide data for predicting long-term performance and subsequent conformance to regulations. The repository simulation tests also help to bridge the gap in interpreting results derived from tests performed under the control of the laboratory to the uncertainity and variability of field tests. In these experiments, site-specific repository components and conditions are emphasized and only the site specific materials contact the waste forms. An important feature of these tests is that both actual and simulated waste glasses are tested identically. 7 figures, 2 tables

Effects of simulant mixed waste on EPDM and butyl rubber

International Nuclear Information System (INIS)

Nigrey, P.J.; Dickens, T.G.

1997-11-01

The authors have developed a Chemical Compatibility Testing Program for the evaluation of plastic packaging components which may be used in transporting mixed waste forms. In this program, they have screened 10 plastic materials in four liquid mixed waste simulants. These plastics were butadiene-acrylonitrile copolymer (Nitrile) rubber, cross-linked polyethylene, epichlorohydrin rubber, ethylene-propylene (EPDM) rubber, fluorocarbons (Viton and Kel-F trademark), polytetrafluoro-ethylene (Teflon), high-density polyethylene, isobutylene-isoprene copolymer (Butyl) rubber, polypropylene, and styrene-butadiene (SBR) rubber. The selected simulant mixed wastes were (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; and (4) a mixture of ketones. The screening testing protocol involved exposing the respective materials to approximately 3 kGy of gamma radiation followed by 14-day exposures to the waste simulants at 60 C. The rubber materials or elastomers were tested using Vapor Transport Rate measurements while the liner materials were tested using specific gravity as a metric. The authors have developed a chemical compatibility program for the evaluation of plastic packaging components which may be incorporated in packaging for transporting mixed waste forms. From the data analyses performed to date, they have identified the thermoplastic, polychlorotrifluoroethylene, as having the greatest chemical compatibility after having been exposed to gamma radiation followed by exposure to the Hanford Tank simulant mixed waste. The most striking observation from this study was the poor performance of polytetrafluoroethylene under these conditions. In the evaluation of the two elastomeric materials they have concluded that while both materials exhibit remarkable resistance to these environmental conditions, EPDM has a greater resistance to this corrosive simulant mixed waste

Secondary Waste Simulant Development for Cast Stone Formulation Testing

Energy Technology Data Exchange (ETDEWEB)

Russell, Renee L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rinehart, Donald E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Swanberg, David J. [Washington River Protection Solutions, Richland, WA (United States); Mahoney, J. [Washington River Protection Solutions, Richland, WA (United States)

2015-04-01

Washington River Protection Solutions, LLC (WRPS) funded Pacific Northwest National Laboratory (PNNL) to conduct a waste form testing program to implement aspects of the Secondary Liquid Waste Treatment Cast Stone Technology Development Plan (Ashley 2012) and the Hanford Site Secondary Waste Roadmap (PNNL 2009) related to the development and qualification of Cast Stone as a potential waste form for the solidification of aqueous wastes from the Hanford Site after the aqueous wastes are treated at the Effluent Treatment Facility (ETF). The current baseline is that the resultant Cast Stone (or grout) solid waste forms would be disposed at the Integrated Disposal Facility (IDF). Data and results of this testing program will be used in the upcoming performance assessment of the IDF and in the design and operation of a solidification treatment unit planned to be added to the ETF. The purpose of the work described in this report is to 1) develop simulants for the waste streams that are currently being fed and future WTP secondary waste streams also to be fed into the ETF and 2) prepare simulants to use for preparation of grout or Cast Stone solid waste forms for testing.

Simulation of uranium transport with variable temperature and oxidation potential: The computer program THCC [Thermo-Hydro-Chemical Coupling

International Nuclear Information System (INIS)

Carnahan, C.L.

1986-12-01

A simulator of reactive chemical transport has been constructed with the capabilities of treating variable temperatures and variable oxidation potentials within a single simulation. Homogeneous and heterogeneous chemical reactions are simulated at temperature-dependent equilibrium, and changes of oxidation states of multivalent elements can be simulated during transport. Chemical mass action relations for formation of complexes in the fluid phase are included explicitly within the partial differential equations of transport, and a special algorithm greatly simplifies treatment of reversible precipitation of solid phases. This approach allows direct solution of the complete set of governing equations for concentrations of all aqueous species and solids affected simultaneously by chemical and physical processes. Results of example simulations of transport, along a temperature gradient, of uranium solution species under conditions of varying pH and oxidation potential and with reversible precipitation of uraninite and coffinite are presented. The examples illustrate how inclusion of variable temperature and oxidation potential in numerical simulators can enhance understanding of the chemical mechanisms affecting migration of multivalent waste elements

Fluidized-bed calcination of simulated commercial high-level radioactive wastes

International Nuclear Information System (INIS)

Freeby, W.A.

1975-11-01

Work is in progress at the Idaho Chemical Processing Plant to verify process flowsheets for converting simulated commercial high-level liquid wastes to granular solids using the fluidized-bed calcination process. Primary emphasis in the series of runs reported was to define flowsheets for calcining simulated Allied-General Nuclear Services (AGNS) waste and to evaluate product properties significant to calcination, solids storage, or post treatment. Pilot-plant studies using simulated high-level acid wastes representative of those to be produced by Nuclear Fuel Services, Inc. (NFS) are also included. Combined AGNS high-level and intermediate-level waste (0.26 M Na in blend) was successfully calcined when powdered iron was added (to result in a Na/Fe mole ratio of 1.0) to the feed to prevent particle agglomeration due to sodium nitrate. Long-term runs (approximately 100 hours) showed that calcination of the combined waste is practical. Concentrated AGNS waste containing sodium at concentrations less than 0.2 M were calcined successfully; concentrated waste containing 1.13 M Na calcined successfully when powdered iron was added to the feed to suppress sodium nitrate formation. Calcination of dilute AGNS waste by conventional fluid-bed techniques was unsuccessful due to the inability to control bed particle size--both particle size and bed level decreased. Fluid-bed solidification of AGNS dilute waste at conditions in which most of the calcined solids left the calciner vessel with the off-gas was successful. In such a concept, the steady-state composition of the bed material would be approximately 22 wt percent calcined solids deposited on inert particles. Calcination of simulated NFS acid waste indicated that solidification by the fluid-bed process is feasible

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).

Energy Technology Data Exchange (ETDEWEB)

Schultz, Peter Andrew

2011-12-01

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC)

International Nuclear Information System (INIS)

Schultz, Peter Andrew

2011-01-01

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M and S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V and V) is required throughout the system to establish evidence-based metrics for the level of confidence in M and S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V and V challenge at the subcontinuum scale, an approach to incorporate V and V concepts into subcontinuum scale modeling and simulation (M and S), and a plan to incrementally incorporate effective V and V into subcontinuum scale M and S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

Mixed Waste Treatment Project: Computer simulations of integrated flowsheets

International Nuclear Information System (INIS)

Dietsche, L.J.

1993-12-01

The disposal of mixed waste, that is waste containing both hazardous and radioactive components, is a challenging waste management problem of particular concern to DOE sites throughout the United States. Traditional technologies used for the destruction of hazardous wastes need to be re-evaluated for their ability to handle mixed wastes, and in some cases new technologies need to be developed. The Mixed Waste Treatment Project (MWTP) was set up by DOE's Waste Operations Program (EM30) to provide guidance on mixed waste treatment options. One of MWTP's charters is to develop flowsheets for prototype integrated mixed waste treatment facilities which can serve as models for sites developing their own treatment strategies. Evaluation of these flowsheets is being facilitated through the use of computer modelling. The objective of the flowsheet simulations is to provide mass and energy balances, product compositions, and equipment sizing (leading to cost) information. The modelled flowsheets need to be easily modified to examine how alternative technologies and varying feed streams effect the overall integrated process. One such commercially available simulation program is ASPEN PLUS. This report contains details of the Aspen Plus program

Hanford Waste Simulants Created to Support the Research and Development on the River Protection Project - Waste Treatment Plant

Energy Technology Data Exchange (ETDEWEB)

Eibling, R.E.

2001-07-26

The development of nonradioactive waste simulants to support the River Protection Project - Waste Treatment Plant bench and pilot-scale testing is crucial to the design of the facility. The report documents the simulants development to support the SRTC programs and the strategies used to produce the simulants.

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

Effects of simulant mixed waste on EPDM and butyl rubber

International Nuclear Information System (INIS)

Nigrey, P.J.; Dickens, T.G.

1998-01-01

We have developed a Chemical Compatibility Testing Program for the evaluation of plastic packaging components which may be used in transporting mixed waste forms. In this program, we have screened 10 plastic materials in four liquid mixed waste simulants. These plastics were butadiene-acrylonitrile copolymer (Nitrile) rubber, cross-linked polyethylene, epi-chloro-hydrin rubber, ethylene-propylene (EPDM) rubber, fluorocarbons (Viton and Kel-F), poly-tetrafluoroethylene (Teflon), high-density polyethylene, isobutylene-isoprene copolymer (Butyl) rubber, polypropylene, and styrene-butadiene (SBR) rubber. The selected simulant mixed wastes were (1) an aqueous alkaline mixture of sodium nitrate and sodium nitrite; (2) a chlorinated hydrocarbon mixture; (3) a simulant liquid scintillation fluid; and (4) a mixture of ketones. The screening testing protocol involved exposing the respective materials to ∼3 kGy of gamma radiation followed by 14-day exposures to the waste simulants at 60 deg. C. The rubber materials or elastomers were tested using VTR measurements while the liner materials were tested using specific gravity as a metric. For these tests, screening criteria of ∼1 g/hr/m 2 for VTR and specific gravity change of 10% were used. Those materials that failed to meet these criteria were judged to have failed the screening tests and were excluded from the next phase of this experimental program. We have completed the comprehensive testing phase of liner materials in a simulant Hanford Tank waste consisting of an aqueous alkaline mixture of sodium nitrate and sodium nitrite. From the data analyses performed, we have identified the chloro-fluorocarbon Kel-F as having the greatest chemical durability after having been exposed to gamma radiation followed by exposure to the aqueous alkaline simulant mixed waste. The most striking observation from this study was the extremely poor performance of Teflon under these conditions. We have also completed the comprehensive

Physical and Liquid Chemical Simulant Formulations for Transuranic Waste in Hanford Single-Shell Tanks

International Nuclear Information System (INIS)

Rassat, Scot D.; Bagaasen, Larry M.; Mahoney, Lenna A.; Russell, Renee L.; Caldwell, Dustin D.; Mendoza, Donaldo P.

2003-01-01

CH2M HILL Hanford Group, Inc. (CH2M HILL) is in the process of identifying and developing supplemental process technologies to accelerate the tank waste cleanup mission. A range of technologies is being evaluated to allow disposal of Hanford waste types, including transuranic (TRU) process wastes. Ten Hanford single-shell tanks (SSTs) have been identified whose contents may meet the criteria for designation as TRU waste: the B-200 series (241-B-201, -B-202, -B 203, and B 204), the T-200 series (241-T-201, T 202, -T-203, and -T-204), and Tanks 241-T-110 and -T-111. CH2M HILL has requested vendor proposals to develop a system to transfer and package the contact-handled TRU (CH-TRU) waste retrieved from the SSTs for subsequent disposal at the Waste Isolation Pilot Plant (WIPP). Current plans call for a modified ''dry'' retrieval process in which a liquid stream is used to help mobilize the waste for retrieval and transfer through lines and vessels. This retrieval approach requires that a significant portion of the liquid be removed from the mobilized waste sludge in a ''dewatering'' process such as centrifugation prior to transferring to waste packages in a form suitable for acceptance at WIPP. In support of CH2M HILL's effort to procure a TRU waste handling and packaging process, Pacific Northwest National Laboratory (PNNL) developed waste simulant formulations to be used in evaluating the vendor's system. For the SST CH-TRU wastes, the suite of simulants includes (1) nonradioactive chemical simulants of the liquid fraction of the waste, (2) physical simulants that reproduce the important dewatering properties of the waste, and (3) physical simulants that can be used to mimic important rheological properties of the waste at different points in the TRU waste handling and packaging process. To validate the simulant formulations, their measured properties were compared with the limited data for actual TRU waste samples. PNNL developed the final simulant formulations

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

Hanford tank waste simulants specification and their applicability for the retrieval, pretreatment, and vitrification processes

Energy Technology Data Exchange (ETDEWEB)

GR Golcar; NG Colton; JG Darab; HD Smith

2000-04-04

A wide variety of waste simulants were developed over the past few years to test various retrieval, pretreatment and waste immobilization technologies and unit operations. Experiments can be performed cost-effectively using non-radioactive waste simulants in open laboratories. This document reviews the composition of many previously used waste simulants for remediation of tank wastes at the Hanford reservation. In this review, the simulants used in testing for the retrieval, pretreatment, and vitrification processes are compiled, and the representative chemical and physical characteristics of each simulant are specified. The retrieval and transport simulants may be useful for testing in-plant fluidic devices and in some cases for filtration technologies. The pretreatment simulants will be useful for filtration, Sr/TRU removal, and ion exchange testing. The vitrification simulants will be useful for testing melter, melter feed preparation technologies, and for waste form evaluations.

Hanford tank waste simulants specification and their applicability for the retrieval, pretreatment, and vitrification processes

International Nuclear Information System (INIS)

GR Golcar; NG Colton; JG Darab; HD Smith

2000-01-01

A wide variety of waste simulants were developed over the past few years to test various retrieval, pretreatment and waste immobilization technologies and unit operations. Experiments can be performed cost-effectively using non-radioactive waste simulants in open laboratories. This document reviews the composition of many previously used waste simulants for remediation of tank wastes at the Hanford reservation. In this review, the simulants used in testing for the retrieval, pretreatment, and vitrification processes are compiled, and the representative chemical and physical characteristics of each simulant are specified. The retrieval and transport simulants may be useful for testing in-plant fluidic devices and in some cases for filtration technologies. The pretreatment simulants will be useful for filtration, Sr/TRU removal, and ion exchange testing. The vitrification simulants will be useful for testing melter, melter feed preparation technologies, and for waste form evaluations

ALLIANCES: simulation platform for radioactive waste disposal

International Nuclear Information System (INIS)

Deville, E.; Montarnal, Ph.; Loth, L.; Chavant, C.

2009-01-01

CEA, ANDRA and EDF are jointly developing the software platform ALLIANCES whose aim is to produce a tool for the simulation of nuclear waste storage and disposal. This type of simulations deals with highly coupled thermo-hydro-mechanical-chemical and radioactive (T-H-M-C-R) processes. ALLIANCES' aim is to accumulate within the same simulation environment the already acquired knowledge and to gradually integrate new knowledge. The current version of ALLIANCES contains the following modules: - Hydraulics and reactive transport in unsaturated and saturated media; - Multi-phase flow; - Mechanical thermal-hydraulics; - Thermo-Aeraulics; - Chemistry/Transport coupling in saturated media; - Alteration of waste package coupled with the environment; - Sensitivity analysis tools. The next releases will include more physical phenomena like: reactive transport in unsaturated flow and multicomponent multiphase flow; incorporation of responses surfaces in sensitivity analysis tools; integration of parallel numerical codes for flow and transport. Since the distribution of the first release of ALLIANCES (December 2003), the platform was used by ANDRA for his safety simulation program and by CEA for reactive transport simulations (migration of uranium in a soil, diffusion of different reactive species on laboratory samples, glass/iron/clay interaction). (authors)

Mathematical simulation of a waste rock heap

International Nuclear Information System (INIS)

Scharer, J.M.; Pettit, C.M.; Chambers, D.B.; Kwong, E.C.

1994-01-01

A computer model has been developed to simulate the generation of acidic drainage in waste rock piles. The model considers the kinetic rates of biological and chemical oxidation of sulfide minerals (pyrite, pyrrhotite) present as fines and rock particles, as well as chemical processes such as dissolution (kinetic or equilibrium controlled), complexation (from equilibrium and stoichiometry of several complexes), and precipitation (formation of complexes and secondary minerals). Through mass balance equations and solubility constraints (e.g., pH, phase equilibria) the model keeps track of the movement of chemical species through the waste pile and provides estimates of the quality of seepage (pH, sulfate, iron, acidity, etc.) leaving the heap. The model has been expanded to include the dissolution (thermodynamic and sorption equilibrium), adsorption and coprecipitation of uranium and radium. The model was applied to simulate waste rock heaps in British Columbia, Canada and in Thueringia, Germany. To improve the accuracy and confidence of long-term predictions of seepage quality, the entire history of the heaps was simulated. Cumulative acidity loads and water treatment considerations were used as a basis for evaluation of various decommissioning alternatives. Simulation of the technical leaching history of a heap in Germany showed it will generate contaminated leachate requiring treatment for acidity and radioactivity for several hundred years; cover installation was shown to provide a significant reduction of potential burdens, although chemical treatment would still be required beyond 100 years

GENERAL REQUIREMENTS FOR SIMULATION MODELS IN WASTE MANAGEMENT

International Nuclear Information System (INIS)

Miller, Ian; Kossik, Rick; Voss, Charlie

2003-01-01

Most waste management activities are decided upon and carried out in a public or semi-public arena, typically involving the waste management organization, one or more regulators, and often other stakeholders and members of the public. In these environments, simulation modeling can be a powerful tool in reaching a consensus on the best path forward, but only if the models that are developed are understood and accepted by all of the parties involved. These requirements for understanding and acceptance of the models constrain the appropriate software and model development procedures that are employed. This paper discusses requirements for both simulation software and for the models that are developed using the software. Requirements for the software include transparency, accessibility, flexibility, extensibility, quality assurance, ability to do discrete and/or continuous simulation, and efficiency. Requirements for the models that are developed include traceability, transparency, credibility/validity, and quality control. The paper discusses these requirements with specific reference to the requirements for performance assessment models that are used for predicting the long-term safety of waste disposal facilities, such as the proposed Yucca Mountain repository

Study on cementation of simulated radioactive borated liquid wastes

International Nuclear Information System (INIS)

Sun Qina; Li Junfeng; Wang Jianlong

2010-01-01

To compare sulfoaluminate cement with ordinary Portland cement on their cementation of radioactive borated liquid waste and to provide more data for formula optimization, simulated radioactive borated liquid waste were solidified by the two cements. 28 d compressive strength and strength losses after water/freezing/irradiation resistance tests were investigated. Leaching test and X-ray diffraction analysis were also conducted. The results show that it is feasible to solidify borated liquid wastes with sulfoaluminate cement and ordinary Portland cement with formulas used in the study. The 28 d compressive strengths, strength losses after tests and simulated nuclides leaching rates of the solidified waste forms meet the demand of GB 14569.1-93. The sulfoaluminate cement formula show better retention of Cs + than ordinary Portland cement formula. Boron, in form of B (OH) 4 - , incorporate in ettringite as solid solutions. (authors)

System Planning With The Hanford Waste Operations Simulator

International Nuclear Information System (INIS)

Crawford, T.W.; Certa, P.J.; Wells, M.N.

2010-01-01

At the U. S. Department of Energy's Hanford Site in southeastern Washington State, 216 million liters (57 million gallons) of nuclear waste is currently stored in aging underground tanks, threatening the Columbia River. The River Protection Project (RPP), a fully integrated system of waste storage, retrieval, treatment, and disposal facilities, is in varying stages of design, construction, operation, and future planning. These facilities face many overlapping technical, regulatory, and financial hurdles to achieve site cleanup and closure. Program execution is ongoing, but completion is currently expected to take approximately 40 more years. Strategic planning for the treatment of Hanford tank waste is by nature a multi-faceted, complex and iterative process. To help manage the planning, a report referred to as the RPP System Plan is prepared to provide a basis for aligning the program scope with the cost and schedule, from upper-tier contracts to individual facility operating plans. The Hanford Tank Waste Operations Simulator (HTWOS), a dynamic flowsheet simulation and mass balance computer model, is used to simulate the current planned RPP mission, evaluate the impacts of changes to the mission, and assist in planning near-term facility operations. Development of additional modeling tools, including an operations research model and a cost model, will further improve long-term planning confidence. The most recent RPP System Plan, Revision 4, was published in September 2009.

Description of waste pretreatment and interfacing systems dynamic simulation model

International Nuclear Information System (INIS)

Garbrick, D.J.; Zimmerman, B.D.

1995-05-01

The Waste Pretreatment and Interfacing Systems Dynamic Simulation Model was created to investigate the required pretreatment facility processing rates for both high level and low level waste so that the vitrification of tank waste can be completed according to the milestones defined in the Tri-Party Agreement (TPA). In order to achieve this objective, the processes upstream and downstream of the pretreatment facilities must also be included. The simulation model starts with retrieval of tank waste and ends with vitrification for both low level and high level wastes. This report describes the results of three simulation cases: one based on suggested average facility processing rates, one with facility rates determined so that approximately 6 new DSTs are required, and one with facility rates determined so that approximately no new DSTs are required. It appears, based on the simulation results, that reasonable facility processing rates can be selected so that no new DSTs are required by the TWRS program. However, this conclusion must be viewed with respect to the modeling assumptions, described in detail in the report. Also included in the report, in an appendix, are results of two sensitivity cases: one with glass plant water recycle steams recycled versus not recycled, and one employing the TPA SST retrieval schedule versus a more uniform SST retrieval schedule. Both recycling and retrieval schedule appear to have a significant impact on overall tank usage

Date palm waste gasification in downdraft gasifier and simulation using ASPEN HYSYS

International Nuclear Information System (INIS)

Bassyouni, M.; Waheed ul Hasan, Syed; Abdel-Aziz, M.H.; Abdel-hamid, S.M.-S.; Naveed, Shahid; Hussain, Ahmed; Ani, Farid Nasir

2014-01-01

Highlights: • Simulation of date palm waste gasification using ASPEN HYSYS was studied. • A steady state simulation of downdraft gasifier has been developed. • The results were used to predict synthesis gas composition. • Simulation results and experimental results are in good agreement. - Abstract: The present research aims to study the simulation of date palm waste gasification using ASPEN HYSYS. A steady state simulation of downdraft gasifier firing date palm leaves has been developed. The model is able to predict syngas composition with sound accuracy and can be used to find optimal operating conditions of the gasifier. Biomass is defined as an unconventional hypothetical solid component in HYSYS. A set of six reactor models simulates various reaction zones of the downdraft gasifier in accordance with its hydrodynamics. Biomass decomposition into constituents in the pyrolysis zone is modeled with a conversion reactor. The combustion of char and volatiles in the combustion zone are modeled with equilibrium and Gibbs reactor models respectively. The gasification zone is modeled with a Gibbs and equilibrium reactor. The results of simulation are validated against experimental results of a parametric variability study on a lab scale gasifier. The proportion of synthesis gas increase as temperature increases (concentration, molar fraction, and partial pressure). CO 2 and CH 4 in the product gases were also found to decrease with increasing temperature. At 800 °C, the exit gas reaches a stable molar composition (H 2 = 56.27%, CO = 21.71%, CO 2 = 18.24%, CH 4 = 3.78%). Increasing steam to biomass ratio increases CO 2 and H 2 at the expense of CO, governed by shift reaction. Steam induction increases the methane contents, thereby improves the heating value of the product gas

Date palm waste gasification in downdraft gasifier and simulation using ASPEN HYSYS

Energy Technology Data Exchange (ETDEWEB)

Bassyouni, M. [Department of Chemical and Materials Engineering, King Abdulaziz University, Rabigh 21911 (Saudi Arabia); Department of Chemical Engineering, Higher Technological Institute, Tenth of Ramdan City (Egypt); Waheed ul Hasan, Syed [Department of Chemical and Materials Engineering, King Abdulaziz University, Rabigh 21911 (Saudi Arabia); Abdel-Aziz, M.H., E-mail: helmy2002@gmail.com [Department of Chemical and Materials Engineering, King Abdulaziz University, Rabigh 21911 (Saudi Arabia); Chemical Engineering Department, Faculty of Engineering, Alexandria University, Alexandria (Egypt); Abdel-hamid, S. M.-S. [Department of Chemical Engineering, Higher Technological Institute, Tenth of Ramdan City (Egypt); Naveed, Shahid [Punjab Institute of Contemporary Sciences, 5.5 KM Raiwind Road, Lahore (Pakistan); Hussain, Ahmed [Department of Nuclear Engineering, King Abdulaziz University, Jeddah 21589 (Saudi Arabia); Ani, Farid Nasir [Faculty of Mechanical Engineering, Universiti Teknologi Malaysia, UTM 81310 Johor Bahru (Malaysia)

2014-12-15

Highlights: • Simulation of date palm waste gasification using ASPEN HYSYS was studied. • A steady state simulation of downdraft gasifier has been developed. • The results were used to predict synthesis gas composition. • Simulation results and experimental results are in good agreement. - Abstract: The present research aims to study the simulation of date palm waste gasification using ASPEN HYSYS. A steady state simulation of downdraft gasifier firing date palm leaves has been developed. The model is able to predict syngas composition with sound accuracy and can be used to find optimal operating conditions of the gasifier. Biomass is defined as an unconventional hypothetical solid component in HYSYS. A set of six reactor models simulates various reaction zones of the downdraft gasifier in accordance with its hydrodynamics. Biomass decomposition into constituents in the pyrolysis zone is modeled with a conversion reactor. The combustion of char and volatiles in the combustion zone are modeled with equilibrium and Gibbs reactor models respectively. The gasification zone is modeled with a Gibbs and equilibrium reactor. The results of simulation are validated against experimental results of a parametric variability study on a lab scale gasifier. The proportion of synthesis gas increase as temperature increases (concentration, molar fraction, and partial pressure). CO{sub 2} and CH{sub 4} in the product gases were also found to decrease with increasing temperature. At 800 °C, the exit gas reaches a stable molar composition (H{sub 2} = 56.27%, CO = 21.71%, CO{sub 2} = 18.24%, CH{sub 4} = 3.78%). Increasing steam to biomass ratio increases CO{sub 2} and H{sub 2} at the expense of CO, governed by shift reaction. Steam induction increases the methane contents, thereby improves the heating value of the product gas.

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

A coupled approach for the three-dimensional simulation of pipe leakage in variably saturated soil

Science.gov (United States)

Peche, Aaron; Graf, Thomas; Fuchs, Lothar; Neuweiler, Insa

2017-12-01

In urban water pipe networks, pipe leakage may lead to subsurface contamination or to reduced waste water treatment efficiency. The quantification of pipe leakage is challenging due to inaccessibility and unknown hydraulic properties of the soil. A novel physically-based model for three-dimensional numerical simulation of pipe leakage in variably saturated soil is presented. We describe the newly implemented coupling between the pipe flow simulator HYSTEM-EXTRAN and the groundwater flow simulator OpenGeoSys and its validation. We further describe a novel upscaling of leakage using transfer functions derived from numerical simulations. This upscaling enables the simulation of numerous pipe defects with the benefit of reduced computation times. Finally, we investigate the response of leakage to different time-dependent pipe flow events and conclude that larger pipe flow volume and duration lead to larger leakage while the peak position in time has a small effect on leakage.

Modeling by GASP-IV simulation of high-level nuclear waste disposal

International Nuclear Information System (INIS)

Kurstedt, H.A. Jr.; DePorter, E.L.; Turek, J.L.; Funk, S.K.; Rasbach, C.E.

1981-01-01

High-level nuclear waste generated by defense-oriented and commercial nuclear energy activities are to be stored ultimately in underground repositories. Research continues on the waste-form and waste-form processing. DOE managers must coordinate the results of this research, the capacities and availability times of the permanent geologic storage repositories, and the capacities and availability times of interim storage facilities (pending availability of permanent repositories). Comprehensive and active DOE program-management information systems contain predicted generation of nuclear wastes from defense and commercial activities; milestones on research on waste-forms; and milestones on research and development, design, acquisition, and construction of facilities and repositories. A GASP IV simulation model is presented which interfaces all of these data. The model accepts alternate management decisions; relates all critical milestones, all research and development data, and the generation of waste nuclear materials; simulates the passage of time; then, predicts the impact of those alternate decisions on the availability of storage capacity for waste nuclear materials. 3 references, 3 figures

Simulation used to qualify nuclear waste glass for disposal

International Nuclear Information System (INIS)

Reimus, T.W.; Kuhn, W.L.

1987-07-01

A hypothetical vitrification system was simulated errors associated with controlling and predicting the composition of the nuclear waste glass produced in the system. The composition of the glass must fall within certain limits to qualify for permanent geologic disposal. The estimated error in predicting the concentrations of various constituents in the glass was 2% to 8%, depending on the strategy for sampling and analyzing the feed and on the assumed magnitudes of the process uncertainties. The estimated error in controlling the glass composition was 2% to 9%, depending on the strategy for sampling and analyzing the waste and on the assumed magnitudes of the uncertainties. This work demonstrates that simulation techniques can be used to assist in qualifying nuclear waste glass for disposal. 3 refs., 2 figs., 4 tabs

Rheological evaluation of simulated neutralized current acid waste

International Nuclear Information System (INIS)

Fow, C.L.; McCarthy, D.; Thornton, G.T.

1986-06-01

A byproduct of the Purex process is an aqueous waste stream that contains fission products. This waste stream, called current acid waste, is chemically neutralized and stored in double shell tanks on the Hanford Site. This neutralized current acid waste (NCAW) will be transported by pipe to B-Plant, a processing plant on the Hanford Site. Rheological and transport properties of NCAW slurry were evaluated. First, researchers conducted lab rheological evaluations of simulated NCAW. The results of these evaluations were then correlated with classical rheological models and scaled up to predict the performance that is likely to occur in the full-scale system. The NCAW in the tank will either be retrieved as is, i.e., no change in the concentration presently in the tank, or will be slightly concentrated before retrieval. Sluicing may be required to retrieve the solids. Three concentrations of simulated NCAW were evaluated that would simulate the different retrieval options: NCAW in the concentration that is presently in the tank; a slightly concentrated NCAW, called NCAW5.5; and equal parts of NCAW settled solids and water (simulating the sluicing stage), called NCAW1:1. The physical and rheological properties of three samples of each concentration at 25 and 100 0 C were evaluated in the laboratory. The properties displayed by NCAW and NCAW5.5 at 25 and 100 0 C allowed it to be classified as a pseudoplastic non-Newtonian fluid. NCAW1:1 at 25 and 100 0 C displayed properties of a yield-pseudoplastic non-Newtonian fluid. The classical non-Newtonian models for pseudoplastic and yield-pseudoplastic fluids were used with the laboratory data to predict the full-scale pump-pipe network parameters

A PC-based discrete event simulation model of the civilian radioactive waste management system

International Nuclear Information System (INIS)

Airth, G.L.; Joy, D.S.; Nehls, J.W.

1992-01-01

This paper discusses a System Simulation Model which has been developed for the Department of Energy to simulate the movement of individual waste packages (spent fuel assemblies and fuel containers) through the Civilian Radioactive Waste Management System (CRWMS). A discrete event simulation language, GPSS/PC, which runs on an IBM/PC and operates under DOS 5.0, mathematically represents the movement and processing of radioactive waste packages through the CRWMS and the interaction of these packages with the equipment in the various facilities. The major features of the System Simulation Model are: the ability to reference characteristics of the different types of radioactive waste (age, burnup, etc.) in order to make operational and/or system design decisions, the ability to place stochastic variations on operational parameters such as processing time and equipment outages, and the ability to include a rigorous simulation of the transportation system. Output from the model includes the numbers, types, and characteristics of waste packages at selected points in the CRWMS and the extent to which various resources will be utilized in order to transport, process, and emplace the waste

Corrosion studies of carbon steel under impinging jets of simulated slurries of neutralized current acid waste (NCAW) and neutralized cladding removal waste (NCRW)

International Nuclear Information System (INIS)

Smith, H.D.; Elmore, M.R.

1992-01-01

Plans for the disposal of radioactive liquid and solid wastes presently stored in double-shell tanks at the Hanford Site call for retrieval and processing of the waste to create forms suitable for permanent disposal. Waste will be retrieved from a tank using a submerged slurry pump in conjunction with one or more rotating slurry jet mixer pumps. Pacific Northwest Laboratory (PNL) has conducted tests using simulated waste slurries to assess the effects of a impinging slurry jet on the corrosion rate of the tank wall and floor, an action that could potentially compromise the tank's structural integrity. Corrosion processes were investigated on a laboratory scale with a simulated neutralized cladding removal waste (NCRW) slurry and in a subsequent test with simulated neutralized current acid waste (NCAW) slurry. The test slurries simulated the actual NCRW and NCAW both chemically and physically. The tests simulated those conditions expected to exist in the respective double-shell tanks during waste retrieval operations. Results of both tests indicate that, because of the action of the mixer pump slurry jets, the waste retrieval operations proposed for NCAW and NCRW will moderately accelerate corrosion of the tank wall and floor. Based on the corrosion of initially unoxidized test specimens, and the removal of corrosion products from those specimens, the maximum time-averaged corrosion rates of carbon steel in both waste simulants for the length of the test was ∼4 mil/yr. The protective oxide layer that exists in each storage tank is expected to inhibit corrosion of the carbon steel

Simulating the structure of gypsum composites using pulverized basalt waste

Directory of Open Access Journals (Sweden)

Buryanov Аleksandr

2017-01-01

Full Text Available This paper examines the possibility of simulating the structure of gypsum composite modified with basalt dust waste to make materials and products based on it. Structural simulating of the topological space in gypsum modified composite by optimizing its grain-size composition highly improves its physical and mechanical properties. Strength and density tests have confirmed the results of the simulation. The properties of modified gypsum materials are improved by obtaining of denser particle packing in the presence of hemihydrate of finely dispersed basalt and plasticizer particles in the system, and by engaging basalt waste in the structuring process of modified gypsum stone.

Transport of nuclear waste flows - a modelling and simulation approach - 59136

International Nuclear Information System (INIS)

Adams, Jonathan F.W.; Biggs, Simon R.; Fairweather, Michael; Yao, Jun; Young, James

2012-01-01

The task of implementing safer and more efficient processing and transport techniques in the handling of nuclear wastes made up of liquid-solid mixtures provides a challenging and interesting area of research. The radioactive nature of nuclear waste means that it is difficult to perform experimental studies of its transport. In contrast, the use of modelling and simulation techniques can help to elucidate the physics that underpin such flows and provide valuable insights into common problems associated with their transport, as well as assisting in the focusing of experimental research. Two phase solid-liquid waste-forms are commonplace within the nuclear reprocessing industry. Currently, there is waste, e.g., in the form of a solid-liquid slurry in cooling ponds and liquid flows containing suspensions of solid particles feature heavily in the treatment and disposal of this waste. With nuclear waste in the form of solid-liquid sludges it is important to understand the nature of the flow, with particular interest in the settling characteristics of the particulate waste material. Knowledge of the propensity of pipe flows to form solid beds is important in avoiding unwanted blockages in pipelines and pumping systems. In cases where the formation of a solid bed is unavoidable, it is similarly important to know how the modified cross-sectional area of the pipe, due to the presence of a bed, will affect particle behaviour through the creation of secondary flows effects that are also common to square duct flows. A greater understanding of particle deposition in square ducts and pipes of circular cross-section is also of significant and broad industrial relevance, with flows containing particulates prevalent throughout the nuclear, pharmaceutical, chemical, mining and agricultural industries. A greater understanding of particle behaviour in square ducts and circular pipes with variable bed height is the focus of this current work. The more computationally expensive but

Powder technological vitrification of simulated high-level waste

International Nuclear Information System (INIS)

Gahlert, S.

1988-03-01

High-level waste simulate from the reprocessing of light water reactor and fast breeder fuel was vitrified by powder technology. After denitration with formaldehyde, the simulated HLW is mixed with glass frit and simultaneously dried in an oil-heated mixer. After 'in-can calcination' for at least 24 hours at 850 or 950 K (depending on the type of waste and glass), the mixture is hot-pressed in-can for several hours at 920 or 1020 K respectively, at pressures between 0.4 and 1.0 MPa. The technology has been demonstrated inactively up to diameters of 30 cm. Leach resistance is significantly enhanced when compared to common borosilicate glasses by the utilization of glasses with higher silicon and aluminium content and lower sodium content. (orig.) [de

Defense Waste Processing Facility Process Simulation Package Life Cycle

International Nuclear Information System (INIS)

Reuter, K.

1991-01-01

The Defense Waste Processing Facility (DWPF) will be used to immobilize high level liquid radioactive waste into safe, stable, and manageable solid form. The complexity and classification of the facility requires that a performance based operator training to satisfy Department of Energy orders and guidelines. A major portion of the training program will be the application and utilization of Process Simulation Packages to assist in training the Control Room Operators on the fluctionality of the process and the application of the Distribution Control System (DCS) in operating and managing the DWPF process. The packages are being developed by the DWPF Computer and Information Systems Simulation Group. This paper will describe the DWPF Process Simulation Package Life Cycle. The areas of package scope, development, validation, and configuration management will be reviewed and discussed in detail

Quantitative measurement of cyanide species in simulated ferrocyanide Hanford waste

International Nuclear Information System (INIS)

Bryan, S.A.; Pool, K.H.; Matheson, J.D.

1993-02-01

Analytical methods for the quantification of cyanide species in Hanford simulated high-level radioactive waste were pursued in this work. Methods studied include infrared spectroscopy (solid state and solution), Raman spectroscopy, Moessbauer spectroscopy, X-ray diffraction, scanning electron microscopy-electron dispersive spectroscopy (SEM-EDS), and ion chromatography. Of these, infrared, Raman, X-ray diffraction, and ion chromatography techniques show promise in the concentration range of interest. Quantitation limits for these latter four techniques were demonstrated to be approximately 0.1 wt% (as cyanide) using simulated Hanford wastes

Simulation of construction and demolition waste leachate

Energy Technology Data Exchange (ETDEWEB)

Townsend, T.G.; Jang, Y.; Thurn, L.G.

1999-11-01

Solid waste produced from construction and demolition (C and D) activities is typically disposed of in unlined landfills. Knowledge of C{ampersand}D debris landfill leachate is limited in comparison to other types of wastes. A laboratory study was performed to examine leachate resulting from simulated rainfall infiltrating a mixed C and D waste stream consisting of common construction materials (e.g., concrete, wood, drywall). Lysimeters (leaching columns) filled with the mixed C and D waste were operated under flooded and unsaturated conditions. Leachate constituent concentrations in the leachate from specific waste components were also examined. Leachate samples were collected and analyzed for a number of conventional water quality parameters including pH, alkalinity, total organic carbon, total dissolved solids, and sulfate. In experiments with the mixed C and D waste, high concentrations of total dissolved solids (TDS) and sulfate were detected in the leachate. C and D leachates produced as a result of unsaturated conditions exhibited TDS concentrations in the range of 570--2,200 mg/L. The major contributor to the TDS was sulfate, which ranged in concentration between 280 and 930 mg/L. The concentrations of sulfate in the leachate exceeded the sulfate secondary drinking water standard of 250 mg/L.

Using simulation to assess the opportunities of dynamic waste collection

NARCIS (Netherlands)

Mes, Martijn R.K.; Bangsow, S.

2012-01-01

In this chapter, we illustrate the use of discrete event simulation to evaluate how dynamic planning methodologies can be best applied for the collection of waste from underground containers. We present a case study that took place at the waste collection company Twente Milieu, located in The

Using Simulation to Assess the Opportunities of Dynamic Waste Collection

NARCIS (Netherlands)

Mes, Martijn R.K.

In this paper, we illustrate the use of discrete event simulation to evaluate how dynamic planning methodologies can be best applied for the collection of waste from underground containers. We present a case study that took place at the waste collection company Twente Milieu, located in The

Temporal Variability of Observed and Simulated Hyperspectral Earth Reflectance

Science.gov (United States)

Roberts, Yolanda; Pilewskie, Peter; Kindel, Bruce; Feldman, Daniel; Collins, William D.

2012-01-01

The Climate Absolute Radiance and Refractivity Observatory (CLARREO) is a climate observation system designed to study Earth's climate variability with unprecedented absolute radiometric accuracy and SI traceability. Observation System Simulation Experiments (OSSEs) were developed using GCM output and MODTRAN to simulate CLARREO reflectance measurements during the 21st century as a design tool for the CLARREO hyperspectral shortwave imager. With OSSE simulations of hyperspectral reflectance, Feldman et al. [2011a,b] found that shortwave reflectance is able to detect changes in climate variables during the 21st century and improve time-to-detection compared to broadband measurements. The OSSE has been a powerful tool in the design of the CLARREO imager and for understanding the effect of climate change on the spectral variability of reflectance, but it is important to evaluate how well the OSSE simulates the Earth's present-day spectral variability. For this evaluation we have used hyperspectral reflectance measurements from the Scanning Imaging Absorption Spectrometer for Atmospheric Cartography (SCIAMACHY), a shortwave spectrometer that was operational between March 2002 and April 2012. To study the spectral variability of SCIAMACHY-measured and OSSE-simulated reflectance, we used principal component analysis (PCA), a spectral decomposition technique that identifies dominant modes of variability in a multivariate data set. Using quantitative comparisons of the OSSE and SCIAMACHY PCs, we have quantified how well the OSSE captures the spectral variability of Earth?s climate system at the beginning of the 21st century relative to SCIAMACHY measurements. These results showed that the OSSE and SCIAMACHY data sets share over 99% of their total variance in 2004. Using the PCs and the temporally distributed reflectance spectra projected onto the PCs (PC scores), we can study the temporal variability of the observed and simulated reflectance spectra. Multivariate time

Long-term durability experiments with concrete-based waste packages in simulated repository conditions

International Nuclear Information System (INIS)

Ipatti, A.

1993-03-01

Two extensive experiments on long-term durability of waste packages in simulated repository conditions are described. The first one is a 'half-scale experiment' comprising radioactive waste product and half-scale concrete containers in site specific groundwater conditions. The second one is 'full-scale experiment' including simulated inactive waste product and full-scale concrete container stored in slowly flowing fresh water. The scope of the experiments is to demonstrate long-term behaviour of the designed waste packages in contact with moderately concrete aggressive groundwater, and to evaluate the possible interactions between the waste product, concrete container and ground water. As the waste packages are made of high-quality concrete, provisions have been made to continue the experiments for several years

Application of stochastic dynamic simulation to waste form qualification for the HWVP vitrification process

International Nuclear Information System (INIS)

Kuhn, W.L.; Westsik, J.H. Jr.

1989-01-01

Processing steps during the conversion of high-level nuclear waste into borosilicate glass in the Hanford Waste Vitrification Plant are being simulated on a computer by addressing transient mass balances. The results are being used to address the US Department of Energy's Waste Form Qualification requirements. The simulated addresses discontinuous (batch) operations and perturbations in the transient behavior of the process caused by errors in measurements and control actions. A collection of tests, based on process measurements, is continually checked and used to halt the simulated process when specified conditions are met. An associated set of control actions is then implemented in the simulation. The results for an example simulation are shown. 8 refs

Internal Interdecadal Variability in CMIP5 Control Simulations

Science.gov (United States)

Cheung, A. H.; Mann, M. E.; Frankcombe, L. M.; England, M. H.; Steinman, B. A.; Miller, S. K.

2015-12-01

Here we make use of control simulations from the CMIP5 models to quantify the amplitude of the interdecadal internal variability component in Atlantic, Pacific, and Northern Hemisphere mean surface temperature. We compare against estimates derived from observations using a semi-empirical approach wherein the forced component as estimated using CMIP5 historical simulations is removed to yield an estimate of the residual, internal variability. While the observational estimates are largely consistent with those derived from the control simulations for both basins and the Northern Hemisphere, they lie in the upper range of the model distributions, suggesting the possibility of differences between the amplitudes of observed and modeled variability. We comment on some possible reasons for the disparity.

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

Development of simulated tank wastes for the US Department of Energy's Underground Storage Tank Integrated Demonstration

International Nuclear Information System (INIS)

Elmore, M.R.; Colton, N.G.; Jones, E.O.

1992-08-01

The purpose of the Underground Storage Tank Integrated Demonstration (USTID) is to identify and evaluate technologies that may be used to characterize, retrieve, treat, and dispose of hazardous and radioactive wastes contained in tanks on US Department of Energy sites. Simulated wastes are an essential component of the evaluation process because they provide controlled samples for technology assessment, and minimize costs and risks involved when working with radioactive wastes. Pacific Northwest Laboratory has developed a recipe to simulate Hanford single-shell tank, (SST) waste. The recipe is derived from existing process recipes, and elemental concentrations are based on characterization data from 18 SSTs. In this procedure, salt cake and metal oxide/hydroxide sludge are prepared individually, and mixed together at varying ratios depending on the specific tank, waste to be simulated or the test being conducted. Elemental and physical properties of the stimulant are comparable with analyzed tank samples, and chemical speciation in the simulant is being improved as speciation data for actual wastes become available. The nonradioactive chemical waste simulant described here is useful for testing technologies on a small scale

Response of ethylene propylene diene monomer rubber (EPDM) to simulant Hanford tank waste

Energy Technology Data Exchange (ETDEWEB)

NIGREY,PAUL J.

2000-02-01

This report presents the findings of the Chemical Compatibility Program developed to evaluate plastic packaging components that may be incorporated in packaging mixed-waste forms for transportation. Consistent with the methodology outlined in this report, the author performed the second phase of this experimental program to determine the effects of simulant Hanford tank mixed wastes on packaging seal materials. That effort involved the comprehensive testing of five plastic liner materials in an aqueous mixed-waste simulant. The testing protocol involved exposing the materials to {approximately}143, 286, 571, and 3,670 krad of gamma radiation and was followed by 7-, 14-, 28-, 180-day exposures to the waste simulant at 18, 50, and 60 C. Ethylene propylene diene monomer (EPDM) rubber samples subjected to the same protocol were then evaluated by measuring seven material properties: specific gravity, dimensional changes, mass changes, hardness, compression set, vapor transport rates, and tensile properties. The author has determined that EPDM rubber has excellent resistance to radiation, this simulant, and a combination of these factors. These results suggest that EPDM is an excellent seal material to withstand aqueous mixed wastes having similar composition to the one used in this study.

A system dynamics-based environmental performance simulation of construction waste reduction management in China.

Science.gov (United States)

Ding, Zhikun; Yi, Guizhen; Tam, Vivian W Y; Huang, Tengyue

2016-05-01

A huge amount of construction waste has been generated from increasingly higher number of construction activities than in the past, which has significant negative impacts on the environment if they are not properly managed. Therefore, effective construction waste management is of primary importance for future sustainable development. Based on the theory of planned behaviors, this paper develops a system dynamic model of construction waste reduction management at the construction phase to simulate the environmental benefits of construction waste reduction management. The application of the proposed model is shown using a case study in Shenzhen, China. Vensim is applied to simulate and analyze the model. The simulation results indicate that source reduction is an effective waste reduction measure which can reduce 27.05% of the total waste generation. Sorting behaviors are a premise for improving the construction waste recycling and reuse rates which account for 15.49% of the total waste generated. The environmental benefits of source reduction outweigh those of sorting behaviors. Therefore, to achieve better environmental performance of the construction waste reduction management, attention should be paid to source reduction such as low waste technologies and on-site management performance. In the meantime, sorting behaviors encouragement such as improving stakeholders' waste awareness, refining regulations, strengthening government supervision and controlling illegal dumping should be emphasized. Copyright © 2016 Elsevier Ltd. All rights reserved.

TRANSPORT OF WASTE SIMULANTS IN PJM VENT LINES

Energy Technology Data Exchange (ETDEWEB)

Qureshi, Z

2007-02-21

The experimental work was conducted to determine whether there is a potential for waste simulant to transport or 'creep' up the air link line and contaminate the pulse jet vent system, and possibly cause long term restriction of the air link line. Additionally, if simulant creep occurred, establish operating parameters for washing down the line. The amount of the addition of flush fluids and mixer downtime must be quantified.

Pyrolysis of plastic packaging waste: A comparison of plastic residuals from material recovery facilities with simulated plastic waste

International Nuclear Information System (INIS)

Adrados, A.; Marco, I. de; Caballero, B.M.; López, A.; Laresgoiti, M.F.; Torres, A.

2012-01-01

Highlights: ► Pyrolysis of plastic waste. ► Comparison of different samples: real waste, simulated and real waste + catalyst. ► Study of the effects of inorganic components in the pyrolysis products. - Abstract: Pyrolysis may be an alternative for the reclamation of rejected streams of waste from sorting plants where packing and packaging plastic waste is separated and classified. These rejected streams consist of many different materials (e.g., polyethylene (PE), polypropylene (PP), polystyrene (PS), polyvinyl chloride (PVC), polyethylene terephthalate (PET), acrylonitrile butadiene styrene (ABS), aluminum, tetra-brik, and film) for which an attempt at complete separation is not technically possible or economically viable, and they are typically sent to landfills or incinerators. For this study, a simulated plastic mixture and a real waste sample from a sorting plant were pyrolyzed using a non-stirred semi-batch reactor. Red mud, a byproduct of the aluminum industry, was used as a catalyst. Despite the fact that the samples had a similar volume of material, there were noteworthy differences in the pyrolysis yields. The real waste sample resulted, after pyrolysis, in higher gas and solid yields and consequently produced less liquid. There were also significant differences noted in the compositions of the compared pyrolysis products.

An MCNP simulation for API applications to waste management issues

International Nuclear Information System (INIS)

Tunnell, L.N.

1994-01-01

Issues associated with waste management have increasingly become a focal point of attention for both the government and private sector since the end of the cold war. The problem are difficult to solve; the solutions are expensive to implement. Consequently, the development of a data simulation system capable of predicting the performance of a real system can save many thousands of dollars in travel expenses, optimization of experimental parameters, etc.. In this effort, computer codes were developed to simulate the production of associated particle imaging data so that its performance in a typical waste management application can be assessed

A PC-based discrete event simulation model of the Civilian Radioactive Waste Management System

International Nuclear Information System (INIS)

Airth, G.L.; Joy, D.S.; Nehls, J.W.

1991-01-01

A System Simulation Model has been developed for the Department of Energy to simulate the movement of individual waste packages (spent fuel assemblies and fuel containers) through the Civilian Radioactive Waste Management System (CRWMS). A discrete event simulation language, GPSS/PC, which runs on an IBM/PC and operates under DOS 5.0, mathematically represents the movement and processing of radioactive waste packages through the CRWMS and the interaction of these packages with the equipment in the various facilities. This model can be used to quantify the impacts of different operating schedules, operational rules, system configurations, and equipment reliability and availability considerations on the performance of processes comprising the CRWMS and how these factors combine to determine overall system performance for the purpose of making system design decisions. The major features of the System Simulation Model are: the ability to reference characteristics of the different types of radioactive waste (age, burnup, etc.) in order to make operational and/or system design decisions, the ability to place stochastic variations on operational parameters such as processing time and equipment outages, and the ability to include a rigorous simulation of the transportation system. Output from the model includes the numbers, types, and characteristics of waste packages at selected points in the CRWMS and the extent to which various resources will be utilized in order to transport, process, and emplace the waste

Leaching behavior of a simulated bituminized radioactive waste form under deep geological conditions

International Nuclear Information System (INIS)

Nakayama, Shinichi; Iida, Yoshihisa; Nagano, Tetsushi; Akimoto, Toshiyuki

2003-01-01

The leaching behavior of a simulated bituminized waste form was studied to acquire data for the performance assessment of the geologic disposal of bituminized radioactive waste. Laboratory-scale leaching tests were performed for radioactive and non-radioactive waste specimens simulating bituminized waste of a French reprocessing company, COGEMA. The simulated waste was contacted with deionized water, an alkaline solution (0.03-mol/l KOH), and a saline solution (0.5-mol/l KCl) under atmospheric and anoxic conditions. The concentrations of Na, Ba, Cs, Sr, Np, Pu, NO 3 , SO 4 and I in the leachates were determined. Swelling of the bituminized waste progressed in deionized water and KOH. The release of the soluble components, Na and Cs, was enhanced by the swelling, and considered to be diffusion-controlled in the swelled layers of the specimens. The release of sparingly soluble components such as Ba and Np was solubility-limited in addition to the progression of leaching. Neptunium, a redox-sensitive element, showed a distinct difference in release between anoxic and atmospheric conditions. The elemental release from the bituminized waste specimens leached in the KCl was very low, which is likely due to the suppression of swelling of the specimens at high ionic strength. (author)

Rheological properties of kaolin and chemically simulated waste

International Nuclear Information System (INIS)

Selby, C.L.

1981-12-01

The Savannah River Laboratory is conducting tests to determine the best operating conditions of pumps used to transfer insoluble radioactive sludges from old to new waste tanks. Because it is not feasible to conduct these tests with real or chemically simulated sludges, kaolin clay is being used as a stand-in for the solid waste. The rheology tests described herein were conducted to determine whether the properties of kaolin were sufficiently similar to those of real sludge to permit meaningful pump tests. The rheology study showed that kaolin can be substituted for real waste to accurately determine pump performance. Once adequately sheared, kaolin properties were found to remain constant. Test results determined that kaolin should not be allowed to settle more than two weeks between pump tests. Water or supernate from the waste tanks can be used to dilute sludge on an equal volume basis because they identically affect the rheological properties of sludge. It was further found that the fluid properties of kaolin and waste are insensitive to temperature

MCNP efficiency calculations of INEEL passive active neutron assay system for simulated TRU waste assays

International Nuclear Information System (INIS)

Yoon, W.Y.; Meachum, T.R.; Blackwood, L.G.; Harker, Y.D.

2000-01-01

The Idaho National Engineering and Environmental Laboratory Stored Waste Examination Pilot Plant (SWEPP) passive active neutron (PAN) radioassay system is used to certify transuranic (TRU) waste drums in terms of quantifying plutonium and other TRU element activities. Depending on the waste form involved, significant systematic and random errors need quantification in addition to the counting statistics. To determine the total uncertainty of the radioassay results, a statistical sampling and verification approach has been developed. In this approach, the total performance of the PAN nondestructive assay system is simulated using the computer models of the assay system, and the resultant output is compared with the known input to assess the total uncertainty. The supporting steps in performing the uncertainty analysis for the passive assay measurements in particular are as follows: (1) Create simulated waste drums and associated conditions; (2) Simulate measurements to determine the basic counting data that would be produced by the PAN assay system under the conditions specified; and (3) Apply the PAN assay system analysis algorithm to the set of counting data produced by simulating measurements to determine the measured plutonium mass. The validity of this simulation approach was verified by comparing simulated output against results from actual measurements using known plutonium sources and surrogate waste drums. The computer simulation of the PAN system performance uses the Monte Carlo N-Particle (MCNP) Code System to produce a neutron transport calculation for a simulated waste drum. Specifically, the passive system uses the neutron coincidence counting technique, utilizing the spontaneous fission of 240 Pu. MCNP application to the SWEPP PAN assay system uncertainty analysis has been very useful for a variety of waste types contained in 208-ell drums measured by a passive radioassay system. The application of MCNP to the active radioassay system is also feasible

From mineral processing to waste treatment: an open-mind process simulator

International Nuclear Information System (INIS)

Guillaneau, J.C.; Brochot, S.; Durance, M.V.; Villeneuve, J.; Fourniguet, G.; Vedrine, H.; Sandvik, K.; Reuter, M.

1999-01-01

More than two hundred companies are using the USIM PAC process simulator within the mineral industry world-wide. Either for design or plant adaptation, simulation is increasingly supporting the process Engineer in his activities. From the mineral field, new domains have been concerned by this model-based approach as new models are developed and new applications involving solid waste appears. Examples are presented in bio-processing, steel-making flue dust treatment for zinc valorisation, soil decontamination or urban waste valorisation (sorting, composting and incineration). (author)

Processing of high-temperature simulated waste glass in a continuous ceramic melter

International Nuclear Information System (INIS)

Barnes, S.M.; Brouns, R.A.; Hanson, M.S.

1980-01-01

Recent operations have demonstrated that high-melting-point glasses and glass-ceramics can be successfully processed in joule-heated, ceramic-lined melters with minor modifications to the existing technology. Over 500 kg of simulated waste glasses have been processed at temperatures up to 1410 0 C. The processability of the two high-temperature waste forms tested is similar to existing borosilicate waste glasses. High-temperature waste glass formulations produced in the bench-scale melter exhibit quality comparing favorably to standard waste glass formulations

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

Simulation analysis and ternary diagram of municipal solid waste pyrolysis and gasification based on the equilibrium model.

Science.gov (United States)

Deng, Na; Zhang, Awen; Zhang, Qiang; He, Guansong; Cui, Wenqian; Chen, Guanyi; Song, Chengcai

2017-07-01

A self-sustained municipal solid waste (MSW) pyrolysis-gasification process with self-produced syngas as heat source was proposed and an equilibrium model was established to predict the syngas reuse rate considering variable MSW components. Simulation results indicated that for constant moisture (ash) content, with the increase of ash (moisture) content, syngas reuse rate gradually increased, and reached the maximum 100% when ash (moisture) content was 73.9% (60.4%). Novel ternary diagrams with moisture, ash and combustible as axes were proposed to predict the adaptability of the self-sustained process and syngas reuse rate for waste. For wastes of given components, its position in the ternary diagram can be determined and the syngas reuse rate can be obtained, which will provide guidance for system design. Assuming that the MSW was composed of 100% combustible content, ternary diagram shows that there was a minimum limiting value of 43.8% for the syngas reuse rate in the process. Copyright © 2017. Published by Elsevier Ltd.

Results Of The Extraction-Scrub-Strip Testing Using An Improved Solvent Formulation And Salt Waste Processing Facility Simulated Waste

International Nuclear Information System (INIS)

Peters, T.; Washington, A.; Fink, S.

2012-01-01

The Office of Waste Processing, within the Office of Technology Innovation and Development, is funding the development of an enhanced solvent - also known as the next generation solvent (NGS) - for deployment at the Savannah River Site to remove cesium from High Level Waste. The technical effort is a collaborative effort between Oak Ridge National Laboratory (ORNL) and Savannah River National Laboratory (SRNL). As part of the program, the Savannah River National Laboratory (SRNL) has performed a number of Extraction-Scrub-Strip (ESS) tests. These batch contact tests serve as first indicators of the cesium mass transfer solvent performance with actual or simulated waste. The test detailed in this report used simulated Tank 49H material, with the addition of extra potassium. The potassium was added at 1677 mg/L, the maximum projected (i.e., a worst case feed scenario) value for the Salt Waste Processing Facility (SWPF). The results of the test gave favorable results given that the potassium concentration was elevated (1677 mg/L compared to the current 513 mg/L). The cesium distribution value, DCs, for extraction was 57.1. As a comparison, a typical D Cs in an ESS test, using the baseline solvent formulation and the typical waste feed, is ∼15. The Modular Caustic Side Solvent Extraction Unit (MCU) uses the Caustic-Side Solvent Extraction (CSSX) process to remove cesium (Cs) from alkaline waste. This process involves the use of an organic extractant, BoBCalixC6, in an organic matrix to selectively remove cesium from the caustic waste. The organic solvent mixture flows counter-current to the caustic aqueous waste stream within centrifugal contactors. After extracting the cesium, the loaded solvent is stripped of cesium by contact with dilute nitric acid and the cesium concentrate is transferred to the Defense Waste Processing Facility (DWPF), while the organic solvent is cleaned and recycled for further use. The Salt Waste Processing Facility (SWPF), under

Cross-flow filtration during the washing of a simulated radioactive waste stream

International Nuclear Information System (INIS)

MARK R., DUIGNAN

2005-01-01

Bechtel National, Inc. has been contracted by the Department of Energy to design a Waste Treatment and Immobilization Plant (WTP) to stabilize liquid radioactive waste that is stored at the Hanford Site as part of the River Protection Project (RPP). Because of its experience with radioactive waste stabilization, the Savannah River National Laboratory (SRNL) of the Westinghouse Savannah River Company is working with Bechtel and Washington Group International, to help design and test certain parts of the waste treatment facility. One part of the process is the separation of radioactive solids from the liquid wastes by cross-flow ultrafiltration. To test this process a cross-flow filter was used that was prototypic in porosity, length, and diameter, along with a simulated radioactive waste slurry, made to prototypically represent the chemical and physical characteristics of a Hanford waste in tank 241-AY-102/C-106. To mimic the filtration process the waste slurry undergoes several steps, including dewatering and washing. During dewatering the concentration of undissolved solids (UDS) of the simulated AY102/C106 waste is increased from 12 wt percent to at least 20 wt percent. Once at the higher concentration the waste must be washed to prepare for its eventual receipt in a High Level Radioactive Waste Melter to be vitrified. This paper describes the process of washing and filtering a batch of concentrated simulated waste in two cycles, which each containing 22 washing steps that used approximately 7.7 liters of a solution of 0.01 M NaOH per step. This will be the method used by the full-scale WTP to prepare the waste for vitrification. The first washing cycle started with the simulated waste that had a solids concentration of 20 wt percent UDS. This cycle began with a permeate filter flux of 0.015 gpm/ft2 (3.68 cm/hr) at 19.6 wt percent UDS with a density of 1.33 kg/L, and yield stress of 8.5 Pa. At the end of the 22 washing steps the permeate filter flux increased to

Nuclear waste repository simulation experiments

International Nuclear Information System (INIS)

Rothfuchs, T.; Wieczorek, K.; Feddersen, H.K.; Staupendahl, G.; Coyle, A.J.; Kalia, H.; Eckert, J.

1986-12-01

This document is the third joint annual report on the Cooperative German-American 'Brine Migration Tests' that are in progress at the Asse salt mine in the Federal Republic of Germany (FRG). This Government supported mine serves as an underground test facility for research and development (R and D)-work in the field of nuclear waste repository research and simulation experiments. The tests are designed to simulate a nuclear waste repository to measure the effects of heat and gamma radiation on brine migration, salt decrepitation, disassociation of brine, and gases collected. The thermal mechanical behavior of salt, such as room closure, stresses and changes of the properties of salt are measured and compared with predicted behavior. This document covers the following sections: Issues and test objectives: This section presents issues that are investigated by the Brine Migration Test, and the test objectives derived from these issues; test site: This section describes the test site location and geology in the Asse mine; test description: A description of the test configuration, procedures, equipment, and instrumentation is given in this section; actual test chronology: The actual history of the test, in terms of the dates at which major activities occured, is presented in this section. Test results: This section presents the test results observed to data and the planned future work that is needed to complete the test; conclusions and recommendations: This section summarizes the conclusions derived to date regarding the Brine Migration Test. Additional work that would be useful to resolve the issues is discussed. (orig.)

Process simulation and uncertainty analysis of plasma arc mixed waste treatment

International Nuclear Information System (INIS)

Ferrada, J.J.; Welch, T.D.

1994-01-01

Innovative mixed waste treatment subsystems have been analyzed for performance, risk, and life-cycle cost as part of the U.S. Department of Energy's (DOE)'s Mixed Waste Integrated Program (MWIP) treatment alternatives development and evaluation process. This paper concerns the analysis of mixed waste treatment system performance. Performance systems analysis includes approximate material and energy balances and assessments of operability, effectiveness, and reliability. Preliminary material and energy balances of innovative processes have been analyzed using FLOW, an object-oriented, process simulator for waste management systems under development at Oak Ridge National Laboratory. The preliminary models developed for FLOW provide rough order-of-magnitude calculations useful for sensitivity analysis. The insight gained from early modeling of these technologies approximately will ease the transition to more sophisticated simulators as adequate performance and property data become available. Such models are being developed in ASPEN by DOE's Mixed Waste Treatment Project (MWTP) for baseline and alternative flow sheets based on commercial technologies. One alternative to the baseline developed by the MWIP support groups in plasma arc treatment. This process offers a noticeable reduction in the number of process operations as compared to the baseline process because a plasma arc melter is capable of accepting a wide variety of waste streams as direct inputs (without sorting or preprocessing). This innovative process for treating mixed waste replaces several units from the baseline process and, thus, promises an economic advantage. The performance in the plasma arc furnace will directly affect the quality of the waste form and the requirements of the off-gas treatment units. The ultimate objective of MWIP is to reduce the amount of final waste produced, the cost, and the environmental impact

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.

Quantifying and mapping spatial variability in simulated forest plots

Science.gov (United States)

Gavin R. Corral; Harold E. Burkhart

2016-01-01

We used computer simulations to test the efficacy of multivariate statistical methods to detect, quantify, and map spatial variability of forest stands. Simulated stands were developed of regularly-spaced plantations of loblolly pine (Pinus taeda L.). We assumed no affects of competition or mortality, but random variability was added to individual tree characteristics...

A testing program to evaluate the effects of simulant mixed wastes on plastic transportation packaging components

International Nuclear Information System (INIS)

Nigrey, P.J.; Dickens, T.G.

1997-01-01

Based on regulatory requirements for Type A and B radioactive material packaging, a Testing Program was developed to evaluate the effects of mixed wastes on plastic materials which could be used as liners and seals in transportation containers. The plastics evaluated in this program were butadiene-acrylonitrile copolymer (Nitrile rubber), cross-linked polyethylene, epichlorohydrin, ethylene-propylene rubber (EPDM), fluorocarbons, high-density polyethylene (HDPE), butyl rubber, polypropylene, polytetrafluoroethylene, and styrene-butadiene rubber (SBR). These plastics were first screened in four simulant mixed wastes. The liner materials were screened using specific gravity measurements and seal materials by vapor transport rate (VTR) measurements. For the screening of liner materials, Kel-F, HDPE, and XLPE were found to offer the greatest resistance to the combination of radiation and chemicals. The tests also indicated that while all seal materials passed exposure to the aqueous simulant mixed waste, EPDM and SBR had the lowest VTRs. In the chlorinated hydrocarbon simulant mixed waste, only Viton passed the screening tests. In both the simulant scintillation fluid mixed waste and the ketone mixture waste, none of the seal materials met the screening criteria. Those materials which passed the screening tests were subjected to further comprehensive testing in each of the simulant wastes. The materials were exposed to four different radiation doses followed by exposure to a simulant mixed waste at three temperatures and four different exposure times (7, 14, 28, 180 days). Materials were tested by measuring specific gravity, dimensional, hardness, stress cracking, VTR, compression set, and tensile properties. The second phase of this Testing Program involving the comprehensive testing of plastic liner has been completed and for seal materials is currently in progress

Fixing of various simulated radioactive wastes in urea-formaldehyde resin

International Nuclear Information System (INIS)

Du Dahai; Wei Peng

1986-01-01

This paper outlines the results of the fixing of a variety of simulated radioactive wastes in the urea-formaldehyde resin. The radioactive waste materials fixed include spent ion exchange resin, concentrates of NaNO 3 -NaBO 2 as well as NaBO 2 and sludge. The performance of the fixed products has been improved by means of selecting the synthetic conditions of resin, a suitable hardener and an inorganic additive

Applying fluid dynamics simulations to improve processing and remediation of nuclear waste - 59172

International Nuclear Information System (INIS)

Knight, Kelly J.; Peltier, Joel; Berkoe, Jon; Rosendall, Brigette; Kennedy, Chris

2012-01-01

Transport and processing of nuclear waste for treatment and storage can involve unique and complex thermal and fluid dynamic conditions that pose potential for safety risk and/or design uncertainty and also are likely to be subjected to more precise performance requirements than in other industries. From an engineering analysis perspective, certainty of outcome is essential. Advanced robust methods for engineering analysis and simulation of critical processes can help reduce risk of design uncertainty and help mitigate or reduce the amount of expensive full-scale demonstration testing. This paper will discuss experience gained in applying computational fluid dynamics models to key processes for mixing, transporting, and thermal treatment of nuclear waste as part of designing a massive vitrification process plant that will convert high and low level nuclear waste into glass for permanent storage. Examples from industrial scale simulations will be presented. The computational models have shown promise in replicating several complex physical processes such as solid-liquid flows in suspension, blending of slurries, and cooling of materials at extremely high temperature. Knowledge gained from applying simulation has provided detailed insight into determining the most critical aspects of these complex processes that can ultimately be used to help guide the optimum design of waste handling equipment based on credible calculations while ensuring risk of design uncertainty is minimized. The WTP Project is faced with complex technical challenges that must have solutions that enable the successful operation of the plant for its 30+ year operating life. The Project chose to reduce those risks by employing an experienced team that applied CFD in a disciplined manner and adhered to an established guideline with the following benefits: - Gained an improvement in accuracy of predictions for complex physical situations; - Gained an improvement of the quality of experimental

Compatibility of packaging components with simulant mixed waste

International Nuclear Information System (INIS)

Nigrey, P.J.; Dickens, T.G.

1996-01-01

The purpose of hazardous and radioactive materials packaging is to enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations in the US have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified by the US Department of Transportation (US DOT, 49 CFR 173) and the US Nuclear Regulatory Commission (NRC, 10 CFR 71). Based on these national requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program provides a basis to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. In this paper, the authors present the results of the second phase of this testing program. The first phase screened five liner materials and six seal materials towards four simulant mixed wastes. This phase involved the comprehensive testing of five candidate liner materials to an aqueous Hanford Tank simulant mixed waste. The comprehensive testing protocol involved exposing the respective materials a matrix of four gamma radiation doses (∼ 1, 3, 6, and 40 kGy), three temperatures (18, 50, and 60 C), and four exposure times (7, 14, 28, and 180 days). Following their exposure to these combinations of conditions, the materials were evaluated by measuring five material properties. These properties were specific gravity, dimensional changes, hardness, stress cracking, and mechanical properties

Design of Stirrer Impeller with Variable Operational Speed for a Food Waste Homogenizer

Directory of Open Access Journals (Sweden)

Idris A. Kayode

2016-05-01

Full Text Available A conceptualized impeller called KIA is designed for impact agitation of food waste in a homogenizer. A comparative analysis of the performance of KIA is made with three conventional impeller types, Rushton, Anchor, and Pitched Blade. Solid–liquid mixing of a moisture-rich food waste is simulated under various operational speeds, in order to compare the dispersions and thermal distributions at homogenous slurry conditions. Using SolidWorks, the design of the impellers employs an Application Programming Interface (API which acts as the canvas for creating a graphical user interface (GUI for automation of its assembly. A parametric analysis of the homogenizer, at varying operational speeds, enables the estimation of the critical speed of the mixing shaft diameter and the deflection under numerous mixing conditions and impeller configurations. The numerical simulation of the moisture-rich food waste (approximated as a Newtonian carrot–orange soup is performed with ANSYS CFX v.15.0. The velocity and temperature field distribution of the homogenizer for various impeller rotational speeds are analyzed. It is anticipated that the developed model will help in the selection of a suitable impeller for efficient mixing of food waste in the homogenizer.

Metals and polybrominated diphenyl ethers leaching from electronic waste in simulated landfills

Energy Technology Data Exchange (ETDEWEB)

Kiddee, Peeranart [Centre for Environmental Risk Assessment and Remediation, University of South Australia, Mawson Lakes Campus, Adelaide, 5095 (Australia); Cooperative Research Centre for Contamination Assessment and Remediation of the Environment, Mawson Lakes Campus, Adelaide, 5095 (Australia); Naidu, Ravi, E-mail: ravi.naidu@crccare.com [Centre for Environmental Risk Assessment and Remediation, University of South Australia, Mawson Lakes Campus, Adelaide, 5095 (Australia); Cooperative Research Centre for Contamination Assessment and Remediation of the Environment, Mawson Lakes Campus, Adelaide, 5095 (Australia); Wong, Ming H. [Croucher Institute for Environmental Sciences, and Department of Biology, Hong Kong Baptist University, Kowloon Tong (China)

2013-05-15

Highlights: • Simulated landfill columns provided realistic results than lab based column study. • Column leachates showed significant seasonal effect on toxic substances. • Toxic substances in the landfill leachates pose environmental and health hazards. • A better management of e-waste is urgently needed. -- Abstract: Landfills established prior to the recognition of potential impacts from the leaching of heavy metals and toxic organic compounds often lack appropriate barriers and pose significant risks of contamination of groundwater. In this study, bioavailable metal(oids) and polybrominated diphenyl ethers (PBDEs) in leachates from landfill columns that contained intact or broken e-waste were studied under conditions that simulate landfills in terms of waste components and methods of disposal of e-wastes, and with realistic rainfall. Fourteen elements and PBDEs were analysed in leachates over a period of 21 months. The results demonstrate that the average concentrations of Al, Ba, Be, Cd, Co, Cr, Cu, Ni, Pb, Sb and V in leachates from the column that contained broken e-waste items were significantly higher than the column without e-waste. BDE-153 was the highest average PBDEs congener in all columns but the average of ∑PBDEs levels in columns that contained intact e-waste were (3.7 ng/l) and were not significantly higher than that in the leachates from the control column.

Evaporation Of Hanford Waste Treatment Plant Direct Feed Low Activity Waste Effluent Management Facility Core Simulant

Energy Technology Data Exchange (ETDEWEB)

Adamson, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Mcclane, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-09-01

stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfates in the recycled Condensate and is a key outcome of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of this task was to demonstrate evaporation of a simulant of the LAW Melter Off-gas Condensate expected during DFLAW operations, in order to predict the composition of the effluents from the EMF evaporator to aid in planning for their disposition. This document describes the results of that test using the core simulant. This simulant formulation is designated as the “core simulant”; other additives will be included for specific testing, such as volatiles for evaporation or hazardous metals for measuring leaching properties of waste forms. The results indicate that the simulant can easily be concentrated via evaporation. During that the pH adjustment step in simulant preparation, ammonium is quickly converted to ammonia, and most of the ammonia was stripped from the simulated waste and partitioned to the condensate. Additionally, it was found that after concentrating (>12x) and cooling that a small amount of LiF and Na₃(SO₄)F precipitate out of solution. With the exception of ammonia, analysis of the condensate indicated very low to below detectable levels of many of the constituents in the simulant, yielding very high decontamination factors (DF).

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

Formulation development for PREPP concreted waste forms

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Welch, J.M.

1984-05-01

Analysis of variance and logistic regression techniques have been used to develop models describing the effects of formulation variables and their interactions on compressive strength, solidification, free-standing water, and workability of hydraulic cement grouts incorporating simulated Process Experimental Pilot Plant (PREPP) wastes. These models provide the basis for specifications of grout formulations to solidify these wastes. The experimental test matrix, formulation preparation, and test methods employed are described. The development of analytical models for formulation behavior and the conclusions drawn regarding appropriate formulation variable ranges are discussed. 13 references, 9 figures, 15 tables

Fuzzy Simulation-Optimization Model for Waste Load Allocation

Directory of Open Access Journals (Sweden)

Motahhare Saadatpour

2006-01-01

Full Text Available This paper present simulation-optimization models for waste load allocation from multiple point sources which include uncertainty due to vagueness of the parameters and goals. This model employs fuzzy sets with appropriate membership functions to deal with uncertainties due to vagueness. The fuzzy waste load allocation model (FWLAM incorporate QUAL2E as a water quality simulation model and Genetic Algorithm (GA as an optimization tool to find the optimal combination of the fraction removal level to the dischargers and pollution control agency (PCA. Penalty functions are employed to control the violations in the system.  The results demonstrate that the goal of PCA to achieve the best water quality and the goal of the dischargers to use the full assimilative capacity of the river have not been satisfied completely and a compromise solution between these goals is provided. This fuzzy optimization model with genetic algorithm has been used for a hypothetical problem. Results demonstrate a very suitable convergence of proposed optimization algorithm to the global optima.

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

Waste Evaporator Accident Simulation Using RELAP5 Computer Code

International Nuclear Information System (INIS)

POLIZZI, L.M.

2004-01-01

An evaporator is used on liquid waste from processing facilities to reduce the volume of the waste through heating the waste and allowing some of the water to be separated from the waste through boiling. This separation process allows for more efficient processing and storage of liquid waste. Commonly, the liquid waste consists of an aqueous solution of chemicals that over time could induce corrosion, and in turn weaken the tubes in the steam tube bundle of the waste evaporator that are used to heat the waste. This chemically induced corrosion could escalate into a possible tube leakage and/or the severance of a tube(s) in the tube bundle. In this paper, analyses of a waste evaporator system for the processing of liquid waste containing corrosive chemicals are presented to assess the system response to this accident scenario. This accident scenario is evaluated since its consequences can propagate to a release of hazardous material to the outside environment. It is therefore important to ensure that the evaporator system component structural integrity is not compromised, i.e. the design pressure and temperature of the system is not exceeded during the accident transient. The computer code used for the accident simulation is RELAP5-MOD31. The accident scenario analyzed includes a double-ended guillotine break of a tube in the tube bundle of the evaporator. A mitigated scenario is presented to evaluate the excursion of the peak pressure and temperature in the various components of the evaporator system to assess whether the protective actions and controls available are adequate to ensure that the structural integrity of the evaporator system is maintained and that no atmospheric release occurs

Variable-density ground-water flow and paleohydrology in the Waste Isolation Pilot Plant (WIPP) region, southeastern New Mexico

International Nuclear Information System (INIS)

Davies, P.B.

1989-01-01

Variable-density groundwater flow was studied near the Waste Isolation Pilot Plant in southeastern New Mexico. An analysis of the relative magnitude of pressure-related and density-related flow-driving forces indicates that density-related gravity effects are not significant at the plant and to the west but are significant in areas to the north, northeast, and south. A regional-scale model of variable-density groundwater flow in the Culebra Dolomite member of the Rustler Formation indicates that the flow velocities are relatively rapid west of the site and extremely slow east and northeast of the site. In the transition zone between those two extremes, which includes the plant, velocities are highly variable. Sensitivity simulations indicates that the central and western parts of the region, including the plant, are fairly well isolated from the eastern and northeastern boundaries. Vertical-flux simulations indicate that as much as 25% of total inflow to the Culebra could be entering as vertical flow, with most of this flow occurring west of the plant. A simple cross-sectional model was developed to examine the flow system as it drains through time following recharge during a past glacial pluvial. This model indicates that the system as a whole drains very slowly and that it apparently could have sustained flow from purely transient drainage following recharge of the system during the Pleistocene

Extraction of technetium from simulated Hanford tank wastes

International Nuclear Information System (INIS)

Chaiko, D.J.; Vojta, Y.; Takeuchi, M.

1993-01-01

Aqueous biphasic separation systems are being developed for the treatment of liquid radioactive wastes. These extraction systems are based on the use of polyethylene glycols (PEGs) for the selective extraction and recovery of long-lived radionuclides, such as 129 I, 75 Se, and 99 Tc, from caustic solutions containing high concentrations of nitrate, nitrite, and carbonate. Because of the high ionic strengths of supernatant liquids in Hanford underground storage tanks, aqueous biphasic systems can be generated by simply adding aqueous PEG solutions directly to the waste solution. In the process, anionic species like I - and TcO 4 - are selectively transferred to the less dense PEG phase. The partition coefficient for a wide range of inorganic cations and anions, such as sodium, potassium, aluminum, nitrate, nitrate, and carbonate, are all less than one. The authors present experimental data on extraction of technetium from several simulated Hanford tank wastes at 25 degree and 50 degree C

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

Waste Isolation Pilot Plant simulated RH TRU waste experiments: Data and interpretation pilot

International Nuclear Information System (INIS)

Molecke, M.A.; Argueello, G.J.; Beraun, R.

1993-04-01

The simulated, i.e., nonradioactive remote-handled transuranic waste (RH TRU) experiments being conducted underground in the Waste Isolation Pilot Plant (WIPP) were emplaced in mid-1986 and have been in heated test operation since 9/23/86. These experiments involve the in situ, waste package performance testing of eight full-size, reference RH TRU containers emplaced in horizontal, unlined test holes in the rock salt ribs (walls) of WIPP Room T. All of the test containers have internal electrical heaters; four of the test emplacements were filled with bentonite and silica sand backfill materials. We designed test conditions to be ''near-reference'' with respect to anticipated thermal outputs of RH TRU canisters and their geometrical spacing or layout in WIPP repository rooms, with RH TRU waste reference conditions current as of the start date of this test program. We also conducted some thermal overtest evaluations. This paper provides a: detailed test overview; comprehensive data update for the first 5 years of test operations; summary of experiment observations; initial data interpretations; and, several status; experimental objectives -- how these tests support WIPP TRU waste acceptance, performance assessment studies, underground operations, and the overall WIPP mission; and, in situ performance evaluations of RH TRU waste package materials plus design details and options. We provide instrument data and results for in situ waste container and borehole temperatures, pressures exerted on test containers through the backfill materials, and vertical and horizontal borehole-closure measurements and rates. The effects of heat on borehole closure, fracturing, and near-field materials (metals, backfills, rock salt, and intruding brine) interactions were closely monitored and are summarized, as are assorted test observations. Predictive 3-dimensional thermal and structural modeling studies of borehole and room closures and temperature fields were also performed

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

FFSM, Long-Term Nuclear Waste Repository Site Simulation by Monte-Carlo

International Nuclear Information System (INIS)

Hadlock, L.R.; Hellstrom, D.I.; Mikulis, M.J.B.; Little, A.D.; Golis, M.J.

1988-01-01

1 - Description of program or function: FFSM (Far Field State Model) predicts the approximate geologic and climatic state of a site for a nuclear waste repository over relatively long periods of time. The purpose of FFSM is to represent quantitatively certain events and processes that could alter the effectiveness of one or more natural barriers in a waste isolation system. The barriers treated by the model are primarily components of the geologic environment surrounding the repository, although biosphere components (e.g. climate parameters) that could affect the impact of radionuclide releases are also considered. These components are treated outside the realm of wastes or repository-induced effects, which is indicated by use of the term f ar field . The model treats both natural and man-induced changes in these barriers within a probabilistic framework, and it accounts for cumulative and interactive effects of multiple phenomena. 2 - Method of solution: Fifteen submodels are included in FFSM to account for phenomena that may be of importance individually or in combination in evaluating sites for repositories. These submodels include: undetected features, climate, worldwide glaciation, local glaciation, folding, salt dispersion, magmatic events, faulting, biosphere state, regional deformation, geomorphic processes, dissolution fronts, localized dissolution (breccia pipes), solution mining, and drilling. FFSM can be used in both a deterministic mode, to evaluate interactions or to calculate point values, and a probabilistic mode, to make statistical estimates of future changes. In the probabilistic mode, Monte Carlo simulation is used to generate output probabilities, based on user-supplied input, largely in the form of probability density functions for variable or uncertain parameters

Numerical simulations of waste forms from the reprocessing of nuclear fuel

International Nuclear Information System (INIS)

Schneider, Stephan

2014-01-01

The usage of fissile material for nuclear fuel causes that alongside radioactive wastes are produced. These waste materials are created during all handling or usage operations within the nuclear fuel cycle. The main source of radiotoxicity is produced during the usage of nuclear fuel within the reactor. Energy is released by neutron induced fission reactions in heavy isotopes. Parts of the created fission products have large radiotoxicities. Due to neutron capture within the nuclear fuel the radiotoxicity is furthermore increased. These waste streams from the nuclear fuel cycle must be stored in a safe way to prevent any contamination of the biosphere and any harm to the civilization or the environment. The waste packages must be treated and conditioned for the final disposal. These created packages are subject to an independent product control to ensure there acceptability for transport, interim and final storage. The independent product control is a significant component of an effective waste management system. The aim of this work is the development of a software system used for the assessment of radioactive waste packages. The software shall permit the auditor to perform scenario analysis to forecast the product properties of a certain waste stream and therefore optimize the needed inspection scope in preparation of a new campaign. The software is designed as a modular library this permits the most flexible use of the software components and a high reusability of written analysis software. The software system is used for coupling of established and well-known simulation programs used for nuclear systems. The results of Monte-Carlo simulations and burn-up calculations are automatically imported and prepared for user interaction. The usage of simulation programs cause different challenges to the computing infrastructure. The scenario analyses need a large number of parameter variations which are bound to the computing time. For this reason additional to the

Constant extension rate testing of Type 304L stainless steel in simulated waste tank environments

International Nuclear Information System (INIS)

Wiersma, B.J.

1992-01-01

New tanks for storage of low level radioactive wastes will be constructed at the Savannah River Site (SRS) of AISI Type 304L stainless steel (304L). The presence of chlorides and fluorides in the wastes may induce Stress Corrosion Cracking (SCC) in 304L. Constant Extension Rate Tests (CERT) were performed to determine the susceptibility of 304L to SCC in simulated wastes. In five of the six tests conducted thus far 304L was not susceptible to SCC in the simulated waste environments. Conflicting results were obtained in the final test and will be resolved by further tests. For comparison purposes the CERT tests were also performed with A537 carbon steel, a material similar to that utilized for the existing nuclear waste storage tanks at SRS

A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests

International Nuclear Information System (INIS)

Thien, Mike G.; Barnes, Steve M.

2013-01-01

The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described

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

Effect of fluidization number on the combustion of simulated municipal solid waste in a fluidized bed

International Nuclear Information System (INIS)

Anwar Johari; Mutahharah, M.M.; Abdul, A.; Salema, A.; Kalantarifard, A.; Rozainee, M.

2010-01-01

The effect of fluidization number on the combustion of simulated municipal solid was in a fluidized bed was investigated. Simulated municipal solid waste was used a sample and it was formulated from major waste composition found in Malaysia which comprised of food waste, paper, plastic and vegetable waste. Proximate and ultimate analyses of the simulated were conducted and results showed its composition was similar to the actual Malaysian municipal solid waste composition. Combustion study was carried out in a rectangular fluidized bed with sand of mean particle size of 0.34 mm as a fluidising medium. The range of fluidization numbers investigated was 3 to 11 U mf . The combustion was carried out at stoichiometric condition (Air Factor = 1). Results showed that the best fluidization number was in the range of 5 to 7 U mf with 5 U mf being the most optimum in which the bed temperature was sustained in a much longer period. (author)

Program for simulation of reduction of effluents

NARCIS (Netherlands)

Grenier, P.; Vilette, P.; Mignonac, J.M.; Leclercq, P.A.; Gayda, A.; Cosson, J.L.; Moletta, R.; Racault, Y.

1994-01-01

Food industry and wineries in particular are responsible for a significant part of the org. industrial pollution in France. This pollution has been characterized at the Caves du Sieur d'Arques winery. Dynamic simulation of waste rejection should permit to est. the variability of inputs at the waste

Practical utilization of modeling and simulation in laboratory process waste assessments

International Nuclear Information System (INIS)

Lyttle, T.W.; Smith, D.M.; Weinrach, J.B.; Burns, M.L.

1993-01-01

At Los Alamos National Laboratory (LANL), facility waste streams tend to be small but highly diverse. Initial characterization of such waste streams is difficult in part due to a lack of tools to assist the waste generators in completing such assessments. A methodology has been developed at LANL to allow process knowledgeable field personnel to develop baseline waste generation assessments and to evaluate potential waste minimization technology. This process waste assessment (PWA) system is an application constructed within the process modeling system. The Process Modeling System (PMS) is an object-oriented, mass balance-based, discrete-event simulation using the common LISP object system (CLOS). Analytical capabilities supported within the PWA system include: complete mass balance specifications, historical characterization of selected waste streams and generation of facility profiles for materials consumption, resource utilization and worker exposure. Anticipated development activities include provisions for a best available technologies (BAT) database and integration with the LANL facilities management Geographic Information System (GIS). The environments used to develop these assessment tools will be discussed in addition to a review of initial implementation results

CAPE-OPEN simulation of waste-to-energy technologies for urban cities

Science.gov (United States)

Andreadou, Christina; Martinopoulos, Georgios

2018-01-01

Uncontrolled waste disposal and unsustainable waste management not only damage the environment, but also affect human health. In most urban areas, municipal solid waste production is constantly increasing following the everlasting increase in energy consumption. Technologies aim to exploit wastes in order to recover energy, decrease the depletion rate of fossil fuels, and reduce waste disposal. In this paper, the annual amount of municipal solid waste disposed in the greater metropolitan area of Thessaloniki is taken into consideration, in order to size and model a combined heat and power facility for energy recovery. From the various waste-to-energy technologies available, a fluidised bed combustion boiler combined heat and power plant was selected and modelled through the use of COCO, a CAPE-OPEN simulation software, to estimate the amount of electrical and thermal energy that could be generated for different boiler pressures. Although average efficiency was similar in all cases, providing almost 15% of Thessaloniki's energy needs, a great variation in the electricity to thermal energy ratio was observed.

Design of Stirrer Impeller with Variable Operational Speed for a Food Waste Homogenizer

OpenAIRE

Idris A. Kayode; Emmanuel O. B. Ogedengbe; Marc A. Rosen

2016-01-01

A conceptualized impeller called KIA is designed for impact agitation of food waste in a homogenizer. A comparative analysis of the performance of KIA is made with three conventional impeller types, Rushton, Anchor, and Pitched Blade. Solid–liquid mixing of a moisture-rich food waste is simulated under various operational speeds, in order to compare the dispersions and thermal distributions at homogenous slurry conditions. Using SolidWorks, the design of the impellers employs an Application P...

Characterisation of Plasma Vitrified Simulant Plutonium Contaminated Material Waste

International Nuclear Information System (INIS)

Hyatt, Neil C.; Morgan, Suzy; Stennett, Martin C.; Scales, Charlie R.; Deegan, David

2007-01-01

The potential of plasma vitrification for the treatment of a simulant Plutonium Contaminated Material (PCM) was investigated. It was demonstrated that the PuO 2 simulant, CeO 2 , could be vitrified in the amorphous calcium iron aluminosilicate component of the product slag with simultaneous destruction of the organic and polymer waste fractions. Product Consistency Tests conducted at 90 deg. C in de-ionised water and buffered pH 11 solution show the PCM slag product to be durable with respect to release of Ce. (authors)

Comparison of existing models to simulate anaerobic digestion of lipid-rich waste.

Science.gov (United States)

Béline, F; Rodriguez-Mendez, R; Girault, R; Bihan, Y Le; Lessard, P

2017-02-01

Models for anaerobic digestion of lipid-rich waste taking inhibition into account were reviewed and, if necessary, adjusted to the ADM1 model framework in order to compare them. Experimental data from anaerobic digestion of slaughterhouse waste at an organic loading rate (OLR) ranging from 0.3 to 1.9kgVSm -3 d -1 were used to compare and evaluate models. Experimental data obtained at low OLRs were accurately modeled whatever the model thereby validating the stoichiometric parameters used and influent fractionation. However, at higher OLRs, although inhibition parameters were optimized to reduce differences between experimental and simulated data, no model was able to accurately simulate accumulation of substrates and intermediates, mainly due to the wrong simulation of pH. A simulation using pH based on experimental data showed that acetogenesis and methanogenesis were the most sensitive steps to LCFA inhibition and enabled identification of the inhibition parameters of both steps. Copyright © 2016 Elsevier Ltd. All rights reserved.

Chemical durability and degradation mechanisms of HT9 based alloy waste forms with variable Zr content

Energy Technology Data Exchange (ETDEWEB)

Olson, L. N. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-10-30

In Corrosion studies were undertaken on alloy waste forms that can result from advanced electrometallurgical processing techniques to better classify their durability and degradation mechanisms. The waste forms were based on the RAW3-(URe) composition, consisting primarily of HT9 steel and other elemental additions to simulate nuclear fuel reprocessing byproducts. The solution conditions of the corrosion studies were taken from an electrochemical testing protocol, and meant to simulate conditions in a repository. The alloys durability was examined in alkaline and acidic brines.

Virtual reality in simulation of operational procedures in radioactive waste deposits

International Nuclear Information System (INIS)

Freitas, Victor Goncalves Gloria

2016-01-01

One of the biggest problems in the nuclear area are still the radioactive waste generated in the various applications of this form of energy, all these tailings are stored in warehouses that often are monitored and restructured for better allocation of then. These tailings are stored until it is safe to release into the environment. This work presents a methodology based on virtual reality, for the development of virtual deposits of radioactive waste in order to enable virtual simulations in these deposits. As application will be developed virtually the nuclear waste repository located at the Institute of Nuclear Engineering IEN/CNEN. The development of a virtual warehouse, more specifically, makes it possible to simulate/train the allocation and reallocation of materials with low and medium level of radioactivity, seen the possibility of locomotion of virtual objects and dynamic calculation of the rate of radiation in this environment. Using this methodology it also possible know the accumulated dose, by the virtual character, during the procedures run in the virtual environment. (author)

Simulations for the transmutation of nuclear wastes with hybrid reactors

International Nuclear Information System (INIS)

Vuillier, St.

1998-06-01

A Monte Carlo simulation, devoted to the spallation, has been built in the framework of the hybrid systems proposed for the nuclear wastes incineration. This system GSPARTE, described the reactions evolution. It takes into account and improves the nuclear codes and the low and high energy particles transport in the GEANT code environment, adapted to the geometry of the hybrid reactors. Many applications and abacus useful for the wastes transmutation, have been realized with this system: production of thick target neutrons, source definition, material damages. (A.L.B.)

Simulating variable source problems via post processing of individual particle tallies

International Nuclear Information System (INIS)

Bleuel, D.L.; Donahue, R.J.; Ludewigt, B.A.; Vujic, J.

2000-01-01

Monte Carlo is an extremely powerful method of simulating complex, three dimensional environments without excessive problem simplification. However, it is often time consuming to simulate models in which the source can be highly varied. Similarly difficult are optimization studies involving sources in which many input parameters are variable, such as particle energy, angle, and spatial distribution. Such studies are often approached using brute force methods or intelligent guesswork. One field in which these problems are often encountered is accelerator-driven Boron Neutron Capture Therapy (BNCT) for the treatment of cancers. Solving the reverse problem of determining the best neutron source for optimal BNCT treatment can be accomplished by separating the time-consuming particle-tracking process of a full Monte Carlo simulation from the calculation of the source weighting factors which is typically performed at the beginning of a Monte Carlo simulation. By post-processing these weighting factors on a recorded file of individual particle tally information, the effect of changing source variables can be realized in a matter of seconds, instead of requiring hours or days for additional complete simulations. By intelligent source biasing, any number of different source distributions can be calculated quickly from a single Monte Carlo simulation. The source description can be treated as variable and the effect of changing multiple interdependent source variables on the problem's solution can be determined. Though the focus of this study is on BNCT applications, this procedure may be applicable to any problem that involves a variable source

Composition, preparation, and gas generation results from simulated wastes of Tank 241-SY-101

International Nuclear Information System (INIS)

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

1994-08-01

This document reviews the preparation and composition of simulants that have been developed to mimic the wastes temporarily stored in Tank 241-SY-101 at Hanford. The kinetics and stoichiometry of gases that are generated using these simulants are also compared, considering the roles of hydroxide, chloride, and transition metal ions; the identities of organic constituents; and the effects of dilution, radiation, and temperature. Work described in this report was conducted for the Flammable Gas Safety Program at Pacific Northwest Laboratory, (a) whose purpose is to develop information that is necessary to mitigate potential safety hazards associated with waste tanks at the Hanford Site. The goal of this research and of related efforts at the Georgia Institute of Technology (GIT), Argonne National Laboratory (ANL), and Westinghouse Hanford Company (WHC) is to determine the thermal and thermal/radiolytic mechanisms by which flammable and other gases are produced in Hanford wastes, emphasizing those stored in Tank 241-SY-101. A variety of Tank 241-SY-101 simulants have been developed to date. The use of simulants in laboratory testing activities provides a number of advantages, including elimination of radiological risks to researchers, lower costs associated with experimentation, and the ability to systematically alter simulant compositions to study the chemical mechanisms of reactions responsible for gas generation. The earliest simulants contained the principal inorganic components of the actual waste and generally a single complexant such as N-(2-hydroxyethyl) ethylenediaminetriacetic acid (HEDTA) or ethylenediaminetriacetic acid (EDTA). Both homogeneous and heterogeneous compositional forms were developed. Aggressive core sampling and analysis activities conducted during Windows C and E provided information that was used to design new simulants that more accurately reflected major and minor inorganic components

Regional simulation of interannual variability over South America

Science.gov (United States)

Misra, V.; Dirmeyer, P. A.; Kirtman, B. P.; Juang, H.-M. Henry; Kanamitsu, M.

2002-08-01

Three regional climate simulations covering the austral summer season during three contrasting phases of the El Niño-Southern Oscillation cycle were conducted with the Regional Spectral Model (RSM) developed at the National Centers for Environmental Prediction (NCEP). The simulated interannual variability of precipitation over the Amazon River Basin, the Intertropical Convergence Zone, the Pacific and Atlantic Ocean basins, and extratropical South America compare reasonably well with observations. The RSM optimally filters the peturbations about a time-varying base field, thereby enhancing the information content of the global NCEP reanalysis. The model is better than the reanalysis in reproducing the observed interannual variability of outgoing longwave radiation at both high frequencies (3-30 days) and intraseasonal (30-60 days) scales. The low-level jet shows a peak in its speed in 1998 and a minimum in the 1999 simulations. The lag correlation of the jet index with convection over various areas in continental South America indicates that the jet induces precipitation over the Pampas region downstream. A detailed moisture budget was conducted over various subregions. This budget reveals that moisture flux convergence determines most of the interannual variability of precipitation over the Amazon Basin, the Atlantic Intertropical Convergence Zone, and the Nordeste region of Brazil. However, both surface evaporation and surface moisture flux convergence were found to be critical in determining the interannual variability of precipitation over the southern Pampas, Gran Chaco area, and the South Atlantic Convergence Zone.

Efficient Simulation Modeling of an Integrated High-Level-Waste Processing Complex

International Nuclear Information System (INIS)

Gregory, Michael V.; Paul, Pran K.

2000-01-01

An integrated computational tool named the Production Planning Model (ProdMod) has been developed to simulate the operation of the entire high-level-waste complex (HLW) at the Savannah River Site (SRS) over its full life cycle. ProdMod is used to guide SRS management in operating the waste complex in an economically efficient and environmentally sound manner. SRS HLW operations are modeled using coupled algebraic equations. The dynamic nature of plant processes is modeled in the form of a linear construct in which the time dependence is implicit. Batch processes are modeled in discrete event-space, while continuous processes are modeled in time-space. The ProdMod methodology maps between event-space and time-space such that the inherent mathematical discontinuities in batch process simulation are avoided without sacrificing any of the necessary detail in the batch recipe steps. Modeling the processes separately in event- and time-space using linear constructs, and then coupling the two spaces, has accelerated the speed of simulation compared to a typical dynamic simulation. The ProdMod simulator models have been validated against operating data and other computer codes. Case studies have demonstrated the usefulness of the ProdMod simulator in developing strategies that demonstrate significant cost savings in operating the SRS HLW complex and in verifying the feasibility of newly proposed processes

DEMONSTRATION OF THE NEXT-GENERATION CAUSTIC-SIDE SOLVENT EXTRACTION SOLVENT WITH 2-CM CENTRIFUGAL CONTRACTORS USING TANK 49H WASTE AND WASTE SIMULANT

Energy Technology Data Exchange (ETDEWEB)

Pierce, R.; Peters, T.; Crowder, M.; Caldwell, T.; Pak, D; Fink, S.; Blessing, R.; Washington, A.

2011-09-27

Researchers successfully demonstrated the chemistry and process equipment of the Caustic-Side Solvent Extraction (CSSX) flowsheet using MaxCalix for the decontamination of high level waste (HLW). The demonstration was completed using a 12-stage, 2-cm centrifugal contactor apparatus at the Savannah River National Laboratory (SRNL). This represents the first CSSX process demonstration of the MaxCalix solvent system with Savannah River Site (SRS) HLW. Two tests lasting 24 and 27 hours processed non-radioactive simulated Tank 49H waste and actual Tank 49H HLW, respectively. Conclusions from this work include the following. The CSSX process is capable of reducing {sup 137}Cs in high level radioactive waste by a factor of more than 40,000 using five extraction, two scrub, and five strip stages. Tests demonstrated extraction and strip section stage efficiencies of greater than 93% for the Tank 49H waste test and greater than 88% for the simulant waste test. During a test with HLW, researchers processed 39 liters of Tank 49H solution and the waste raffinate had an average decontamination factor (DF) of 6.78E+04, with a maximum of 1.08E+05. A simulant waste solution ({approx}34.5 liters) with an initial Cs concentration of 83.1 mg/L was processed and had an average DF greater than 5.9E+03, with a maximum DF of greater than 6.6E+03. The difference may be attributable to differences in contactor stage efficiencies. Test results showed the solvent can be stripped of cesium and recycled for {approx}25 solvent turnovers without the occurrence of any measurable solvent degradation or negative effects from minor components. Based on the performance of the 12-stage 2-cm apparatus with the Tank 49H HLW, the projected DF for MCU with seven extraction, two scrub, and seven strip stages operating at a nominal efficiency of 90% is {approx}388,000. At 95% stage efficiency, the DF in MCU would be {approx}3.2 million. Carryover of organic solvent in aqueous streams (and aqueous in organic

Multiscale thermohydrologic model: addressing variability and uncertainty at Yucca Mountain

International Nuclear Information System (INIS)

Buscheck, T; Rosenberg, N D; Gansemer, J D; Sun, Y

2000-01-01

Performance assessment and design evaluation require a modeling tool that simultaneously accounts for processes occurring at a scale of a few tens of centimeters around individual waste packages and emplacement drifts, and also on behavior at the scale of the mountain. Many processes and features must be considered, including non-isothermal, multiphase-flow in rock of variable saturation and thermal radiation in open cavities. Also, given the nature of the fractured rock at Yucca Mountain, a dual-permeability approach is needed to represent permeability. A monolithic numerical model with all these features requires too large a computational cost to be an effective simulation tool, one that is used to examine sensitivity to key model assumptions and parameters. We have developed a multi-scale modeling approach that effectively simulates 3D discrete-heat-source, mountain-scale thermohydrologic behavior at Yucca Mountain and captures the natural variability of the site consistent with what we know from site characterization and waste-package-to-waste-package variability in heat output. We describe this approach and present results examining the role of infiltration flux, the most important natural-system parameter with respect to how thermohydrologic behavior influences the performance of the repository

Hydration products and mechanical properties of hydroceramics solidified waste for simulated Non-alpha low and intermediate level radioactive wastes

International Nuclear Information System (INIS)

Wang Jin; Hong Ming; Wang Junxia; Li Yuxiang; Teng Yuancheng; Wu Xiuling

2011-01-01

In this paper, simulated non-alpha low and intermediate level radioactive wastes was handled as curing object and that of 'alkali-slag-coal fly ash-metakaolin' hydroceramics waste forms were prepared by hydrothermal synthesis method. The hydration products were analyzed by X ray diffraction. The composition of hydrates and the compressive strength of waste forms were determined and measured. The results indicate that the main crystalline phase of hydration products were analcite when the temperature was 150 to 180 degree C and the salt content ratio was 0.10 to 0.30. Analcite diffraction peaks in hydration products is increasing when the temperature was raised and the reaction time prolonged. Strength test results show that the solidified waste forms have superior compressive strength. The compressive strength gradually decreased with the increase in salt content ratio in waste forms. (authors)

A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests - 13342

Energy Technology Data Exchange (ETDEWEB)

Thien, Mike G. [Washington River Protection Solutions, LLC, P.O Box 850, Richland WA, 99352 (United States); Barnes, Steve M. [Waste Treatment Plant, 2435 Stevens Center Place, Richland WA 99354 (United States)

2013-07-01

The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described. (authors)

A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests - 13342

International Nuclear Information System (INIS)

Thien, Mike G.; Barnes, Steve M.

2013-01-01

The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described. (authors)

Performance of aged cement - polymer composite immobilizing borate waste simulates during flooding scenarios

International Nuclear Information System (INIS)

Eskander, S.B.; Bayoumi, T.A.; Saleh, H.M.

2012-01-01

An advanced composite of cement and water extended polyester based on the recycled Poly(ethylene terephthalate) waste was developed to incorporate the borate waste. Previous studies have reported the characterizations of the waste composite (cement-polymer composite immobilizing borate waste simulates) after 28 days of curing time. The current work studied the performance of waste composite aged for seven years and subjected to flooding scenario during 260 days using three types of water. The state of waste composite was assessed at the end of each definite interval of the water infiltration through visual examination and mechanical measurement. Scanning electron microscopy, infrared spectroscopy, X-ray diffraction and thermal analyses were used to investigate the changes that may occur in the microstructure of the waste composite under aging and flooding effects. The actual experimental results indicated reasonable evidence for the waste composite. Acceptable consistency was confirmed for the waste composite even after aging seven years and exposure to flooding scenario for 260 days.

Leaching behavior of simulated high-level waste glass

International Nuclear Information System (INIS)

Kamizono, Hiroshi

1987-03-01

The author's work in the study on the leaching behavior of simulated high-level waste (HLW) glass were summarized. The subjects described are (1) leach rates at high temperatures, (2) effects of cracks on leach rates, (3) effects of flow rate on leach rates, and (4) an in-situ burial test in natural groundwater. In the following section, the leach rates obtained by various experiments were summarized and discussed. (author)

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) verification and validation plan. version 1.

Energy Technology Data Exchange (ETDEWEB)

Bartlett, Roscoe Ainsworth; Arguello, Jose Guadalupe, Jr.; Urbina, Angel; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Knupp, Patrick Michael; Wang, Yifeng; Schultz, Peter Andrew; Howard, Robert (Oak Ridge National Laboratory, Oak Ridge, TN); McCornack, Marjorie Turner

2011-01-01

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. To meet this objective, NEAMS Waste IPSC M&S capabilities will be applied to challenging spatial domains, temporal domains, multiphysics couplings, and multiscale couplings. A strategic verification and validation (V&V) goal is to establish evidence-based metrics for the level of confidence in M&S codes and capabilities. Because it is economically impractical to apply the maximum V&V rigor to each and every M&S capability, M&S capabilities will be ranked for their impact on the performance assessments of various components of the repository systems. Those M&S capabilities with greater impact will require a greater level of confidence and a correspondingly greater investment in V&V. This report includes five major components: (1) a background summary of the NEAMS Waste IPSC to emphasize M&S challenges; (2) the conceptual foundation for verification, validation, and confidence assessment of NEAMS Waste IPSC M&S capabilities; (3) specifications for the planned verification, validation, and confidence-assessment practices; (4) specifications for the planned evidence information management system; and (5) a path forward for the incremental implementation of this V&V plan.

Volumetric change of simulated radioactive waste glass irradiated by electron accelerator. [Silica glass

Energy Technology Data Exchange (ETDEWEB)

Sato, Seichi; Furuya, Hirotaka; Inagaki, Yaohiro; Kozaka, Tetsuo; Sugisaki, Masayasu

1987-11-01

Density changes of simulated radioactive waste glasses, silica glass and Pyrex glass irradiated by an electron accelerator were measured by a ''sink-float'' technique. The density changes of the waste and silica glasses were less than 0.05 %, irradiated at 2.0 MeV up to the fluence of 1.7 x 10/sup 17/ ecm/sup 2/, while were remarkably smaller than that of Pyrex glass of 0.18 % shrinkage. Precision of the measurements in the density changes of the waste glass was lower than that of Pyrex glass possibly because of the inhomogeneity of the waste glass

Regulatory off-gas analysis from the evaporation of Hanford simulated waste spiked with organic compounds.

Science.gov (United States)

Saito, Hiroshi H; Calloway, T Bond; Ferrara, Daro M; Choi, Alexander S; White, Thomas L; Gibson, Luther V; Burdette, Mark A

2004-10-01

After strontium/transuranics removal by precipitation followed by cesium/technetium removal by ion exchange, the remaining low-activity waste in the Hanford River Protection Project Waste Treatment Plant is to be concentrated by evaporation before being mixed with glass formers and vitrified. To provide a technical basis to permit the waste treatment facility, a relatively organic-rich Hanford Tank 241-AN-107 waste simulant was spiked with 14 target volatile, semi-volatile, and pesticide compounds and evaporated under vacuum in a bench-scale natural circulation evaporator fitted with an industrial stack off-gas sampler at the Savannah River National Laboratory. An evaporator material balance for the target organics was calculated by combining liquid stream mass and analytical data with off-gas emissions estimates obtained using U.S. Environmental Protection Agency (EPA) SW-846 Methods. Volatile and light semi-volatile organic compounds (1 mm Hg vapor pressure) in the waste simulant were found to largely exit through the condenser vent, while heavier semi-volatiles and pesticides generally remain in the evaporator concentrate. An OLI Environmental Simulation Program (licensed by OLI Systems, Inc.) evaporator model successfully predicted operating conditions and the experimental distribution of the fed target organics exiting in the concentrate, condensate, and off-gas streams, with the exception of a few semi-volatile and pesticide compounds. Comparison with Henry's Law predictions suggests the OLI Environmental Simulation Program model is constrained by available literature data.

Effects of tuff waste package components on release from 76-68 simulated waste glass: Final report

International Nuclear Information System (INIS)

McVay, G.L.; Robinson, G.R.

1984-04-01

An experimental matrix has been conducted that will allow evaluation of the effects of waste package constituents on the waste form release behavior in a tuff repository environment. Tuff rock and groundwater were used along with 304L, 316, and 1020M ferrous metals to evaluate release from uranium-doped MCC 76-68 simulated waste glass. One of the major findings was that in the absence of 1020M mild steel, tuff rock powder dominates the system. However, when 1020M mild steel is present, it appears to dominate the system. The rock-dominated system results in suppressed glass-water reaction and leaching while the 1020M-dominated system results in enhanced leaching - but the metal effectively scavenges uranium from solution. The 300-series stainless steels play no significant role in affecting glass leaching characteristics. 6 refs., 28 figs., 5 tabs

Recycle Waste Collection Tank (RWCT) simulant testing in the PVTD feed preparation system

International Nuclear Information System (INIS)

Abrigo, G.P.; Daume, J.T.; Halstead, S.D.; Myers, R.L.; Beckette, M.R.; Freeman, C.J.; Hatchell, B.K.

1996-03-01

(This is part of the radwaste vitrification program at Hanford.) RWCT was to routinely receive final canister decontamination sand blast frit and rinse water, Decontamination Waste Treatment Tank bottoms, and melter off-gas Submerged Bed Scrubber filter cake. In order to address the design needs of the RWCT system to meet performance levels, the PNL Vitrification Technology (PVTD) program used the Feed Preparation Test System (FPTS) to evaluate its equipment and performance for a simulant of RWCT slurry. (FPTS is an adaptation of the Defense Waste Processing Facility feed preparation system and represents the initially proposed Hanford Waste Vitrification Plant feed preparation system designed by Fluor-Daniel, Inc.) The following were determined: mixing performance, pump priming, pump performance, simulant flow characterization, evaporator and condenser performance, and ammonia dispersion. The RWCT test had two runs, one with and one without tank baffles

Simplified analytical model to simulate radionuclide release from radioactive waste trenches

International Nuclear Information System (INIS)

Sa, Bernardete Lemes Vieira de

2001-01-01

In order to evaluate postclosure off-site doses from low-level radioactive waste disposal facilities, a computer code was developed to simulate the radionuclide released from waste form, transport through vadose zone and transport in the saturated zone. This paper describes the methodology used to model these process. The radionuclide released from the waste is calculated using a model based on first order kinetics and the transport through porous media was determined using semi-analytical solution of the mass transport equation, considering the limiting case of unidirectional convective transport with three-dimensional dispersion in an isotropic medium. The results obtained in this work were compared with other codes, showing good agreement. (author)

Simulating sanitation and waste flows and their environmental impacts in East African urban centres

NARCIS (Netherlands)

Oyoo, R.

2014-01-01

Simulating Sanitation and Waste Flows and their Environmental Impacts in East African Urban Centres

Abstract

If improperly managed, urban waste flows can pose a significant threat to the quality of both the natural environment and public health.

FRACTIONAL CRYSTALLIZATION LABORATORY TESTS WITH SIMULATED TANK WASTE

International Nuclear Information System (INIS)

HERTING DL

2007-01-01

Results are presented for several simulated waste tests related to development of the fractional crystallization process. Product salt dissolution rates were measured to support pilot plant equipment design. Evaporation tests were performed to evaluate the effects of organics on slurry behavior and to determine optimum antifoam addition levels. A loss-of-power test was performed to support pilot plant accident scenario analysis. Envelope limit tests were done to address variations in feed composition

Alternative Chemical Cleaning Methods for High Level Waste Tanks: Simulant Studies

Energy Technology Data Exchange (ETDEWEB)

Rudisill, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); King, W. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hay, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jones, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2015-11-19

Solubility testing with simulated High Level Waste tank heel solids has been conducted in order to evaluate two alternative chemical cleaning technologies for the dissolution of sludge residuals remaining in the tanks after the exhaustion of mechanical cleaning and sludge washing efforts. Tests were conducted with non-radioactive pure phase metal reagents, binary mixtures of reagents, and a Savannah River Site PUREX heel simulant to determine the effectiveness of an optimized, dilute oxalic/nitric acid cleaning reagent and pure, dilute nitric acid toward dissolving the bulk non-radioactive waste components. A focus of this testing was on minimization of oxalic acid additions during tank cleaning. For comparison purposes, separate samples were also contacted with pure, concentrated oxalic acid which is the current baseline chemical cleaning reagent. In a separate study, solubility tests were conducted with radioactive tank heel simulants using acidic and caustic permanganate-based methods focused on the “targeted” dissolution of actinide species known to be drivers for Savannah River Site tank closure Performance Assessments. Permanganate-based cleaning methods were evaluated prior to and after oxalic acid contact.

Rheological evaluation of simulated neutralized current acid waste - transuranics

International Nuclear Information System (INIS)

Fow, C.L.; McCarthy, D.; Thornton, G.T.; Scott, P.A.; Bray, L.A.

1986-09-01

At the Hanford Plutonium and Uranium Extraction Plant (PUREX), in Richland, Washington, plutonium and uranium products are recovered from irradiated fuel by a solvent extraction process. A byproduct of this process is an aqueous waste stream that contains fission products. This waste stream, called current acid waste (CAW), is chemically neutralized and stored in double shell tanks (DSTs) on the Hanford Site. This neutralized current acid waste (NCAW) will be transported by pipe to B-Plant, a processing plant located nearby. In B-Plant, the transuranic (TRU) elements in NCAW are separated from the non-TRU elements. The majority of the TRU elements in NCAW are in the solids. Therefore, the primary processing operation is to separate the NCAW solids (NCAW-TRU) from the NCAW liquid. These two waste streams will be pumped to suitable holding tanks before being further processed for permanent disposal. To ensure that the retrieval and transportation of NCAW and NCAW-TRU are successful, researchers at Pacific Northwest Laboratory (PNL) evaluated the rheological and transport properties of the slurries. This evaluation had two phases. First, researchers conducted laboratory rheological evaluations of simulated NCAW and NCAW-TRU. The results of these evaluations were then correlated with classical rheological models and scaled up to predict the performance that is likely to occur in the full-scale system. This scale-up procedure has already been successfully used to predict the critical transport properties of a slurry (Neutralized Cladding Removal Waste) with rheological properties similar to those displayed by NCAW and NCAW-TRU

Corrosion of inconel in high-temperature borosilicate glass melts containing simulant nuclear waste

Science.gov (United States)

Mao, Xianhe; Yuan, Xiaoning; Brigden, Clive T.; Tao, Jun; Hyatt, Neil C.; Miekina, Michal

2017-10-01

The corrosion behaviors of Inconel 601 in the borosilicate glass (MW glass) containing 25 wt.% of simulant Magnox waste, and in ZnO, Mn2O3 and Fe2O3 modified Mg/Ca borosilicate glasses (MZMF and CZMF glasses) containing 15 wt.% of simulant POCO waste, were evaluated by dimensional changes, the formation of internal defects and changes in alloy composition near corrosion surfaces. In all three kinds of glass melts, Cr at the inconel surface forms a protective Cr2O3 scale between the metal surface and the glass, and alumina precipitates penetrate from the metal surface or formed in-situ. The corrosion depths of inconel 601 in MW waste glass melt are greater than those in the other two glass melts. In MW glass, the Cr2O3 layer between inconel and glass is fragmented because of the reaction between MgO and Cr2O3, which forms the crystal phase MgCr2O4. In MZMF and CZMF waste glasses the layers are continuous and a thin (Zn, Fe, Ni, B)-containing layer forms on the surface of the chromium oxide layer and prevents Cr2O3 from reacting with MgO or other constituents. MgCr2O4 was observed in the XRD analysis of the bulk MW waste glass after the corrosion test, and ZrSiO4 in the MZMF waste glass, and ZrSiO4 and CaMoO4 in the CZMF waste glass.

Vitrification testing of simulated high-level radioactive waste at Hanford

International Nuclear Information System (INIS)

Perez, J.M. Jr.; Nakaoka, R.R.

1986-03-01

The Hanford Waste Vitrification Plant may apply vitrification technology, being developed at Pacific Northwest Laboratory, to solidify selected Hanford waste streams prior to disposal in a federal repository. Based on the first stage of flowsheet development and laboratory testing, a reference working glass and two candidate simulated feed slurries were recommended for vitrification testing. Over 500 hours of melter testing were performed in 1985 during prototype vitrification experiments. Testing demonstrated that the slurry compositions had acceptable processing characteristics in a ceramic melter. A pre-made glass-former frit was determined to be preferred as the method of glass-former addition. Due to a high chromium content in the waste, spinal crystal formation and settling occurred in the glass tank. The nature and extent of off-gas effluents were consistent with past experiments processing slurries containing formic acid

Bitumen coating as a tool for improving the porosity and chemical stability of simulated cement-waste forms

International Nuclear Information System (INIS)

Saleh, H.M.

2010-01-01

Coating process of simulated cement-based waste form with bitumen was evaluated by performing physical and chemical experimental tests. X-ray diffraction (X-RD), Fourier transform infrared spectroscopy (FT-IR) and electron microscope investigations were applied on coated and non-coated simulated waste forms. Experimental results indicated that coating process improved the applicable properties of cement-based waste form such as porosity and leachability. Diffusion coefficients and leach indecies of coated specimens were calculated and show acceptable records. It could be stated that coating cemented waste form by bitumen emulsion, isolate the radioactive contaminants, thus reduces their back release to surrounding and in consequently save the environment proper and safe

A dynamic simulation model of the Savannah River Site high level waste complex

International Nuclear Information System (INIS)

Gregory, M.V.; Aull, J.E.; Dimenna, R.A.

1994-01-01

A detailed, dynamic simulation entire high level radioactive waste complex at the Savannah River Site has been developed using SPEEDUP(tm) software. The model represents mass transfer, evaporation, precipitation, sludge washing, effluent treatment, and vitrification unit operation processes through the solution of 7800 coupled differential and algebraic equations. Twenty-seven discrete chemical constituents are tracked through the unit operations. The simultaneous simultaneous simulation of concurrent batch and continuous processes is achieved by several novel, customized SPEEDUP(tm) algorithms. Due to the model's computational burden, a high-end work station is required: simulation of a years operation of the complex requires approximately three CPU hours on an IBM RS/6000 Model 590 processor. The model will be used to develop optimal high level waste (HLW) processing strategies over a thirty year time horizon. It will be employed to better understand the dynamic inter-relationships between different HLW unit operations, and to suggest strategies that will maximize available working tank space during the early years of operation and minimize overall waste processing cost over the long-term history of the complex. Model validation runs are currently underway with comparisons against actual plant operating data providing an excellent match

Fundamental Science-Based Simulation of Nuclear Waste Forms

Energy Technology Data Exchange (ETDEWEB)

Devanathan, Ramaswami; Gao, Fei; Sun, Xin; Khaleel, Mohammad A.

2010-10-04

This report presents a hierarchical multiscale modeling scheme based on two-way information exchange. To account for all essential phenomena in waste forms over geological time scales, the models have to span length scales from nanometer to kilometer and time scales from picoseconds to millenia. A single model cannot cover this wide range and a multi-scale approach that integrates a number of different at-scale models is called for. The approach outlined here involves integration of quantum mechanical calculations, classical molecular dynamics simulations, kinetic Monte Carlo and phase field methods at the mesoscale, and continuum models. The ultimate aim is to provide science-based input in the form of constitutive equations to integrated codes. The atomistic component of this scheme is demonstrated in the promising waste form xenotime. Density functional theory calculations have yielded valuable information about defect formation energies. This data can be used to develop interatomic potentials for molecular dynamics simulations of radiation damage. Potentials developed in the present work show a good match for the equilibrium lattice constants, elastic constants and thermal expansion of xenotime. In novel waste forms, such as xenotime, a considerable amount of data needed to validate the models is not available. Integration of multiscale modeling with experimental work is essential to generate missing data needed to validate the modeling scheme and the individual models. Density functional theory can also be used to fill knowledge gaps. Key challenges lie in the areas of uncertainty quantification, verification and validation, which must be performed at each level of the multiscale model and across scales. The approach used to exchange information between different levels must also be rigorously validated. The outlook for multiscale modeling of wasteforms is quite promising.

A users guide for the radioactive waste management code 'SIMULATION 2'

International Nuclear Information System (INIS)

Moore, D.; Tymons, B.J.

1984-09-01

This report is a users' guide to the radioactive waste management program SIMULATION. It gives a complete description of the calculational method used (with worked examples) a specification of the input data requirements, and samples of printout from the program. (author)

Pyrolysis of Municipal Green Waste: A Modelling, Simulation and Experimental Analysis

Directory of Open Access Journals (Sweden)

Mohammed J. Kabir

2015-07-01

Full Text Available Pyrolysis is the thermo-chemical conversion of carbonaceous feedstock in the absence of oxygen to produce bio-fuel (bio-oil, bio-char and syn-gas. Bio-fuel production from municipal green waste (MGW through the pyrolysis process has attracted considerable attention recently in the renewable energy sector because it can reduce greenhouse gas emissions and contribute to energy security. This study analyses properties of MGW feedstock available in Rockhampton city of Central Queensland, Australia, and presents an experimental investigation of producing bio-fuel from that MGW through the pyrolysis process using a short sealed rotary furnace. It was found from the experiment that about 19.97% bio-oil, 40.83% bio-char and 29.77% syn-gas can be produced from the MGW. Then, a four-stage steady state simulation model is developed for pyrolysis process performance simulation using Aspen Plus software. In the first stage, the moisture content of the MGW feed is reduced. In the second stage, the MGW is decomposed according to its elemental constituents. In the third stage, condensate material is separated and, finally, the pyrolysis reactions are modelled using the Gibbs free energy minimisation approach. The MGW’s ultimate and proximate analysis data were used in the Aspen Plus simulation as input parameters. The model is validated with experimentally measured data. A good agreement between simulation and experimental results was found. More specifically, the variation of modelling and experimental elemental compositions of the MGW was found to be 7.3% for carbon, 15.82% for hydrogen, 7.04% for nitrogen and 5.56% for sulphur. The validated model is used to optimise the biofuel production from the MGW as a function of operating variables such as temperature, moisture content, particle size and process heat air–fuel ratio. The modelling and optimisation results are presented, analysed and discussed.

Spatial variability in alluvium properties at a low-level nuclear waste site

International Nuclear Information System (INIS)

Istok, J.D.; Blout, D.O.; Barker, L.; Johnejack, K.R.; Hammermeister, D.P.

1994-01-01

Geological and statistical models for the spatial variability of soil properties are needed to predict field-scale water flow and solute transport but only limited information is currently available on unsaturated soils below the root zone. Spatial variability of selected physical and hydrologic properties was quantified for fine- and coarse-grained alluvial deposits at a low-level nuclear waste disposal site on the Nevada Test Site. Gravimetric water content (w), bulk density (ρ b ), saturated hydraulic conductivity (K a ), and particle-size distribution were determined for vertical and horizontal core specimens and bulk samples collected from 183-m-long horizontal transects in two existing waste disposal trenches located on a single alluvial fan. The transects were approximately aligned parallel and perpendicular to the principal direction of sediment transport. Properties were modeled as either normally or lognormally distributed random variables. Sample coefficients of variation were smallest for ρ b and largest for log(K a ); a weak correlation was identified between log(K a ) and the grain-size parameter d 10 . Particle-size distributions for the fine- and coarse-grained materials were different and significant differences in the natural logarithm of saturated hydraulic conductivity, log(K a ), existed between coarse and fine layers in an excavation aligned with the principal direction of alluvium deposition but not in a perpendicular direction. 37 refs., 7 figs., 11 tabs

Read margin analysis of crossbar arrays using the cell-variability-aware simulation method

Science.gov (United States)

Sun, Wookyung; Choi, Sujin; Shin, Hyungsoon

2018-02-01

This paper proposes a new concept of read margin analysis of crossbar arrays using cell-variability-aware simulation. The size of the crossbar array should be considered to predict the read margin characteristic of the crossbar array because the read margin depends on the number of word lines and bit lines. However, an excessively high-CPU time is required to simulate large arrays using a commercial circuit simulator. A variability-aware MATLAB simulator that considers independent variability sources is developed to analyze the characteristics of the read margin according to the array size. The developed MATLAB simulator provides an effective method for reducing the simulation time while maintaining the accuracy of the read margin estimation in the crossbar array. The simulation is also highly efficient in analyzing the characteristic of the crossbar memory array considering the statistical variations in the cell characteristics.

Dynamic simulation of variable capacity refrigeration systems under abnormal conditions

International Nuclear Information System (INIS)

Liang Nan; Shao Shuangquan; Tian Changqing; Yan Yuying

2010-01-01

There are often abnormal working conditions at evaporator outlet of a refrigeration system, such as two-phase state in transient process, and it is essential to investigate such transient behaviours for system design and control strategy. In this paper, a dynamic lumped parameter model is developed to simulate the transient behaviours of refrigeration system with variable capacity in both normal and abnormal working conditions. The appropriate discriminant method is adopted to switch the normal and abnormal conditions smoothly and to eliminate the simulated data oscillation. In order to verify the dynamic model, we built a test system with variable frequency compressor, water-cooling condenser, evaporator and electronic expansion valve. Calculated values from the mathematical model show reasonable agreement with the experimental data. The simulation results show that the transient behaviours of the variable capacity refrigeration system in the abnormal working conditions can be calculated reliably with the dynamic model when the compressor rotary speed or the opening of electronic expansion valve changes abruptly.

Coupled Multi-physical Simulations for the Assessment of Nuclear Waste Repository Concepts: Modeling, Software Development and Simulation

Science.gov (United States)

Massmann, J.; Nagel, T.; Bilke, L.; Böttcher, N.; Heusermann, S.; Fischer, T.; Kumar, V.; Schäfers, A.; Shao, H.; Vogel, P.; Wang, W.; Watanabe, N.; Ziefle, G.; Kolditz, O.

2016-12-01

As part of the German site selection process for a high-level nuclear waste repository, different repository concepts in the geological candidate formations rock salt, clay stone and crystalline rock are being discussed. An open assessment of these concepts using numerical simulations requires physical models capturing the individual particularities of each rock type and associated geotechnical barrier concept to a comparable level of sophistication. In a joint work group of the Helmholtz Centre for Environmental Research (UFZ) and the German Federal Institute for Geosciences and Natural Resources (BGR), scientists of the UFZ are developing and implementing multiphysical process models while BGR scientists apply them to large scale analyses. The advances in simulation methods for waste repositories are incorporated into the open-source code OpenGeoSys. Here, recent application-driven progress in this context is highlighted. A robust implementation of visco-plasticity with temperature-dependent properties into a framework for the thermo-mechanical analysis of rock salt will be shown. The model enables the simulation of heat transport along with its consequences on the elastic response as well as on primary and secondary creep or the occurrence of dilatancy in the repository near field. Transverse isotropy, non-isothermal hydraulic processes and their coupling to mechanical stresses are taken into account for the analysis of repositories in clay stone. These processes are also considered in the near field analyses of engineered barrier systems, including the swelling/shrinkage of the bentonite material. The temperature-dependent saturation evolution around the heat-emitting waste container is described by different multiphase flow formulations. For all mentioned applications, we illustrate the workflow from model development and implementation, over verification and validation, to repository-scale application simulations using methods of high performance computing.

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

Treatment of an Anonymous Recipient: Solid-Waste Management Simulation Game

Science.gov (United States)

Wu, Ko-Chiu; Huang, Po-Yuan

2015-01-01

This study developed a game simulation based on problem solving in the management of urban waste. We then investigated the factors affecting the decisions made by players. During gameplay, the players sought to guide the development of a city via management strategies involving a balance of economic growth and environmental protection. Nature…

Laboratory development of wind turbine simulator using variable ...

African Journals Online (AJOL)

user

1*Department of Electronics Engineering, Prof. ... In this paper variable speed induction motor drive using scalar control is interfaced in wind energy conversion ... the wind turbine simulator is used as a necessary tool in research laboratories.

Chemical stability of seven years aged cement-PET composite waste form containing radioactive borate waste simulates

Energy Technology Data Exchange (ETDEWEB)

Saleh, H.M., E-mail: hosamsaleh70@yahoo.com [Radioisotope Department, Atomic Energy Authority, Dokki (Egypt); Tawfik, M.E. [Department of Polymers and Pigments, National Research Center, Dokki (Egypt); Bayoumi, T.A. [Radioisotope Department, Atomic Energy Authority, Dokki (Egypt)

2011-04-15

Different samples of radioactive borate waste simulate [originating from pressurized water reactors (PWR)] have been prepared and solidified after mixing with cement-water extended polyester composite (CPC). The polymer-cement composite samples were prepared from recycled poly (ethylene terephthalate) (PET) waste and cement paste (water/cement ratio of 40%). The prepared samples were left to set at room temperature (25 deg. C {+-} 5) under humid conditions. After 28 days curing time the obtained specimens were kept in their molds to age for 7 years under ambient conditions. Cement-polymer composite waste form specimens (CPCW) have been subjected to leach tests for both {sup 137}Cs and {sup 60}Co radionuclides according to the method proposed by the International Atomic Energy Agency (IAEA). Leaching tests were justified under various factors that may exist within the disposal site (e.g. type of leachant, surrounding temperature, leachant behavior, the leachant volume to CPCW surface area...). The obtained data after 260 days of leaching revealed that after 7 years of aging the candidate cement-polymer composite (CPC) containing radioactive borate waste samples are characterized by adequate chemical stability required for the long-term disposal process.

Durability of cemented waste in repository and under simulated conditions

International Nuclear Information System (INIS)

Dragolici, F.; Nicu, M.; Lungu, L.; Turcanu, C.; Rotarescu, Gh.

2000-01-01

The Romanian Radioactive Waste National Repository for low level and intermediate level radioactive waste was built in Baita - Bihor county, in an extinct uranium exploitation. The site is at 840 m above sea level and the host rock is crystalline with a low porosity, a good chemical homogeneity and impermeability, keeping these qualities over a considerable horizontal and vertical spans. To obtain the experimental data necessary for the waste form and package characterization together with the back-filling material behaviour in the repository environment, a medium term research programme (1996 - 2010) was implemented. The purpose of this experimental programme is to obtain a part of the data base necessary for the approach of medium and long term assessment of the safety and performance of Baita - Bihor Repository. The programme will provide: a deeper knowledge of the chemical species and reaction mechanisms, the structure, properties and performances of the final products. For safety reasons the behaviour of waste package, which is a main barrier, must be properly known in terms of long term durability in real repository conditions. Characterization of the behaviour includes many interactions between the waste package itself and the surrounding near field conditions such as mineralogy, hydrogeology and groundwater chemistry. To obtain a more deeper knowledge of the species and physical-chemical reactions participating in the matrix formation, as well as their future behaviour during the disposal period, a thorough XRD study started in 1998. For Romanian Radioactive Waste National Repository (DNDR) Baita - Bihor the following steps are planned for the conditioned waste matrix characterization in simulated and real conditions: - preparation and characterization of normal reference matrices based on different cement formulations; - preparation of reference simulated sludge cemented matrices containing iron hydroxide and iron phosphate; - selection of real and

Impact of operating variables on re-refining of vehicle waste oil to base-oil by acid-clay process

International Nuclear Information System (INIS)

Durrani, H.A.; Panhwar, M.I.; Kazi, R.A.

2009-01-01

A large volume of waste oil is generated each year as the number of vehicles in the country is increasing every day. This used lubricant oil constitutes a serious pollution problem. It easily be converted as a resource depending on the manners of utilization and management. This paper compares various end uses of waste oil and develops technology by which basic properties of the base oil can be regained. An experimental setup (Experimental rig) of acid-clay method was used to regain waste oil collected and number of experiments were conducted by varying different process variables. The impacts of the operating variables on the quality of re-refined oil are also discussed. The recovery of re-refined base oil was found in between 44-49%. (author)

Stream-simulation experiments for waste-repository investigations

International Nuclear Information System (INIS)

Seitz, M.G.

1980-01-01

The potential for radionuclide migration by groundwater flow from a breached-water repository depends on the leaching process and on chemical changes that might occur as the radionuclide moves away from the repository. Therefore, migration involves the interactions of leached species with (1) the waste and canister, (2) the engineered barrier, and (3) the geologic materials surrounding the repository. Rather than attempt to synthesize each species and study it individually, another approach is to integrate all species and interactions using stream-simulation experiments. Interactions identified in these studies can then be investigated in detail in simpler experiments

Bituminization of simulated waste, spent resins, evaporator concentrates and animal ashes by extrusion process

International Nuclear Information System (INIS)

Grosche Filho, C.E.; Chandra, U.

1986-01-01

The results of the study of simulated radwaste, spent ion-exchange resins, borates/evaporator-concentrates and animal ashes, in bituminized form, are presented and discussed. Distilled and oxidized bitumen were used for characterizing the crude material and simulated wastes-bitumen mixtures of varying weight composition 30, 40, 50, 60% by weight the dry waste material. The asphaltine and parafin contents in the bitumens were determined. Some additives and clays were used aiming best characteristics of solidified wastes. For leaching studies, granular ion-exchange resins were loaded with Cs 134 and mixtures of resins-bitumens were prepared. The leaching studies were executed using the IAEA recommendation and the ISO method. It was used a conventional screw-extruder, used in plastic industry, to determine operational conditions and process difficulties. Mixtures resins-bitumen and concentrate-bitumen in differents operational condition were prepared and analysed. (Author) [pt

Converting Simulated Sodium-bearing Waste into a Single Solid Waste Form by Evaporation: Laboratory- and Pilot-Scale Test Results on Recycling Evaporator Overheads

Energy Technology Data Exchange (ETDEWEB)

Griffith, D.; D. L. Griffith; R. J. Kirkham; L. G. Olson; S. J. Losinski

2004-01-01

Conversion of Idaho National Engineering and Environmental Laboratory radioactive sodium-bearing waste into a single solid waste form by evaporation was demonstrated in both flask-scale and pilot-scale agitated thin film evaporator tests. A sodium-bearing waste simulant was adjusted to represent an evaporator feed in which the acid from the distillate is concentrated, neutralized, and recycled back through the evaporator. The advantage to this flowsheet is that a single remote-handled transuranic waste form is produced in the evaporator bottoms without the generation of any low-level mixed secondary waste. However, use of a recycle flowsheet in sodium-bearing waste evaporation results in a 50% increase in remote-handled transuranic volume in comparison to a non-recycle flowsheet.

Simulation of radioactive waste transmutation on the t.node parallel computer

International Nuclear Information System (INIS)

Bacha, F.; Maillard, J.; Silva, J.

1995-01-01

Before any experiment on reactor driven by an accelerator, computer simulation supplies tools for optimization. Some of the key parameters are neutron production on a heavy target and neutronic distribution flux in the core. During two code benchmarks organized by the NEA-OECD, simulations of energetic incident proton collisions on a thin lead target for the first one, on a thick lead target for the second one, are described. One validation of the numeric codes is based on these results. A preliminary design of a burning waste system using benchmark result analysis and fission focused simulations is proposed

Simulation of radioactive waste transmutation on the T. Node parallel computer

International Nuclear Information System (INIS)

Bacha, F.; Maillard, J.; Silva, J.

1995-01-01

Before any experiment on reactor driven by an accelerator, computer simulation supplies tools for optimization. Some of the key parameters are neutron production on a heavy target and neutronic distribution flux in the core. During two code benchmarks organized by the NEA-OECD, simulations of energetic incident proton collisions on a thin lead target for the first one, on a thick lead target for the second one, are described. One validation of our numeric codes is based on these results. A preliminary design of a burning waste system using benchmark result analysis and fission focused simulations is proposed

Simulation of radioactive waste transmutation on the t.node parallel computer

Energy Technology Data Exchange (ETDEWEB)

Bacha, F.; Maillard, J.; Silva, J. [LPC College de France, Paris (France)

1995-10-01

Before any experiment on reactor driven by an accelerator, computer simulation supplies tools for optimization. Some of the key parameters are neutron production on a heavy target and neutronic distribution flux in the core. During two code benchmarks organized by the NEA-OECD, simulations of energetic incident proton collisions on a thin lead target for the first one, on a thick lead target for the second one, are described. One validation of the numeric codes is based on these results. A preliminary design of a burning waste system using benchmark result analysis and fission focused simulations is proposed.

Effects of feed process variables on Hanford Vitrification Plant performance

International Nuclear Information System (INIS)

Farnsworth, R.K.; Peterson, M.E.; Wagner, R.N.

1987-01-01

As a result of nuclear defense activities, high-level liquid radioactive wastes have been generated at the Hanford Site for over 40 yr. The Hanford Waste Vitrification Plant (HWVP) is being proposed to immobilize these wastes in a waste form suitable for disposal in a geologic repository. Prior to vitrification, the waste will undergo several conditioning steps before being fed to the melter. The effect of certain process variables on the resultant waste slurry properties must be known to assure processability of the waste slurry during feed preparation. Of particular interest are the rheological properties, which include the yield stress and apparent viscosity. Identification of the rheological properties of the slurry is required to adequately design the process equipment used for feed preparation (agitators, mixing tanks, concentrators, etc.). Knowledge of the slurry rheological properties is also necessary to establish processing conditions and operational limits for maximum plant efficiency and reliability. A multivariable study was performed on simulated HWVP feed to identify the feed process variables that have a significant impact on rheology during processing. Two process variables were evaluated in this study: (a) the amount of formic acid added to the feed and (b) the degree of shear encountered by the feed during processing. The feed was physically and rheologically characterized at various stages during feed processing

Grout treatment facility dangerous waste permit application

International Nuclear Information System (INIS)

1992-07-01

The Grout Treatment Facility (GTF) will provide permanent disposal for approximately 43 Mgal of radioactive liquid waste currently being stored in underground tanks on the Hanford Site. The first step in permanent disposal is accomplished by solidifying the low-level liquid waste with cementitious dry materials. The resulting grout is cast within underground vaults. This report on the GTF contains information on the following: Hanford Site Maps, road evaluation for the grout treatment facility, Department of Ecology certificate of non-designation for centralia fly ash, double-shell tank waste compositional modeling, laboratory analysis reports for double-shell tank waste, stored in tanks 241-AN-103, 241-AN-106, and 241-AW-101, grout vault heat transfer results for M-106 grout formulation, test results for extraction procedure toxicity testing, test results for toxicity testing of double-shell tank grout, pilot-scale grout production test with a simulated low-level waste, characterization of simulated low-level waste grout produced in a pilot-scale test, description of the procedure for sampling nonaging waste storage tanks, description of laboratory procedures, grout campaign waste composition verification, variability in properties of grouted phosphate/sulfate N-reactor waste, engineering drawings, description of operating procedures, equipment list--transportable grout equipment, grout treatment facility--tank integrity assessment plan, long-term effects of waste solutions on concrete and reinforcing steel, vendor information, grout disposal facilities construction quality assurance plan, and flexible membrane liner/waste compatibility test results

Simulated variability of the Atlantic meridional overturning circulation

Science.gov (United States)

Bentsen, M.; Drange, H.; Furevik, T.; Zhou, T.

To examine the multi-annual to decadal scale variability of the Atlantic Meridional Overturning Circulation (AMOC) we conducted a four-member ensemble with a daily reanalysis forced, medium-resolution global version of the isopycnic coordinate ocean model MICOM, and a 300-years integration with the fully coupled Bergen Climate Model (BCM). The simulations of the AMOC with both model systems yield a long-term mean value of 18 Sv and decadal variability with an amplitude of 1-3 Sv. The power spectrum of the inter-annual to decadal scale variability of the AMOC in BCM generally follows the theoretical red noise spectrum, with indications of increased power near the 20-years period. Comparison with observational proxy indices for the AMOC, e.g. the thickness of the Labrador Sea Water, the strength of the baroclinic gyre circulation in the North Atlantic Ocean, and the surface temperature anomalies along the mean path of the Gulf Stream, shows similar trends and phasing of the variability, indicating that the simulated AMOC variability is robust and real. Mixing indices have been constructed for the Labrador, the Irminger and the Greenland-Iceland-Norwegian (GIN) seas. While convective mixing in the Labrador and the GIN seas are in opposite phase, and linked to the NAO as observations suggest, the convective mixing in the Irminger Sea is in phase with or leads the Labrador Sea. Newly formed deep water is seen as a slow, anomalous cold and fresh, plume flowing southward along the western continental slope of the Atlantic Ocean, with a return flow of warm and saline water on the surface. In addition, fast-travelling topographically trapped waves propagate southward along the continental slope towards equator, where they go east and continue along the eastern rim of the Atlantic. For both types of experiments, the Northern Hemisphere sea level pressure and 2 m temperature anomaly patterns computed based on the difference between climate states with strong and weak AMOC

Simulating Radionuclide Migrations of Low-level Wastes in Nearshore Environment

Science.gov (United States)

Lu, C. C.; Li, M. H.; Chen, J. S.; Yeh, G. T.

2016-12-01

Tunnel disposal into nearshore mountains was tentatively selected as one of final disposal sites for low-level wastes in Taiwan. Safety assessment on radionuclide migrations in far-filed may involve geosphere processes under coastal environments and into nearshore ocean. In this study the 3-D HYDROFEOCHE5.6 numerical model was used to perform simulations of groundwater flow and radionuclide transport with decay chains. Domain of interest on the surface includes nearby watersheds delineated by digital elevation models and nearshore seabed. As deep as 800 m below the surface and 400 m below sea bed were considered for simulations. The disposal site was located at 200m below the surface. Release rates of radionuclides from near-field was estimated by analytical solutions of radionuclide diffusion with decay out of engineered barriers. Far-field safety assessments were performed starting from the release of radionuclides out of engineered barriers to a time scale of 10,000 years. Sensitivity analyses of geosphere and transport parameters were performed to improve our understanding of safety on final disposal of low-level waste in nearshore environments.

Evaluation of high-level waste vitrification feed preparation chemistry for an NCAW simulant, FY 1994: Alternate flowsheets (DRAFT)

International Nuclear Information System (INIS)

Smith, H.D.; Merz, M.D.; Wiemers, K.D.; Smith, G.L.

1996-02-01

High-level radioactive waste stored in tanks at the U.S. Department of Energy's (DOE's) Hanford Site will be pretreated to concentrate radioactive constituents and fed to the vitrification plant A flowsheet for feed preparation within the vitrification plant (based on the Hanford Waste Vitrification Plant (HWVP) design) called for HCOOH addition during the feed preparation step to adjust rheology and glass redox conditions. However, the potential for generating H 2 and NH 3 during treatment of high-level waste (HLW) with HCOOH was identified at Pacific Northwest Laboratory (PNL). Studies at the University of Georgia, under contract with Savannah River Technology Center (SRTC) and PNL, have verified the catalytic role of noble metals (Pd, Rh, Ru), present in the waste, in the generation of H 2 and NH 3 . Both laboratory-scale and pilot-scale studies at SRTC have documented the H 2 and NH 3 generation phenomenal Because H 2 and NH 3 may create hazardous conditions in the vessel vapor space and offgas system of a vitrification plant, reducing the H 2 generation rate and the NH 3 generation to the lowest possible levels consistent with desired melter feed characteristics is important. The Fiscal Year 1993 and 1994 studies were conducted with simulated (non-radioactive), pre-treated neutralized current acid waste (NCAW). Neutralized current acid waste is a high-level waste originating from the plutonium/uranium extraction (PUREX) plant that has been partially denitrated with sugar, neutralized with NaOH, and is presently stored in double-shell tanks. The non-radioactive simulant used for the present study includes all of the trace components found in the waste, or substitutes a chemically similar element for radioactive or very toxic species. The composition and simulant preparation steps were chosen to best simulate the chemical processing characteristics of the actual waste

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

Fate of selected microorganisms when introduced as cross-contamination inocula into simulated food trash compartment waste

Science.gov (United States)

Strayer, Richard; Hummerick, Mary; Richards, Jeffrey; Birmele, Michele; Roberts, Michael

AdHocReviewCycleID-309796538 NewReviewCycle EmailSubjectPlease review this (?today?) AuthorEm Richard F. (KSC)[DYNAMAC CORP] ReviewingToolsShownOnceurn:schemas-microsoft-com:office:smart One goal of Exploration Life Support solid waste processing is to stabilize wastes for storage, mitigate crew risks, and enable resource recovery. Food and crew fecal wastes contain easily biodegraded organic components that support microbial growth. Our objective is to determine a baseline for the fate of selected microbes in wastes prior to processing treatments. Challenge microbes, including human-associated pathogens, were added to unsterilized, simulated food trash solid waste containing a mixed microbial community. The fate of the microbial community and challenge microbes was determined over a 6 week time course of waste storage. Challenge microbes were selected from a list of microorganisms common to residual food or fecal wastes and included: Escherichia coli, Salmonella enterica serovar typhimurium, Staphylococcus aureus, Pseudomonas aeruginosa, Aspergillus niger (a common mold), and Bacillus pumilus SAFR-032, a spore-forming bacterium previously isolated from spacecraft assembly facilities selected for its resistance to heat, uv, and desiccation. The trash model simulant contained 80% food trash (food waste and containers) and 20% hygiene wipes. Cultures of challenge microbes were grown overnight on Nutrient Agar (Difco), harvested, re-suspended in physiological saline, and diluted to achieve the desired optical density for inoculation. The six organisms were pooled and inoculated into the simulated food wastes and packaging before manual mixing. Inoculated simulated waste was stored in custom FlexfoilTM gas sampling bags (SKC, Inc.) which were then connected to a gas analysis system designed to supply fresh air to each bag to maintain O2 above 1%. Bag headspace was monitored for CO2 (PP Systems) and O2 (Maxtec). Total microbes were quantified by microscopic direct

Impact of radionuclide spatial variability on groundwater quality downstream from a shallow waste burial in the Chernobyl Exclusion Zone

Science.gov (United States)

Nguyen, H. L.; de Fouquet, C.; Courbet, C.; Simonucci, C. A.

2016-12-01

The effects of spatial variability of hydraulic parameters and initial groundwater plume localization on the possible extent of groundwater pollution plumes have already been broadly studied. However, only a few studies, such as Kjeldsen et al. (1995), take into account the effect of source term spatial variability. We explore this question with the 90Sr migration modeling from a shallow waste burial located in the Chernobyl Exclusion Zone to the underlying sand aquifer. Our work is based upon groundwater sampled once or twice a year since 1995 until 2015 from about 60 piezometers and more than 3,000 137Cs soil activity measurements. These measurements were taken in 1999 from one of the trenches dug after the explosion of the Chernobyl nuclear power plant, the so-called "T22 Trench", where radioactive waste was buried in 1987. The geostatistical analysis of 137Cs activity data in soils from Bugai et al. (2005) is first reconsidered to delimit the trench borders using georadar data as a covariable and to perform geostatistical simulations in order to evaluate the uncertainties of this inventory. 90Sr activity in soils is derived from 137Cs/154Eu and 90Sr/154Eu activity ratios in Chernobyl hot fuel particles (Bugai et al., 2003). Meanwhile, a coupled 1D non saturated/3D saturated transient transport model is constructed under the MELODIE software (IRSN, 2009). The previous 90Sr transport model developed by Bugai et al. (2012) did not take into account the effect of water table fluctuations highlighted by Van Meir et al. (2007) which may cause some discrepancies between model predictions and field observations. They are thus reproduced on a 1D vertical non saturated model. The equiprobable radionuclide localization maps produced by the geostatistical simulations are selected to illustrate different heterogeneities in the radionuclide inventory and are implemented in the 1D model. The obtained activity fluxes from all the 1D vertical models are then injected in a 3D

Immobilization of simulated high-level radioactive waste in borosilicate glass: Pilot scale demonstrations

International Nuclear Information System (INIS)

Ritter, J.A.; Hutson, N.D.; Zamecnik, J.R.; Carter, J.T.

1991-01-01

The Integrated DWPF Melter System (IDMS), operated by the Savannah River Laboratory, is a pilot scale facility used in support of the start-up and operation of the Department of Energy's Defense Waste Processing Facility. The IDMS has successfully demonstrated, on an engineering scale (one-fifth), that simulated high level radioactive waste (HLW) sludge can be chemically treated with formic acid to adjust both its chemical and physical properties, and then blended with simulated precipitate hydrolysis aqueous (PHA) product and borosilicate glass frit to produce a melter feed which can be processed into a durable glass product. The simulated sludge, PHA and frit were blended, based on a product composition program, to optimize the loading of the waste glass as well as to minimize those components which can cause melter processing and/or glass durability problems. During all the IDMS demonstrations completed thus far, the melter feed and the resulting glass that has been produced met all the required specifications, which is very encouraging to future DWPF operations. The IDMS operations also demonstrated that the volatile components of the melter feed (e.g., mercury, nitrogen and carbon, and, to a lesser extent, chlorine, fluorine and sulfur) did not adversely affect the melter performance or the glass product

Changes in soil hydraulic properties caused by construction of a simulated waste trench at the Idaho National Engineering Laboratory, Idaho

International Nuclear Information System (INIS)

Shakofsky, S.

1995-03-01

In order to assess the effect of filled waste disposal trenches on transport-governing soil properties, comparisons were made between profiles of undisturbed soil and disturbed soil in a simulated waste trench. The changes in soil properties induced by the construction of a simulated waste trench were measured near the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory (INEL) in the semiarid southeast region of Idaho. The soil samples were collected, using a hydraulically-driven sampler to minimize sample disruption, from both a simulated waste trench and an undisturbed area nearby. Results show that the undisturbed profile has distinct layers whose properties differ significantly, whereas the soil profile in the simulated waste trench is, by comparison, homogeneous. Porosity was increased in the disturbed cores, and, correspondingly, saturated hydraulic conductivities were on average three times higher. With higher soil-moisture contents (greater than 0.32), unsaturated hydraulic conductivities for the undisturbed cores were typically greater than those for the disturbed cores. With lower moisture contents, most of the disturbed cores had greater hydraulic conductivities. The observed differences in hydraulic conductivities are interpreted and discussed as changes in the soil pore geometry

Mechanisms of gas generation from simulated SY tank farm wastes: FY 1995 progress report

International Nuclear Information System (INIS)

Barefield, E.K.; Boatright, D.; Deshpande, A.; Doctorovich, F.; Liotta, C.L.; Neumann, H.M.; Seymore, S.

1996-07-01

The objective of this work is to develop a better understanding of the mechanism of formation of flammable gases in the thermal decomposition of metal complexants such as HEDTA and sodium glycolate in simulated SY tank farm waste mixtures. This report summarizes the results of work done at the Georgia Institute of Technology in fiscal year 1995. Topics discussed are (1) long-term studies of the decomposition of HEDTA in simulated waste mixtures under an argon atmosphere at 90 and 120 degrees C, including time profiles for disappearance of HEDTA and appearance of products and the quantitative analysis of the kinetic behavior; (2) considerations of hydroxylamine as an intermediate in the production of nitrogen containing gases by HEDTA decomposition; (3) some thoughts on the revision of the global mechanism for thermal decomposition of HEDTA under argon; (4) preliminary long-term studies of the decomposition of HEDTA in simulated waste under an oxygen atmosphere at 120 degrees C; (5) estimation of the amount of NH 3 in the gas phase above HEDTA reaction mixtures; and (6) further, examination of the interaction of aluminum with nitrite ion using 27 Al NMR spectroscopy. Section 2 of this report describes the work conducted over the last three years at GIT. Section 3 contains a discussion of the kinetic behavior of HEDTA under argon; Section 4 discusses the role of hydroxylamine. Thermal decomposition of HEDTA to ED3A is the subject of Section 5, and decomposition of HEDTA in simulated waste mixtures under oxygen is covered in Section 6. In Section 7 we estimate ammonia in the gas phase; the role of aluminum is discussed in Section 8

Corrosion of a carbon steel in simulated liquid nuclear wastes

International Nuclear Information System (INIS)

Saenz Gonzalez, Eduardo

2005-01-01

This work is part of a collaboration agreement between CNEA (National Atomic Energy Commission of Argentina) and USDOE (Department of Energy of the United States of America), entitled 'Tank Corrosion Chemistry Cooperation', to study the corrosion behavior of carbon steel A537 class 1 in different simulated non-radioactive wastes in order to establish the safety concentration limits of the tank waste chemistry at Hanford site (Richland-US). Liquid high level nuclear wastes are stored in tanks made of carbon steel A537 (ASTM nomenclature) that were designed for a service life of 20 to 50 years. A thickness reduction of some tank walls, due to corrosion processes, was detected at Hanford site, beyond the existing predicted values. Two year long-term immersion tests were started using non radioactive simulated liquid nuclear waste solutions at 40 C degrees. This work extends throughout the first year of immersion. The simulated solutions consist basically in combinations of the 10 most corrosion significant chemical components: 5 main components (NaNO 3 , NaCl, NaF, NaNO 2 and NaOH) at three concentration levels and 5 secondary components at two concentration levels. Measurements of the general corrosion rate with time were performed for carbon steel coupons, both immersed in the solutions and in the vapor phases, using weight loss and electrochemistry impedance spectroscopy techniques. Optic and scanning electron microscopy examination, analysis of U-bend samples and corrosion potential measurements, were also done. Localized corrosion susceptibility (pitting and crevice corrosion) was assessed in isolated short-term tests by means of cyclic potentiodynamic polarization curves. The effect of the simulated waste composition on the corrosion behavior of A537 steel was studied based on statistical analyses. The Surface Response Model could be successfully applied to the statistical analysis of the A537 steel corrosion in the studied solutions. General corrosion was not

Groundwater-stream-simulation experiments for the evaluation of the safety of proposed nuclear waste repositories

International Nuclear Information System (INIS)

Seitz, M.G.

1981-01-01

A bench-scale experimental design which integrates repository components to simulate a groundwater stream infiltrating a breached repository is described in this paper. An experiment performed with a nuclear waste solid and one rock core is briefly summarized. The nuclear waste solid consists of borosilicate glass containing formulated nuclear waste and is the source of the leached radionuclides. The rock core used is of granite and serves as the adsorption medium for the leached radionuclides

Geochemical Modeling Of F Area Seepage Basin Composition And Variability

International Nuclear Information System (INIS)

Millings, M.; Denham, M.; Looney, B.

2012-01-01

From the 1950s through 1989, the F Area Seepage Basins at the Savannah River Site (SRS) received low level radioactive wastes resulting from processing nuclear materials. Discharges of process wastes to the F Area Seepage Basins followed by subsequent mixing processes within the basins and eventual infiltration into the subsurface resulted in contamination of the underlying vadose zone and downgradient groundwater. For simulating contaminant behavior and subsurface transport, a quantitative understanding of the interrelated discharge-mixing-infiltration system along with the resulting chemistry of fluids entering the subsurface is needed. An example of this need emerged as the F Area Seepage Basins was selected as a key case study demonstration site for the Advanced Simulation Capability for Environmental Management (ASCEM) Program. This modeling evaluation explored the importance of the wide variability in bulk wastewater chemistry as it propagated through the basins. The results are intended to generally improve and refine the conceptualization of infiltration of chemical wastes from seepage basins receiving variable waste streams and to specifically support the ASCEM case study model for the F Area Seepage Basins. Specific goals of this work included: (1) develop a technically-based 'charge-balanced' nominal source term chemistry for water infiltrating into the subsurface during basin operations, (2) estimate the nature of short term and long term variability in infiltrating water to support scenario development for uncertainty quantification (i.e., UQ analysis), (3) identify key geochemical factors that control overall basin water chemistry and the projected variability/stability, and (4) link wastewater chemistry to the subsurface based on monitoring well data. Results from this study provide data and understanding that can be used in further modeling efforts of the F Area groundwater plume. As identified in this study, key geochemical factors affecting basin

Evaluation of FTIR-based analytical methods for the analysis of simulated wastes

International Nuclear Information System (INIS)

Rebagay, T.V.; Cash, R.J.; Dodd, D.A.; Lockrem, L.L.; Meacham, J.E.; Winkelman, W.D.

1994-01-01

Three FTIR-based analytical methods that have potential to characterize simulated waste tank materials have been evaluated. These include: (1) fiber optics, (2) modular transfer optic using light guides equipped with non-contact sampling peripherals, and (3) photoacoustic spectroscopy. Pertinent instrumentation and experimental procedures for each method are described. The results show that the near-infrared (NIR) region of the infrared spectrum is the region of choice for the measurement of moisture in waste simulants. Differentiation of the NIR spectrum, as a preprocessing steps, will improve the analytical result. Preliminary data indicate that prominent combination bands of water and the first overtone band of the ferrocyanide stretching vibration may be utilized to measure water and ferrocyanide species simultaneously. Both near-infrared and mid-infrared spectra must be collected, however, to measure ferrocyanide species unambiguously and accurately. For ease of sample handling and the potential for field or waste tank deployment, the FTIR-Fiber Optic method is preferred over the other two methods. Modular transfer optic using light guides and photoacoustic spectroscopy may be used as backup systems and for the validation of the fiber optic data

Challenges of characterization of radioactive waste with High composition variability and their consequences measurement methodology

International Nuclear Information System (INIS)

Lexa, D.

2014-01-01

Radioactive waste characterization is a key step in every nuclear decommissioning project. It normally relies on a combination of facility operational history with results of destructive and non-destructive analysis. A particularly challenging situation arises when historical waste from a nuclear research facility is to be characterized, meaning little or no radiological information is available and the composition of the waste is highly variable. The nuclide vector concept is of limited utility, resulting in increased requirements placed on both the extent and performance of destructive and non-destructive analysis. Specific challenges are illustrated on an example of the decommissioning project underway at the Joint Research Center of the European Commission in Ispra. (author)

Bituminization of simulated waste, spent resins, evaporator concentrates and animal ashes by extrusion process

International Nuclear Information System (INIS)

Grosche Filho, C.E.; Chandra, U.

1987-01-01

The results of the study of bituminization of simulated radwaste - spennt ion-exchange resins, borate evaporator/concentrates and animal ashes, are presented and discussed. Distilled and oxidizer bitumen were used. Characterization of the crude material and simulated wastes-bitumen mixtures of varying weigt composition (30, 40, 50, 60% by weight of dry waste material) was carried out. The asphaltene and parafin contents in the bitumens were also determined. Some additives and were used with an aim to improve the characteristcs of solidified wastes. For leaching studies, granular ion-exchange resins were with Cs - 134 and mixtures of resin-bitumen were prepared. The leaching studies were executed using the IAEA recommendation and the ISO method. A conventional screw-extruder, common in plastic industry, was used determine operational parameters and process difficulties. Mixtures of resin-bitumen and evaporator concentrate-bitumen obtained from differents operational conditions were characterized. (Author) [pt

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.

Energy Technology Data Exchange (ETDEWEB)

Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.; Dewers, Thomas A.; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Wang, Yifeng; Schultz, Peter Andrew

2011-02-01

This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.

Accelerated damage studies of titanate ceramics containing simulated PW-4b and JW-A waste

International Nuclear Information System (INIS)

Hart, K.P.; Vance, E.R.; Lumpkin, G.R.; Mitamura, H.; Matsumoto, S.; Banba, T.

1999-01-01

Ceramic waste forms are affected by radiation damage, primarily arising from aloha-decay processes that can lead to volume expansion and amorphization of the component crystalline phases. The understanding of the extent and impact of these effects on the overall durability of the waste form is critical to the prediction of their long-term performance under repository conditions. Since 1985 ANSTO and JAERI have carried out joint studies on the use of 244 Cm to simulate alpha-radiation damage in ceramic waste forms. These studies have focussed on synroc formulations doped with simulated PW-4b and JW-A wastes. The studies have established the relationship between density change and irradiation levels for Synroc containing JW-A and PW-4b wastes. The storage of samples at 200 C halves the rate of decrease in the density of the samples compared to that measured at room temperature. This effect is consistent with that found for natural samples where the amorphization of natural samples stored under crustal conditions is lower, by factors between 2 and 4, than that measured for samples from accelerated doping experiments stored at room temperature. (J.P.N.)

Design and performance of feed-delivery systems for simulated radioactive waste slurries

International Nuclear Information System (INIS)

Perez, J.M. Jr.

1983-02-01

Processes for vitrifying simulated high-level radioactive waste have been developed at the Pacific Northwest Laboratory (PNL) over the last several years. Paralleling this effort, several feed systems used to deliver the simulated waste slurry to the melter have been tested. Because there had been little industrial experience in delivering abrasive slurries at feed rates of less than 10 L/min, early experience helped direct the design of more-dependable systems. Also, as feed delivery requirements changed, the feed system was modified to meet these new requirements. The various feed systems discussed in this document are part of this evolutionary process, so they have not been ranked against each other. The four slurry feed systems discussed are: (1) vertical-cantilevered centrifugal pump system; (2) airlift feed systems; (3) pressurized-loop systems; and (4) positive-displacement pump system. 20 figures, 11 tables

Variable slip wind generator modeling for real-time simulation

Energy Technology Data Exchange (ETDEWEB)

Gagnon, R.; Brochu, J.; Turmel, G. [Hydro-Quebec, Varennes, PQ (Canada). IREQ

2006-07-01

A model of a wind turbine using a variable slip wound-rotor induction machine was presented. The model was created as part of a library of generic wind generator models intended for wind integration studies. The stator winding of the wind generator was connected directly to the grid and the rotor was driven by the turbine through a drive train. The variable resistors was synthesized by an external resistor in parallel with a diode rectifier. A forced-commutated power electronic device (IGBT) was connected to the wound rotor by slip rings and brushes. Simulations were conducted in a Matlab/Simulink environment using SimPowerSystems blocks to model power systems elements and Simulink blocks to model the turbine, control system and drive train. Detailed descriptions of the turbine, the drive train and the control system were provided. The model's implementation in the simulator was also described. A case study demonstrating the real-time simulation of a wind generator connected at the distribution level of a power system was presented. Results of the case study were then compared with results obtained from the SimPowerSystems off-line simulation. Results showed good agreement between the waveforms, demonstrating the conformity of the real-time and the off-line simulations. The capability of Hypersim for real-time simulation of wind turbines with power electronic converters in a distribution network was demonstrated. It was concluded that hardware-in-the-loop (HIL) simulation of wind turbine controllers for wind integration studies in power systems is now feasible. 5 refs., 1 tab., 6 figs.

Preliminary results from uranium/americium affinity studies under experimental conditions for cesium removal from NPP ''Kozloduy'' simulated wastes solutions

International Nuclear Information System (INIS)

Nikiforova, A.; Kinova, L.; Peneva, C.; Taskaeva, I.; Petrova, P.

2005-01-01

We use the approach described by Westinghouse Savannah River Company using ammonium molybdophosphate (AMP) to remove elevated concentrations of radioactive cesium to facilitate handling waste samples from NPP K ozloduy . Preliminary series of tests were carried out to determine the exact conditions for sufficient cesium removal from five simulated waste solutions with concentrations of compounds, whose complexing power complicates any subsequent processing. Simulated wastes solutions contain high concentrations of nitrates, borates, H 2 C 2 O 4 , ethylenediaminetetraacetate (EDTA) and Citric acid, according to the composition of the real waste from the NPP. On this basis a laboratory treatment protocol was created. This experiment is a preparation for the analysis of real waste samples. In this sense the results are preliminary. Unwanted removal of non-cesium radioactive species from simulated waste solutions was studied with gamma spectrometry with the aim to find a compromise between on the one hand the AMP effectiveness and on the other hand unwanted affinity to AMP of Uranium and Americium. Success for the treatment protocol is defined by proving minimal uptake of U and Am, while at the same time demonstrating good removal effectiveness through the use of AMP. Uptake of U and Am were determined as influenced by oxidizing agents at nitric acid concentrations, proposed by Savannah River National laboratory. It was found that AMP does not significantly remove U and Am when concentration of oxidizing agents is more than 0.1M for simulated waste solutions and for contact times inherent in laboratory treatment protocol. Uranium and Americium affinity under experimental conditions for cesium removal were evaluated from gamma spectrometric data. Results are given for the model experiment and an approach for the real waste analysis is chosen. Under our experimental conditions simulated wastes solutions showed minimal affinity to AMP when U and Am are most probably in

Couplex1 test case nuclear - Waste disposal far field simulation

International Nuclear Information System (INIS)

2001-01-01

This first COUPLEX test case is to compute a simplified Far Field model used in nuclear waste management simulation. From the mathematical point of view the problem is of convection diffusion type but the parameters are highly varying from one layer to another. Another particularity is the very concentrated nature of the source, both in space and in time. (author)

Waste and Simulant Precipitation Issues

International Nuclear Information System (INIS)

Steele, W.V.

2000-01-01

As Savannah River Site (SRS) personnel have studied methods of preparing high-level waste for vitrification in the Defense Waste Processing Facility (DWPF), questions have arisen with regard to the formation of insoluble waste precipitates at inopportune times. One option for decontamination of the SRS waste streams employs the use of an engineered form of crystalline silicotitanate (CST). Testing of the process during FY 1999 identified problems associated with the formation of precipitates during cesium sorption tests using CST. These precipitates may, under some circumstances, obstruct the pores of the CST particles and, hence, interfere with the sorption process. In addition, earlier results from the DWPF recycle stream compatibility testing have shown that leaching occurs from the CST when it is stored at 80 C in a high-pH environment. Evidence was established that some level of components of the CST, such as silica, was leached from the CST. This report describes the results of equilibrium modeling and precipitation studies associated with the overall stability of the waste streams, CST component leaching, and the presence of minor components in the waste streams

Pyrolysis of plastic packaging waste: A comparison of plastic residuals from material recovery facilities with simulated plastic waste.

Science.gov (United States)

Adrados, A; de Marco, I; Caballero, B M; López, A; Laresgoiti, M F; Torres, A

2012-05-01

Pyrolysis may be an alternative for the reclamation of rejected streams of waste from sorting plants where packing and packaging plastic waste is separated and classified. These rejected streams consist of many different materials (e.g., polyethylene (PE), polypropylene (PP), polystyrene (PS), polyvinyl chloride (PVC), polyethylene terephthalate (PET), acrylonitrile butadiene styrene (ABS), aluminum, tetra-brik, and film) for which an attempt at complete separation is not technically possible or economically viable, and they are typically sent to landfills or incinerators. For this study, a simulated plastic mixture and a real waste sample from a sorting plant were pyrolyzed using a non-stirred semi-batch reactor. Red mud, a byproduct of the aluminum industry, was used as a catalyst. Despite the fact that the samples had a similar volume of material, there were noteworthy differences in the pyrolysis yields. The real waste sample resulted, after pyrolysis, in higher gas and solid yields and consequently produced less liquid. There were also significant differences noted in the compositions of the compared pyrolysis products. Copyright © 2011 Elsevier Ltd. All rights reserved.

Damaris: Addressing performance variability in data management for post-petascale simulations

International Nuclear Information System (INIS)

Dorier, Matthieu; Antoniu, Gabriel; Cappello, Franck; Snir, Marc; Sisneros, Robert

2016-01-01

With exascale computing on the horizon, reducing performance variability in data management tasks (storage, visualization, analysis, etc.) is becoming a key challenge in sustaining high performance. Here, this variability significantly impacts the overall application performance at scale and its predictability over time. In this article, we present Damaris, a system that leverages dedicated cores in multicore nodes to offload data management tasks, including I/O, data compression, scheduling of data movements, in situ analysis, and visualization. We evaluate Damaris with the CM1 atmospheric simulation and the Nek5000 computational fluid dynamic simulation on four platforms, including NICS’s Kraken and NCSA’s Blue Waters. Our results show that (1) Damaris fully hides the I/O variability as well as all I/O-related costs, thus making simulation performance predictable; (2) it increases the sustained write throughput by a factor of up to 15 compared with standard I/O approaches; (3) it allows almost perfect scalability of the simulation up to over 9,000 cores, as opposed to state-of-the-art approaches that fail to scale; and (4) it enables a seamless connection to the VisIt visualization software to perform in situ analysis and visualization in a way that impacts neither the performance of the simulation nor its variability. In addition, we extended our implementation of Damaris to also support the use of dedicated nodes and conducted a thorough comparison of the two approaches—dedicated cores and dedicated nodes—for I/O tasks with the aforementioned applications.

Status of test results of electrochemical organic oxidation of a tank 241-SY-101 simulated waste

International Nuclear Information System (INIS)

Colby, S.A.

1994-06-01

This report presents scoping test results of an electrochemical waste pretreatment process to oxidize organic compounds contained in the Hanford Site's radioactive waste storage tanks. Electrochemical oxidation was tested on laboratory scale to destroy organics that are thought to pose safety concerns, using a nonradioactive, simulated tank waste. Minimal development work has been applied to alkaline electrochemical organic destruction. Most electrochemical work has been directed towards acidic electrolysis, as in the metal purification industry, and silver catalyzed oxidation. Alkaline electrochemistry has traditionally been associated with the following: (1) inefficient power use, (2) electrode fouling, and (3) solids handling problems. Tests using a laboratory scale electrochemical cell oxidized surrogate organics by applying a DC electrical current to the simulated tank waste via anode and cathode electrodes. The analytical data suggest that alkaline electrolysis oxidizes the organics into inorganic carbonate and smaller carbon chain refractory organics. Electrolysis treats the waste without adding chemical reagents and at ambient conditions of temperature and pressure. Cell performance was not affected by varying operating conditions and supplemental electrolyte additions

PROCESS VARIABLES OPTIMIZATION FOR PREPARATION AND CHARACTERIZATION OF NOVEL ADSORBENT FROM LIGNOCELLULOSIC WASTE

Directory of Open Access Journals (Sweden)

Zaira Zaman Chowdhury,

2012-07-01

Full Text Available In this study, powdered activated carbon (PAC was produced from agricultural waste biomass of kenaf (Hibiscus cannabinus L. core or stalk, and the process variables were optimized by applying central composite design (CCD. The influences of carbonization temperature, duration of activation, and impregnation ratio on removal percentage of copper Cu(II ions from aqueous solution and carbon yield were investigated. A quadratic model for adsorption percentage and a 2FI model for carbon yield were developed. The models were used to determine optimum process variables for production of activated carbon from the woody core of kenaf. From the analysis of variance (ANOVA, the most significant factor on each response of adsorption capacity and yield were identified. An activation temperature of (568±1°C, activation time of 2.02 hours, and impregnation ratio of 1.75 were found as optimum production conditions. The experimental results showed excellent agreement with the predicted results obtained from a designed experiment that included the variables and responses of sorption capacity and yield. Textural properties of the prepared activated carbons were determined. The performance of the prepared activated carbon was further evaluated by fitting the equilibrium data with regression equations of Langmuir, Freundlich, and Temkin models at different temperatures. The research concluded that the prepared sorbent has good potential to remove Cu (II ions from waste water.

Viscoelastic Earthquake Cycle Simulation with Memory Variable Method

Science.gov (United States)

Hirahara, K.; Ohtani, M.

2017-12-01

There have so far been no EQ (earthquake) cycle simulations, based on RSF (rate and state friction) laws, in viscoelastic media, except for Kato (2002), who simulated cycles on a 2-D vertical strike-slip fault, and showed nearly the same cycles as those in elastic cases. The viscoelasticity could, however, give more effects on large dip-slip EQ cycles. In a boundary element approach, stress is calculated using a hereditary integral of stress relaxation function and slip deficit rate, where we need the past slip rates, leading to huge computational costs. This is a cause for almost no simulations in viscoelastic media. We have investigated the memory variable method utilized in numerical computation of wave propagation in dissipative media (e.g., Moczo and Kristek, 2005). In this method, introducing memory variables satisfying 1st order differential equations, we need no hereditary integrals in stress calculation and the computational costs are the same order of those in elastic cases. Further, Hirahara et al. (2012) developed the iterative memory variable method, referring to Taylor et al. (1970), in EQ cycle simulations in linear viscoelastic media. In this presentation, first, we introduce our method in EQ cycle simulations and show the effect of the linear viscoelasticity on stick-slip cycles in a 1-DOF block-SLS (standard linear solid) model, where the elastic spring of the traditional block-spring model is replaced by SLS element and we pull, in a constant rate, the block obeying RSF law. In this model, the memory variable stands for the displacement of the dash-pot in SLS element. The use of smaller viscosity reduces the recurrence time to a minimum value. The smaller viscosity means the smaller relaxation time, which makes the stress recovery quicker, leading to the smaller recurrence time. Second, we show EQ cycles on a 2-D dip-slip fault with the dip angel of 20 degrees in an elastic layer with thickness of 40 km overriding a Maxwell viscoelastic half

Hanford Waste Vitrification Plant full-scale feed preparation testing with water and process simulant slurries

International Nuclear Information System (INIS)

Gaskill, J.R.; Larson, D.E.; Abrigo, G.P.

1996-03-01

The Hanford Waste Vitrification Plant was intended to convert selected, pretreated defense high-level waste and transuranic waste from the Hanford Site into a borosilicate glass. A full-scale testing program was conducted with nonradioactive waste simulants to develop information for process and equipment design of the feed-preparation system. The equipment systems tested included the Slurry Receipt and Adjustment Tank, Slurry Mix Evaporator, and Melter-Feed Tank. The areas of data generation included heat transfer (boiling, heating, and cooling), slurry mixing, slurry pumping and transport, slurry sampling, and process chemistry. 13 refs., 129 figs., 68 tabs

Monte Carlo simulations of radioactive waste encapsulated by bisphenol-A polycarbonate and effect of bismuth-III oxide filler material

International Nuclear Information System (INIS)

Özdemir, Tonguç

2017-01-01

Radioactive waste generated from the nuclear industry and non-power applications should carefully be treated, conditioned and disposed according to the regulations set by the competent authority(ies). Bisphenol-a polycarbonate (BPA-PC), a very widely used polymer, might be considered as a potential candidate material for low level radioactive waste encapsulation. In this work, the dose rate distribution in the radioactive waste drum (containing radioactive waste and the BPA-PC polymer matrix) was determined using Monte Carlo simulations. Moreover, the change of mechanical properties of BPA-PC was estimated and their variation within the waste drum was determined for the periods of 15, 30 and 300 years after disposal to the final disposal site. The change of the dose rate within the waste drum with different contents of bismuth-III oxide were also simulated. It was concluded that addition of bismuth-III oxide filler decreases the dose delivered to the polymeric matrix due to photoelectric effect. - Highlights: • Bisphenol-a polycarbonate (BPA-PC) is a widely used polymeric material and have a considerable gamma radiation stability. • BPA-PC could have a potential candidate material for radioactive waste embedding. • Activity of the radioactive waste that could be embedded into the BPA-PC matrix was simulated. • Effect of bismuth-III-oxide filler to the BPA-PC matrix was determined.

MIIT: International in-situ testing of nuclear-waste glasses: Performance of SRS simulated waste glass after five years of burial at the Waste Isolation Pilot Plant (WIPP)

International Nuclear Information System (INIS)

Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Clark, D.E.

1991-01-01

In July of 1986, the first in-situ test involving burial of simulated high-level waste (HLW) forms conducted in the United States was started. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest, most cooperative field-testing venture in the international waste management community. In July of 1991, the experimental portion of the 5-year MIIT study was completed on schedule. During this time interval, many in-situ measurements were performed, thousands of brine analyses conducted, and hundreds of waste glass and package components exhumed and evaluated after 6 mo., 1 yr., 2 yr. and 5 yr. burial periods. Although analyses are still in progress, the performance of SRS waste glass based on all data currently available has been seen to be excellent thus far. Initial analyses and assessment of Savannah River (SR) waste glass after burial in WIPP at 90 degrees C for 5 years are presented in this document

Investigation of variable compositions on the removal of technetium from Hanford Waste Treatment Plant low activity waste melter off-gas condensate simulant

Energy Technology Data Exchange (ETDEWEB)

Taylor-Pashow, Kathryn M. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, John M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2017-03-29

The Low Activity Waste (LAW) vitrification facility at the Hanford Waste Treatment and Immobilization Plant (WTP) will generate an aqueous condensate recycle stream (LAW Off-Gas Condensate) from the offgas system. The plan for disposition of this stream during baseline operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility again. The primary reason to recycle this stream is so that the semi-volatile ⁹⁹Tc isotope eventually becomes incorporated into the glass. This stream also contains non-radioactive salt components that are problematic in the melter, so diversion of this stream to another process would eliminate recycling of these salts and would enable simplified operation of the LAW melter and the Pretreatment Facilities. This diversion from recycling this stream within WTP would have the effect of decreasing the LAW vitrification mission duration and quantity of glass waste. The concept being tested here involves removing the ⁹⁹Tc so that the decontaminated aqueous stream, with the problematic salts, can be disposed elsewhere.

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

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.

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.

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

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

Analyst-to-Analyst Variability in Simulation-Based Prediction

Energy Technology Data Exchange (ETDEWEB)

Glickman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Romero, Vicente J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-02-01

This report describes findings from the culminating experiment of the LDRD project entitled, "Analyst-to-Analyst Variability in Simulation-Based Prediction". For this experiment, volunteer participants solving a given test problem in engineering and statistics were interviewed at different points in their solution process. These interviews are used to trace differing solutions to differing solution processes, and differing processes to differences in reasoning, assumptions, and judgments. The issue that the experiment was designed to illuminate -- our paucity of understanding of the ways in which humans themselves have an impact on predictions derived from complex computational simulations -- is a challenging and open one. Although solution of the test problem by analyst participants in this experiment has taken much more time than originally anticipated, and is continuing past the end of this LDRD, this project has provided a rare opportunity to explore analyst-to-analyst variability in significant depth, from which we derive evidence-based insights to guide further explorations in this important area.

Simulated HLLW compositions for cold test of waste management development

International Nuclear Information System (INIS)

Banba, Tsunetaka; Kimura, Hideo; Kamizono, Hiroshi; Tashiro, Shingo

1982-07-01

Three grades of simulated high-level liquid waste (HLLW)-JW-A, JW-B, and JW-C - were proposed to be used respectively according to stages of various cold tests for safety assessment of HLW management. The composition of HLLW was estimated taking into account the spectrum of fission products and actinides, waste volume, corrosion products, and chemical additives. One of conditions, the spectrum of fission products and actinides of LWR spent fuels, was calculated by DCHAIN-code. Fuel burn-up of 28,000 MWD/tUO 2 and 33,000 MWD/tUO 2 were adopted as normal and maximum values of Japanese LWR power plants. The other conditions were estimated using the data obtained at Marcoule plant in France. (author)

G189A analytical simulation of the RITE Integrated Waste Management-Water System

Science.gov (United States)

Coggi, J. V.; Clonts, S. E.

1974-01-01

This paper discusses the computer simulation of the Integrated Waste Management-Water System Using Radioisotopes for Thermal Energy (RITE) and applications of the simulation. Variations in the system temperature and flows due to particular operating conditions and variations in equipment heating loads imposed on the system were investigated with the computer program. The results were assessed from the standpoint of the computed dynamic characteristics of the system and the potential applications of the simulation to system development and vehicle integration.

Challenge problem and milestones for : Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC).

Energy Technology Data Exchange (ETDEWEB)

Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe, Jr.

2010-09-01

This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

Challenge problem and milestones for: Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC)

International Nuclear Information System (INIS)

Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe Jr.

2010-01-01

This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

Long-term interactions of full-scale cemented waste simulates with salt brines

Energy Technology Data Exchange (ETDEWEB)

Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

2016-07-01

Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO{sub 3} solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

Long-term interactions of full-scale cemented waste simulates with salt brines

International Nuclear Information System (INIS)

Kienzler, B.; Borkel, C.; Metz, V.; Schlieker, M.

2016-01-01

Since 1967 radioactive wastes have been disposed of in the Asse II salt mine in Northern Germany. A significant part of these wastes originated from the pilot reprocessing plant WAK in Karlsruhe and consisted of cemented NaNO 3 solutions bearing fission products, actinides, as well as process chemicals. With respect to the long-term behavior of these wastes, the licensing authorities requested leaching experiments with full scale samples in relevant salt solutions which were performed since 1979. The experiments aimed at demonstrating the transferability of results obtained with laboratory samples to real waste forms and at the investigation of the effects of the industrial cementation process on the properties of the waste forms. This research program lasted until 2013. The corroding salt solutions were sampled several times and after termination of the experiments, the solid materials were analyzed by various methods. The results presented in this report cover the evolution of the solutions and the chemical and mineralogical characterization of the solids including radionuclides and waste components, and the paragenesis of solid phases (corrosion products). The outcome is compared to the results of model calculations. For safety analysis, conclusions are drawn on radionuclide retention, evolution of the geochemical environment, evolution of the density of solutions, and effects of temperature and porosity of the cement waste simulates on cesium mobilization.

Composition variability of the organic fraction of municipal solid waste and effects on hydrogen and methane production potentials.

Science.gov (United States)

Alibardi, Luca; Cossu, Raffaello

2015-02-01

The composition of the Organic Fraction of Municipal Solid Waste (OFMSW) strongly depends on the place and time of collection for a specific municipality or area. Moreover synthetic food waste or organic waste from cafeterias and restaurants may not be representative of the overall OFMSW received at treatment facilities for source-separated waste. This work is aimed at evaluating the composition variability of OFMSW, the potential productions of hydrogen and methane from specific organic waste fractions typically present in MSW and the effects of waste composition on overall hydrogen and methane yields. The organic waste fractions considered in the study were: bread-pasta, vegetables, fruits, meat-fish-cheese and undersieve 20mm. Composition analyses were conducted on samples of OFMSW that were source segregated at household level. Batch tests for hydrogen and methane productions were carried out under mesophilic conditions on selected fractions and OFMSW samples. Results indicated that the highest production of hydrogen was achieved by the bread-pasta fraction while the lowest productions were measured for the meat-fish-cheese fraction. The results indicated that the content of these two fractions in organic waste had a direct influence on the hydrogen production potentials of OFMSW. The higher the content of bread-pasta fraction, the higher the hydrogen yields were while the contrary was observed for the meat-fish-cheese fraction. The definition of waste composition therefore represents fundamental information to be reported in scientific literature to allow data comparison. The variability of OFMSW and its effects on hydrogen potentials might also represents a problematic issue in the management of pilot or full-scale plants for the production of hydrogen by dark fermentation. Copyright © 2014 Elsevier Ltd. All rights reserved.

Microbial aspects of gas generation from low level radioactive waste simulant

International Nuclear Information System (INIS)

Kidby, D.W.; Billington, R.S.

1992-01-01

This report details the experimental work undertaken to further the understanding of the kinetics of methanogenesis associated with radioactive LLW disposal. A series of treatments were established by inoculating a LLW simulant and investigating the kinetics of methanogenesis in small Wheaton bottles. Treatments were set up to study the effects of waste compaction, the addition of metal to the simulant, the initial aerobic phase, pH and temperature on gas production. A separate experiment was also established to determine whether cellulose in the simulant acted as a biogas precursor. Results are presented from the head space gas analysis and the solid and liquid phase analyses undertaken over a 600 day period. (Author)

Separation of technetium from nuclear waste stream simulants. Final report

International Nuclear Information System (INIS)

Strauss, S.H.

1995-01-01

The author studied liquid anion exchangers, such as Aliquat-336 nitrate, various pyridinium nitrates, and related salts, so that they may be applied toward a specific process for extracting (partitioning) and recovering 99 TcO 4 - from nuclear waste streams. Many of the waste streams are caustic and contain a variety of other ions. For this reason, the author studied waste stream simulants that are caustic and contain appropriate concentrations of selected, relevant ions. Methods of measuring the performance of the exchangers and extractant systems included contact experiments. Batch contact experiments were used to determine the forward and reverse extraction parameters as a function of temperature, contact time, phase ratio, concentration, solvent (diluent), and other physical properties. They were also used for stability and competition studies. Specifically, the author investigated the solvent extraction behavior of salts of perrhenate (ReO 4 - ), a stable (non-radioactive) chemical surrogate for 99 TcO 4 - . Results are discussed for alternate organic solvents; metalloporphyrins, ferrocenes, and N-cetyl pyridium nitrate as alternate extractant salts; electroactive polymers; and recovery of ReO 4 - and TcO 4 -

Effect of Particle Size Distribution on Slurry Rheology: Nuclear Waste Simulant Slurries

International Nuclear Information System (INIS)

Chun, Jaehun; Oh, Takkeun; Luna, Maria L.; Schweiger, Michael J.

2011-01-01

Controlling the rheological properties of slurries has been of great interest in various industries such as cosmetics, ceramic processing, and nuclear waste treatment. Many physicochemical parameters, such as particle size, pH, ionic strength, and mass/volume fraction of particles, can influence the rheological properties of slurry. Among such parameters, the particle size distribution of slurry would be especially important for nuclear waste treatment because most nuclear waste slurries show a broad particle size distribution. We studied the rheological properties of several different low activity waste nuclear simulant slurries having different particle size distributions under high salt and high pH conditions. Using rheological and particle size analysis, it was found that the percentage of colloid-sized particles in slurry appears to be a key factor for rheological characteristics and the efficiency of rheological modifiers. This behavior was shown to be coupled with an existing electrostatic interaction between particles under a low salt concentration. Our study suggests that one may need to implement the particle size distribution as a critical factor to understand and control rheological properties in nuclear waste treatment plants, such as the U.S. Department of Energy's Hanford and Savannah River sites, because the particle size distributions significantly vary over different types of nuclear waste slurries.

Induction melting of simulated transuranic waste

International Nuclear Information System (INIS)

Tenaglia, R.D.; McCall, J.L.

1983-06-01

Coreless induction melting was investigated as a method to melt and consolidate waste material representative of the transuranic waste (TRU) stored at the Idaho National Engineering Laboratory (INEL). Waste material was introduced onto the surface of a molten cast iron bath in a coreless induction furnace. Waste metallics were incorporated into the bath. Noncombustibles formed a slag which was poured or skimmed from the bath surface. Stack sampling was performed to characterize the off-gas and particulate matter evolved. Experimental melting tests were performed for a variety of types of wastes including metallics, chemical sludge, soil, concrete, and glass. Each test also included a representative level of combustible materials consisting of paper, wood, cloth, polyvinyl chloride and polyethylene. Metallic wastes were readily processed by induction melting with a minimum of slag production. Test waste consisting primarily of chemical sludge provided fluid slags which could be poured from the bath surface. Processing of wastes consisting of soil, concrete, or glass was limited by the inability to achieve fluid slags. It appears from test results that coreless induction melting is a feasible method to process INEL-type waste materials if two problems can be resolved. First, slag fluidity must be improved to facilitate the collection of slags formed from soil, concrete, or glass containing wastes. Secondly, refractory life must be further optimized to permit prolonged processing of the waste materials. The use of a chrome-bearing high-alumina refractory was found to resist slag line attach much better than a magnesia refractory, although some attack was still noted

Complexant Identification in Hanford Waste Simulant Sr/TRU Filtrate

International Nuclear Information System (INIS)

Bannochie, C.J.

2003-01-01

This project was designed to characterize the available multidentate ligand species and metal ion complexes of iron, strontium and manganese formed with the parent chelators, complexing agents and their fragment products. Complex identification was applied to AN-102 and AN-107 filtrate simulants for Hanford waste after an oxidation reaction with sodium permanganate to create a freshly precipitated manganese dioxide solid for adsorption of transuranic elements. Separation efficiency of different ligands was investigated based on the exchange capability of different ion exchange and ion exclusion analytical columns including Dionex IonPac AS-5A, AS-10, AS-11 and AS-6. The elution programs developed with different mobile phase concentrations were based on the change in the effective charge of the anionic species and therefore the retention on the stationary phase. In the present work, qualitative and quantitative assessments of multidentate ligands were investigated. Identification methods for the metal ion complexes responsible for solubilizing Fe, Mn and Sr were applied to aged and fresh simulant waste filtrates. Although concentration measurements of both fresh and 3-week aged filtrates showed that the degradation process occurs mainly due to the harsh chemical environment, it was found that the concentration of iron and manganese did not increase, within the error of the analytical measurements, after three weeks when compared with fresh filtrate

Liquid Secondary Waste Grout Formulation and Waste Form Qualification

Energy Technology Data Exchange (ETDEWEB)

Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Williams, B. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Snyder, Michelle M. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-05-23

This report describes the results from liquid secondary waste (LSW) grout formulation and waste form qualification tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate new formulations for preparing a grout waste form with high-sulfate secondary waste simulants and the release of key constituents from these grout monoliths. Specific objectives of the LSW grout formulation and waste form qualification tests described in this report focused on five activities: 1.preparing new formulations for the LSW grout waste form with high-sulfate LSW simulants and solid characterization of the cured LSW grout waste form; 2.conducting the U.S. Environmental Protection Agency (EPA) Method 1313 leach test (EPA 2012) on the grout prepared with the new formulations, which solidify sulfate-rich Hanford Tank Waste Treatment and Immobilization Plant (WTP) off-gas condensate secondary waste simulant, using deionized water (DIW); 3.conducting the EPA Method 1315 leach tests (EPA 2013) on the grout monoliths made with the new dry blend formulations and three LSW simulants (242-A evaporator condensate, Environmental Restoration Disposal Facility (ERDF) leachate, and WTP off-gas condensate) using two leachants, DIW and simulated Hanford Integrated Disposal Facility (IDF) Site vadose zone pore water (VZPW); 4.estimating the ⁹⁹Tc desorption K_d (distribution coefficient) values for ⁹⁹Tc transport in oxidizing conditions to support the IDF performance assessment (PA); 5.estimating the solubility of ⁹⁹Tc(IV)-bearing solid phases for ⁹⁹Tc transport in reducing conditions to support the IDF PA.

Cesium uptake capacity of simulated ferrocyanide tank waste. Interim report FY 1994, Ferrocyanide Safety Project

International Nuclear Information System (INIS)

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

1994-09-01

The objective of this project is to determine the capacity for 137 CS uptake by mixed metal ferrocyanides present in Hanford waste tanks, and to assess the potential for aggregation of these 137 CS exchanged materials to form tank ''hot-spots.'' This research, performed at the Pacific Northwest Laboratory (PNL) for the Westinghouse Hanford Company (WHC), stems from concerns of possible localized radiolytic heating within the tanks. If radioactive cesium is exchanged and concentrated by the remaining nickel ferrocyanide present in the tanks, this heating could cause temperatures to rise above the safety limits specified for the ferrocyanide tanks. For the purposes of this study, two simulants, In-Farm-2 and U-Plant-2, were chosen to represent the wastes generated by the scavenging processes. These simulants were formulated using protocols from the original cesium scavenging campaign. Later additions of cesium-rich wastes from various processes also were considered. The simulants were prepared and centrifuged to obtain a moist ferrocyanide sludge. The centrifuged sludges were treated with the original supernate spiked with a known amount of cesium nitrate. After analysis by flame atomic absorption spectrometry, distribution coefficients (K d ) were calculated. The capacity of solid waste simulants to exchange radioactive cesium from solution was examined. Initial results showed that the greater the molar ratio of cesium to cesium nickel ferrocyanide, the less effective the exchange of cesium from solution. The theoretical capacity of 2 mol cesium per mol of nickel ferrocyanide was not observed. The maximum capacity under experimental conditions was 0.35 mol cesium per mol nickel ferrocyanide. Future work on this project will examine the layering tendency of the cesium nickel ferrocyanide species

Selection of Steady-State Process Simulation Software to Optimize Treatment of Radioactive and Hazardous Waste

International Nuclear Information System (INIS)

Nichols, T. T.; Barnes, C. M.; Lauerhass, L.; Taylor, D. D.

2001-01-01

The process used for selecting a steady-state process simulator under conditions of high uncertainty and limited time is described. Multiple waste forms, treatment ambiguity, and the uniqueness of both the waste chemistries and alternative treatment technologies result in a large set of potential technical requirements that no commercial simulator can totally satisfy. The aim of the selection process was two-fold. First, determine the steady-state simulation software that best, albeit not completely, satisfies the requirements envelope. And second, determine if the best is good enough to justify the cost. Twelve simulators were investigated with varying degrees of scrutiny. The candidate list was narrowed to three final contenders: ASPEN Plus 10.2, PRO/II 5.11, and CHEMCAD 5.1.0. It was concluded from ''road tests'' that ASPEN Plus appears to satisfy the project's technical requirements the best and is worth acquiring. The final software decisions provide flexibility: they involve annual rather than multi-year licensing, and they include periodic re-assessment

Selection of Steady-State Process Simulation Software to Optimize Treatment of Radioactive and Hazardous Waste

Energy Technology Data Exchange (ETDEWEB)

Nichols, Todd Travis; Barnes, Charles Marshall; Lauerhass, Lance; Taylor, Dean Dalton

2001-06-01

The process used for selecting a steady-state process simulator under conditions of high uncertainty and limited time is described. Multiple waste forms, treatment ambiguity, and the uniqueness of both the waste chemistries and alternative treatment technologies result in a large set of potential technical requirements that no commercial simulator can totally satisfy. The aim of the selection process was two-fold. First, determine the steady-state simulation software that best, albeit not completely, satisfies the requirements envelope. And second, determine if the best is good enough to justify the cost. Twelve simulators were investigated with varying degrees of scrutiny. The candidate list was narrowed to three final contenders: ASPEN Plus 10.2, PRO/II 5.11, and CHEMCAD 5.1.0. It was concluded from "road tests" that ASPEN Plus appears to satisfy the project's technical requirements the best and is worth acquiring. The final software decisions provide flexibility: they involve annual rather than multi-year licensing, and they include periodic re-assessment.

IMPLEMENTING THE DC MODE IN COSMOLOGICAL SIMULATIONS WITH SUPERCOMOVING VARIABLES

International Nuclear Information System (INIS)

Gnedin, Nickolay Y.; Kravtsov, Andrey V.; Rudd, Douglas H.

2011-01-01

As emphasized by previous studies, proper treatment of the density fluctuation on the fundamental scale of a cosmological simulation volume-the D C mode - is critical for accurate modeling of spatial correlations on scales ∼> 10% of simulation box size. We provide further illustration of the effects of the DC mode on the abundance of halos in small boxes and show that it is straightforward to incorporate this mode in cosmological codes that use the 'supercomoving' variables. The equations governing evolution of dark matter and baryons recast with these variables are particularly simple and include the expansion factor, and hence the effect of the DC mode, explicitly only in the Poisson equation.

Iron phosphate glass containing simulated fast reactor waste: Characterization and comparison with pristine iron phosphate glass

International Nuclear Information System (INIS)

Joseph, Kitheri; Asuvathraman, R.; Venkata Krishnan, R.; Ravindran, T.R.; Govindaraj, R.; Govindan Kutty, K.V.; Vasudeva Rao, P.R.

2014-01-01

Detailed characterization was carried out on an iron phosphate glass waste form containing 20 wt.% of a simulated nuclear waste. High temperature viscosity measurement was carried out by the rotating spindle method. The Fe 3+ /Fe ratio and structure of this waste loaded iron phosphate glass was investigated using Mössbauer and Raman spectroscopy respectively. Specific heat measurement was carried out in the temperature range of 300–700 K using differential scanning calorimeter. Isoconversional kinetic analysis was employed to understand the crystallization behavior of the waste loaded iron phosphate glass. The glass forming ability and glass stability of the waste loaded glass were also evaluated. All the measured properties of the waste loaded glass were compared with the characteristics of pristine iron phosphate glass

Molecular Monte Carlo Simulations Using Graphics Processing Units: To Waste Recycle or Not?

Science.gov (United States)

Kim, Jihan; Rodgers, Jocelyn M; Athènes, Manuel; Smit, Berend

2011-10-11

In the waste recycling Monte Carlo (WRMC) algorithm, (1) multiple trial states may be simultaneously generated and utilized during Monte Carlo moves to improve the statistical accuracy of the simulations, suggesting that such an algorithm may be well posed for implementation in parallel on graphics processing units (GPUs). In this paper, we implement two waste recycling Monte Carlo algorithms in CUDA (Compute Unified Device Architecture) using uniformly distributed random trial states and trial states based on displacement random-walk steps, and we test the methods on a methane-zeolite MFI framework system to evaluate their utility. We discuss the specific implementation details of the waste recycling GPU algorithm and compare the methods to other parallel algorithms optimized for the framework system. We analyze the relationship between the statistical accuracy of our simulations and the CUDA block size to determine the efficient allocation of the GPU hardware resources. We make comparisons between the GPU and the serial CPU Monte Carlo implementations to assess speedup over conventional microprocessors. Finally, we apply our optimized GPU algorithms to the important problem of determining free energy landscapes, in this case for molecular motion through the zeolite LTA.

Operating Range for High Temperature Borosilicate Waste Glasses: (Simulated Hanford Enveloped)

International Nuclear Information System (INIS)

Mohammad, J.; Ramsey, W. G.; Toghiani, R. K.

2003-01-01

The following results are a part of an independent thesis study conducted at Diagnostic Instrumentation and Analysis Laboratory-Mississippi State University. A series of small-scale borosilicate glass melts from high-level waste simulant were produced with waste loadings ranging from 20% to 55% (by mass). Crushed glass was allowed to react in an aqueous environment under static conditions for 7 days. The data obtained from the chemical analysis of the leachate solutions were used to test the durability of the resulting glasses. Studies were performed to determine the qualitative effects of increasing the B2O3 content on the overall waste glass leaching behavior. Structural changes in a glass arising due to B2O3 were detected indirectly by its chemical durability, which is a strong function of composition and structure. Modeling was performed to predict glass durability quantitatively in an aqueous environment as a direct function of oxide composition

Research over the extraction process in order to simulate 226Ra and Unat radionuclides migration in uranium waste dumps

International Nuclear Information System (INIS)

Bragea, Mihaela; Toro, Laszlo; Cristache, Carmen

2008-01-01

To simulate the phenomenon of migration availability of radioactive elements in uranium waste product from waste dump into the environment it was taken into account the Ciudanovita, Romania area with its uranium waste dumps. It were made studies through experiments of static extraction in agitation pot (static regime), which allow to pursue the evolution of contaminants along a specific period of time. It was accomplish a study about sorption and adsorption process on the geological environment (clay). It was observed that 226 Ra and U nat transition through the sorption and adsorption processes were induced by the type of clay (due to variable solubility of clay and the physics and chemical status of 226 Ra and U nat into the clay) and the time interval until the water contacts the clay. The concentration measurements of radionuclides into different dimension of clay particles were made to give pertinent information about the way of chosen the repair strategies for the studied site. The present work is based mainly on this research direction: optimization and improvement of the performances obtained from the geo-chemical processes investigation methods in the waste dumps and the surrounding geological structures to isolate them, in order to reduce the negative environmental impact. Measurements of the two radionuclides concentrations in particles with different dimensions were made to supply accurate information regarding the selection of remediation strategies for the dumpsite in study. (author)

Assessment study for multi-barrier system used in radioactive borate waste isolation based on Monte Carlo simulations.

Science.gov (United States)

Bayoumi, T A; Reda, S M; Saleh, H M

2012-01-01

Radioactive waste generated from the nuclear applications should be properly isolated by a suitable containment system such as, multi-barrier container. The present study aims to evaluate the isolation capacity of a new multi-barrier container made from cement and clay and including borate waste materials. These wastes were spiked by (137)Cs and (60)Co radionuclides to simulate that waste generated from the primary cooling circuit of pressurized water reactors. Leaching of both radionuclides in ground water was followed and calculated during ten years. Monte Carlo (MCNP5) simulations computed the photon flux distribution of the multi-barrier container, including radioactive borate waste of specific activity 11.22KBq/g and 4.18KBq/g for (137)Cs and (60)Co, respectively, at different periods of 0, 15.1, 30.2 and 302 years. The average total flux for 100cm radius of spherical cell was 0.192photon/cm(2) at initial time and 2.73×10(-4)photon/cm(2) after 302 years. Maximum waste activity keeping the surface radiation dose within the permissible level was calculated and found to be 56KBq/g with attenuation factors of 0.73cm(-1) and 0.6cm(-1) for cement and clay, respectively. The average total flux was 1.37×10(-3)photon/cm(2) after 302 years. Monte Carlo simulations revealed that the proposed multi-barrier container is safe enough during transportation, evacuation or rearrangement in the disposal site for more than 300 years. Copyright © 2011 Elsevier Ltd. All rights reserved.

Hydrologic transport of radionuclides from low-level waste burial grounds

International Nuclear Information System (INIS)

Duguid, J.O.

1979-01-01

The physical characteristics of the virgin site and of the disturbed site after burial drastically affect the transport of radionuclides from buried waste. The disturbance of the land surface during the waste burial operation causes changes in the local ground-water regimen. These changes can increase the water table elevation and cause the occurrence of perched water in burial trenches. The combination of these changes may lead to submersion of the waste and to increased radionuclide transport from the burial site in both surface and groundwater. Factors such as ion exchange can retard or in some cases, with competing ions, can also mobilize radionuclides and increase their discharge into ground and surface water. Because of complexing agents (organics) contained in the waste, increased mobility of some radionuclides can be expected. The chemical form of radionuclides in the water, the ground-water quality, and the chemistry of the geologic formation in which the waste is buried all influence the movement of radionuclides in the hydrologic system. For the assessment of the environmental impact of low-level waste burial, models capable of simulating both the chemical and the physical factors that affect hydrologic transport must be available. Several models for conducting such simulation are presently available. However, the input parameters used in these models are highly variable; and the accuracy of parameter measurement must be considered in evaluating the reliability of simulated results

Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

Energy Technology Data Exchange (ETDEWEB)

Adamson, Duane J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, Charles A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Crawford, Charles L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Wilmarth, William R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-01-01

(chloride, fluoride, sulfur), will have high ammonia, and will contain carryover particulates of glass-former chemicals. These species have potential to cause corrosion of tanks and equipment, precipitation of solids, release of ammonia gas vapors, and scale in the tank farm evaporator. Routing this stream to the tank farms does not permanently divert it from recycling into the WTP, only temporarily stores it prior to reprocessing. Testing is normally performed to demonstrate acceptable conditions and limits for these compounds in wastes sent to the tank farms. The primary parameter of this phase of the test program was measuring the formation of solids during evaporation in order to assess the compatibility of the stream with the evaporator and transfer and storage equipment. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW facility melter offgas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and, thus, the composition will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. This report discusses results of evaporation testing of the simulant. Two conditions were tested, one with the simulant at near neutral pH, and a second at alkaline pH. The neutral pH test is comparable to the conditions in the Hanford Effluent Treatment Facility (ETF) evaporator, although that evaporator operates at near atmospheric pressure and tests were done under vacuum. For the alkaline test, the target pH was based on the tank farm corrosion control program requirements, and the test protocol and equipment was comparable to that

Laboratory Evaporation Testing Of Hanford Waste Treatment Plant Low Activity Waste Off-Gas Condensate Simulant

International Nuclear Information System (INIS)

Adamson, Duane J.; Nash, Charles A.; McCabe, Daniel J.; Crawford, Charles L.; Wilmarth, William R.

2014-01-01

(chloride, fluoride, sulfur), will have high ammonia, and will contain carryover particulates of glass-former chemicals. These species have potential to cause corrosion of tanks and equipment, precipitation of solids, release of ammonia gas vapors, and scale in the tank farm evaporator. Routing this stream to the tank farms does not permanently divert it from recycling into the WTP, only temporarily stores it prior to reprocessing. Testing is normally performed to demonstrate acceptable conditions and limits for these compounds in wastes sent to the tank farms. The primary parameter of this phase of the test program was measuring the formation of solids during evaporation in order to assess the compatibility of the stream with the evaporator and transfer and storage equipment. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW facility melter offgas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet, and, thus, the composition will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. This report discusses results of evaporation testing of the simulant. Two conditions were tested, one with the simulant at near neutral pH, and a second at alkaline pH. The neutral pH test is comparable to the conditions in the Hanford Effluent Treatment Facility (ETF) evaporator, although that evaporator operates at near atmospheric pressure and tests were done under vacuum. For the alkaline test, the target pH was based on the tank farm corrosion control program requirements, and the test protocol and equipment was comparable to that

Morphologies of CaMoO sub 4 crystals in simulated nuclear waste disposal glass

Energy Technology Data Exchange (ETDEWEB)

Li, D.; Sengers, E.G.F.; Janssen, F.J.J.G. (KEMA, Arnhem (Netherlands). Chemical Technology and Material Research Dept.); Waal, H. de (TPO-TNO Glass Technology, Eindhoven (Netherlands))

1992-07-01

Fission products can diffuse through nuclear waste disposal glass due to the action of the temperature gradient caused by radioactive decay and the small thermal conductivity of the glass. Diffusion may eventually lead to crystallization. Because the densities of the products of glass crystallization may differ from that of the parent glass, crystallization causes stresses to develop, which can lead to fracture and exposure of increased surface area to environmental attack. Several kinds of crystals including, CaMoO{sub 4}, in the simulated nuclear waste disposal glass K{sub 3}, which consists of Na{sub 2}O, B{sub 2}O{sub 3}, SiO{sub 2} and other oxides, were identified previously after heat treatment. Recently it was found that CaMoO{sub 4} crystals have two kinds of morphologies in glass K{sub 3} heat-treated at temperatures between 870 and 1120 K. One kind of morphology, which is rather special, has not previously been reported. In this letter the morphologies of CaMoO{sub 4} crystals in this simulated nuclear waste disposal glass are discussed. (author).

Essays of leaching in cemented products containing simulated waste from evaporator concentrated of PWR reactor

International Nuclear Information System (INIS)

Haucz, Maria Judite A.; Calabria, Jaqueline A. Almeida; Tello, Cledola Cassia O.; Candido, Francisco Donizete; Seles, Sandro Rogerio Novaes

2011-01-01

This paper evaluated the results from leaching resistance essays of cemented products, prepared from three distinct formulations, containing simulated waste of concentrated from the PWR reactor evaporator. The leaching rate is a parameter of qualification of solidified products containing radioactive waste and is determined in accordance with regulation ISO 6961. This procedure evaluates the capacity of transfer organic and inorganic substances presents in the waste through dissolution in the extractor medium. For the case of radioactive waste it is reached the more retention of contaminants in the cemented product, i.e.the lesser value of lixiviation rate. Therefore, this work evaluated among the proposed formulation that one which attend the criterion established in the regulation CNEN-NN-6.09

A feasibility study for the design of a simulated radioactive waste repository facility

International Nuclear Information System (INIS)

1986-10-01

The paper contains the text and references of a feasibility study for the design of a simulated radioactive waste repository facility (final report). The work was commissioned by the Department of the Environment, United Kingdom, as part of its radioactive waste management research programme. The nature of the candidate buffer materials, the factors defining their behaviour, and the nature of a buffer material selection and testing programme, are examined. A description is given of the properties and modelling of host materials. The complex interactions between host materials, and between buffer and host materials, are discussed, along with the instrumentation requirements for measuring the interactions. Finally, the temperature field around a waste package, and modelling a host continuum with a segmental block, are both investigated. (U.K.)

Distributions of 12 elements on 64 absorbers from simulated Hanford Neutralized Current Acid Waste (NCAW)

International Nuclear Information System (INIS)

Svitra, Z.V.; Bowen, S.M.; Marsh, S.F.

1994-12-01

As part of the Hanford Tank Waste Remediation System program at Los Alamos, we evaluated 64 commercially available or experimental absorber materials for their ability to remove hazardous components from high-level waste. These absorbers included cation and anion exchange resins, inorganic exchangers, composite absorbers, and a series of liquid extractants sorbed on porous support-beads. We tested these absorbers with a solution that simulates Hanford neutralized current acid waste (NCAW) (pH 14.2). To this simulant solution we added the appropriate radionuclides and used gamma spectrometry to measure fission products (Cs, Sr, Tc, and Y) and matrix elements (Cr, Co, Fe, Mn, Ni, V, Zn, and Zr). For each of 768 element/absorber combinations, we measured distribution coefficients for dynamic contact periods of 30 min, 2 h, and 6 h to obtain information about sorption kinetics. On the basis of these 2304 measured distribution coefficients, we determined that many of the tested absorbers may be suitable for processing NCAW solutions

Does industrial waste taxation contribute to reduction of landfilled waste? Dynamic panel analysis considering industrial waste category in Japan.

Science.gov (United States)

Sasao, Toshiaki

2014-11-01

Waste taxes, such as landfill and incineration taxes, have emerged as a popular option in developed countries to promote the 3Rs (reduce, reuse, and recycle). However, few studies have examined the effectiveness of waste taxes. In addition, quite a few studies have considered both dynamic relationships among dependent variables and unobserved individual heterogeneity among the jurisdictions. If dependent variables are persistent, omitted variables cause a bias, or common characteristics exist across the jurisdictions that have introduced waste taxes, the standard fixed effects model may lead to biased estimation results and misunderstood causal relationships. In addition, most existing studies have examined waste in terms of total amounts rather than by categories. Even if significant reductions in total waste amounts are not observed, some reduction within each category may, nevertheless, become evident. Therefore, this study analyzes the effects of industrial waste taxation on quantities of waste in landfill in Japan by applying the bias-corrected least-squares dummy variable (LSDVC) estimators; the general method of moments (difference GMM); and the system GMM. In addition, the study investigates effect differences attributable to industrial waste categories and taxation types. This paper shows that industrial waste taxes in Japan have minimal, significant effects on the reduction of final disposal amounts thus far, considering dynamic relationships and waste categories. Copyright © 2014 Elsevier Ltd. All rights reserved.

Retrospection-Simulation-Revision: Approach to the Analysis of the Composition and Characteristics of Medical Waste at a Disaster Relief Site.

Science.gov (United States)

Zhang, Li; Wu, Lihua; Tian, Feng; Wang, Zheng

2016-01-01

A large amount of medical waste is produced during disaster relief, posing a potential hazard to the habitat and the environment. A comprehensive understanding of the composition and characteristics of medical waste that requires management is one of the most basic steps in the development of a plan for medical waste management. Unfortunately, limited reliable information is available in the open literature on the characteristics of the medical waste that is generated at disaster relief sites. This paper discusses the analysis of the composition and characteristics of medical waste at a disaster relief site using the retrospection-simulation-revision method. For this study, we obtained 35 medical relief records of the Wenchuan Earthquake, Sichuan, May 2008 from a field cabin hospital. We first present a retrospective analysis of the relief medical records, and then, we simulate the medical waste generated in the affected areas. We ultimately determine the composition and characteristics of medical waste in the affected areas using untreated medical waste to revise the composition of the simulated medical waste. The results from 35 cases showed that the medical waste generated from disaster relief consists of the following: plastic (43.2%), biomass (26.3%), synthetic fiber (15.3%), rubber (6.6%), liquid (6.6%), inorganic salts (0.3%) and metals (1.7%). The bulk density of medical relief waste is 249 kg/m3, and the moisture content is 44.75%. The data should be provided to assist the collection, segregation, storage, transportation, disposal and contamination control of medical waste in affected areas. In this paper, we wish to introduce this research method of restoring the medical waste generated in disaster relief to readers and researchers. In addition, we hope more disaster relief agencies will become aware of the significance of medical case recording and storing. This may be very important for the environmental evaluation of medical waste in disaster areas, as

A Summary of Rheology Data For SB3 and SB2/3 Blend Simulant Savannah River Site Wastes

International Nuclear Information System (INIS)

KOOPMAN, DAVIDC.

2004-01-01

The purpose of this study was to document the rheological measurements made for Sludge Batch 3 (SB3) and the blend of SB3 with Sludge Batch 2 (SB2). These measurements were primarily made on Slurry Mix Evaporator (SME) products, i.e. melter feeds. Some measurements were made on SB2/3 blend Sludge Receipt and Adjustment Tank (SRAT) products. Measurements on radioactive SB3 and SB2/3 samples have been limited to sludge characterization. SB2/3 measurements studied the impact of changing the SRAT acid addition strategy on the SRAT and SME product rheology. SB2/3 measurements also studied the impact of changing the waste loading target (sludge oxides content in glass) of the SME product. SB3 measurements studied the impact of changes in the wash end point and acid addition strategy on the SME product (melter feed) rheology. A summary of the significant findings is given below: SB3 radioactive sludge and blended SB2/3 radioactive sludge were less viscous than SB2 radioactive sludge. SB2/3 b lend sludge is more viscous than SB3 sludge. SB3 simulant SME product rheology was strongly impacted by changing the noble metal concentrations to more closely match those of the qualification sample. This reduction in noble metals produced a lower pH product that was also considerably less viscous. Increased acid addition in the SB2/3 SRAT generally led to less viscous simulant SRAT products. This trend did not persist in the SME products. SME products became more viscous when increased acid was used in the SRAT cycle from 135 per cent up to 170 per cent of stoichiometry, then became less viscous as total acid was increased further to 185 per cent. A significant increase in hydrogen generation occurred between 170 per cent and 185 per cent. The impact of acid addition on SB3 SME products was also variable. The impact of waste loading changes from 31 to 35 to 40 per cent on SB2/3 simulant SME products led to more viscous melter feeds as waste loading increased at constant wt. per cent

Separation of technetium from nuclear waste stream simulants. Final report

Energy Technology Data Exchange (ETDEWEB)

Strauss, S.H. [Colorado State Univ., Fort Collins, CO (United States). Dept. of Chemistry

1995-09-11

The author studied liquid anion exchangers, such as Aliquat-336 nitrate, various pyridinium nitrates, and related salts, so that they may be applied toward a specific process for extracting (partitioning) and recovering {sup 99}TcO{sub 4}{sup {minus}} from nuclear waste streams. Many of the waste streams are caustic and contain a variety of other ions. For this reason, the author studied waste stream simulants that are caustic and contain appropriate concentrations of selected, relevant ions. Methods of measuring the performance of the exchangers and extractant systems included contact experiments. Batch contact experiments were used to determine the forward and reverse extraction parameters as a function of temperature, contact time, phase ratio, concentration, solvent (diluent), and other physical properties. They were also used for stability and competition studies. Specifically, the author investigated the solvent extraction behavior of salts of perrhenate (ReO{sub 4}{sup {minus}}), a stable (non-radioactive) chemical surrogate for {sup 99}TcO{sub 4}{sup {minus}}. Results are discussed for alternate organic solvents; metalloporphyrins, ferrocenes, and N-cetyl pyridium nitrate as alternate extractant salts; electroactive polymers; and recovery of ReO{sub 4}{sup {minus}} and TcO{sub 4}{sup {minus}}.

Analytical technology in support of the Waste Isolation Pilot Plant (WIPP)

International Nuclear Information System (INIS)

Villareal, R.

1994-01-01

The need for long-term disposal of defense-related transuranic (TRU) wastes became apparent as the DOE recognized the environmental consequences of maintaining waste storage facilities designed for short or interim storage periods, not long-term storage. In 1979, Congress authorized the Waste Isolation Pilot Plant (WIPP), a research and development facility and full-scale pilot plant, to demonstrate the safe management, storage, and disposal of TRU wastes. Environmental Protection Agency (EPA) regulations governing disposal of TRU wastes in 40 CFR 191 require that TRU waste disposal systems be designed to limit migration of radionuclides to the accessible environment for 10,000 years based on performance assessment results. The actinide source-term waste test program (STTP) is an experiment designed to quantitatively measure the time-dependent concentrations of plutonium, uranium, neptunium, thorium, and americium in TRU wastes immersed in brines that simulate the chemistry that may occur in WIPP disposal rooms, partially or completely contacted with brines. The total concentration of each actinide in brine is the sum of its dissolved and colloidally suspended components, as determined by variables including pcH, oxidation-reduction potential (Eh), chelating and complexing agents, sorption capacity, and colloidal suspension capabilities. To determine the effect of influencing variables on the concentration of actinides in WIPP brines, several TRU waste types will be characterized and loaded into specially designed noncorrosive test containers filled with brine containing additives to enhance the action of each influencing variable. The test container brine and headspace gases will be analyzed

Pilot-scale grout production test with a simulated low-level waste

International Nuclear Information System (INIS)

Fow, C.L.; Mitchell, D.H.; Treat, R.L.; Hymas, C.R.

1987-05-01

Plans are underway at the Hanford Site near Richland, Washington, to convert the low-level fraction of radioactive liquid wastes to a grout form for permanent disposal. Grout is a mixture of liquid waste and grout formers, including portland cement, fly ash, and clays. In the plan, the grout slurry is pumped to subsurface concrete vaults on the Hanford Site, where the grout will solidify into large monoliths, thereby immobilizing the waste. A similar disposal concept is being planned at the Savannah River Laboratory site. The underground disposal of grout was conducted at Oak Ridge National Laboratory between 1966 and 1984. Design and construction of grout processing and disposal facilities are underway. The Transportable Grout Facility (TGF), operated by Rockwell Hanford Operations (Rockwell) for the Department of Energy (DOE), is scheduled to grout Phosphate/Sulfate N Reactor Operations Waste (PSW) in FY 1988. Phosphate/Sulfate Waste is a blend of two low-level waste streams generated at Hanford's N Reactor. Other wastes are scheduled to be grouted in subsequent years. Pacific Northwest Laboratory (PNL) is verifying that Hanford grouts can be safely and efficiently processed. To meet this objective, pilot-scale grout process equipment was installed. On July 29 and 30, 1986, PNL conducted a pilot-scale grout production test for Rockwell. During the test, 16,000 gallons of simulated nonradioactive PSW were mixed with grout formers to produce 22,000 gallons of PSW grout. The grout was pumped at a nominal rate of 15 gpm (about 25% of the nominal production rate planned for the TGF) to a lined and covered trench with a capacity of 30,000 gallons. Emplacement of grout in the trench will permit subsequent evaluation of homogeneity of grout in a large monolith. 12 refs., 34 figs., 5 tabs

Immobilization of simulated low and intermediate level waste in alkali-activated slag-fly ash-metakaolin hydroceramics

Energy Technology Data Exchange (ETDEWEB)

Wang, Jin, E-mail: wjin761026@163.com [State Key Laboratory Cultivation Base for Nonmetal Composite and Functional Materials, Southwest University of Science and Technology, Mianyang 621010, Sichuan (China); School of Materials Science and Engineering, Southwest University of Science and Technology, Mianyang 621010, Sichuan (China); Wang, Jun-xia; Zhang, Qin [School of Materials Science and Engineering, Southwest University of Science and Technology, Mianyang 621010, Sichuan (China); Li, Yu-xiang [State Key Laboratory Cultivation Base for Nonmetal Composite and Functional Materials, Southwest University of Science and Technology, Mianyang 621010, Sichuan (China); School of Materials Science and Engineering, Southwest University of Science and Technology, Mianyang 621010, Sichuan (China)

2016-04-15

Highlights: • Evaluation of the suitability of ASFMH for solidifying simulated S-LILW. • The introduction of S-LILW avails forming zeolitic phases of ASFMH waste forms. • The ASFMH waste forms have low leachability and high compressive strength. - Abstract: In the current study, the alkali-activated slag-fly ash-metakaolin hydroceramic (ASFMH) waste forms for immobilizing simulated low and intermediate level waste (S-LILW) were prepared by hydrothermal process. The crystalline phase compositions, morphology, compressive strength and aqueous stability of S-LILW ASFMH waste forms were investigated. The results showed that the main crystalline phases of S-LILW ASFMH waste forms were analcime and zeolite NaP1. The changes of Si/Al molar ratio (from 1.7 to 2.2) and Ca/Al molar ratio (from 0.15 to 0.35) had little effect on the phase compositions of S-LILW ASFMH waste forms. However, the hydrothermal temperature, time as well as the content of S-LILW (from 12.5 to 37.5 wt%) had a major impact on the phase compositions. The compressive strength of S-LILW ASFMH waste forms was not less than 20 MPa when the content of S-LILW reached 37.5 wt%. In addition, the aqueous stability testing was carried out using the standard MCC-1 static leach test method; the normalized elemental leach rates of Sr and Cs were fairly constant in a low value below 5 × 10{sup −4} g m{sup −2} d{sup −1} and 3 × 10{sup −4} g m{sup −2} d{sup −1} after 28 days, respectively. It is indicated that ASFMH waste form could be a potential host for safely immobilizing LILW.

Biochemistry Oxidation Process for Treatment the Simulation of Organic Liquid Radioactive Waste

International Nuclear Information System (INIS)

Gunandjar; Zainus Salimin; Sugeng Purnomo; Ratiko

2010-01-01

The nuclear industry activities generate the organic liquid wastes such as detergent waste from laundry, solvent waste of 30% TBP (tri-n-butyl phosphate) in kerosene from purification or recovery of uranium from rejection of nuclear fuel element fabrication, and solvent waste containing D 2 EHPA (di-2-ethyl hexyl phosphoric acid) and TOPO (trioctyl phospine oxide) in kerosene from phosphoric acid purification. The wastes are included in category of the hazard and poison materials which also radioactive, so that the wastes have to be treated to detoxification of the hazard and poison materials and decontamination of the radionuclides. The research of biochemistry oxidation process for treatment the simulation of organic liquid radioactive waste from laundry using mixture of aerobe bacteria of bacillus sp, pseudomonas sp, arthrobacter sp, and aeromonas sp have been carried out. The waste containing detergent 1,496 g/Litre, activity 10 -1 Ci/m 3 , with COD (Chemical Oxygen Demand) 128, BOD (Biological Oxygen Demand) 68 and TSS (Total Suspended Solid) 1000 ppm, it is treated by biochemistry oxidation with addition of bacteria which be fed nutrition of nitrogen and phosphor, and aeration. The result show that the bacteria can decompose the detergent to become carbon dioxyde and water so that can fulfill the quality standard of water group-B with content of BOD and COD are 6 and 10 ppm respectively, the time of decomposition is needed 106 hours to be fulfill the quality standard of water. The longer of process time will give bigger the total solid content in sludge, because the biomass generated from the colony of bacteria which life and dead to so much. (author)

HANFORD TANK WASTE OPERATIONS SIMULATOR VERSION DESCRIPTION DOCUMENT

International Nuclear Information System (INIS)

ALLEN, G.K.

2003-01-01

This document describes the software version controls established for the Hanford Tank Waste Operations Simulator (HTWOS). It defines: the methods employed to control the configuration of HTWOS; the version of each of the 26 separate modules for the version 1.0 of HTWOS; the numbering rules for incrementing the version number of each module; and a requirement to include module version numbers in each case results documentation. Version 1.0 of HTWOS is the first version under formal software version control. HTWOS contains separate revision numbers for each of its 26 modules. Individual module version numbers do not reflect the major release HTWOS configured version number

Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

Energy Technology Data Exchange (ETDEWEB)

Digby Macdonald; Brian Marx; Balaji Soundararajan; Morgan Smith

2005-07-28

The different tasks that have been carried out under the current program are as follows: (1) Theoretical and experimental assessment of general corrosion of iron/steel in borate buffer solutions by using electrochemical impedance spectroscopy (EIS), ellipsometry and XPS techniques; (2) Development of a damage function analysis (DFA), which would help in predicting the accumulation of damage due to pitting corrosion in an environment prototypical of DOE liquid waste systems; (3) Experimental measurement of crack growth rate, acoustic emission signals, and coupling currents for fracture in carbon and low alloy steels as functions of mechanical (stress intensity), chemical (conductivity), electrochemical (corrosion potential, ECP), and microstructural (grain size, precipitate size, etc) variables in a systematic manner, with particular attention being focused on the structure of the noise in the current and its correlation with the acoustic emissions; (4) Development of fracture mechanisms for carbon and low alloy steels that are consistent with the crack growth rate, coupling current data and acoustic emissions; (5) Inserting advanced crack growth rate models for SCC into existing deterministic codes for predicting the evolution of corrosion damage in DOE liquid waste storage tanks; (6) Computer simulation of the anodic and cathodic activity on the surface of the steel samples in order to exactly predict the corrosion mechanisms; (7) Wavelet analysis of EC noise data from steel samples undergoing corrosion in an environment similar to that of the high level waste storage containers, to extract data pertaining to general, pitting and stress corrosion processes, from the overall data. The work has yielded a number of important findings, including an unequivocal demonstration of the role of chloride ion in passivity breakdown on nickel in terms of cation vacancy generation within the passive film, the first detection and characterization of individual micro fracture

Selection of Steady-State Process Simulation Software to Optimize Treatment of Radioactive and Hazardous Waste

Energy Technology Data Exchange (ETDEWEB)

Nichols, T. T.; Barnes, C. M.; Lauerhass, L.; Taylor, D. D.

2001-06-01

The process used for selecting a steady-state process simulator under conditions of high uncertainty and limited time is described. Multiple waste forms, treatment ambiguity, and the uniqueness of both the waste chemistries and alternative treatment technologies result in a large set of potential technical requirements that no commercial simulator can totally satisfy. The aim of the selection process was two-fold. First, determine the steady-state simulation software that best, albeit not completely, satisfies the requirements envelope. And second, determine if the best is good enough to justify the cost. Twelve simulators were investigated with varying degrees of scrutiny. The candidate list was narrowed to three final contenders: ASPEN Plus 10.2, PRO/II 5.11, and CHEMCAD 5.1.0. It was concluded from ''road tests'' that ASPEN Plus appears to satisfy the project's technical requirements the best and is worth acquiring. The final software decisions provide flexibility: they involve annual rather than multi-year licensing, and they include periodic re-assessment.

Discrete event simulation of the Defense Waste Processing Facility (DWPF) analytical laboratory

International Nuclear Information System (INIS)

Shanahan, K.L.

1992-02-01

A discrete event simulation of the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) analytical laboratory has been constructed in the GPSS language. It was used to estimate laboratory analysis times at process analytical hold points and to study the effect of sample number on those times. Typical results are presented for three different simultaneous representing increasing levels of complexity, and for different sampling schemes. Example equipment utilization time plots are also included. SRS DWPF laboratory management and chemists found the simulations very useful for resource and schedule planning

Comments on ''Use of conditional simulation in nuclear waste site performance assessment'' by Carol Gotway

International Nuclear Information System (INIS)

Downing, D.J.

1993-01-01

This paper discusses Carol Gotway's paper, ''The Use of Conditional Simulation in Nuclear Waste Site Performance Assessment.'' The paper centers on the use of conditional simulation and the use of geostatistical methods to simulate an entire field of values for subsequent use in a complex computer model. The issues of sampling designs for geostatistics, semivariogram estimation and anisotropy, turning bands method for random field generation, and estimation of the comulative distribution function are brought out

Formulation and preparation of Hanford Waste Treatment Plant direct feed low activity waste Effluent Management Facility core simulant

Energy Technology Data Exchange (ETDEWEB)

McCabe, Daniel J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nash, Charles A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL; Adamson, Duane J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL

2016-05-01

The Hanford Waste Treatment and Immobilization Plant (WTP) Low Activity Waste (LAW) vitrification facility will generate an aqueous condensate recycle stream (LAW Melter Off-Gas Condensate, LMOGC) from the off-gas system. The baseline plan for disposition of this stream during full WTP operations is to send it to the WTP Pretreatment Facility, where it will be blended with LAW, concentrated by evaporation and recycled to the LAW vitrification facility. However, during the Direct Feed LAW (DFLAW) scenario, planned disposition of this stream is to evaporate it in a new evaporator in the Effluent Management Facility (EMF) and then return it to the LAW melter. It is important to understand the composition of the effluents from the melter and new evaporator so that the disposition of these streams can be accurately planned and accommodated. Furthermore, alternate disposition of the LMOGC stream would eliminate recycling of problematic components, and would enable less integrated operation of the LAW melter and the Pretreatment Facilities. Alternate disposition would also eliminate this stream from recycling within WTP when it begins operations and would decrease the LAW vitrification mission duration and quantity of glass waste, amongst the other problems such a recycle stream present. This LAW Melter Off-Gas Condensate stream will contain components that are volatile at melter temperatures and are problematic for the glass waste form, such as halides and sulfate. Because this stream will recycle within WTP, these components accumulate in the Melter Condensate stream, exacerbating their impact on the number of LAW glass containers that must be produced. Diverting the stream reduces the halides and sulfate in the recycled Condensate and is a key outcome of this work. This overall program examines the potential treatment and immobilization of this stream to enable alternative disposal. The objective of this task was to formulate and prepare a simulant of the LAW Melter

Hydrologic transport of radionuclides from low-level waste burial grounds

International Nuclear Information System (INIS)

Duguid, J.O.

1977-01-01

The physical characteristics of the virgin site and of the disturbed site after burial drastically affect the transport of radionuclides from buried waste. The disturbance of the land surface during the waste burial operation causes changes in the local ground-water regimen. These changes can increase the water table elevation and cause the occurrence of perched water in burial trenches. The combination of these changes may lead to submersion of the waste and to increased radionuclide transport from the burial site in both surface and ground water. Factors such as ion exchange can retard or in some cases, with competing ions, can also mobilize radionuclides and increase their discharge into ground and surface water. Because of complexing agents (organics) contained in the waste, increased mobility of some radionuclides can be expected. The chemical form of radionuclides in the water, the ground-water quality, and the chemistry of the geologic formation in which the waste is buried all influence the movement of radionuclides in the hydrologic system. For the assessment of the environmental impact of low-level waste burial, models capable of simulating both the chemical and the physical factors that affect hydrologic transport must be available. Several models for conducting such simulation are presently available. However,the input parameters used in these models are highly variable, and the accuracy of parameter measurement must be considered in evaluating the reliability of simulated results

Shallow-land burial of low-level radioactive wastes: preliminary simulations of long-term health risks

International Nuclear Information System (INIS)

Fields, D.E.; Little, C.A.; Emerson, C.J.; Hiromoto, G.

1982-01-01

PRESTO, a computer code developed for the Environmental Protection Agency for the evaluation of possible health effects associated with shallow-land rad-waste burial areas, has been used to perform simulations for three such sites. Preliminary results for the 1000 y period following site closure suggest that shallow burial, at properly chosen sites, is indeed an appropriate disposal practice for low-level wastes. Periods of maximum risk to subject populations are also inferred

Analysis of a waste-heat boiler by CFD simulation

Energy Technology Data Exchange (ETDEWEB)

Yang, Yongziang; Jokilaakso, A. [Helsinki Univ. of Technology, Otaniemi (Finland)

1996-12-31

Waste-heat boilers play important roles in the continuous operation of a smelter and in the conservation of energy. However, the fluid flow and heat transfer behaviour has not been well studied, concerning the boiler performance and design. This presentation describes simulated gas flow and heat transfer of a waste-heat boiler in the Outokumpu copper flash smelting process. The governing transport equations for the conservation of mass, momentum and enthalpy were solved with a commercial CFD-code PHOENICS. The standard k-{epsilon} turbulence model and a composite-flux radiation model were used in the computations. The computational results show that the flow is strongly recirculating and distinctly three-dimensional in most part of the boiler, particularly in the radiation section. The predicted flow pattern and temperature distribution were in a good agreement with laboratory models and industrial measurements. The results provide detailed information of flow pattern, the temperature distribution and gas cooling efficiency. The CFD proved to be a useful tool in analysing the boiler operation. (author)

Analysis of a waste-heat boiler by CFD simulation

Energy Technology Data Exchange (ETDEWEB)

Yang, Yongziang; Jokilaakso, A [Helsinki Univ. of Technology, Otaniemi (Finland)

1997-12-31

Waste-heat boilers play important roles in the continuous operation of a smelter and in the conservation of energy. However, the fluid flow and heat transfer behaviour has not been well studied, concerning the boiler performance and design. This presentation describes simulated gas flow and heat transfer of a waste-heat boiler in the Outokumpu copper flash smelting process. The governing transport equations for the conservation of mass, momentum and enthalpy were solved with a commercial CFD-code PHOENICS. The standard k-{epsilon} turbulence model and a composite-flux radiation model were used in the computations. The computational results show that the flow is strongly recirculating and distinctly three-dimensional in most part of the boiler, particularly in the radiation section. The predicted flow pattern and temperature distribution were in a good agreement with laboratory models and industrial measurements. The results provide detailed information of flow pattern, the temperature distribution and gas cooling efficiency. The CFD proved to be a useful tool in analysing the boiler operation. (author)

Laboratory-scale vitrification and leaching of Hanford high-level waste for the purpose of simulant and glass property models validation

International Nuclear Information System (INIS)

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

1993-02-01

The Hanford Waste Vitrification Plant (HWVP) is being built to process the high-level and TRU waste into canistered glass logs for disposal in a national repository. Testing programs have been established within the Project to verify process technology using simulated waste. A parallel testing program with actual radioactive waste is being performed to confirm the validity of using simulates and glass property models for waste form qualification and process testing. The first feed type to be processed by and the first to be tested on a laboratory-scale is pretreated neutralized current acid waste (NCAW). The NCAW is a neutralized high-level waste stream generated from the reprocessing of irradiated nuclear fuel in the Plutonium and Uranium Extraction (PUREX) Plant at Hanford. As part of the fuel reprocessing, the high-level waste generated in PUREX was denitrated with sugar to form current acid waste (CAW). Sodium hydroxide and sodium nitrite were added to the CAW to minimize corrosion in the tanks, thus yielding neutralized CAW. The NCAW contains small amounts of plutonium, fission products from the irradiated fuel, stainless steel corrosion products, and iron and sulfate from the ferrous sulfamate reductant used in the PUREX process. This paper will discuss the results and status of the laboratory-scale radioactive testing

Uncertainty analysis of a nondestructive radioassay system for transuranic waste

International Nuclear Information System (INIS)

Harker, Y.D.; Blackwood, L.G.; Meachum, T.R.; Yoon, W.Y.

1996-01-01

Radioassay of transuranic waste in 207 liter drums currently stored at the Idaho National Engineering Laboratory is achieved using a Passive Active Neutron (PAN) nondestructive assay system. In order to meet data quality assurance requirements for shipping and eventual permanent storage of these drums at the Waste Isolation Pilot Plant in Carlsbad, New Mexico, the total uncertainty of the PAN system measurements must be assessed. In particular, the uncertainty calculations are required to include the effects of variations in waste matrix parameters and related variables on the final measurement results. Because of the complexities involved in introducing waste matrix parameter effects into the uncertainty calculations, standard methods of analysis (e.g., experimentation followed by propagation of errors) could not be implemented. Instead, a modified statistical sampling and verification approach was developed. In this modified approach the total performance of the PAN 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 describes the simulation process and illustrates its application to waste comprised of weapons grade plutonium-contaminated graphite molds

BAYESIAN TECHNIQUES FOR COMPARING TIME-DEPENDENT GRMHD SIMULATIONS TO VARIABLE EVENT HORIZON TELESCOPE OBSERVATIONS

Energy Technology Data Exchange (ETDEWEB)

Kim, Junhan; Marrone, Daniel P.; Chan, Chi-Kwan; Medeiros, Lia; Özel, Feryal; Psaltis, Dimitrios, E-mail: junhankim@email.arizona.edu [Department of Astronomy and Steward Observatory, University of Arizona, 933 N. Cherry Avenue, Tucson, AZ 85721 (United States)

2016-12-01

The Event Horizon Telescope (EHT) is a millimeter-wavelength, very-long-baseline interferometry (VLBI) experiment that is capable of observing black holes with horizon-scale resolution. Early observations have revealed variable horizon-scale emission in the Galactic Center black hole, Sagittarius A* (Sgr A*). Comparing such observations to time-dependent general relativistic magnetohydrodynamic (GRMHD) simulations requires statistical tools that explicitly consider the variability in both the data and the models. We develop here a Bayesian method to compare time-resolved simulation images to variable VLBI data, in order to infer model parameters and perform model comparisons. We use mock EHT data based on GRMHD simulations to explore the robustness of this Bayesian method and contrast it to approaches that do not consider the effects of variability. We find that time-independent models lead to offset values of the inferred parameters with artificially reduced uncertainties. Moreover, neglecting the variability in the data and the models often leads to erroneous model selections. We finally apply our method to the early EHT data on Sgr A*.

Noble metal behavior during melting of simulated high-level nuclear waste glass feeds

International Nuclear Information System (INIS)

Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

1993-04-01

Noble metals and their oxides can settle in waste glass melters and cause electrical shorting. Simulated waste feeds from Hanford, Savannah River, and Germany were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C--1000 degrees C and examined by electron microscopy to determine shapes, sizes, and distribution of noble metal particles as a function of temperature. Individual noble metal particles and agglomerates of rhodium (Rh), ruthenium (RuO 2 ), and palladium (Pd), as well as their alloys, were seen. the majority of particles and agglomerates were generally less than 10 microns; however, large agglomerations (up to 1 mm) were found in the German feed. Detailed particle distribution and characterization was performed for a Hanford waste to provide input to computer modeling of particle settling in the melter

Selective partitioning of mercury from co-extracted actinides in a simulated acidic ICPP waste stream

International Nuclear Information System (INIS)

Brewer, K.N.; Herbst, R.S.; Tranter, T.J.

1995-01-01

The TRUEX process is being evaluated at the Idaho Chemical Processing Plant (ICPP) as a means to partition the actinides from acidic sodium-bearing waste (SBW). The mercury content of this waste averages 1 g/l. Because the chemistry of mercury has not been extensively evaluated in the TRUEX process, mercury was singled out as an element of interest. Radioactive mercury, 203 Hg, was spiked into a simulated solution of SBW containing 1 g/l mercury. Successive extraction batch contacts with the mercury spiked waste simulant and successive scrubbing and stripping batch contacts of the mercury loaded TRUEX solvent (0.2 M CMPO-1.4 M TBP in dodecane) show that mercury will extract into and strip from the solvent. The extraction distribution coefficient for mercury, as HgCl 2 from SBW having a nitric acid concentration of 1.4 M and a chloride concentration of 0.035 M was found to be 3. The stripping distribution coefficient was found to be 0.5 with 5 M HNO 3 and 0.077 with 0.25 M Na 2 CO 3 . An experimental flowsheet was designed from the batch contact tests and tested counter-currently using 5.5 cm centrifugal contactors. Results from the counter-current test show that mercury can be removed from the acidic mixed SBW simulant and recovered separately from the actinides

Numerical simulation of air- and water-flow experiments in a block of variably saturated, fractured tuff from Yucca Mountain, Nevada

International Nuclear Information System (INIS)

Kwicklis, E.M.; Healy, R.W.; Thamir, F.; Hampson, D.

1998-01-01

Numerical models of water movement through variably saturated, fractured tuff have undergone little testing against experimental data collected from relatively well-controlled and characterized experiments. This report used the results of a multistage experiment on a block of variably saturated, fractured, welded tuff and associated core samples to investigate if those results could be explained using models and concepts currently used to simulate water movement in variably saturated, fractured tuff at Yucca Mountain, Nevada, the potential location of a high-level nuclear-waste repository. Aspects of the experiment were modeled with varying degrees of success. Imbibition experiments performed on cores of various lengths and diameters were adequately described by models using independently measured permeabilities and moisture-characteristic curves, provided that permeability reductions resulting from the presence of entrapped air were considered. Entrapped gas limited maximum water saturations during imbibition to approximately 0.70 to 0,80 of the fillable porosity values determined by vacuum saturation. A numerical simulator developed for application to fluid flow problems in fracture networks was used to analyze the results of air-injection tests conducted within the tuff block through 1.25-cm-diameter boreholes. These analyses produced estimates of transmissivity for selected fractures within the block. Transmissivities of other fractures were assigned on the basis of visual similarity to one of the tested fractures. The calibrated model explained 53% of the observed pressure variance at the monitoring boreholes (with the results for six outliers omitted) and 97% of the overall pressure variance (including monitoring and injection boreholes) in the subset of air-injection tests examined

Development of an immobilisation technique by cementation for non-radioactive simulated liquid waste, from Mo-99 production process

International Nuclear Information System (INIS)

Arva, E A; Marabini, S G; Varani, J L

2012-01-01

The Argentine Atomic Energy Commission (CNEA) is the responsible for developing a management nuclear waste disposal programme. This programme contemplates the strictly environmental safe and efficient management of the radioactive waste from different sources. Since 1985, CNEA has been producing commercially Mo-99 for medical use. In this process two types of liquid waste are produced. One of them has high alkaline (NaOH 3,5M) and aluminate contents. Since Mo-99 production started, such liquid waste was stored in specially designed containers during production, and after a decay period in smaller containers in interim storage conditions. As this waste is still a liquid, development of an immobilisation technique is required. Immobilisation of radioactive liquid waste by cementation is a frequently used technique, and will be studied in the present work using Mo-99 non-radioactive simulated liquid waste. In this second stage, a full scale (200 liters drum) cementation test using simulated non radioactive waste was carried out. Such test included: using the BEBA 201 mixing machine - the same that will be used with real waste in the future for 'tuning up' the process, construction of a specially designed temperature sensor for measuring the maximum temperature value (five different positions, four inside the drum and one outside) and the time elapsed after all components mixing. Finally, standard specimens (IRAM 1622) were made for mechanical resistance tests after cement setting at 28 days. The results show values of temperature not above 40 o C with the maximum at 12 hours before component mixing and compression strength of 14 MPa. Such values are compatible for a waste immobilisation process by cementation (author)

FDS3 simulations of indoor hydrocarbon fires engulfing radioactive waste packages

International Nuclear Information System (INIS)

Bruecher, W.; Roewekamp, M.; Kunze, V.

2004-01-01

The thermal environment of a hypothetical large indoor hydrocarbon pool fire is more complex compared to outdoor fires and can be more severe for engulfed objects. In order to analyze potential thermal environments for interim storage of spent fuel casks or low-level radioactive waste packages engulfed in pool fires numerical simulations with the CFD fire code FDS3 were carried out for different storage configurations. In addition, data of indoor pool fire experiments were used to validate the model for this type of application. A series of pool fire experiments under different ventilation conditions and varied pool surface (1 m 2 - 4 m 2 ) inside a compartment of 3.6 m x 3.6 m x 5.7 m was conducted at iBMB (Institut fuer Baustoffe, Massivbau und Brandschutz) of Braunschweig University of Technology, Germany. The instrumentation included thermocouples, heatflux and pressure gauges, bi-directional flow probes and gas concentration measurements. A mock low-level waste drum equipped with outside and inside thermocouples was positioned as an additional heat sink near the fire source. Two of these experiments have recently been used for benchmarking a number of fire simulation codes within the International Collaborative Fire Model Project (ICFMP). FDS3 simulations by GRS of some of the above mentioned experiments will be presented showing the ability of the model to sufficiently well represent the fire environment in most cases. Further simulations were performed for hypothetical pool fire environments in interim storage facilities for German spent fuel transport and storage casks. The resulting temperature curves were then used for the thermomechanical analysis of the cask reaction performed by BAM (Bundesanstalt fuer Materialforschung und -pruefung, see corresponding conference paper by Wieser et al.). The FDS3 pool fire simulations show that the fire environment is strongly influenced by the ventilation conditions and cooling effects depending on the number and

Final report on LDRD project: Simulation/optimization tools for system variability analysis

Energy Technology Data Exchange (ETDEWEB)

R. L. Bierbaum; R. F. Billau; J. E. Campbell; K. D. Marx; R. J. Sikorski; B. M. Thompson; S. D. Wix

1999-10-01

>This work was conducted during FY98 (Proposal Number 98-0036) and FY99 (Proposal Number 99-0818) under the auspices of the Sandia National Laboratories Laboratory-Directed Research and Development (LDRD) program. Electrical simulation typically treats a single data point in the very large input space of component properties. For electrical simulation to reach its full potential as a design tool, it must be able to address the unavoidable variability and uncertainty in component properties. Component viability is strongly related to the design margin (and reliability) of the end product. During the course of this project, both tools and methodologies were developed to enable analysis of variability in the context of electrical simulation tools. Two avenues to link relevant tools were also developed, and the resultant toolset was applied to a major component.

Hot-wall corrosion testing of simulated high level nuclear waste

International Nuclear Information System (INIS)

Chandler, G.T.; Zapp, P.E.; Mickalonis, J.I.

1995-01-01

Three materials of construction for steam tubes used in the evaporation of high level radioactive waste were tested under heat flux conditions, referred to as hot-wall tests. The materials were type 304L stainless steel alloy C276, and alloy G3. Non-radioactive acidic and alkaline salt solutions containing halides and mercury simulated different high level waste solutions stored or processed at the United States Department of Energy's Savannah River Site. Alloy C276 was also tested for corrosion susceptibility under steady-state conditions. The nickel-based alloys C276 and G3 exhibited excellent corrosion resistance under the conditions studied. Alloy C276 was not susceptible to localized corrosion and had a corrosion rate of 0.01 mpy (0.25 μm/y) when exposed to acidic waste sludge and precipitate slurry at a hot-wall temperature of 150 degrees C. Type 304L was susceptible to localized corrosion under the same conditions. Alloy G3 had a corrosion rate of 0.1 mpy (2.5 μm/y) when exposed to caustic high level waste evaporator solution at a hot-wall temperature of 220 degrees C compared to 1.1 mpy (28.0 μ/y) for type 304L. Under extreme caustic conditions (45 weight percent sodium hydroxide) G3 had a corrosion rate of 0.1 mpy (2.5 μm/y) at a hot-wall temperature of 180 degrees C while type 304L had a high corrosion rate of 69.4 mpy (1.8 mm/y)

Characterization Of Actinides In Simulated Alkaline Tank Waste Sludges And Leachates

International Nuclear Information System (INIS)

Nash, Kenneth L.

2008-01-01

In this project, both the fundamental chemistry of actinides in alkaline solutions (relevant to those present in Hanford-style waste storage tanks), and their dissolution from sludge simulants (and interactions with supernatants) have been investigated under representative sludge leaching procedures. The leaching protocols were designed to go beyond conventional alkaline sludge leaching limits, including the application of acidic leachants, oxidants and complexing agents. The simulant leaching studies confirm in most cases the basic premise that actinides will remain in the sludge during leaching with 2-3 M NaOH caustic leach solutions. However, they also confirm significant chances for increased mobility of actinides under oxidative leaching conditions. Thermodynamic data generated improves the general level of experiemental information available to predict actinide speciation in leach solutions. Additional information indicates that improved Al removal can be achieved with even dilute acid leaching and that acidic Al(NO3)3 solutions can be decontaminated of co-mobilized actinides using conventional separations methods. Both complexing agents and acidic leaching solutions have significant potential to improve the effectiveness of conventional alkaline leaching protocols. The prime objective of this program was to provide adequate insight into actinide behavior under these conditions to enable prudent decision making as tank waste treatment protocols develop.

CHARACTERIZATION OF ACTINIDES IN SIMULATED ALKALINE TANK WASTE SLUDGES AND LEACHATES

Energy Technology Data Exchange (ETDEWEB)

Nash, Kenneth L.

2008-11-20

In this project, both the fundamental chemistry of actinides in alkaline solutions (relevant to those present in Hanford-style waste storage tanks), and their dissolution from sludge simulants (and interactions with supernatants) have been investigated under representative sludge leaching procedures. The leaching protocols were designed to go beyond conventional alkaline sludge leaching limits, including the application of acidic leachants, oxidants and complexing agents. The simulant leaching studies confirm in most cases the basic premise that actinides will remain in the sludge during leaching with 2-3 M NaOH caustic leach solutions. However, they also confirm significant chances for increased mobility of actinides under oxidative leaching conditions. Thermodynamic data generated improves the general level of experiemental information available to predict actinide speciation in leach solutions. Additional information indicates that improved Al removal can be achieved with even dilute acid leaching and that acidic Al(NO3)3 solutions can be decontaminated of co-mobilized actinides using conventional separations methods. Both complexing agents and acidic leaching solutions have significant potential to improve the effectiveness of conventional alkaline leaching protocols. The prime objective of this program was to provide adequate insight into actinide behavior under these conditions to enable prudent decision making as tank waste treatment protocols develop.

Numerical and experimental studies on effects of moisture content on combustion characteristics of simulated municipal solid wastes in a fixed bed

International Nuclear Information System (INIS)

Sun, Rui; Ismail, Tamer M.; Ren, Xiaohan; Abd El-Salam, M.

2015-01-01

Highlights: • The effects of moisture content on the burning process of MSW are investigated. • A two-dimensional mathematical model was built to simulate the combustion process. • Temperature distributions, process rates, gas species were measured and simulated. • The The conversion ratio of C/CO and N/NO in MSW are inverse to moisture content. - Abstract: In order to reveal the features of the combustion process in the porous bed of a waste incinerator, a two-dimensional unsteady state model and experimental study were employed to investigate the combustion process in a fixed bed of municipal solid waste (MSW) on the combustion process in a fixed bed reactor. Conservation equations of the waste bed were implemented to describe the incineration process. The gas phase turbulence was modeled using the k–ε turbulent model and the particle phase was modeled using the kinetic theory of granular flow. The rate of moisture evaporation, devolatilization rate, and char burnout was calculated according to the waste property characters. The simulation results were then compared with experimental data for different moisture content of MSW, which shows that the incineration process of waste in the fixed bed is reasonably simulated. The simulation results of solid temperature, gas species and process rate in the bed are accordant with experimental data. Due to the high moisture content of fuel, moisture evaporation consumes a vast amount of heat, and the evaporation takes up most of the combustion time (about 2/3 of the whole combustion process). The whole bed combustion process reduces greatly as MSW moisture content increases. The experimental and simulation results provide direction for design and optimization of the fixed bed of MSW

Numerical and experimental studies on effects of moisture content on combustion characteristics of simulated municipal solid wastes in a fixed bed

Energy Technology Data Exchange (ETDEWEB)

Sun, Rui, E-mail: Sunsr@hit.edu.cn [School of Energy Science and Engineering, Harbin Institute of Technology, 92, West Dazhi Street, Harbin 150001 (China); Ismail, Tamer M., E-mail: temoil@aucegypt.edu [Department of Mechanical Engineering, Suez Canal University, Ismailia (Egypt); Ren, Xiaohan [School of Energy Science and Engineering, Harbin Institute of Technology, 92, West Dazhi Street, Harbin 150001 (China); Abd El-Salam, M. [Department of Basic Science, Cairo University, Giza (Egypt)

2015-05-15

Highlights: • The effects of moisture content on the burning process of MSW are investigated. • A two-dimensional mathematical model was built to simulate the combustion process. • Temperature distributions, process rates, gas species were measured and simulated. • The The conversion ratio of C/CO and N/NO in MSW are inverse to moisture content. - Abstract: In order to reveal the features of the combustion process in the porous bed of a waste incinerator, a two-dimensional unsteady state model and experimental study were employed to investigate the combustion process in a fixed bed of municipal solid waste (MSW) on the combustion process in a fixed bed reactor. Conservation equations of the waste bed were implemented to describe the incineration process. The gas phase turbulence was modeled using the k–ε turbulent model and the particle phase was modeled using the kinetic theory of granular flow. The rate of moisture evaporation, devolatilization rate, and char burnout was calculated according to the waste property characters. The simulation results were then compared with experimental data for different moisture content of MSW, which shows that the incineration process of waste in the fixed bed is reasonably simulated. The simulation results of solid temperature, gas species and process rate in the bed are accordant with experimental data. Due to the high moisture content of fuel, moisture evaporation consumes a vast amount of heat, and the evaporation takes up most of the combustion time (about 2/3 of the whole combustion process). The whole bed combustion process reduces greatly as MSW moisture content increases. The experimental and simulation results provide direction for design and optimization of the fixed bed of MSW.

Biochemical process of low level radioactive liquid simulation waste containing detergent

International Nuclear Information System (INIS)

Kundari, Noor Anis; Putra, Sugili; Mukaromah, Umi

2015-01-01

Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10 −5 Ci/m 3 . The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod’s model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour −1

Biochemical process of low level radioactive liquid simulation waste containing detergent

Energy Technology Data Exchange (ETDEWEB)

Kundari, Noor Anis, E-mail: nooranis@batan.go.id; Putra, Sugili; Mukaromah, Umi [Sekolah Tinggi Teknologi Nuklir – Badan Tenaga Nuklir Nasional Jl. Babarsari P.O. BOX 6101 YKBB Yogyakarta 55281 Telp : (0274) 48085, 489716, Fax : (0274) 489715 (Indonesia)

2015-12-29

Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10{sup −5} Ci/m{sup 3}. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod’s model and the decreasing of COD and BOD were first order with the rate constant of 0

Biochemical process of low level radioactive liquid simulation waste containing detergent

Science.gov (United States)

Kundari, Noor Anis; Putra, Sugili; Mukaromah, Umi

2015-12-01

Research of biochemical process of low level radioactive liquid waste containing detergent has been done. Thse organic liquid wastes are generated in nuclear facilities such as from laundry. The wastes that are cotegorized as hazard and poison materials are also radioactive. It must be treated properly by detoxification of the hazard and decontamination of the radionuclides to ensure that the disposal of the waste meets the requirement of standard quality of water. This research was intended to determine decontamination factor and separation efficiensies, its kinetics law, and to produce a supernatant that ensured the environmental quality standard. The radioactive element in the waste was thorium with activity of 5.10-5 Ci/m3. The radioactive liquid waste which were generated in simulation plant contains detergents that was further processed by aerobic biochemical process using SGB 103 bacteria in a batch reactor equipped with aerators. Two different concentration of samples were processed and analyzed for 212 hours and 183 hours respectively at a room temperature. The product of this process is a liquid phase called as supernatant and solid phase material called sludge. The chemical oxygen demand (COD), biological oxygen demand (BOD), suspended solid (SS), and its alpha activity were analyzed. The results show that the decontamination factor and the separation efficiency of the lower concentration samples are higher compared to the samples with high concentration. Regarding the decontamination factor, the result for 212 hours processing of waste with detergent concentration of 1.496 g/L was 3.496 times, whereas at the detergent concentration of 0.748 g/L was 15.305 times for 183 hours processing. In case of the separation efficiency, the results for both samples were 71.396% and 93.465% respectively. The Bacterial growth kinetics equation follow Monod's model and the decreasing of COD and BOD were first order with the rate constant of 0.01 hour-1.

In vitro studies evaluating leaching of mercury from mine waste calcine using simulated human body fluids

Science.gov (United States)

Gray, John E.; Plumlee, Geoffrey S.; Morman, Suzette A.; Higueras, Pablo L.; Crock, James G.; Lowers, Heather A.; Witten, Mark L.

2010-01-01

In vitro bioaccessibility (IVBA) studies were carried out on samples of mercury (Hg) mine-waste calcine (roasted Hg ore) by leaching with simulated human body fluids. The objective was to estimate potential human exposure to Hg due to inhalation of airborne calcine particulates and hand-to-mouth ingestion of Hg-bearing calcines. Mine waste calcines collected from Hg mines at Almadén, Spain, and Terlingua, Texas, contain Hg sulfide, elemental Hg, and soluble Hg compounds, which constitute primary ore or compounds formed during Hg retorting. Elevated leachate Hg concentrations were found during calcine leaching using a simulated gastric fluid (as much as 6200 μg of Hg leached/g sample). Elevated Hg concentrations were also found in calcine leachates using a simulated lung fluid (as much as 9200 μg of Hg leached/g), serum-based fluid (as much as 1600 μg of Hg leached/g), and water of pH 5 (as much as 880 μg of Hg leached/g). The leaching capacity of Hg is controlled by calcine mineralogy; thus, calcines containing soluble Hg compounds contain higher leachate Hg concentrations. Results indicate that ingestion or inhalation of Hg mine-waste calcine may lead to increased Hg concentrations in the human body, especially through the ingestion pathway.

Simulation of spontaneous and variable global dust storms with the GFDL Mars GCM

OpenAIRE

Basu, Shabari; Wilson, John; Richardson, Mark; Ingersoll, Andrew

2006-01-01

We report on the successful simulation of global dust storms in a general circulation model. The simulated storms develop spontaneously in multiyear simulations and exhibit significant interannual variability. The simulated storms produce dramatic increases in atmospheric dustiness, global-mean air temperatures, and atmospheric circulation intensity, in accord with observations. As with observed global storms, spontaneous initiation of storms in the model occurs in southern spring and summer,...

Performance Variables and Professional Experience in Simulated Laparoscopy: A Two-Group Learning Curve Study

NARCIS (Netherlands)

Luursema, J.M.; Rovers, Maroeska M.; Groenier, Marleen; van Goor, Harry

2014-01-01

Objective Virtual reality simulators are increasingly used in laparoscopy training. Such simulators allow objective assessment of performance. However, both low-level variables and overall scores generated by the simulator can be hard to interpret. We present a method to generate intermediate

Performance variables and professional experience in simulated laparoscopy: a two-group learning curve study

NARCIS (Netherlands)

Luursema, J.M.; Rovers, M.M.; Groenier, M.; Goor, H. van

2014-01-01

OBJECTIVE: Virtual reality simulators are increasingly used in laparoscopy training. Such simulators allow objective assessment of performance. However, both low-level variables and overall scores generated by the simulator can be hard to interpret. We present a method to generate intermediate

Quantitative Comparison of the Variability in Observed and Simulated Shortwave Reflectance

Science.gov (United States)

Roberts, Yolanda, L.; Pilewskie, P.; Kindel, B. C.; Feldman, D. R.; Collins, W. D.

2013-01-01

The Climate Absolute Radiance and Refractivity Observatory (CLARREO) is a climate observation system that has been designed to monitor the Earth's climate with unprecedented absolute radiometric accuracy and SI traceability. Climate Observation System Simulation Experiments (OSSEs) have been generated to simulate CLARREO hyperspectral shortwave imager measurements to help define the measurement characteristics needed for CLARREO to achieve its objectives. To evaluate how well the OSSE-simulated reflectance spectra reproduce the Earth s climate variability at the beginning of the 21st century, we compared the variability of the OSSE reflectance spectra to that of the reflectance spectra measured by the Scanning Imaging Absorption Spectrometer for Atmospheric Cartography (SCIAMACHY). Principal component analysis (PCA) is a multivariate decomposition technique used to represent and study the variability of hyperspectral radiation measurements. Using PCA, between 99.7%and 99.9%of the total variance the OSSE and SCIAMACHY data sets can be explained by subspaces defined by six principal components (PCs). To quantify how much information is shared between the simulated and observed data sets, we spectrally decomposed the intersection of the two data set subspaces. The results from four cases in 2004 showed that the two data sets share eight (January and October) and seven (April and July) dimensions, which correspond to about 99.9% of the total SCIAMACHY variance for each month. The spectral nature of these shared spaces, understood by examining the transformed eigenvectors calculated from the subspace intersections, exhibit similar physical characteristics to the original PCs calculated from each data set, such as water vapor absorption, vegetation reflectance, and cloud reflectance.

Management of solid wastes during decommissioning of research reactors. Evaluation of gross clearance levels and mathematical simulation of solid waste assay techniques

International Nuclear Information System (INIS)

Gopalakrishnan, R.K.; Sobhan Babu, K.; Sharma, D.N.

2008-01-01

Full text: Decommissioning of nuclear facilities constitute a challenge mainly due to the huge and complex nature of radioactive waste generated during this process. In the context of management and disposal of waste and reuse/recycle of usable materials during decommissioning of reactors, clearance levels for relevant radio nuclides are of vital importance. Radionuclide specific clearance levels are developed by IAEA and such levels allow the facility for free release of materials to the environment without further regulatory consideration. An effort has been made in this paper to establish clearance levels for radionuclides associated with various system and structural components of a research reactor and rather than radionuclide specific clearance levels, these values are derived for gross activity concentration, which is more practical for radioactive waste categorization, disposal and reuse or recycle of usable materials. The first step towards the derivation of clearance levels is the calculation of annual doses relating to unit activity concentration for each nuclide using various enveloping scenarios. After the estimation of doses, the limiting enveloping scenario (the one that gives the highest dose) is identified. The clearance levels are then derived by dividing the reference dose level (10 μSv/y) by the annual dose calculated per unit activity concentration for the limiting enveloping scenario The clearance level for gross beta-gamma activity concentration is then evaluated as the product of the limiting clearance level and the number of radionuclides characterized for the structural components. Simulation studies were also carried out for the design of a monitoring system for estimation of activity concentration of the decommissioned materials, especially rubbles/ concrete, using mathematical models. Conventional solid waste assay techniques would not suffice to the requirement of decommissioning waste categorization since very low level activity

Process simulation and economic analysis of biodiesel production from waste cooking oil with membrane bioreactor

Science.gov (United States)

Abdurakhman, Yuanita Budiman; Putra, Zulfan Adi; Bilad, Muhammad Roil

2017-10-01

Pollution and shortage of clean energy supply are among major problems that are caused by rapid population growth. Due to this growth, waste cooking oil is one of the pollution sources. On the other hand, biodiesel appears to be one of the most promising and feasible energy sources as it emits less toxic pollutants and greenhouse gases than petroleum diesel. Thus, biodiesel production using waste cooking oil offers a two-in-one solution to cater pollution and energy issues. However, the conventional biodiesel production process using homogeneous base catalyst and stirred tank reactor is unable to produce high purity of biodiesel from waste cooking oil. It is due its sensitivity to free fatty acid (FFA) content in waste cooking oil and purification difficulties. Therefore, biodiesel production using heterogeneous acid catalyst in membrane reactor is suggested. The product of this process is fatty acid methyl esters (FAME) or biodiesel with glycerol as by-product. This project is aimed to study techno-economic feasibility of biodiesel production from waste cooking oil via heterogeneous acid catalyst in membrane reactor. Aspen HYSYS is used to accomplish this aim. Several cases, such as considering different residence times and the production of pharmaceutical (USP) grade glycerol, are evaluated and compared. Economic potential of these cases is calculated by considering capital expenditure, utilities cost, product and by-product sales, as well as raw material costs. Waste cooking oil, inorganic pressure-driven membrane and WAl is used as raw material, type of membrane and heterogeneous acid catalyst respectively. Based on literature data, FAME yield formulation is developed and used in the reactor simulation. Simulation results shows that economic potential increases by 30% if pharmaceutical (USP) grade glycerol is produced regardless the residence time of the reactor. In addition, there is no significant effect of residence time on the economic potential.

Waste isolation safety assessment program. Collection and generation of transport data

International Nuclear Information System (INIS)

Apps, J.A.

1977-01-01

A project devoted to evaluation of mechanisms and rates of radioactive waste transport in igneous, metamorphic and sedimentary rocks is described. The research effort includes/ (1) calculation of the range of concentration expected for different radionuclides in given geologic environments by computer simulation of the groundwater chemistry; (2) development of a comprehensive theory relating exchange constants (K/sub D/s) to significant variables; (3) fabrication of test equipment to measure waste radionuclide transport rates in rock samples; (4) identification of transport rate controlling mechanisms; (5) experiments to determine K/sub D/ values for important radioactive waste elements for a variety of rock types and environmental conditions

Variable Ratio Hydrostatic Transmission Simulator for Optimal Wind Power Drivetrains

Directory of Open Access Journals (Sweden)

Jose M. Garcia-Bravo

2017-01-01

Full Text Available This work presents a hydromechanical transmission coupled to an electric AC motor and DC generator to simulate a wind power turbine drive train. The goal of this project was to demonstrate and simulate the ability of a hydrostatic variable ratio system to produce constant electric power at varying wind speeds. The experimental results show that the system can maintain a constant voltage when a 40% variation in input speed is produced. An accompanying computer simulation of the system was built and experimentally validated showing a discrete error no larger than 12%. Both the simulation and the experimental results show that the electrical power output can be regulated further if an energy storage device is used to absorb voltage spikes produced by abrupt changes in wind speed or wind direction.

Variety and variability of bentonites as buffer materials in radioactive waste disposal

International Nuclear Information System (INIS)

Sato, Tsutomu

1994-01-01

Bentonite which is considered to be most promising as the buffer material in the formation disposal of high level radioactive waste is the clay having Montmorillonite of smectite group as the main component mineral. The clay of smectite group shows different properties, and its range of variability is wide. In this report, the clay minerals of smectite group and their variety are explained from the viewpoint of crystal chemistry, and the difference in expansion property, water recovery property and long period stability, which are expected for the buffer material in formation disposal, in various smectite clays is described. The trend of the investigation of the buffer materials and the importance of making their standard are referred to. In the formation disposal of high level radioactive waste, multiple barrier concept is investigated. The expectation for the development of intelligent materials is large. Bentonite established the position as one of the intelligent materials. The factors controlling the properties of clay are the compositions of clay minerals, nonclay minerals, organic substances, exchangeable cations and soluble salts and texture. (K.I.)

Simulation studies for quantification of solid waste during decommissioning of nuclear reactors

International Nuclear Information System (INIS)

Sobhan Babu, K.; Gopalakrishnan, R.K.; Gupta, P.C.

2007-01-01

Decommissioning is the final phase in the lifecycle of a nuclear installation and in the area of occupational radiation protection, decommissioning constitute a challenge mainly due to the huge and complex radioactive waste generation. In the context of management and disposal of waste and reuse/recycle of usable materials during decommissioning of reactors, clearance levels for relevant radionuclides are of vital importance. During the process of decommissioning radionuclide-specific clearance levels allow the release of a major quantity of materials to the environment, without regulatory considerations. These levels may also be used to declare the usable materials for reuse or recycle. Assessment of activity concentration in huge quantities of material, for the purpose of clearance, is a challenge in decommissioning process. This paper describes the simulation studies being carried out for the design of a monitoring system for the estimation of activity concentration of the decommissioned materials, especially rubbles/concrete, using mathematical models. Several designs were studied using simulation and it was observed that for the estimation of very low levels of activity concentration, to satisfy the conditions of unrestricted releases, detection system using the principle of Emission Computed Tomography (ECT) is the best suitable method. (author)

Biochar Preparation from Simulated Municipal Solid Waste Employing Low Temperature Carbonization Process

Science.gov (United States)

Areeprasert, C.; Leelachaikul, P.; Jangkobpattana, G.; Phumprasop, K.; Kiattiwat, T.

2018-02-01

This paper presents an investigation on carbonization process of simulated municipal solid waste (MSW). Simulated MSW consists of a representative of food residue (68%), plastic waste (20%), paper (8%), and textile (4%). Laboratory-scale carbonization was performed in this study using a vertical-type pyrolyzer varying carbonization temperature (300, 350, 400, and 450 °C) and heating rate (5, 10, 15, and 20 °C/min). Appearance of the biochar product was in black and the volume was significantly reduced. Low carbonization temperature (300 °C) might not completely decompose plastic materials in MSW. Results showed that the carbonization at the temperature of 400 °C with the heating rate of 5 °C/min was the optimal condition. The yield of biochar from the optimal process was 50.6% with the heating value of 26.85 MJ/kg. Energy input of the process was attributed to water evaporation and the decomposition of plastics and paper. Energy output of the process was highest at the optimal condition. Energy output and input ratio was around 1.3-1.7 showing the feasibility of the carbonization process in all heating rate condition.

Molecular Simulation Study of Montmorillonite in Contact with Variably Wet Supercritical Carbon Dioxide

KAUST Repository

Kadoura, Ahmad Salim; Nair, Arun Kumar Narayanan; Sun, Shuyu

2017-01-01

We perform grand canonical Monte Carlo simulations to study the detailed molecular mechanism of intercalation behavior of CO2 in Na-, Ca-, and Mg- montmorillonite exposed to variably hydrated supercritical CO2 at 323.15 K and 90 bar, The simulations

Isothermal crystallization kinetics in simulated high-level nuclear waste glass

International Nuclear Information System (INIS)

Vienna, J.D.; Hrma, P.; Smith, D.E.

1997-01-01

Crystallization kinetics of a simulated high-level waste (HLW) glass were measured and modelled. Kinetics of acmite growth in the standard HW39-4 glass were measured using the isothermal method. A time-temperature-transformation (TTT) diagram was generated from these data. Classical glass-crystal transformation kinetic models were empirically applied to the crystallization data. These models adequately describe the kinetics of crystallization in complex HLW glasses (i.e., RSquared = 0.908). An approach to measurement, fitting, and use of TTT diagrams for prediction of crystallinity in a HLW glass canister is proposed

Graphical models for simulation and control of robotic systems for waste handling

International Nuclear Information System (INIS)

Drotning, W.D.; Bennett, P.C.

1992-01-01

This paper discusses detailed geometric models which have been used within a graphical simulation environment to study transportation cask facility design and to perform design and analyses of robotic systems for handling of nuclear waste. The models form the basis for a robot control environment which provides safety, flexibility, and reliability for operations which span the spectrum from autonomous control to tasks requiring direct human intervention

Using Aspen simulation package to determine solubility of mixed salts in TRU waste evaporator bottoms

Energy Technology Data Exchange (ETDEWEB)

Hatchell, J.L.

1998-03-01

Nitric acid from plutonium process waste is a candidate for waste minimization by recycling. Process simulation software packages, such as Aspen, are valuable tools to estimate how effective recovery processes can be, however, constants in equations of state for many ionic components are not in their data libraries. One option is to combine single salt solubility`s in the Aspen model for mixed salt system. Single salt solubilities were regressed in Aspen within 0.82 weight percent of literature values. These were combined into a single Aspen model and used in the mixed salt studies. A simulated nitric acid waste containing mixed aluminum, calcium, iron, magnesium and sodium nitrate was tested to determine points of solubility between 25 and 100 C. Only four of the modeled experimental conditions, at 50 C and 75 C, produced a saturated solution. While experimental results indicate that sodium nitrate is the first salt to crystallize out, the Aspen computer model shows that the most insoluble salt, magnesium nitrate, the first salt to crystallize. Possible double salt formation is actually taking place under experimental conditions, which is not captured by the Aspen model.

Areal variability of the mineral soil cover in a reclaimed soda waste dumping site

Directory of Open Access Journals (Sweden)

Klatka Sławomir

2017-03-01

Full Text Available Areal variability of the mineral soil cover in a reclaimed soda waste dumping site. This paper provides an analysis of the areal variability of the thickness and selected physical and chemical properties of the mineral cover formed in the process of settling ponds reclamation at the former Krakow Soda Plant “Solvay”. The topsoil is intended to provide a substrate for plants, therefore, its quality is the main determinant of the development for herbaceous and woody vegetation. Areal variability of the topsoil parameters was determined by kriging. In the context of the envisaged direction of management of the settling ponds, the analysis showed that electrical conductivity, thickness of the soil cover and the sand fraction content have potentially the highest impact on the diversification of vegetation. Understanding the spatial variability of the soil cover parameters, that are essential for vegetation, may contribute to increasing the efficiency of biological reclamation and also to cost reduction. Precise selection of the areas unsuitable for plant growth makes it possible to improve soil parameters on limited areas similarly as in the precision agriculture.

Rotary kiln incinerator engineering tests on simulated transuranic wastes from the Idaho National Engineering Laboratory. Final report

International Nuclear Information System (INIS)

Pattengill, M.G.; Brunner, F.A.; Fasso, J.L.; Mitchel, S.R.; Praskac, R.T.

1982-09-01

Nine rotary kiln incineration tests were performed at Colorado School of Mines Research Institute on simulated transuranic waste materials. The rotary kiln incinerator used as 3 ft ID and 30 ft long and was included in an incineration system that also included an afterburner and a baghouse. The purpose of the incineration test program was to determine the applicability and operating characteristics of the rotary kiln with relation to the complete incineration of the simulated waste materials. The results of the study showed that the rotary kiln did completely incinerate the waste materials. Off-gas determinations showed emission levels of SO 2 , NO/sub x/, H 2 SO 4 , HC1, particulate loading, and hydrocarbons, as well as exhaust gas volume, to be within reasonable controllable ranges in a production operation. Included in the report are the results of materials and energy balances, based upon data collected, and design recommendations based upon the data and upon observations during the incineration operation

Chemical analysis of simulated high level waste glasses to support stage III sulfate solubility modeling

Energy Technology Data Exchange (ETDEWEB)

Fox, K. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-03-17

The U.S. Department of Energy (DOE), Office of Environmental Management (EM) is sponsoring an international, collaborative project to develop a fundamental model for sulfate solubility in nuclear waste glass. The solubility of sulfate has a significant impact on the achievable waste loading for nuclear waste forms within the DOE complex. These wastes can contain relatively high concentrations of sulfate, which has low solubility in borosilicate glass. This is a significant issue for low-activity waste (LAW) glass and is projected to have a major impact on the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Sulfate solubility has also been a limiting factor for recent high level waste (HLW) sludge processed at the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF). The low solubility of sulfate in glass, along with melter and off-gas corrosion constraints, dictate that the waste be blended with lower sulfate concentration waste sources or washed to remove sulfate prior to vitrification. The development of enhanced borosilicate glass compositions with improved sulfate solubility will allow for higher waste loadings and accelerate mission completion.The objective of the current scope being pursued by SHU is to mature the sulfate solubility model to the point where it can be used to guide glass composition development for DWPF and WTP, allowing for enhanced waste loadings and waste throughput at these facilities. A series of targeted glass compositions was selected to resolve data gaps in the model and is identified as Stage III. SHU fabricated these glasses and sent samples to SRNL for chemical composition analysis. SHU will use the resulting data to enhance the sulfate solubility model and resolve any deficiencies. In this report, SRNL provides chemical analyses for the Stage III, simulated HLW glasses fabricated by SHU in support of the sulfate solubility model development.

In Vitro Studies Evaluating Leaching of Mercury from Mine Waste Calcine Using Simulated Human Body Fluids

OpenAIRE

Gray, John E.; Plumlee, Geoffrey S.; Morman, Suzette A.; Higueras, Pablo L.; Crock, James G.; Lowers, Heather A.; Witten, Mark L.

2010-01-01

In vitro bioaccessibility (IVBA) studies were carried out on samples of mercury (Hg) mine-waste calcine (roasted Hg ore) by leaching with simulated human body fluids. The objective was to estimate potential human exposure to Hg due to inhalation of airborne calcine particulates and hand-to-mouth ingestion of Hg-bearing calcines. Mine waste calcines collected from Hg mines at Almad?n, Spain, and Terlingua, Texas, contain Hg sulfide, elemental Hg, and soluble Hg compounds, which constitute prim...

Investigating the effect of variable gutter technique as a novel method on vertical flow of material in closed die forging processes

Energy Technology Data Exchange (ETDEWEB)

Pourbashiri, M.; Sedighi, M. [Iran University, Tehran (Iran, Islamic Republic of)

2016-04-15

Recently, Variable gutter technique has been introduced as a novel method in order to reduce waste materials in closed-die forging processes. In this paper, the capability of this method is investigated for a family of forged parts that the vertical flow of material is the last stage of forming process. As a case study, using the variable gutter technique, the amount of waste material is decreased about 50% for a sample forged part with a local rising. The results of FVM simulations and experiments confirmed the effectiveness of the variable gutter technique in such forging processes. The vertical flow of material in the die cavity (h parameter), as a criterion, for different gutter width and thickness dimensions was examined by FVM simulations. The results shown that the gutter thickness has more effect on vertical flow of material than the gutter width. By decreasing the gutter thickness and increasing the gutter width, the amount of vertical flow of material is increased about 120% and 29%, respectively. Finally, A/H ratio (A = Max width of sectional area of a forged part, H = Max height of a forged part) is proposed as shape complexity factor of a forged part. The results of FVM simulations are indicated that for the ratio of A/H > 2, the variable gutter thickness technique is more effective and can be successfully used to reduce the amount of waste materials.

Evaluating Residence Time for Cesium Removal from Simulated Hanford Tank Wastes Using SuperLig(R) 644 Resin

International Nuclear Information System (INIS)

Hassan, N.M.

2003-01-01

Batch contact and column experiments were performed to evaluate the effect of residence time on cesium (Cs) removal from two simulated Hanford tank wastes using SuperLig(R) 644 resin. The two waste simulants mimic the compositions of tanks 241-AZ-102 and 241-AN-107 at the U.S. Department of Energy (DOE) Hanford site. A single column made of glass tube (2.7-cm i.d.), which contained approximately 100 mL of H-form SuperLig(R) 644 resin was used in the column experiments. The experiments each consisted of loading, elution, and regeneration steps were performed at flow rates ranging from 0.64 to 8.2 BV/h for AZ-102 and from 1.5 to 18 BV/h for AN-107 simulant. The lowest flow rates of 0.64 and 1.5 BV/h were selected to evaluate less than optimal flow conditions in the plant. The range of the flow rates is consistent with the River Protection Project design for the waste treatment plant (WTP) columns, which will operate at a flow rate between 1.5 to 3 BV/h. Batch contact experiments were also performed for two batches of SuperLig(R) 644 to determine the equilibrium distribution coefficients (Kds) as a function of Cs concentration

Studies on disposal of low-level radioactive wastes in Turkey

International Nuclear Information System (INIS)

Uslu, I.; Fields, D.E.; Yalcintas, M.G.

1989-08-01

The Turkish Government is in the process of planning two nuclear reactors in Turkey. Studies have begun for improved control of low level wastes (LLW) in Turkey before establishment of these reactors. In this study, the PRESTO-II (Prediction of Radiation Exposures form Shallow Trench Operations) computer code is used to assess the risk associated with the shallow land disposal of low level waste (LLW) in various sites in Turkey. PRESTO-II is a computer code developed under the United States Environmental Protection Agency, Department of Energy and Nuclear Regulatory Commission funding to evaluate possible health effects from radioactive releases from shallow, radioactive waste disposal trenches and from areas contaminated with operational spillage. A preliminary simulation using the PRESTO-II computer code has been run for the site in Koteyli, Balikesir, Turkey. This example simulation was performed using the same radionuclide data set believed representative of the LLW disposal facility in Barnwell, South Carolina. Site environmental variables were selected to typify credible worst case exposure scenarios. Radionuclide inventories are primarily based on estimated waste composition rather than measured values. 9 refs., 4 figs., 1 tab

Models for recurrent gas release event behavior in hazardous waste tanks

International Nuclear Information System (INIS)

Anderson, D.N.; Arnold, B.C.

1994-08-01

Certain radioactive waste storage tanks at the United States Department of Energy Hanford facilities continuously generate gases as a result of radiolysis and chemical reactions. The congealed sludge in these tanks traps the gases and causes the level of the waste within the tanks to rise. The waste level continues to rise until the sludge becomes buoyant and ''rolls over'', changing places with heavier fluid on top. During a rollover, the trapped gases are released, resulting, in a sudden drop in the waste level. This is known as a gas release event (GRE). After a GRE, the wastes leading to another GRE. We present nonlinear time waste re-congeals and gas again accumulates leading to another GRE. We present nonlinear time series models that produce simulated sample paths that closely resemble the temporal history of waste levels in these tanks. The models also imitate the random GRE, behavior observed in the temporal waste level history of a storage tank. We are interested in using the structure of these models to understand the probabilistic behavior of the random variable ''time between consecutive GRE's''. Understanding the stochastic nature of this random variable is important because the hydrogen and nitrous oxide gases released from a GRE, are flammable and the ammonia that is released is a health risk. From a safety perspective, activity around such waste tanks should be halted when a GRE is imminent. With credible GRE models, we can establish time windows in which waste tank research and maintenance activities can be safely performed

Internal and external variability in regional simulations of the Iberian Peninsula climate over the last millennium

Directory of Open Access Journals (Sweden)

J. J. Gómez-Navarro

2012-01-01

Full Text Available In this study we analyse the role of internal variability in regional climate simulations through a comparison of two regional paleoclimate simulations for the last millennium. They share the same external forcings and model configuration, differing only in the initial condition used to run the driving global model simulation. A comparison of these simulations allows us to study the role of internal variability in climate models at regional scales, and how it affects the long-term evolution of climate variables such as temperature and precipitation. The results indicate that, although temperature is homogeneously sensitive to the effect of external forcings, the evolution of precipitation is more strongly governed by random unpredictable internal dynamics. There are, however, some areas where the role of internal variability is lower than expected, allowing precipitation to respond to the external forcings. In this respect, we explore the underlying physical mechanisms responsible for it. This study identifies areas, depending on the season, in which a direct comparison between model simulations of precipitation and climate reconstructions would be meaningful, but also other areas where good agreement between them should not be expected even if both are perfect.

MASBAL: A computer program for predicting the composition of nuclear waste glass produced by a slurry-fed ceramic melter

International Nuclear Information System (INIS)

Reimus, P.W.

1987-07-01

This report is a user's manual for the MASBAL computer program. MASBAL's objectives are to predict the composition of nuclear waste glass produced by a slurry-fed ceramic melter based on a knowledge of process conditions; to generate simulated data that can be used to estimate the uncertainty in the predicted glass composition as a function of process uncertainties; and to generate simulated data that can be used to provide a measure of the inherent variability in the glass composition as a function of the inherent variability in the feed composition. These three capabilities are important to nuclear waste glass producers because there are constraints on the range of compositions that can be processed in a ceramic melter and on the range of compositions that will be acceptable for disposal in a geologic repository. MASBAL was developed specifically to simulate the operation of the West Valley Component Test system, a commercial-scale ceramic melter system that will process high-level nuclear wastes currently stored in underground tanks at the site of the Western New York Nuclear Services Center (near West Valley, New York). The program is flexible enough, however, to simulate any slurry-fed ceramic melter system. 4 refs., 16 figs., 5 tabs

Analysis by simulation of the disposition of nuclear-fuel waste

International Nuclear Information System (INIS)

Turek, J.L.

1980-09-01

To achieve the non-proliferation objectives of the United States, the reprocessing of spent nuclear fuel was discontinued in 1977. Since current at-reactor storage capacity is based upon a nuclear fuel cycle which includes reprocessing, this halt in reprocessing is causing large quantities of non-storable spent fuel. Permanent nuclear waste storage repositories will not be available until the end of the century. Present Department of Energy policy calls for sufficient interim Away-From-Reactor (AFR) Storage capacity to insure that no commercial reactor has to shutdown due to inadequate storage space for discharged spent fuel. A descriptive simulation model is developed which includes all aspects of nuclear waste disposition. The model is comprised of two systems, the second system orchestrated by GASP IV. A spent fuel generation prediction module is interfaced with the AFR Program Management Information System and a repository scheduling information module. The user is permitted a wide range of options with which to tailor the simulation to any desired storage scenario. The model projects storage requirements through the year 2020. The outputs are evaluations of the impact that alternative decision policies and milestone date changes have on the demand for, the availability of, and the utilization of spent fuel storage capacities. Both graphs and detailed listings are available. These outputs give a comprehensive view of the particular scenario under observation, including the tracking, by year, of each discharge from every reactor. Included within the work is a review of the status of spent fuel disposition based on input data accurate as of August 1980

Simulation Analysis and Experiment of Variable-Displacement Asymmetric Axial Piston Pump

Directory of Open Access Journals (Sweden)

Youshan Gao

2017-03-01

Full Text Available The variable displacement pump control system has greater energy-saving advantages and application prospects than the valve control system. However, the variable displacement pump control of differential cylinder is not concurrent with the existing technologies. The asymmetric pump-controlled cylinder is, therefore, used to balance the unequal volume flow through a single rod cylinder in closed-circuit system. This is considered to be an effective method. Nevertheless, the asymmetric axial piston pump (AAPP is a constant displacement pump. In this study, variable-displacement asymmetric axial piston pump (VAPP is investigated according to the same principle used in investigating AAPP. This study, therefore, aims at investigating the characteristics of VAPP. The variable-displacement output of VAPP is implemented by controlling the swash plate angle with angle feedback control circuit, which is composed of a servo proportional valve and an angular displacement sensor. The angular displacement sensor is connected to the swash plate. The simulation model of VAPP, which is set up through the ITI-SimulationX simulation platform, is used to predict VAPP’s characteristics. The purpose of implementing the experiment is to verify the theoretical results. Both the simulation and the experiment results demonstrated that the swash plate angle is controlled by a variable mechanism; when the swash plate angle increases, the flow of Port B and Port T increases while the response speed of Port B and Port T also accelerates. When the swash plate angle is constant, the flow of Port B and Port T increases along with the increase of pump speed, although the pressure-response speed of Port B is faster than that of Port T. Consequently, the flow pulsation of Port B and Port T tends to decrease gradually along with the increase of pump speed. When the pressure loaded on Port B equals to that of Port T, the flow ripple cycle of Port B is longer than that of Port T

Output variability caused by random seeds in a multi-agent transport simulation model

DEFF Research Database (Denmark)

Paulsen, Mads; Rasmussen, Thomas Kjær; Nielsen, Otto Anker

2018-01-01

Dynamic transport simulators are intended to support decision makers in transport-related issues, and as such it is valuable that the random variability of their outputs is as small as possible. In this study we analyse the output variability caused by random seeds of a multi-agent transport...... simulator (MATSim) when applied to a case study of Santiago de Chile. Results based on 100 different random seeds shows that the relative accuracies of estimated link loads tend to increase with link load, but that relative errors of up to 10 % do occur even for links with large volumes. Although...

MIIT: International in-situ testing of nuclear waste glasses-performance of SRS simulated waste glass after 5 years of burial at the waste isolation pilot plant (WIPP)

International Nuclear Information System (INIS)

Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Clark, D.E.

1993-01-01

In July of 1986, the first in-situ test involving burial of simulated high-level waste [HLW] forms conducted in the United States was started. This program, called the Materials Interface Interactions Test or MIIT, comprises the largest, most cooperative field-testing venture in the international waste management community. Included in the study are over 900 waste form samples comprising 15 different systems supplied by seven nations. Also included are about 300 potential canister or overpack metal samples along with more than 500 geologic and backfill specimens. There are almost 2000 relevant interactions that characterize this effort which has been conducted in the bedded salt site at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. The MIIT program represents a joint effort managed by the Savannah River Technology Center (SRTC) in Aiken, S.C., and Sandia National Laboratories (SNL) in Albuquerque, N.M.. and sponsored by the US Department of Energy. Involved in MIIT are participants from national and federal laboratories, universities, and representatives from laboratories in France, Germany, Canada, Belgium, Japan, Sweden, the United Kingdom, and the United States. In July of 1991, the experimental portion of the 5-year MIIT study was completed on schedule. During this time interval, many in-situ measurements were performed, thousands of brine analyses conducted, and hundreds of waste glass and package components exhumed and evaluated after 6 mo., 1 yr., 2 yr. and 5 yr. burial periods. Although analyses are still in progress, the performance of SRS waste glass based on all data currently available has been seen to be excellent thus far. Initial analyses and assessment of Savannah River (SR) waste glass after burial in WIPP at 90 degrees C for 5 years is presented

Laboratory-Scale SuperLig 639 Column Tests With Hanford Waste Simulants

International Nuclear Information System (INIS)

King, William D.; Spencer, William A.; Bussey, Myra Pettis

2003-01-01

This report describes the results of SuperLig 639 column tests conducted at the Savannah River Technology Center (SRTC) in support of the Hanford River Protection Project - Waste Treatment Plant (RPP-WTP). The RPP-WTP contract was awarded to Bechtel National Inc. (BNI) for the design, construction, and initial operation of a plant for the treatment and vitrification of millions of gallons of radioactive waste currently stored in tanks at Hanford, WA. Part of the current treatment process involves the removal of technetium from tank supernate solutions using columns containing SuperLig 639 resin. This report is part of a body of work intended to quantify and optimize the operation of the technetium removal columns with regard to various parameters (such as liquid flow rate, column aspect ratio, resin particle size, loading and elution temperature, etc.). The tests were conducted using nonradioactive simulants of the actual tank waste samples containing rhenium as a surrogate for the technetium in the actual waste. A previous report focused on the impacts of liquid flow rate and column aspect ratio upon performance. More recent studies have focused on the impacts of resin particle size, solution composition, and temperature. This report describes column loading experiments conducted varying temperature and solution composition. Each loading experiment was followed by high temperature elution of the sorbed rhenium. Results from limited testing are also described which were intended to evaluate the physical stability of SuperLig 639 resin during exposure to repeated temperature cycles covering the range of potential processing extremes

Methane production from food waste leachate in laboratory-scale simulated landfill.

Science.gov (United States)

Behera, Shishir Kumar; Park, Jun Mo; Kim, Kyeong Ho; Park, Hung-Suck

2010-01-01

Due to the prohibition of food waste landfilling in Korea from 2005 and the subsequent ban on the marine disposal of organic sludge, including leachate generated from food waste recycling facilities from 2012, it is urgent to develop an innovative and sustainable disposal strategy that is eco-friendly, yet economically beneficial. In this study, methane production from food waste leachate (FWL) in landfill sites with landfill gas recovery facilities was evaluated in simulated landfill reactors (lysimeters) for a period of 90 d with four different inoculum-substrate ratios (ISRs) on volatile solid (VS) basis. Simultaneous biochemical methane potential batch experiments were also conducted at the same ISRs for 30 d to compare CH(4) yield obtained from lysimeter studies. Under the experimental conditions, a maximum CH(4) yield of 0.272 and 0.294 L/g VS was obtained in the batch and lysimeter studies, respectively, at ISR of 1:1. The biodegradability of FWL in batch and lysimeter experiments at ISR of 1:1 was 64% and 69%, respectively. The calculated data using the modified Gompertz equation for the cumulative CH(4) production showed good agreement with the experimental result obtained from lysimeter study. Based on the results obtained from this study, field-scale pilot test is required to re-evaluate the existing sanitary landfills with efficient leachate collection and gas recovery facilities as engineered bioreactors to treat non-hazardous liquid organic wastes for energy recovery with optimum utilization of facilities. 2010 Elsevier Ltd. All rights reserved.

Radiolytic gas generation from cement-based waste hosts for DOE low-level radioactive wastes

International Nuclear Information System (INIS)

Dole, L.R.; Friedman, H.A.

1986-01-01

Using cement-based immobilization binders with simulated radioactive waste containing sulfate, nitrate, nitrite, phosphate, and fluoride anions, the gamma- and alpha-radiolytic gas generation factors (G/sub t/, molecules/100 eV) and gas compositions were measured on specimens of cured grouts. These tests studied the effects of; (1) waste composition; (2) the sample surface-to-volume ratio; (3) the waste slurry particle size; and (4) the water content of the waste host formula. The radiolysis test vessels were designed to minimize the ''dead'' volume and to simulate the configuration of waste packages

Preparation and Characterization of Chemical Plugs Based on Selected Hanford Waste Simulants

International Nuclear Information System (INIS)

Mattigod, Shas V.; Wellman, Dawn M.; Parker, Kent E.; Cordova, Elsa A.; Gunderson, Katie M.; Baum, Steven R.; Crum, Jarrod V.; Poloski, Adam P.

2008-01-01

This report presents the results of preparation and characterization of chemical plugs based on selected Hanford Site waste simulants. Included are the results of chemical plug bench testing conducted in support of the M1/M6 Flow Loop Chemical Plugging/Unplugging Test (TP-RPP-WTP-495 Rev A). These results support the proposed plug simulants for the chemical plugging/ unplugging tests. Based on the available simulant data, a set of simulants was identified that would likely result in chemical plugs. The three types of chemical plugs that were generated and tested in this task consisted of: 1. Aluminum hydroxide (NAH), 2. Sodium aluminosilicate (NAS), and 3. Sodium aluminum phosphate (NAP). While both solvents, namely 2 molar (2 M) nitric acid (HNO3) and 2 M sodium hydroxide (NaOH) at 60 C, used in these tests were effective in dissolving the chemical plugs, the 2 M nitric acid was significantly more effective in dissolving the NAH and NAS plugs. The caustic was only slightly more effecting at dissolving the NAP plug. In the bench-scale dissolution tests, hot (60 C) 2 M nitric acid was the most effective solvent in that it completely dissolved both NAH and NAS chemical plugs much faster (1.5 - 2 x) than 2 M sodium hydroxide. So unless there are operational benefits for the use of caustic verses nitric acid, 2 M nitric acid heated to 60 C should be the solvent of choice for dissolving these chemical plugs. Flow-loop testing was planned to identify a combination of parameters such as pressure, flush solution, composition, and temperature that would effectively dissolve and flush each type of chemical plug from preformed chemical plugs in 3-inch-diameter and 4-feet-long pipe sections. However, based on a review of the results of the bench-top tests and technical discussions, the Waste Treatment Plant (WTP) Research and Technology (R and T), Engineering and Mechanical Systems (EMS), and Operations concluded that flow-loop testing of the chemically plugged pipe sections

Impact of Variable-Resolution Meshes on Regional Climate Simulations

Science.gov (United States)

Fowler, L. D.; Skamarock, W. C.; Bruyere, C. L.

2014-12-01

The Model for Prediction Across Scales (MPAS) is currently being used for seasonal-scale simulations on globally-uniform and regionally-refined meshes. Our ongoing research aims at analyzing simulations of tropical convective activity and tropical cyclone development during one hurricane season over the North Atlantic Ocean, contrasting statistics obtained with a variable-resolution mesh against those obtained with a quasi-uniform mesh. Analyses focus on the spatial distribution, frequency, and intensity of convective and grid-scale precipitations, and their relative contributions to the total precipitation as a function of the horizontal scale. Multi-month simulations initialized on May 1st 2005 using ERA-Interim re-analyses indicate that MPAS performs satisfactorily as a regional climate model for different combinations of horizontal resolutions and transitions between the coarse and refined meshes. Results highlight seamless transitions for convection, cloud microphysics, radiation, and land-surface processes between the quasi-uniform and locally- refined meshes, despite the fact that the physics parameterizations were not developed for variable resolution meshes. Our goal of analyzing the performance of MPAS is twofold. First, we want to establish that MPAS can be successfully used as a regional climate model, bypassing the need for nesting and nudging techniques at the edges of the computational domain as done in traditional regional climate modeling. Second, we want to assess the performance of our convective and cloud microphysics parameterizations as the horizontal resolution varies between the lower-resolution quasi-uniform and higher-resolution locally-refined areas of the global domain.

Numerical simulation on stir system of jet ballast in high level liquid waste storage tank

International Nuclear Information System (INIS)

Lu Yingchun

2012-01-01

The stir system of jet ballast in high level liquid waste storage tank was simulation object. Gas, liquid and solid were air, sodium nitrate liquor and titanium whitening, respectively. The mathematic model based on three-fluid model and the kinetic theory of particles was established for the stir system of jet ballast in high level liquid waste storage tank. The CFD commercial software was used for solving this model. The detail flow parameters as three phase velocity, pressure and phase loadings were gained. The calculated results agree with the experimental results, so they can well define the flow behavior in the tank. And this offers a basic method for the scale-up and optimization design of the stir system of jet ballast in high level liquid waste storage tank. (author)

Summer U.S. Surface Air Temperature Variability: Controlling Factors and AMIP Simulation Biases

Science.gov (United States)

Merrifield, A.; Xie, S. P.

2016-02-01

This study documents and investigates biases in simulating summer surface air temperature (SAT) variability over the continental U.S. in the Coupled Model Intercomparison Project (CMIP5) Atmospheric Model Intercomparison Project (AMIP). Empirical orthogonal function (EOF) and multivariate regression analyses are used to assess the relative importance of circulation and the land surface feedback at setting summer SAT over a 30-year period (1979-2008). In observations, regions of high SAT variability are closely associated with midtropospheric highs and subsidence, consistent with adiabatic theory (Meehl and Tebaldi 2004, Lau and Nath 2012). Preliminary analysis shows the majority of the AMIP models feature high SAT variability over the central U.S., displaced south and/or west of observed centers of action (COAs). SAT COAs in models tend to be concomitant with regions of high sensible heat flux variability, suggesting an excessive land surface feedback in these models modulate U.S. summer SAT. Additionally, tropical sea surface temperatures (SSTs) play a role in forcing the leading EOF mode for summer SAT, in concert with internal atmospheric variability. There is evidence that models respond to different SST patterns than observed. Addressing issues with the bulk land surface feedback and the SST-forced component of atmospheric variability may be key to improving model skill in simulating summer SAT variability over the U.S.

Shift of biome patterns due to simulated climate variability and climate change

International Nuclear Information System (INIS)

Claussen, M.

1993-01-01

The variability of simulated equilibrium-response patterns of biomes caused by simulated climate variability and climate shift is analysed. This investigation is based on various realisations of simulated present-day climate and climate shift. It has been found that the difference between biomes computed from three 10-year climatologies and from the corresponding 30-year climatology, simulated by the Hamburg climate model at T21 resolution, amounts to approximately 6% of the total land area, Antarctica excluded. This difference is mainly due to differences in annual moisture availability and winter temperatures. When intercomparing biomes from the 10-year climatologies a 10% difference is seen, but there is no unique difference pattern. In contrast to the interdecadal variability, the shift of conditions favorable for biomes due to a shift in climate in the next 100 years, caused by an increase in sea-surface temperatures and atmospheric CO 2 , reveals a unique trend pattern. It turns out that the strongest and most significant signal is the north-east shift of conditions for boreal biomes. This signal is caused by an increase of annual temperature sums as well as mean temperatures of the coldest and warmest months. Trends in annual moisture availability are of secondary importance globally. Regionally, a decrease in water availability affects biomes in Central and East Europe and an increase of water availability leads to a potential increase in tropical rain forest. In total, all differences amount to roughly 30% of the total land surface, Antarctica excluded. (orig./KW)

Noble metal behavior during melting of simulated high-level nuclear waste glass feeds

International Nuclear Information System (INIS)

Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

1994-01-01

Noble metals and their oxides can settle in waste glass melters and cause electrical shorting. Simulate waste feeds from Hanford, Savannah River, and Kernforschungszentrum Karlsruhe were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C to 1000 degrees C and examined by electron microscopy to determine shapes, sizes, and distribution of noble metal particles as a function of temperature. Individual noble metal particles and agglomerates of rhodium (Rh), ruthenium (RuO 2 ), and palladium (Pd), as well as their alloys, were seen. The majority of particles and agglomerates were generally less than 10 μm; however, large agglomerations (up to 1 mm) were found in the German feed. 5 refs., 6 figs., 2 tabs

Factors affecting the leachability of caesium and strontium from cemented simulant evaporator wastes

International Nuclear Information System (INIS)

Lee, D.J.; Brown, D.J.

1981-08-01

Leach rates of stable cesium and strontium from a range of simulated evaporator waste/cement formulations have been determined. Important factors in plant operation are assessed for their effect on leach rates. Increasing the curing time and lowering the water/cement ratio has been shown to reduce leach rates by up to a factor of four. Incorporation of additives such as clays and supplementary cementatious materials can reduce leach rates by up to three orders magnitude, and coating the surface of the waste form with a neat cement grout can reduce the cesium leach rate by up to four orders of magnitude. The effects of permeability of the matrix and its cesium absorption capacity on the leach rates have been analysed qualitatively. (U.K.)

Pore solution chemistry of simulated low-level liquid waste incorporated in cement grouts

International Nuclear Information System (INIS)

Kruger, A.A.

1995-12-01

Expressed pore solutions from simulated low level liquid waste cement grouts cured at room temperature, 50 degree C and 90 degree C for various duration were analyzed by standard chemical methods and ion chromatography. The solid portions of the grouts were formulated with portland cement, fly ash, slag, and attapulgite clay in the ratios of 3:3:3:1. Two different solutions simulating off-gas condensates expected from vitrification of Hanford low level tank wastes were made. One is highly alkaline and contains the species Na + , P0 4 3- , N0 2 - , NO 3 - and OH - . The other is carbonated and contains the species, Na + , PO 4 3- , NO 2 - , NO 3 - , and CO 3 2- . In both cases phosphate rapidly disappeared from the pore solution, leaving behind sodium in the form of hydroxide. The carbonates were also removed from the pore solution to form calcium carbonate and possibly calcium monocarboaluminate. These reactions resulted in the increase of hydroxide ion concentration in the early period. Subsequently there was a significant reduction OH - and Na + ion concentrations. In contrast high concentration of N0 2 - and N0 3 - were retained in the pore solution indefinitely

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.

Thermal and Physical Property Determinations for Ionsiv IE-911 Crystalline Silicotitanate and Savannah River Site Waste Simulant Solutions

International Nuclear Information System (INIS)

Bostick, D.T.; Steele, W.V.

1999-01-01

This document describes physical and thermophysical property determinations that were made in order to resolve questions associated with the decontamination of Savannah River Site (SRS) waste streams using ion exchange on crystalline silicotitanate (CST). The research will aid in the understanding of potential issues associated with cooling of feed streams within SRS waste treatment processes. Toward this end, the thermophysical properties of engineered CST, manufactured under the trade name, Ionsivereg s ign IE-911 by UOP, Mobile, AL, were determined. The heating profiles of CST samples from several manufacturers' production runs were observed using differential scanning calorimetric (DSC) measurements. DSC data were obtained over the region of 10 to 215 C to check for the possibility of a phase transition or any other enthalpic event in that temperature region. Finally, the heat capacity, thermal conductivity, density, viscosity, and salting-out point were determined for SRS waste simulants designated as Average, High NO 3 - and High OH - simulants

Development of Advanced Electrochemical Emission Spectroscopy for Monitoring Corrosion in Simulated DOE Liquid Waste

Energy Technology Data Exchange (ETDEWEB)

Digby D. Macdonald; Brian M. Marx; Sejin Ahn; Julio de Ruiz; Balaji Soundararaja; Morgan Smith; and Wendy Coulson

2008-01-15

Various forms of general and localized corrosion represent principal threats to the integrity of DOE liquid waste storage tanks. These tanks, which are of a single wall or double wall design, depending upon their age, are fabricated from welded carbon steel and contain a complex waste-form comprised of NaOH and NaNO{sub 3}, along with trace amounts of phosphate, sulfate, carbonate, and chloride. Because waste leakage can have a profound environmental impact, considerable interest exists in predicting the accumulation of corrosion damage, so as to more effectively schedule maintenance and repair. The different tasks that are being carried out under the current program are as follows: (1) Theoretical and experimental assessment of general corrosion of iron/steel in borate buffer solutions by using electrochemical impedance spectroscopy (EIS), ellipsometry and XPS techniques; (2) Development of a damage function analysis (DFA) which would help in predicting the accumulation of damage due to pitting corrosion in an environment prototypical of DOE liquid waste systems; (3) Experimental measurement of crack growth rate, acoustic emission signals and coupling currents for fracture in carbon and low alloy steels as functions of mechanical (stress intensity), chemical (conductivity), electrochemical (corrosion potential, ECP), and microstructural (grain size, precipitate size, etc) variables in a systematic manner, with particular attention being focused on the structure of the noise in the current and its correlation with the acoustic emissions; (4) Development of fracture mechanisms for carbon and low alloy steels that are consistent with the crack growth rate, coupling current data and acoustic emissions; (5) Inserting advanced crack growth rate models for SCC into existing deterministic codes for predicting the evolution of corrosion damage in DOE liquid waste storage tanks; (6) Computer simulation of the anodic and cathodic activity on the surface of the steel samples

Thermodynamic simulation of ammonia-water absorption refrigeration system

Directory of Open Access Journals (Sweden)

Sathyabhama A.

2008-01-01

Full Text Available The ammonia-water absorption refrigeration system is attracting increasing research interests, since the system can be powered by waste thermal energy, thus reducing demand on electricity supply. The development of this technology demands reliable and effective system simulations. In this work, a thermodynamic simulation of the cycle is carried out to investigate the effects of different operating variables on the performance of the cycle. A computer program in C language is written for the performance analysis of the cycle.

A hydrothermal flow-through apparatus to simulate leaching of nuclear waste forms under quasi-dynamic conditions

International Nuclear Information System (INIS)

Heimann, R.B.

1985-03-01

A hydrothermal flow-through apparatus has been designed that will allow the testing of individual waste package components, as well as combinations of these, under a wide range of environmental conditions. The maximum permissible temperature is 700 degrees C, while the maximum pressure is 300 MPa. Flow rates can be adjusted by sequential operation of a pneumatically operated valve with preset pause and working cycles. The main applications of the apparatus to nuclear fuel waste management research are: (i) the study of migration of ionic species through a rock column at specified hydraulic head, and (ii) the study of the rate of leaching of radionuclides from waste forms under disposal vault conditions in the presence of groundwater with variable flow rates

The effect of the waste separation policy in municipal solid waste management using the system dynamic approach

Directory of Open Access Journals (Sweden)

Ahmad Jamshidi Zanjani

2012-01-01

Full Text Available Aims: In the present study, Vensim was used to simulate waste management system of Tehran, the capital of Iran, with the system dynamic approach. Materials and Methods: The environmental system dynamic modeling is one of the comprehensive simulation tools capable of simulating and analyzing complex systems. In this approach, the model is developed based on the existing realities and userâ€′comments. User participation to develop the model could increase the reliability of the results. Results: The simulation results revealed good conformity with the statistical data. Waste production prediction in the model with real data was more than 95%. Moreover, the effect of applying an encouraging policy for people to separate their waste was considered. The result indicated that applying a new policy, and the economic benefit through this policy would prevent getting a loan from the government after 20 years. Conclusions: It could be concluded that public participation in waste separation was an effective policy to help in the financial independence of the municipality in terms of urban waste management. Moreover, conformity between the simulation results and real data revealed an appropriate capability of the simulated model to predict Tehran waste generation.

Evaluation of the transport and resuspension of a simulated nuclear waste slurry: Nuclear Waste Treatment Program

International Nuclear Information System (INIS)

Carleson, T.E.; Drown, D.C.; Hart, R.E.; Peterson, M.E.

1987-09-01

The Department of Chemical Engineering at the University of Idaho conducted research on the transport and resuspension of a simulated high-level nuclear waste slurry. In the United States, the reference process for treating both defense and civilian HLLW is vitrification using the liquid-fed ceramic melter process. The non-Newtonian behavior of the slurry complicates the evaluation of the transport and resuspension characteristics of the slurry. The resuspension of a simulated (nonradioactive) melter feed slurry was evaluated using a slurry designated as WV-205. The simulated slurry was developed for the West Valley Demonstration Project and was used during a pilot-scale ceramic melter (PSCM) experiment conducted at PNL in July 1985 (PSCM-21). This study involved determining the transport characteristics of a fully suspended slurry and the resuspension characteristics of settled solids in a pilot-scale pipe loop. The goal was to predict the transport and resuspension of a full-scale system based on rheological data for a specific slurry. The rheological behavior of the slurry was evaluated using a concentric cylinder rotational viscometer, a capillary tube viscometer, and the pilot-scale pipe loop. The results obtained from the three approaches were compared. 40 refs., 74 figs., 15 tabs

Integration of a driving simulator and a traffic simulator case study: Exploring drivers' behavior in response to variable message signs

Directory of Open Access Journals (Sweden)

Mansoureh Jeihani

2017-12-01

Full Text Available For the first time, a driving simulator has been integrated with a traffic simulator at the network level to allow subjects to drive in a fairly realistic environment with a realistic traffic flow and density. A 10 mi2 (25 km2 network was developed in a driving simulator and then exported to a traffic simulator. About 30 subjects drove the simulator under different traffic and driving conditions and variable message sign (VMS information, both with and without integration. Route guidance was available for the subjects. The challenges of the integration process are explained and its advantages investigated. The study concluded that traffic density, VMS reliability and compliance behavior are higher when driving and traffic simulators are integrated. To find factors affecting route diversion, researchers applied a binary logistic regression model. The results indicated that the original chosen route, displayed VMS information, subjects' attitude toward VMS information helpfulness, and their level of exposure to VMS affect route diversion. In addition, a multinomial logistic regression model was employed to investigate important factors in route choice. The results revealed that there is a significant correlation with driver route choice behavior and their actual travel time, the need for GPS, VMS exposure and also the designed scenarios. It should be noted that the paper was peer-reviewed by TRB and presented at the TRB Annual Meeting, Washington, D.C., January 2016. Keywords: Integration, Variable message sign, Compliance behavior, Driving simulator, Traffic simulator, Discrete choice analysis

Simulating variable-density flows with time-consistent integration of Navier-Stokes equations

Science.gov (United States)

Lu, Xiaoyi; Pantano, Carlos

2017-11-01

In this talk, we present several features of a high-order semi-implicit variable-density low-Mach Navier-Stokes solver. A new formulation to solve pressure Poisson-like equation of variable-density flows is highlighted. With this formulation of the numerical method, we are able to solve all variables with a uniform order of accuracy in time (consistent with the time integrator being used). The solver is primarily designed to perform direct numerical simulations for turbulent premixed flames. Therefore, we also address other important elements, such as energy-stable boundary conditions, synthetic turbulence generation, and flame anchoring method. Numerical examples include classical non-reacting constant/variable-density flows, as well as turbulent premixed flames.

The use of the virtual reality Helmet Samsung gear VR as interaction interface of a radioactive waste repository simulator

Energy Technology Data Exchange (ETDEWEB)

Santos, Julio A. dos; Mól, Antônio C. de A.; Santo, André C. Do E., E-mail: julio_andrade11@hotmail.com [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Centro Universitário Carioca (UniCarioca), Rio de Janeiro, RJ (Brazil)

2017-07-01

Radioactive waste is all material resulting from human activity that contains elements that emit radiation that can generate risks to health and the environment. In this sense, they are very toxic also for those who perform the storage of radioactive waste in nuclear facilities. On the other hand, the virtual reality (VR) has been destined to the most diverse purposes, like simulations for educational systems, for military purposes as for diverse training. VR can be considered as the junction of three basic principles: immersion, interaction and involvement. Bases on these principles of VR, this work aimed to develop a simulator of a repository of nuclear tailings, for mobile computing, whose interaction interface will be through the Samsung Gear VR helmet. The simulator of the nuclear waste repository was developed in the unity 3D tool and the elements that make up the scenario in the 3D MAX program. In this work we tried to put virtual reality under scrutiny in conjunction with Gear VR, to help in the sensation of immersion, as well as, the possibility of interaction with joysticks. The purpose was to provide greater insight into the operating environment. (author)

The use of the virtual reality Helmet Samsung gear VR as interaction interface of a radioactive waste repository simulator

International Nuclear Information System (INIS)

Santos, Julio A. dos; Mól, Antônio C. de A.; Santo, André C. Do E.

2017-01-01

Radioactive waste is all material resulting from human activity that contains elements that emit radiation that can generate risks to health and the environment. In this sense, they are very toxic also for those who perform the storage of radioactive waste in nuclear facilities. On the other hand, the virtual reality (VR) has been destined to the most diverse purposes, like simulations for educational systems, for military purposes as for diverse training. VR can be considered as the junction of three basic principles: immersion, interaction and involvement. Bases on these principles of VR, this work aimed to develop a simulator of a repository of nuclear tailings, for mobile computing, whose interaction interface will be through the Samsung Gear VR helmet. The simulator of the nuclear waste repository was developed in the unity 3D tool and the elements that make up the scenario in the 3D MAX program. In this work we tried to put virtual reality under scrutiny in conjunction with Gear VR, to help in the sensation of immersion, as well as, the possibility of interaction with joysticks. The purpose was to provide greater insight into the operating environment. (author)

Heat transfer in vitrified radioactive waste

International Nuclear Information System (INIS)

Palancar, M.C.; Luis, M.A.; Luis, P.; Aragon, J.M.; Montero, M.A.

1987-01-01

An experimental method for measuring the thermal conductivity and convection coefficient of borosilicate glass cylinders, containing a simulated high level radioactive waste, is described. A simulation of the thermal behaviour of matrices of solidified waste during the cooling in air, water and a geological repository has been done. The experimental values of the thermal conductivity are ranging from 0.267 to 0.591 w/m K, for matrices with simulated waste contents of 10 to 40% (the waste is simulated by no radioactive isotopes). The convection coefficient for air/cylinders under the operating conditions used is 116 w/m 2 K. The simulated operation of cooling in air shows that about 1-2 days are enough to cool a solidified waste cylinder 0.6m diameter from 900 to 400 0 C. The cooling under water from 400 to near 80 0 C is faster than in air, but sharp temperature gradients within the matrices could be expected. The simulation of geological repositories lead to some criteria of arranging the matrices for avoiding undesirable high temperature points. (author) 1 fig

Soil erodibility variability in laboratory and field rainfall simulations

Science.gov (United States)

Szabó, Boglárka; Szabó, Judit; Jakab, Gergely; Centeri, Csaba; Szalai, Zoltán

2017-04-01

Rainfall simulation experiments are the most common way to observe and to model the soil erosion processes in in situ and ex situ circumstances. During modelling soil erosion, one of the most important factors are the annual soil loss and the soil erodibility which represent the effect of soil properties on soil loss and the soil resistance against water erosion. The amount of runoff and soil loss can differ in case of the same soil type, while it's characteristics determine the soil erodibility factor. This leads to uncertainties regarding soil erodibility. Soil loss and soil erodibility were examined with the investigation of the same soil under laboratory and field conditions with rainfall simulators. The comparative measurement was carried out in a laboratory on 0,5 m2, and in the field (Shower Power-02) on 6 m2 plot size where the applied slope angles were 5% and 12% with 30 and 90 mm/h rainfall intensity. The main idea was to examine and compare the soil erodibility and its variability coming from the same soil, but different rainfall simulator type. The applied model was the USLE, nomograph and other equations which concern single rainfall events. The given results show differences between the field and laboratory experiments and between the different calculations. Concerning for the whole rainfall events runoff and soil loss, were significantly higher at the laboratory experiments, which affected the soil erodibility values too. The given differences can originate from the plot size. The main research questions are that: How should we handle the soil erodibility factors and its significant variability? What is the best solution for soil erodibility determination?

Waste Feed Evaporation Physical Properties Modeling

International Nuclear Information System (INIS)

Daniel, W.E.

2003-01-01

This document describes the waste feed evaporator modeling work done in the Waste Feed Evaporation and Physical Properties Modeling test specification and in support of the Hanford River Protection Project (RPP) Waste Treatment Plant (WTP) project. A private database (ZEOLITE) was developed and used in this work in order to include the behavior of aluminosilicates such a NAS-gel in the OLI/ESP simulations, in addition to the development of the mathematical models. Mathematical models were developed that describe certain physical properties in the Hanford RPP-WTP waste feed evaporator process (FEP). In particular, models were developed for the feed stream to the first ultra-filtration step characterizing its heat capacity, thermal conductivity, and viscosity, as well as the density of the evaporator contents. The scope of the task was expanded to include the volume reduction factor across the waste feed evaporator (total evaporator feed volume/evaporator bottoms volume). All the physical properties were modeled as functions of the waste feed composition, temperature, and the high level waste recycle volumetric flow rate relative to that of the waste feed. The goal for the mathematical models was to predict the physical property to predicted simulation value. The simulation model approximating the FEP process used to develop the correlations was relatively complex, and not possible to duplicate within the scope of the bench scale evaporation experiments. Therefore, simulants were made of 13 design points (a subset of the points used in the model fits) using the compositions of the ultra-filtration feed streams as predicted by the simulation model. The chemistry and physical properties of the supernate (the modeled stream) as predicted by the simulation were compared with the analytical results of experimental simulant work as a method of validating the simulation software

Impact of vegetation variability on potential predictability and skill of EC-Earth simulations

Energy Technology Data Exchange (ETDEWEB)

Weiss, Martina; Hurk, Bart van den; Haarsma, Reindert; Hazeleger, Wilco [Royal Netherlands Meteorological Institute (KNMI), De Bilt (Netherlands)

2012-12-15

Climate models often use a simplified and static representation of vegetation characteristics to determine fluxes of energy, momentum and water vapour between surface and lower atmosphere. In order to analyse the impact of short term variability in vegetation phenology, we use remotely-sensed leaf area index and albedo products to examine the role of vegetation in the coupled land-atmosphere system. Perfect model experiments are carried out to determine the impact of realistic temporal variability of vegetation on potential predictability of evaporation and temperature, as well as model skill of EC-Earth simulations. The length of the simulation period is hereby limited by the availability of satellite products to 2000-2010. While a realistic representation of vegetation positively influences the simulation of evaporation and its potential predictability, a positive impact on 2 m temperature is of smaller magnitude, regionally confined and more pronounced in climatically extreme years. (orig.)

Wind Turbine Generator Modeling and Simulation Where Rotational Speed is the Controlled Variable

DEFF Research Database (Denmark)

Mihet-Popa, Lucian; Blaabjerg, Frede; Boldea, Ion

2004-01-01

the interaction between a wind turbine and the power system. The model is intended to simulate the behaviour of the wind turbine using induction generators both during normal operation. Sample simulation results for two induction generators (2/0.5 MW) validate the fundamental issues.......To optimise the power produced in a wind turbine, the speed of the turbine should vary with the wind speed. A simple control method is proposed that will allow an induction machine to run a turbine at its maximum power coefficient. Various types of power control strategies have been suggested...... for application in variable speed wind turbines. The usual strategy is to control the power or the torque acting on the wind turbine shafts. This paper presents an alternative control strategy, where the rotational speed is the controlled variable. The paper describes a model, which is being developed to simulate...

Temperature simulation of thermal plasma melting furnace for disposal of radioactive waste and preliminary research of vitrification formula

International Nuclear Information System (INIS)

Lin Peng; Lu Yonghong; Xiang Wenyuan; Chen Mingzhou; Liu Xiajie; Qin Yuxin

2013-01-01

Radioactive waste treatment techniques currently used in nuclear power plant increase the volume greatly and bring much pressure on final disposal; Thermal plasma treatment as a crucial technique to reduce the waste volume is introduced. How to improve the efficiency of the plasma energy is the limiting factor of concern. In this paper, the temperature field of thermal plasma melting furnace is simulated, the maximal temperature of fixed bed melting furnace is calculated (about 1445 ℃). According to the optional fire-resistant materials, the feasibility of furnace fabrication is discussed. Vitrification formulas for three typical radioactive wastes are tested with their feasibilities being analyzed then. Finally, the prospect of thermal plasma techniques of radioactive waste is discussed, and issues for future study are raised. (authors)

Effects of mixed waste simulants on transportation packaging plastic components

International Nuclear Information System (INIS)

Nigrey, P.J.; Dickens, T.G.

1994-01-01

The purpose of hazardous and radioactive materials packaging is to, enable these materials to be transported without posing a threat to the health or property of the general public. To achieve this aim, regulations have been written establishing general design requirements for such packagings. While no regulations have been written specifically for mixed waste packaging, regulations for the constituents of mixed wastes, i.e., hazardous and radioactive substances, have been codified. The design requirements for both hazardous and radioactive materials packaging specify packaging compatibility, i.e., that the materials of the packaging and any contents be chemically compatible with each other. Furthermore, Type A and Type B packaging design requirements stipulate that there be no significant chemical, galvanic, or other reaction between the materials and contents of the package. Based on these requirements, a Chemical Compatibility Testing Program was developed in the Transportation Systems Department at Sandia National Laboratories (SNL). The program, supported by the US Department of Energy's (DOE) Transportation Management Division, EM-261 provides the means to assure any regulatory body that the issue of packaging material compatibility towards hazardous and radioactive materials has been addressed. In this paper, we describe the general elements of the testing program and the experimental results of the screening tests. The implications of the results of this testing are discussed in the general context of packaging development. Additionally, we present the results of the first phase of this experimental program. This phase involved the screening of five candidate liner and six seal materials against four simulant mixed wastes

Quantifying uncertainty due to internal variability using high-resolution regional climate model simulations

Science.gov (United States)

Gutmann, E. D.; Ikeda, K.; Deser, C.; Rasmussen, R.; Clark, M. P.; Arnold, J. R.

2015-12-01

The uncertainty in future climate predictions is as large or larger than the mean climate change signal. As such, any predictions of future climate need to incorporate and quantify the sources of this uncertainty. One of the largest sources comes from the internal, chaotic, variability within the climate system itself. This variability has been approximated using the 30 ensemble members of the Community Earth System Model (CESM) large ensemble. Here we examine the wet and dry end members of this ensemble for cool-season precipitation in the Colorado Rocky Mountains with a set of high-resolution regional climate model simulations. We have used the Weather Research and Forecasting model (WRF) to simulate the periods 1990-2000, 2025-2035, and 2070-2080 on a 4km grid. These simulations show that the broad patterns of change depicted in CESM are inherited by the high-resolution simulations; however, the differences in the height and location of the mountains in the WRF simulation, relative to the CESM simulation, means that the location and magnitude of the precipitation changes are very different. We further show that high-resolution simulations with the Intermediate Complexity Atmospheric Research model (ICAR) predict a similar spatial pattern in the change signal as WRF for these ensemble members. We then use ICAR to examine the rest of the CESM Large Ensemble as well as the uncertainty in the regional climate model due to the choice of physics parameterizations.

Model reliability and software quality assurance in simulation of nuclear fuel waste management systems

International Nuclear Information System (INIS)

Oeren, T.I.; Elzas, M.S.; Sheng, G.; Wageningen Agricultural Univ., Netherlands; McMaster Univ., Hamilton, Ontario)

1985-01-01

As is the case with all scientific simulation studies, computerized simulation of nuclear fuel waste management systems can introduce and hide various types of errors. Frameworks to clarify issues of model reliability and software quality assurance are offered. Potential problems with reference to the main areas of concern for reliability and quality are discussed; e.g., experimental issues, decomposition, scope, fidelity, verification, requirements, testing, correctness, robustness are treated with reference to the experience gained in the past. A list comprising over 80 most common computerization errors is provided. Software tools and techniques used to detect and to correct computerization errors are discussed

Process performance of the pilot-scale in situ vitrification of a simulated waste disposal site at the Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Carter, J.G.; Koegler, S.S.; Bates, S.O.

1988-06-01

Process feasibility studies have been successfully performed on three developmental scales to determine the potential for applying in situ vitrification to intermediate-level (low-level) waste placed in seepage pits and trenches at Oak Ridge National Laboratory (ORNL). In the laboratory, testing was performed in crucibles containing a mixture of 50% ORNL soil and 50% limestone. In an engineering-scale test at Pacific Northwest Laboratory a /1/12/-scale simulation of an ORNL waste trench was constructed and vitrified, resulting in a waste product containing soil and limestone concentrations of 68 wt % and 32 wt %, respectively. In the pilot-scale test a /3/8/-scale simulation of the same trench was constructed and vitrified at ORNL, resulting in soil and limestone concentrations of 80% and 20%, respectively, in the waste product. Results of the three scales of testing indicate that the ORNL intermediate-level (low-level) waste sites can be successfully processed by in situ vitrification; the waste form will retain significant quantities of the cesium and strontium. Because cesium-137 and strontium-90 are the major components of the radionuclide inventory in the ORNL seepage pits and trenches, final field process decontamination factors (i.e., losses to the off-gas system relative to the waste inventory) of 1.0 E + 4 are desired to minimize activity buildup in the off-gas system. 17 refs., 34 figs., 13 tabs

Sensitivity analysis and uncertainties simulation of the migration of radionuclide in the system of geological disposal-CRP-GEORC model

International Nuclear Information System (INIS)

Su Rui; Wang Ju; Chen Weiming; Zong Zihua; Zhao Honggang

2008-01-01

CRP-GEORC concept model is an artificial system of geological disposal for High-Level radioactive waste. Sensitivity analysis and uncertainties simulation of the migration of radionuclide Se-79 and I-129 in the far field of this system by using GoldSim Code have been conducted. It can be seen from the simulation results that variables used to describe the geological features and characterization of groundwater flow are sensitive variables of whole geological disposal system. The uncertainties of parameters have remarkable influence on the simulation results. (authors)

Mercury separation from mixed wastes. Annual report

International Nuclear Information System (INIS)

Taylor, P.A.; Klasson, K.T.; Corder, S.L.; Carlson, T.R.; McCandless, K.R.

1995-11-01

This is an assessment of new sorbents for removing Hg from wastes at US DOE sites. Four aqueous wastes were used for the laboratory tests: a simulant of a high-salt, acidic waste currently stored at INEL, a simulant of a high-salt, alkaline waste stored at Savannah River (SRS), a dilute LiOH solution stored at Y-12, and a low-salt, neutral groundwater generated at Y-12. Eight adsorbents covering a wide range of cost and capability were tested. Screening tests identified the most promising adsorbents, and column tests were performed using at least two adsorbents for each waste stream. No one adsorbent is effective in all of these waste streams. Based on loading capacity and compatibility, the most effect adsorbents to date are SuperLig 618 for the INEL tank waste simulant, Mersorb and Ionac SR-3 for the SRS tank waste simulant, Durasil 70 and Ionac SR-3 for the LiOH solution, and Ionac SR-3, followed by Ionac SR-4 and Mersorb, for the Y-12 groundwater

Performance Demonstration Program Plan for Nondestructive Assay of Boxed Wastes for the TRU Waste Characterization Program

International Nuclear Information System (INIS)

2001-01-01

The Performance Demonstration Program (PDP) for nondestructive assay (NDA) consists of a series of tests to evaluate the capability for NDA of transuranic (TRU) waste throughout the Department of Energy (DOE) complex. Each test is termed a PDP cycle. These evaluation cycles provide an objective measure of the reliability of measurements obtained from NDA systems used to characterize the radiological constituents of TRU waste. The primary documents governing the conduct of the PDP are the Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WAC; DOE 1999a) and the Quality Assurance Program Document (QAPD; DOE 1999b). The WAC requires participation in the PDP; the PDP must comply with the QAPD and the WAC. The WAC contains technical and quality requirements for acceptable NDA. This plan implements the general requirements of the QAPD and applicable requirements of the WAC for the NDA PDP for boxed waste assay systems. Measurement facilities demonstrate acceptable performance by the successful testing of simulated waste containers according to the criteria set by this PDP Plan. Comparison among DOE measurement groups and commercial assay services is achieved by comparing the results of measurements on similar simulated waste containers reported by the different measurement facilities. These tests are used as an independent means to assess the performance of measurement groups regarding compliance with established quality assurance objectives (QAO's). Measurement facilities must analyze the simulated waste containers using the same procedures used for normal waste characterization activities. For the boxed waste PDP, a simulated waste container consists of a modified standard waste box (SWB) emplaced with radioactive standards and fabricated matrix inserts. An SWB is a waste box with ends designed specifically to fit the TRUPACT-II shipping container. SWB's will be used to package a substantial volume of the TRU waste for disposal. These PDP sample components

Durability of simulated waste glass: effects of pressure and formation of surface layers

International Nuclear Information System (INIS)

Wicks, G.G.; Mosley, W.C.; Whitkop, P.G.; Saturday, K.A.

1981-01-01

The leaching behavior of simulated Savannah River Plant (SRP) waste glass was studied at elevated pressures and anticipated storage temperatures. An integrated approach, which combined leachate solution analyses with both bulk and surface studies, was used to study the corrosion process. Compositions of leachates were evaluated by colorimetry and atomic absorption. Used in the bulk and surface analyses were optical microscopy, scanning electron microscopy, x-ray energy spectroscopy, wide-angle x-ray, diffraction, electron microprobe analysis, infrared reflectance spectroscopy, electron spectroscopy for chemical analysis, and Auger electron spectroscopy. Results from this study show that there is no significant adverse effect of pressure, up to 1500 psi and 90 0 C, on the chemical durability of simulated SPR waste glass leached for one month in deionized water. In addition, the leached glass surface layer was characterized by an adsorbed film rich in minor constituents from the glass. This film remained on the glass surface even after leaching in relatively alkaline solutions at elevated pressures at 90 0 C for one month. The sample surface area to volume of leachant ratios (SA/V) was 10:1 cm -1 and 1:10 cm -1 . The corrosion mechanisms and surface and subsurface layers produced will be discussed along with the potential importance of these results to repository storage

Regional Community Climate Simulations with variable resolution meshes in the Community Earth System Model

Science.gov (United States)

Zarzycki, C. M.; Gettelman, A.; Callaghan, P.

2017-12-01

Accurately predicting weather extremes such as precipitation (floods and droughts) and temperature (heat waves) requires high resolution to resolve mesoscale dynamics and topography at horizontal scales of 10-30km. Simulating such resolutions globally for climate scales (years to decades) remains computationally impractical. Simulating only a small region of the planet is more tractable at these scales for climate applications. This work describes global simulations using variable-resolution static meshes with multiple dynamical cores that target the continental United States using developmental versions of the Community Earth System Model version 2 (CESM2). CESM2 is tested in idealized, aquaplanet and full physics configurations to evaluate variable mesh simulations against uniform high and uniform low resolution simulations at resolutions down to 15km. Different physical parameterization suites are also evaluated to gauge their sensitivity to resolution. Idealized variable-resolution mesh cases compare well to high resolution tests. More recent versions of the atmospheric physics, including cloud schemes for CESM2, are more stable with respect to changes in horizontal resolution. Most of the sensitivity is due to sensitivity to timestep and interactions between deep convection and large scale condensation, expected from the closure methods. The resulting full physics model produces a comparable climate to the global low resolution mesh and similar high frequency statistics in the high resolution region. Some biases are reduced (orographic precipitation in the western United States), but biases do not necessarily go away at high resolution (e.g. summertime JJA surface Temp). The simulations are able to reproduce uniform high resolution results, making them an effective tool for regional climate studies and are available in CESM2.

LABORATORY OPTIMIZATION TESTS OF TECHNETIUM DECONTAMINATION OF HANFORD WASTE TREATMENT PLANT LOW ACTIVITY WASTE OFF-GAS CONDENSATE SIMULANT

Energy Technology Data Exchange (ETDEWEB)

Taylor-Pashow, K.; Nash, C.; McCabe, D.

2014-09-29

compatible with longterm tank storage and immobilization methods. For this new application, testing is needed to demonstrate acceptable treatment sorbents and precipitating agents and measure decontamination factors for additional radionuclides in this unique waste stream. The origin of this LAW Off-Gas Condensate stream will be the liquids from the Submerged Bed Scrubber (SBS) and the Wet Electrostatic Precipitator (WESP) from the LAW melter off-gas system. The stream is expected to be a dilute salt solution with near neutral pH, and will likely contain some insoluble solids from melter carryover. The soluble components are expected to be mostly sodium and ammonium salts of nitrate, chloride, and fluoride. This stream has not been generated yet and will not be available until the WTP begins operation, but a simulant has been produced based on models, calculations, and comparison with pilot-scale tests. One of the radionuclides that is volatile and expected to be in greatest abundance in this LAW Off-Gas Condensate stream is Technetium-99 ({sup 99}Tc). Technetium will not be removed from the aqueous waste in the Hanford WTP, and will primarily end up immobilized in the LAW glass by repeated recycle of the off-gas condensate into the LAW melter. Other radionuclides that are low but are also expected to be in measurable concentration in the LAW Off-Gas Condensate are {sup 129}I, {sup 90}Sr, {sup 137}Cs, {sup 241}Pu, and {sup 241}Am. These are present due to their partial volatility and some entrainment in the off-gas system. This report discusses results of optimized {sup 99}Tc decontamination testing of the simulant. Testing examined use of inorganic reducing agents for {sup 99}Tc. Testing focused on minimizing the quantity of sorbents/reactants added, and minimizing mixing time to reach the decontamination targets in this simulant formulation. Stannous chloride and ferrous sulfate were tested as reducing agents to determine the minimum needed to convert soluble pertechnetate

Software verification, model validation, and hydrogeologic modelling aspects in nuclear waste disposal system simulations. A paradigm shift

International Nuclear Information System (INIS)

Sheng, G.M.

1994-01-01

This work reviewed the current concept of nuclear waste disposal in stable, terrestrial geologic media with a system of natural and man-made multi-barriers. Various aspects of this concept and supporting research were examined with the emphasis on the Canadian Nuclear Fuel Waste Management Program. Several of the crucial issues and challenges facing the current concept were discussed. These include: The difficulties inherent in a concept that centres around lithologic studies; the unsatisfactory state of software quality assurance in the present computer simulation programs; and the lack of a standardized, comprehensive, and systematic procedure to carry out a rigorous process of model validation and assessment of simulation studies. An outline of such an approach was presented and some of the principles, tools and techniques for software verification were introduced and described. A case study involving an evaluation of the Canadian performance assessment computer program is presented. A new paradigm to nuclear waste disposal was advocated to address the challenges facing the existing concept. The RRC (Regional Recharge Concept) was introduced and its many advantages were described and shown through a modelling exercise. (orig./HP)

Fluid dynamic demonstrations for waste retrieval and treatment

International Nuclear Information System (INIS)

Youngblood, E.L. Jr.; Hylton, T.D.; Berry, J.B.; Cummins, R.L.; Ruppel, F.R.; Hanks, R.W.

1994-02-01

The objective of this study was to develop or identify flow correlations for predicting the flow parameters needed for the design and operation of slurry pipeline systems for transporting radioactive waste of the type stored in the Hanford single-shell tanks and the type stored at the Oak Ridge National Laboratory (ORNL). This was done by studying the flow characteristics of simulated waste with rheological properties similar to those of the actual waste. Chemical simulants with rheological properties similar to those of the waste stored in the Hanford single-shell tanks were developed by Pacific Northwest Laboratories, and simulated waste with properties similar to those of ORNL waste was developed at ORNL for use in the tests. Rheological properties and flow characteristics of the simulated slurry were studied in a test loop in which the slurry was circulated through three pipeline viscometers (constructed of 1/2-, 3/4-, and 1-in. schedule 40 pipe) at flow rates up to 35 gal/min. Runs were made with ORNL simulated waste at 54 wt % to 65 wt % total solids and temperatures of 25 degree C and 55 degree C. Grinding was done prior to one run to study the effect of reduced particle size. Runs were made with simulated Hanford single-shell tank waste at approximately 43 wt % total solids and at temperatures of 25 degree C and 50 degree C. The rheology of simulated Hanford and ORNL waste supernatant liquid was also measured

Medication waste reduction in pediatric pharmacy batch processes.

Science.gov (United States)

Toerper, Matthew F; Veltri, Michael A; Hamrock, Eric; Mollenkopf, Nicole L; Holt, Kristen; Levin, Scott

2014-04-01

To inform pediatric cart-fill batch scheduling for reductions in pharmaceutical waste using a case study and simulation analysis. A pre and post intervention and simulation analysis was conducted during 3 months at a 205-bed children's center. An algorithm was developed to detect wasted medication based on time-stamped computerized provider order entry information. The algorithm was used to quantify pharmaceutical waste and associated costs for both preintervention (1 batch per day) and postintervention (3 batches per day) schedules. Further, simulation was used to systematically test 108 batch schedules outlining general characteristics that have an impact on the likelihood for waste. Switching from a 1-batch-per-day to a 3-batch-per-day schedule resulted in a 31.3% decrease in pharmaceutical waste (28.7% to 19.7%) and annual cost savings of $183,380. Simulation results demonstrate how increasing batch frequency facilitates a more just-in-time process that reduces waste. The most substantial gains are realized by shifting from a schedule of 1 batch per day to at least 2 batches per day. The simulation exhibits how waste reduction is also achievable by avoiding batch preparation during daily time periods where medication administration or medication discontinuations are frequent. Last, the simulation was used to show how reducing batch preparation time per batch provides some, albeit minimal, opportunity to decrease waste. The case study and simulation analysis demonstrate characteristics of batch scheduling that may support pediatric pharmacy managers in redesign toward minimizing pharmaceutical waste.

Simulant composition for the Mixed Waste Management Facility (MWMF) groundwater remediation project

International Nuclear Information System (INIS)

Siler, J.L.

1992-01-01

A project has been initiated at the request of ER to study and remediate the groundwater contamination at the Mixed Waste Management Facility (MWMF). This water contains a wide variety of both inorganics (e.g., sodium) and organics (e.g., benzene, trichloroethylene). Most compounds are present in the ppB range, and certain components (e.g., trichloroethylene, silver) are present at concentrations that exceed the primary drinking water standards (PDWS). These compounds must be reduced to acceptable levels as per RCRA and CERCLA orders. This report gives a listing of the important constituents which are to be included in a simulant to model the MWMF aquifer. This simulant will be used to evaluate the feasibility of various state of the art separation/destruction processes for remediating the aquifer

Valuation of large variable annuity portfolios: Monte Carlo simulation and synthetic datasets

Directory of Open Access Journals (Sweden)

Gan Guojun

2017-12-01

Full Text Available Metamodeling techniques have recently been proposed to address the computational issues related to the valuation of large portfolios of variable annuity contracts. However, it is extremely diffcult, if not impossible, for researchers to obtain real datasets frominsurance companies in order to test their metamodeling techniques on such real datasets and publish the results in academic journals. To facilitate the development and dissemination of research related to the effcient valuation of large variable annuity portfolios, this paper creates a large synthetic portfolio of variable annuity contracts based on the properties of real portfolios of variable annuities and implements a simple Monte Carlo simulation engine for valuing the synthetic portfolio. In addition, this paper presents fair market values and Greeks for the synthetic portfolio of variable annuity contracts that are important quantities for managing the financial risks associated with variable annuities. The resulting datasets can be used by researchers to test and compare the performance of various metamodeling techniques.

Performance Demonstration Program Plan for Nondestructive Assay of Drummed Wastes for the TRU Waste Characterization Program

International Nuclear Information System (INIS)

DOE Carlsbad Field Office

2001-01-01

The Performance Demonstration Program (PDP) for nondestructive assay (NDA) consists of a series of tests to evaluate the capability for NDA of transuranic (TRU) waste throughout the Department of Energy (DOE) complex. Each test is termed a PDP cycle. These evaluation cycles provide an objective measure of the reliability of measurements obtained from NDA systems used to characterize the radiological constituents of TRU waste. The primary documents governing the conduct of the PDP are the Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WAC; DOE 1999a) and the Quality Assurance Program Document (QAPD; DOE 1999b). The WAC requires participation in the PDP; the PDP must comply with the QAPD and the WAC. The WAC contains technical and quality requirements for acceptable NDA. This plan implements the general requirements of the QAPD and applicable requirements of the WAC for the NDA PDP. Measurement facilities demonstrate acceptable performance by the successful testing of simulated waste containers according to the criteria set by this PDP Plan. Comparison among DOE measurement groups and commercial assay services is achieved by comparing the results of measurements on similar simulated waste containers reported by the different measurement facilities. These tests are used as an independent means to assess the performance of measurement groups regarding compliance with established quality assurance objectives (QAO's). Measurement facilities must analyze the simulated waste containers using the same procedures used for normal waste characterization activities. For the drummed waste PDP, a simulated waste container consists of a 55-gallon matrix drum emplaced with radioactive standards and fabricated matrix inserts. These PDP sample components are distributed to the participating measurement facilities that have been designated and authorized by the Carlsbad Field Office (CBFO). The NDA Drum PDP materials are stored at these sites under secure conditions to

Evaluation of Hanford high level waste vitrification chemistry for an NCAW simulant -- FY 1994: Potential exothermic reactions in the presence of formic acid, glycolic acid, and oxalic acid

Energy Technology Data Exchange (ETDEWEB)

Sills, J.A.

1995-07-01

A potential for an uncontrollable exothermic reaction between nitrate and organic salts during preparation of a high level waste melter feed has been identified. In order to examine this potential more closely, the thermal behavior of simulated neutralized current acid waste (NCAW) treated with various organic reductants was studied. Differential scanning calorimetry (DSC) measurements were collected on simulated waste samples and their supernates treated with organics. Organic reductants used were formic acid, glycolic acid, and oxalic acid. For comparison, samples of untreated simulant and untreated simulant with added noble metals were tested. When heated, untreated simulant samples both with and without noble metals showed no exothermic behavior. All of the treated waste simulant samples showed exothermic behavior. Onset temperatures of exothermic reactions were 120 C to 210 C. Many onset temperatures, particularly those for formic acid treated samples, are well below 181 C, the estimated maximum steam coil temperature (considered to be a worst case maximum temperature for chemical process tank contents). The enthalpies of the reactions were {minus}180 {times} 10{sup {minus}3} J/Kg supernate ({minus}181 J/g) for the oxalic acid treated simulant supernate to {minus}1,150 {times} 10{sup {minus}3} J/Kg supernate ({minus}1,153 J/g) for the formic acid treated simulant supernate.

Evaluation of Hanford high level waste vitrification chemistry for an NCAW simulant -- FY 1994: Potential exothermic reactions in the presence of formic acid, glycolic acid, and oxalic acid

International Nuclear Information System (INIS)

Sills, J.A.

1995-07-01

A potential for an uncontrollable exothermic reaction between nitrate and organic salts during preparation of a high level waste melter feed has been identified. In order to examine this potential more closely, the thermal behavior of simulated neutralized current acid waste (NCAW) treated with various organic reductants was studied. Differential scanning calorimetry (DSC) measurements were collected on simulated waste samples and their supernates treated with organics. Organic reductants used were formic acid, glycolic acid, and oxalic acid. For comparison, samples of untreated simulant and untreated simulant with added noble metals were tested. When heated, untreated simulant samples both with and without noble metals showed no exothermic behavior. All of the treated waste simulant samples showed exothermic behavior. Onset temperatures of exothermic reactions were 120 C to 210 C. Many onset temperatures, particularly those for formic acid treated samples, are well below 181 C, the estimated maximum steam coil temperature (considered to be a worst case maximum temperature for chemical process tank contents). The enthalpies of the reactions were -180 x 10 -3 J/Kg supernate (-181 J/g) for the oxalic acid treated simulant supernate to -1,150 x 10 -3 J/Kg supernate (-1,153 J/g) for the formic acid treated simulant supernate

Plasma processing of compacted drums of simulated radioactive waste

International Nuclear Information System (INIS)

Geimer, R.; Batdorf, J.; Larsen, M.M.

1991-01-01

The charter of the Department of Energy (DOE) Office of Technology Development (OTD) is to identify and develop technologies that have potential application in the treatment of DOE wastes. One particular waste of concern within the DOE is transuranic (TRU) waste, which is generated and stored at several DOE sites. High temperature DC arc generated plasma technology is an emerging treatment method for TRU waste, and its use has the potential to provide many benefits in the management of TRU. This paper begins by discussing the need for development of a treatment process for TRU waste, and the potential benefits that a plasma waste treatment system can provide in treating TRU waste. This is followed by a discussion of the results of a project conducted for the DOE to demonstrate the effectiveness of a plasma process for treating supercompacted TRU waste. 1 fig., 1 tab

Synergistic and alkaline stability studies of mixtures of simulated high level waste sludge with selected energetic compounds

International Nuclear Information System (INIS)

Fondeur, F.F.

2000-01-01

This study examined the stability of mercury oxalate and mercury fulminate in alkaline sludge simulating Savannah River Site waste. These compounds represent two classes of energetic compounds previously speculated as potential components in sludge stored without a supernatant liquid

Leaching of vitrified high-level-active-waste in a near reality simulated repository system

International Nuclear Information System (INIS)

Froeschen, W.; Wolf, G.K.

1987-01-01

In the FRG it is planned to vitrify the high level waste from spent fuel reprocessing and to dispose of in a salt-mine. If water penetrates into the repository a highly corrosive brine (Q-brine) will be formed and radioactive material may be leached from the glasses and transported to human environment. The corrosion system of brine, corroded steel containers of the vitrified waste, and waste-glasses was investigated under near reality conditions. Experiments in hydrothermal environment were carried out including gamma radiation of the waste-glasses and ceramic In Can Lining between glasses and metallic containments. Screening experiments by application of external cobalt-gamma-radiation showed no principal changes in leaching behaviour of simulate glasses compared to leaching without radiation. Radiation effects result in pH changes mainly which are diminished by buffer capacity of Q-brine. Lining of steel containments with ceramic fleece does not reduce leaching but retards solution of Mo and Sr into brine. Decreasing of elements Sr, Cs and Mo in the near surface area of the glass and increasing of Zr and Ti has been found to be enhanced considerably in presence of canister corrosion products in Q-brine as well as in NaCl-leaching solution. (orig.) With 13 refs., 22 figs [de

Development of a pyro-partitioning process for long-lived radioactive nuclides. Process test for pretreatment of simulated high-level waste containing uranium

International Nuclear Information System (INIS)

Kurata, Masateru; Hijikata, Takatoshi; Kinoshita, Kensuke; Inoue, Tadashi

2000-01-01

A pyro-partitioning process developed at CRIEPI requires a pre-treatment process to convert high-level liquid waste to chloride. A combination process of denitration and chlorination has been developed for this purpose. Continuous process tests using simulated high-level waste were performed to certify the applicability of the process. Test results indicated a successful material balance sufficient for satisfying pyro-partitioning process criteria. In the present study, process tests using simulated high-level waste containing uranium were also carried out to prove that the pre-treatment process is feasible for uranium. The results indicated that uranium can be converted to chloride appropriate for the pyro-partitioning process. The material balance obtained from the tests is to be used to revise the process flow diagram. (author)

Simulation of Few Bifurcation Phase Diagrams of Belousov-Zhabotinsky Reaction with Eleven Variable Chaotic Model in CSTR

Directory of Open Access Journals (Sweden)

B. Swathi

2009-01-01

Full Text Available Simulation of the Gyorgyi, Rempe and Field eleven variable chaotic model in CSTR [Continuously Stirred Tank Reactor] is performed with respect to the concentrations of malonic acid and [Ce(III]. These simulation studies show steady state, periodic and non-periodic regions. These studies have been presented as two variable bifurcation phase diagrams. We also have observed the bursting phenomenon under different set of constraints. We have given much importance on computer simulation work but not included the experimental methods in this paper.

Conceptualization and software development of a simulation environment for probalistic safety assessment of radioactive waste repositories

Energy Technology Data Exchange (ETDEWEB)

Ghofrani, Javad

2016-05-26

Uncertainty and sensitivity analysis of complex simulation models are prominent issues, both in scientific research and education. ReSUS (Repository Simulation, Uncertainty propagation and Sensitivity analysis) is an integrated platform to perform such analysis with numerical models that simulate the THMC (Thermal Hydraulical Mechanical and Chemical) coupled processes via different programs, in particular in the context of safety assessments for radioactive waste repositories. This thesis presents the idea behind the software platform ReSUS and its working mechanisms. Apart from the idea and the working mechanisms, the thesis describes applications related to the safety assessment of radioactive waste disposal systems. In this thesis, previous simulation tools (including the preceding version of ReSUS) are analyzed in order to provide a comprehensive view of the state of the art. In comparison to this state, a more sophisticated software tool is developed here, which provides features which are not offered by previous simulation tools. To achieve this objective, the software platform ReSUS provides a framework for handling probabilistic data uncertainties using deterministic external simulation tools, thus enhancing uncertainty and sensitivity analysis. This platform performs probabilistic simulations of various models, in particular THMC coupled processes, using stand-alone deterministic simulation software tools. The complete software development process of the ReSUS Platform is discussed in this thesis. ReSUS components are developed as libraries, which are capable of being linked to other code implementations. In addition, ASCII template files are used as means for uncertainty propagation into the input files of deterministic simulation tools. The embedded input sampler and analysis tools allow for sensitivity analysis in several kinds of simulation designs. The novelty of the ReSUS platform consists in the flexibility to assign external stand-alone software

Conceptualization and software development of a simulation environment for probalistic safety assessment of radioactive waste repositories

International Nuclear Information System (INIS)

Ghofrani, Javad

2016-01-01

Uncertainty and sensitivity analysis of complex simulation models are prominent issues, both in scientific research and education. ReSUS (Repository Simulation, Uncertainty propagation and Sensitivity analysis) is an integrated platform to perform such analysis with numerical models that simulate the THMC (Thermal Hydraulical Mechanical and Chemical) coupled processes via different programs, in particular in the context of safety assessments for radioactive waste repositories. This thesis presents the idea behind the software platform ReSUS and its working mechanisms. Apart from the idea and the working mechanisms, the thesis describes applications related to the safety assessment of radioactive waste disposal systems. In this thesis, previous simulation tools (including the preceding version of ReSUS) are analyzed in order to provide a comprehensive view of the state of the art. In comparison to this state, a more sophisticated software tool is developed here, which provides features which are not offered by previous simulation tools. To achieve this objective, the software platform ReSUS provides a framework for handling probabilistic data uncertainties using deterministic external simulation tools, thus enhancing uncertainty and sensitivity analysis. This platform performs probabilistic simulations of various models, in particular THMC coupled processes, using stand-alone deterministic simulation software tools. The complete software development process of the ReSUS Platform is discussed in this thesis. ReSUS components are developed as libraries, which are capable of being linked to other code implementations. In addition, ASCII template files are used as means for uncertainty propagation into the input files of deterministic simulation tools. The embedded input sampler and analysis tools allow for sensitivity analysis in several kinds of simulation designs. The novelty of the ReSUS platform consists in the flexibility to assign external stand-alone software

INVESTIGATING SUSPENSION OF MST, CST, AND SIMULATED SLUDGE SLURRIES IN A PILOT-SCALE WASTE TANK

Energy Technology Data Exchange (ETDEWEB)

Poirier, M.; Qureshi, Z.; Restivo, M.; Steeper, T.; Williams, M.

2011-05-24

The Small Column Ion Exchange (SCIX) process is being developed to remove cesium, strontium, and actinides from Savannah River Site (SRS) Liquid Waste using an existing waste tank (i.e., Tank 41H) to house the process. Savannah River National Laboratory (SRNL) is conducting pilot-scale mixing tests to determine the pump requirements for suspending and resuspending monosodium titanate (MST), crystalline silicotitanate (CST), and simulated sludge. The purpose of this pilot scale testing is for the pumps to resuspend the MST, CST, and simulated sludge particles so that they can be removed from the tank, and to suspend the MST so it can contact strontium and actinides. The pilot-scale tank is a 1/10.85 linear scaled model of Tank 41H. The tank diameter, tank liquid level, pump nozzle diameter, pump elevation, and cooling coil diameter are all 1/10.85 of their dimensions in Tank 41H. The pump locations correspond to the proposed locations in Tank 41H by the SCIX program (Risers B5, B3, and B1). Previous testing showed that three Submersible Mixer Pumps (SMPs) will provide sufficient power to initially suspend MST in an SRS waste tank, and to resuspend MST that has settled in a waste tank at nominal 45 C for four weeks. The conclusions from this analysis are: (1) Three SMPs will be able to resuspend more than 99.9% of the MST and CST that has settled for four weeks at nominal 45 C. The testing shows the required pump discharge velocity is 84% of the maximum discharge velocity of the pump. (2) Three SMPs will be able to resuspend more than 99.9% of the MST, CST, and simulated sludge that has settled for four weeks at nominal 45 C. The testing shows the required pump discharge velocity is 82% of the maximum discharge velocity of the pump. (3) A contact time of 6-12 hours is needed for strontium sorption by MST in a jet mixed tank with cooling coils, which is consistent with bench-scale testing and actinide removal process (ARP) operation.

Leaching of actinides from simulated nuclear waste glass

International Nuclear Information System (INIS)

Pickering, S.; Walker, C.T.; Offermann, P.

1982-01-01

Two types of simulated nuclear waste glass doped with actinides were leached at 200 0 C in distilled water and salt solutions. Am, Np, Pu and U were all preferentially retained in the surface layer on the glass. Leaching ratios of 0.1 to 0.2 for Np and approx. 0.02 for Am were measured. The losses of Am and Np to the leachant were proportional to the total weight loss of the glass and were larger at 10 ml leachant/cm 2 glass than at 5 ml/cm 2 . Weight loss from the glass occurred only at the start of the experiments for periods ranging from 10 h to 10 days according to leachant composition and volume. Wt losses from the C31-3-EC glass were much greater in saturated NaCl solution than in distilled water. Enrichment in the outer surface layer of Al or Ca according to glass type could be correlated with leachant pH, glass composition and weight loss measurements

Variable dead time counters: 2. A computer simulation

International Nuclear Information System (INIS)

Hooton, B.W.; Lees, E.W.

1980-09-01

A computer model has been developed to give a pulse train which simulates that generated by a variable dead time counter (VDC) used in safeguards determination of Pu mass. The model is applied to two algorithms generally used for VDC analysis. It is used to determine their limitations at high counting rates and to investigate the effects of random neutrons from (α,n) reactions. Both algorithms are found to be deficient for use with masses of 240 Pu greater than 100g and one commonly used algorithm is shown, by use of the model and also by theory, to yield a result which is dependent on the random neutron intensity. (author)

Waste glass melting stages

International Nuclear Information System (INIS)

Anderson, L.D.; Dennis, T.; Elliott, M.L.; Hrma, P.

1993-04-01

Three different simulated nuclear waste glass feeds, consisting of dried waste and glass frit, were heat treated for 1 hour in a gradient furnace at temperatures ranging from approximately 600 degrees C--1000 degrees C. Simulated melter feeds from the Hanford Waste Vitrification Plant (HWVP), the Defense Waste Processing Facility (DWPF), and Kernforschungszentrum Karlsruhe (KfK) in Germany were used. The samples were thin-sectioned and examined by optical microscopy to investigate the stages of the conversion from feed to glass. Various phenomena were seen, such as frit softening, bubble formation, foaming, bubble motion and removal, convective mixing, and homogenization. Behavior of different feeds was similar, although the degree of gas generation and melt homogenization varied

Hanford Waste Vitrification Plant hydrogen generation

International Nuclear Information System (INIS)

King, R.B.; King, A.D. Jr.; Bhattacharyya, N.K.

1996-02-01

The most promising method for the disposal of highly radioactive nuclear wastes is a vitrification process in which the wastes are incorporated into borosilicate glass logs, the logs are sealed into welded stainless steel canisters, and the canisters are buried in suitably protected burial sites for disposal. The purpose of the research supported by the Hanford Waste Vitrification Plant (HWVP) project of the Department of Energy through Battelle Pacific Northwest Laboratory (PNL) and summarized in this report was to gain a basic understanding of the hydrogen generation process and to predict the rate and amount of hydrogen generation during the treatment of HWVP feed simulants with formic acid. The objectives of the study were to determine the key feed components and process variables which enhance or inhibit the.production of hydrogen. Information on the kinetics and stoichiometry of relevant formic acid reactions were sought to provide a basis for viable mechanistic proposals. The chemical reactions were characterized through the production and consumption of the key gaseous products such as H 2 . CO 2 , N 2 0, NO, and NH 3 . For this mason this research program relied heavily on analyses of the gases produced and consumed during reactions of the HWVP feed simulants with formic acid under various conditions. Such analyses, used gas chromatographic equipment and expertise at the University of Georgia for the separation and determination of H 2 , CO, CO 2 , N 2 , N 2 O and NO

Improving plot- and regional-scale crop models for simulating impacts of climate variability and extremes

Science.gov (United States)

Tao, F.; Rötter, R.

2013-12-01

Many studies on global climate report that climate variability is increasing with more frequent and intense extreme events1. There are quite large uncertainties from both the plot- and regional-scale models in simulating impacts of climate variability and extremes on crop development, growth and productivity2,3. One key to reducing the uncertainties is better exploitation of experimental data to eliminate crop model deficiencies and develop better algorithms that more adequately capture the impacts of extreme events, such as high temperature and drought, on crop performance4,5. In the present study, in a first step, the inter-annual variability in wheat yield and climate from 1971 to 2012 in Finland was investigated. Using statistical approaches the impacts of climate variability and extremes on wheat growth and productivity were quantified. In a second step, a plot-scale model, WOFOST6, and a regional-scale crop model, MCWLA7, were calibrated and validated, and applied to simulate wheat growth and yield variability from 1971-2012. Next, the estimated impacts of high temperature stress, cold damage, and drought stress on crop growth and productivity based on the statistical approaches, and on crop simulation models WOFOST and MCWLA were compared. Then, the impact mechanisms of climate extremes on crop growth and productivity in the WOFOST model and MCWLA model were identified, and subsequently, the various algorithm and impact functions were fitted against the long-term crop trial data. Finally, the impact mechanisms, algorithms and functions in WOFOST model and MCWLA model were improved to better simulate the impacts of climate variability and extremes, particularly high temperature stress, cold damage and drought stress for location-specific and large area climate impact assessments. Our studies provide a good example of how to improve, in parallel, the plot- and regional-scale models for simulating impacts of climate variability and extremes, as needed for

Quantifying the Effectiveness of Dose Individualization by Simulation for a Drug With Moderate Pharmacokinetic Variability.

Science.gov (United States)

Liefaard, Lia; Chen, Chao

2015-10-01

Dose individualization can reduce variability in exposure. The objective of this work was to quantify, through pharmacokinetic (PK) simulation, the potential for reducing the variability in exposure by dose individualization for a drug with moderate PK variability between subjects and between occasions within a subject, and a narrow therapeutic window. Using a population PK model that includes between-subject and between-occasion variability for apparent clearance, individual PK profiles in a trial of 300 subjects after a test dose were simulated. From the simulated data, datasets were created mimicking various sampling regimens (from single predose sample to full profile samples over 12 hours) on 1 or more occasions (1, 2, 3, 5, or 10 visits). Using these datasets, individual apparent clearance values were estimated, which were then used to calculate an individualized dose for a predefined target area under the concentration-time curve (AUC), based on the available formulation strengths. The proportion of people whose mean AUC was within a predefined therapeutic AUC range was calculated for the test (before) and the individualized dose (after), and compared between the different sampling scenarios. The maximum increase in proportion of subjects with an AUC within the range was 20%. To achieve this benefit, PK samples over 4 hours from 100 dosing occasions were required. As a result of the dose adjustment, the AUC of 7.3% of the subjects moved from inside the therapeutic range to outside of the range. This work shows how modeling and simulation can help assess the benefit and risk of dose individualization for a compound with variability between subjects and between occasions. The framework can be applied to similar situations with a defined set of conditions (eg, therapeutic window, tablet strengths, and PK and/or pharmacodynamic sampling scheme) to inform dose change and to assess the utility of dose individualization against certain success criteria.

Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

International Nuclear Information System (INIS)

Deshingkar, D.S.; Chitnis, R.R.; Wattal, P.K.; Theyyunni, T.K.; Nair, M.K.T.; Ramanujam, A.; Dhami, P.S.; Gopalakrishnan, V.; Rao, M.K.; Mathur, J.N.; Murali, M.S.; Iyer, R.H.; Badheka, L.P.; Banerji, A.

1994-01-01

High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs

Partitioning of actinide from simulated high level wastes arising from reprocessing of PHWR fuels: counter current extraction studies using CMPO

Energy Technology Data Exchange (ETDEWEB)

Deshingkar, D S; Chitnis, R R; Wattal, P K; Theyyunni, T K; Nair, M K.T. [Bhabha Atomic Research Centre, Bombay (India). Process Engineering and Systems Development Div.; Ramanujam, A; Dhami, P S; Gopalakrishnan, V; Rao, M K [Bhabha Atomic Research Centre, Bombay (India). Fuel Reprocessing Group; Mathur, J N; Murali, M S; Iyer, R H [Bhabha Atomic Research Centre, Bombay (India). Radiochemistry Div.; Badheka, L P; Banerji, A [Bhabha Atomic Research Centre, Bombay (India). Bio-organic Div.

1994-12-31

High level wastes (HLW) arising from reprocessing of pressurised heavy water reactor (PHWR) fuels contain actinides like neptunium, americium and cerium which are not extracted in the Purex process. They also contain small quantities of uranium and plutonium in addition to fission products. Removal of these actinides prior to vitrification of HLW can effectively reduce the active surveillance period of final waste form. Counter current studies using indigenously synthesised octyl (phenyl)-N, N-diisobutylcarbamoylmethylphosphine oxide (CMPO) were taken up as a follow-up of successful runs with simulated sulphate bearing low acid HLW solutions. The simulated HLW arising from reprocessing of PHWR fuel was prepared based on presumed burnup of 6500 MWd/Te of uranium, 3 years cooling period and 800 litres of waste generation per tonne of fuel reprocessed. The alpha activity of the HLW raffinate after extraction with the CMPO-TBP mixture could be brought down to near background level. (author). 13 refs., 2 tabs., 12 figs.

The development of WIPPVENT, a windows based interactive mine ventilation simulation software program at the Waste Isolation Pilot Plant

International Nuclear Information System (INIS)

McDaniel, K.H.

1995-01-01

An interactive mine ventilation simulation software program (WIPPVENT) was developed at the Waste Isolation Pilot Plant (WIPP). The WIPP is a US Department of Energy (DOE) research and development project located near Carlsbad, New Mexico. The facility is designed to provide a permanent, safe underground disposal of US defense generated transuranic waste in bedded salt. In addition to it's regular functions, the underground ventilation system is engineered to prevent the uncontrolled spread of radioactive materials in the unlikely event of a release. To enhance the operability system, Westinghouse Electric Corporation has developed an interactive mine ventilation simulation software program (WIPPVENT). While WIPPVENT includes most of the functions of the commercially available simulation program VNETPC (copyright 1991 Mine Ventilation Services, Inc.), the user interface has been completely rewritten as a Windows reg-sign application and screen graphics have been added. WIPPVENT is designed to interact with the WIPP ventilation monitoring systems through the site wide Central Monitoring System

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.

Multiple Regression and Mediator Variables can be used to Avoid Double Counting when Economic Values are Derived using Stochastic Herd Simulation

DEFF Research Database (Denmark)

Østergaard, Søren; Ettema, Jehan Frans; Hjortø, Line

Multiple regression and model building with mediator variables was addressed to avoid double counting when economic values are estimated from data simulated with herd simulation modeling (using the SimHerd model). The simulated incidence of metritis was analyzed statistically as the independent v...... in multiparous cows. The merit of using this approach was demonstrated since the economic value of metritis was estimated to be 81% higher when no mediator variables were included in the multiple regression analysis......Multiple regression and model building with mediator variables was addressed to avoid double counting when economic values are estimated from data simulated with herd simulation modeling (using the SimHerd model). The simulated incidence of metritis was analyzed statistically as the independent...... variable, while using the traits representing the direct effects of metritis on yield, fertility and occurrence of other diseases as mediator variables. The economic value of metritis was estimated to be €78 per 100 cow-years for each 1% increase of metritis in the period of 1-100 days in milk...

The role of SST variability in the simulation of the MJO

Science.gov (United States)

Stan, Cristiana

2017-12-01

The sensitivity of the Madden-Julian Oscillation to high-frequency variability (period 1-5 days) of sea surface temperature (SST) is investigated using numerical experiments with the super-parameterized Community Climate System Model. The findings of this study emphasize the importance of air-sea interactions in the simulation of the MJO, and stress the necessity of an accurate representation of ocean variability on short time scales. Eliminating 1-5-day variability of surface boundary forcing reduces the intraseasonal variability (ISV) of the tropics during the boreal winter. The ISV spectrum becomes close to the red noise background spectrum. The variability of atmospheric circulation shifts to longer time scales. In the absence of high-frequency variability of SST the MJO power gets confined to wavenumbers 1-2 and the magnitude of westward power associated with Rossby waves increases. The MJO convective activity propagating eastward from the Indian Ocean does not cross the Maritime Continent, and convection in the western Pacific Ocean is locally generated. In the Indian Ocean convection tends to follow the meridional propagation of SST anomalies. The response of the MJO to 1-5-day variability in the SST is through the charging and discharging mechanisms contributing to the atmospheric column moist static energy before and after peak MJO convection. Horizontal advection and surface fluxes show the largest sensitivity to SST perturbations.

SPEEDUP simulation of liquid waste batch processing. Revision 1

International Nuclear Information System (INIS)

Shannahan, K.L.; Aull, J.E.; Dimenna, R.A.

1994-01-01

The Savannah River Site (SRS) has accumulated radioactive hazardous waste for over 40 years during the time SRS made nuclear materials for the United States Department of Energy (DOE) and its predecessors. This waste is being stored as caustic slurry in a large number of 1 million gallon steel tanks, some of which were initially constructed in the early 1950's. SRS and DOE intend to clean up the Site and convert this waste into stable forms which then can be safely stored. The liquid waste will be separated into a partially decontaminated low-level and radioactive high-level waste in one feed preparation operation, In-Tank Precipitation. The low-level waste will be used to make a concrete product called saltstone in the Saltstone Facility, a part of the Defense Waste Processing Facility (DWPF). The concrete will be poured into large vaults, where it will be permanently stored. The high-level waste will be added to glass-formers and waste slurry solids from another feed preparation operation, Extended Sludge Processing. The mixture will then be converted to a stable borosilicate glass by a vitrification process that is the other major part of the DWPF. This glass will be poured into stainless steel canisters and sent to a temporary storage facility prior to delivery to a permanent underground storage site

Dynamic bioconversion mathematical modelling and simulation of urban organic waste co-digestion in continuously stirred tank reactor

DEFF Research Database (Denmark)

Fitamo, Temesgen Mathewos; Boldrin, Alessio; Dorini, G.

of this study was to apply a dynamic mathematical model to simulate the co-digestion of different urban organic wastes (UOW). The modelling was based on experimental activities, during which two reactors (R1, R2) were operated at hydraulic retention times (HRT) of 30, 20, 15, 10 days, in thermophilic conditions......The application of anaerobic digestion (AD) as process technology is increasing worldwide: the production of biogas, a versatile form of renewable energy, from biomass and organic waste materials allows mitigating greenhouse gas emission from the energy and transportation sectors while treating...... waste. However, the successful operation of AD processes is challenged by economic and technological issues. To overcome these barriers, mathematical modelling of the bioconversion process can provide support to develop strategies for controlling and optimizing the AD process. The objective...

Fire simulation in radioactive waste disposal and the radiation risk associated

International Nuclear Information System (INIS)

Domingos, Érica Nascimento

2018-01-01

An atmospheric dispersion of radioactive material is one of the possible consequences of an accident scenario in nuclear installations, radiative and radioactive waste deposit. Taking into account a possibility of this release of radioactive material into the atmosphere this work proposes a modeling of the atmospheric dispersion from a fire scenario in a deposit of radioactive waste of low and middle level of radiation varying the amount of inventory released in the fire. For this simulation was adopted the software of physical codes of medical health, the HotSpot Health Physics Codes which uses the Gaussian model to calculate an atmospheric dispersion based on the Pasquill atmospheric stability classes. This software calculates a total effective dose in relation to distance, such as a compromised dose in a list of specific organs, among them the lung, object of work study for calculating the risk of cancer associated with a low dose of radiation. The radiological risk calculation is held by the BEIR V model, Biological Effects of Ionizing Radiations, one of the models to estimate the relative risk of cancer induced by ionizing radiation. (author)

GRMHD Simulations of Visibility Amplitude Variability for Event Horizon Telescope Images of Sgr A*

Science.gov (United States)

Medeiros, Lia; Chan, Chi-kwan; Özel, Feryal; Psaltis, Dimitrios; Kim, Junhan; Marrone, Daniel P.; Sa¸dowski, Aleksander

2018-04-01

The Event Horizon Telescope will generate horizon scale images of the black hole in the center of the Milky Way, Sgr A*. Image reconstruction using interferometric visibilities rests on the assumption of a stationary image. We explore the limitations of this assumption using high-cadence disk- and jet-dominated GRMHD simulations of Sgr A*. We also employ analytic models that capture the basic characteristics of the images to understand the origin of the variability in the simulated visibility amplitudes. We find that, in all simulations, the visibility amplitudes for baselines oriented parallel and perpendicular to the spin axis of the black hole follow general trends that do not depend strongly on accretion-flow properties. This suggests that fitting Event Horizon Telescope observations with simple geometric models may lead to a reasonably accurate determination of the orientation of the black hole on the plane of the sky. However, in the disk-dominated models, the locations and depths of the minima in the visibility amplitudes are highly variable and are not related simply to the size of the black hole shadow. This suggests that using time-independent models to infer additional black hole parameters, such as the shadow size or the spin magnitude, will be severely affected by the variability of the accretion flow.

Numerical simulations of waste forms from the reprocessing of nuclear fuel; Numerische Simulationen von Abfallgebinden aus der Wiederaufarbeitung von Kernbrennstoffen

Energy Technology Data Exchange (ETDEWEB)

Schneider, Stephan