Fixation of calcined waste by bituminization or cementation

International Nuclear Information System (INIS)

Napravnik, J.; Kyrs, M.; Ditl, P.

1983-01-01

The overall concept is given of the combination of calcination with fixation into bitumen, cement etc. The design is shown of a calciner with the capacity of 10 L/h which was tested on real radioactive wastes for 2000 h. The geometrical and operating parameters of the apparatus have been optimized based on a statistical evaluation of the experiments. Wastes containing nitrates are calcined at 300-550 deg. C, yielding oxides. Wastes containing sulphates, carbonates, KMnO 4 , or borates are calcined at 150-330 deg. C, yielding soluble salts. The content of H 3 BO 4 and Na 2 B 4 O 7 and in some cases of sulphates in the calcinate retards hardening of the mixture with cement. Nitrates and detergents also interfere. The effect of the above components on the products mixed with bitumen is much less. Detergents can be decomposed at 200-300 deg. C; organic acids can be reacted with A1 salts to form insoluble substances lowering the leaching rate of Sr and Cs; small amounts of SiO 2 eliminate the effect of borates on cement hardening. The drawbacks of bituminization with bitumen emulsions are the complicated preparation of the emulsion, higher leaching rate of the product and low stability of the emulsion against breaking. The leachability was determined (1-50 days) of different products containing LWR wastes: 33% of concentrated waste in cement of calcination product stabilized with PVA exhibit approx. 8x10 - 3 g/cm 2 per day, 33% of calcine in cement approx. 3x10 - 3 ; 40% concentrate fixed with bitumen emulsion approx. 9x10 - 4 ; 50% calcine stabilized with PVA in bitumen, pilot-plant scale approx. 2x10 - 5 ; the same but on a laboratory scale approx. 1.10 - 5 . (author)

Remotely replaceable fuel and feed nozzles for the new waste calcining facility calciner vessel

International Nuclear Information System (INIS)

Fletcher, R.D.; Carter, J.A.; May, K.W.

1978-01-01

The development and testing of remotely replaceable fuel and feed nozzles for calcination of liquid radioactive wastes in the calciner vessel of the New Waste Calcining Facility being built at the Idaho National Engineering Laboratory is described. A complete fuel nozzle assembly was fabricated and tested at the Remote Maintenance Development Facility to evolve design refinements, identify required support equipment, and develop handling techniques. The design also provided for remote replacement of the nozzle support carriage and adjacent feed and fuel pipe loops using two pairs of master-slave manipulators

New Waste Calcining Facility (NWCF) Waste Streams

International Nuclear Information System (INIS)

K. E. Archibald

1999-01-01

This report addresses the issues of conducting debris treatment in the New Waste Calcine Facility (NWCF) decontamination area and the methods currently being used to decontaminate material at the NWCF

Design and development of a rotary calciner for radiochemical waste

International Nuclear Information System (INIS)

Pande, D.P.; Sutar, V.D.; Sengar, P.B.S.

1997-01-01

Present experience and knowledge in handling of radioactive waste has led to identification of major thrust areas in the development of the treatment processes. In order to reduce evaporation and volatility losses in the vitrification facility, it is advantageous to carry out evaporation and calcination steps in another equipment like rotary calciner. Efforts have been directed for the engineering development of a Rotary Ball Kiln calciner. This paper highlights the important design features of the Rotary Ball Kiln Calciner for the radioactive waste. In this work, an attempt has been made to systematically evaluate the influence of process and design parameters. The results obtained on calcination will provide a design basis and rational methodology for the optimum utilization of these processes and equipment for volume reduction and calcination of the liquid waste

Preparation of plutonium waste forms with ICPP calcined high-level waste

Energy Technology Data Exchange (ETDEWEB)

Staples, B.A.; Knecht, D.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); O`Holleran, T.P. [Argonne National Lab.-West, Idaho Falls, ID (United States)] [and others

1997-05-01

Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce{sup +4}) as a surrogate for plutonium (Pu{sup +4}) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study.

Preparation of plutonium waste forms with ICPP calcined high-level waste

International Nuclear Information System (INIS)

Staples, B.A.; Knecht, D.A.; O'Holleran, T.P.

1997-05-01

Glass and glass-ceramic forms developed for the immobilization of calcined high-level wastes generated by Idaho Chemical Processing Plant (ICPP) fuel reprocessing activities have been investigated for ability to immobilize plutonium and to simultaneously incorporate calcined waste as an anti-proliferation barrier. Within the forms investigated, crystallization of host phases result in an increased loading of plutonium as well as its incorporation into potentially more durable phases than the glass. The host phases were initially formed and characterized with cerium (Ce +4 ) as a surrogate for plutonium (Pu +4 ) and samarium as a neutron absorber for criticality control. Verification of the surrogate testing results were then performed replacing cerium with plutonium. All testing was performed with surrogate calcined high-level waste. The results of these tests indicated that a potentially useful host phase, based on zirconia, can be formed either by devitrification or solid state reaction in the glass studied. This phase incorporates plutonium as well as samarium and the calcined waste becomes part of the matrix. Its ease of formation makes it potentially useful in excess plutonium dispositioning. Other durable host phases for plutonium and samarium, including zirconolite and zircon have been formed from zirconia or alumina calcine through cold press-sintering techniques and hot isostatic pressing. Host phase formation experiments conducted through vitrification or by cold press-sintering techniques are described and the results discussed. Recommendations are given for future work that extends the results of this study

Talc-silicon glass-ceramic waste forms for immobilization of high- level calcined waste

International Nuclear Information System (INIS)

Vinjamuri, K.

1993-06-01

Talc-silicon glass-ceramic waste forms are being evaluated as candidates for immobilization of the high level calcined waste stored onsite at the Idaho Chemical Processing Plant. These glass-ceramic waste forms were prepared by hot isostatically pressing a mixture of simulated nonradioactive high level calcined waste, talc, silicon and aluminum metal additives. The waste forms were characterized for density, chemical durability, and glass and crystalline phase compositions. The results indicate improved density and chemical durability as the silicon content is increased

Process Design Concepts for Stabilization of High Level Waste Calcine

Energy Technology Data Exchange (ETDEWEB)

T. R. Thomas; A. K. Herbst

2005-06-01

The current baseline assumption is that packaging ¡§as is¡¨ and direct disposal of high level waste (HLW) calcine in a Monitored Geologic Repository will be allowed. The fall back position is to develop a stabilized waste form for the HLW calcine, that will meet repository waste acceptance criteria currently in place, in case regulatory initiatives are unsuccessful. A decision between direct disposal or a stabilization alternative is anticipated by June 2006. The purposes of this Engineering Design File (EDF) are to provide a pre-conceptual design on three low temperature processes under development for stabilization of high level waste calcine (i.e., the grout, hydroceramic grout, and iron phosphate ceramic processes) and to support a down selection among the three candidates. The key assumptions for the pre-conceptual design assessment are that a) a waste treatment plant would operate over eight years for 200 days a year, b) a design processing rate of 3.67 m3/day or 4670 kg/day of HLW calcine would be needed, and c) the performance of waste form would remove the HLW calcine from the hazardous waste category, and d) the waste form loadings would range from about 21-25 wt% calcine. The conclusions of this EDF study are that: (a) To date, the grout formulation appears to be the best candidate stabilizer among the three being tested for HLW calcine and appears to be the easiest to mix, pour, and cure. (b) Only minor differences would exist between the process steps of the grout and hydroceramic grout stabilization processes. If temperature control of the mixer at about 80„aC is required, it would add a major level of complexity to the iron phosphate stabilization process. (c) It is too early in the development program to determine which stabilizer will produce the minimum amount of stabilized waste form for the entire HLW inventory, but the volume is assumed to be within the range of 12,250 to 14,470 m3. (d) The stacked vessel height of the hot process vessels

Structural Integrity Program for the Calcined Solids Storage Facilities at the Idaho Nuclear Technology and Engineering Center

International Nuclear Information System (INIS)

Bryant, J.W.; Nenni, J.A.

2003-01-01

This report documents the activities of the structural integrity program at the Idaho Nuclear Technology and Engineering Center relevant to the high-level waste Calcined Solids Storage Facilities and associated equipment, as required by DOE M 435.1-1, ''Radioactive Waste Management Manual.'' Based on the evaluation documented in this report, the Calcined Solids Storage Facilities are not leaking and are structurally sound for continued service. Recommendations are provided for continued monitoring of the Calcined Solids Storage Facilities

Structural Integrity Program for the Calcined Solids Storage Facilities at the Idaho Nuclear Technology and Engineering Center

International Nuclear Information System (INIS)

Jeffrey Bryant

2008-01-01

This report documents the activities of the structural integrity program at the Idaho Nuclear Technology and Engineering Center relevant to the high-level waste Calcined Solids Storage Facilities and associated equipment, as required by DOE M 435.1-1, 'Radioactive Waste Management Manual'. Based on the evaluation documented in this report, the Calcined Solids Storage Facilities are not leaking and are structurally sound for continued service. Recommendations are provided for continued monitoring of the Calcined Solids Storage Facilities

Calcine production and management

International Nuclear Information System (INIS)

Dickey, B.R.; Hogg, G.W.; Berreth, J.R.

1979-01-01

The process technology related to calcination of power reactor wastes is summarized. The primary calcination processes developed are spray calcination, fluidized-bed calcination, and rotary kiln calcination. Calcines from the spray calciner and rotary kiln are fed directly to a glassification process. The fluidized-bed product can either be fed to a waste form conversion process or stored. The major process steps for calcinations are feed preparation, calcination and product handling, and off-gas cleanup. Feed systems for the three processes are basically similar. Gravity flow and pump pressurized systems have been used successfully. The major problems are fatigue failure of feed valve bellows, plugging by undissolved solids, and calibration of flowmeters. Process heat input is by electrical resistance heating for the spray and rotary kiln calciners and in-bed combustion or in-bed heat exchange for the fluidized-bed system. Low-melting solids which can cause scaling or solids agglomeration in any of the processes is a major calcination problem; however, feed blending, process operating conditions, and equipment design have successfully controlled solids agglomeration. Primary off-gas cleanup devices for particulates are cyclones, sintered metal filters, venturi scrubbers, and HEPA filters. Scrubbers, condensers, and solid adsorbents are used successfully for volatile ruthenium removal. The years of pilot-plant and plant-scale calcination testing and operation of the three systems have shown that reactor wastes can be calcined safely and practically. 11 figures, 2 tables

Formulation Efforts for Direct Vitrification of INEEL Blend Calcine Waste Simulate: Fiscal Year 2000

Energy Technology Data Exchange (ETDEWEB)

Crum, Jarrod V.; Vienna, John D.; Peeler, David K.; Reamer, I. A.

2001-03-30

This report documents the results of glass formulation efforts for Idaho National Engineering and Environmental Laboratory (INEEL) high level waste (HWL) calcine. Two waste compositions were used during testing. Testing started by using the Run 78 calcine composition and switched to simulated Blend calcine composition when it became available. The goal of the glass formulation efforts was to develop a frit composition that will accept higher waste loading that satisfies the glass processing and product acceptance constraints. 1. Melting temperature of 1125 ? 25?C 2. Viscosity between 2 and 10 Pa?s at the melting temperature 3. Liquidus temperature at least 100?C below the melting temperature 4. Normalized release of B, Li and Na each below 1 g/m2 (per ASTM C 1285-97) Glass formulation efforts tested several frit compositions with variable waste loadings of Run 78 calcine waste simulant. Frit 107 was selected as the primary candidate for processing since it met all process and performance criteria up to 45 mass% waste loading. When the simulated Blend calcine waste composition became available Frits 107 and 108 compositions were retested and again Frit 107 remained the primary candidate. However, both frits suffered a decrease in waste loading when switching from the Run 78 calcine to simulated Blend calcine waste composition. This was due to increase concentrations of both F and Al2O3 along with a decrease in CaO and Na2O in the simulate Blend calcine waste all of which have strong impacts on the glass properties that limit waste loading of this type of waste.

Spring 2009 Semiannual (III.H. and I.U.) Report for the HWMA/RCRA Post-Closure Permit for the INTEC Waste Calcining Facility at the INL Site

International Nuclear Information System (INIS)

Boehmer, Ann M.

2009-01-01

The Waste Calcining Facility is located at the Idaho Nuclear Technology and Engineering Center. In 1999, the Waste Calcining Facility was closed under and approved Hazardous Waste Management Act/Resource Conservation and Recovery Act Closure plan. Vessels and spaces were grouted and then covered with a concrete cap. This permit sets forth procedural requirements for groundwater characterization and monitoring, maintenance, and inspections of the Waste Calcining Facility to ensure continued protection of human health and the environment.

Spring 2009 Semiannual (III.H. and I.U.) Report for the HWMA/RCRA Post-Closure Permit for the INTEC Waste Calcining Facility at the INL Site

Energy Technology Data Exchange (ETDEWEB)

Boehmer, Ann M.

2009-05-31

The Waste Calcining Facility is located at the Idaho Nuclear Technology and Engineering Center. In 1999, the Waste Calcining Facility was closed under and approved Hazardous Waste Management Act/Resource Conservation and Recovery Act Closure plan. Vessels and spaces were grouted and then covered with a concrete cap. This permit sets forth procedural requirements for groundwater characterization and monitoring, maintenance, and inspections of the Waste Calcining Facility to ensure continued protection of human health and the environment.

Microwave energy for post-calcination treatment of high-level nuclear wastes

International Nuclear Information System (INIS)

Gombert, D.; Priebe, S.J.; Berreth, J.R.

1980-01-01

High-level radioactive wastes generated from nuclear fuel reprocessing require treatment for effective long-term storage. Heating by microwave energy is explored in processing of two possible waste forms: (1) drying of a pelleted form of calcined waste; and (2) vitrification of calcined waste. It is shown that residence times for these processes can be greatly reduced when using microwave energy rather than conventional heating sources, without affecting product properties. Compounds in the waste and in the glass frit additives couple very well with the 2.45 GHz microwave field so that no special microwave absorbers are necessary

Chemistry of proposed calcination/dissolution processing of Hanford Site tank wastes

International Nuclear Information System (INIS)

Delegard, C.H.

1995-01-01

Plans exist to separate radioactive waste stored in underground tanks at the US Department of Energy's Hanford Site in south central Washington State into low-level and high-level fractions, and to immobilize the separate fractions in high-integrity vitrified forms for long-term disposal. Calcination with water dissolution has been proposed as a possible treatment for achieving low/high-level separation. Chemistry development activities conducted since 1992 with simulated and genuine tank waste show that calcination/dissolution destroys organic carbon and converts nitrate and nitrite to hydroxide and benign offgases. The process also dissolves significant quantities of bulk chemicals (aluminum, chromium, and phosphate), allowing their redistribution from the high-level to the low-level fraction. Present studies of the chemistry of calcination/dissolution processing of genuine wastes, conducted in the period October 1993 to September 1994, show the importance of sodium fluoride phosphate double salt in controlling phosphate dissolution. Peptization of waste solids is of concern if extensive washing occurs. Strongly oxidizing conditions imposed by calcination reactions were found to convert transition metals to soluble anions in the order chromate > manganate > > ferrate. In analogy with manganese behavior, plutonium dissolution, presumably by oxidation to more soluble anionic species, also occurs by calcination/dissolution. Methods to remove plutonium from the product low-level solution stream must be developed

Retrieval System for Calcined Waste for the Idaho Cleanup Project - 12104

Energy Technology Data Exchange (ETDEWEB)

Eastman, Randy L.; Johnston, Beau A.; Lower, Danielle E. [CH2M-WG Idaho, LLC. The Idaho Cleanup Project at the Idaho National Laboratory (United States)

2012-07-01

This paper describes the conceptual approach to retrieve radioactive calcine waste, hereafter called calcine, from stainless steel storage bins contained within concrete vaults. The retrieval system will allow evacuation of the granular solids (calcine) from the storage bins through the use of stationary vacuum nozzles. The nozzles will use air jets for calcine fluidization and will be able to rotate and direct the fluidization or displacement of the calcine within the bin. Each bin will have a single retrieval system installed prior to operation to prevent worker exposure to the high radiation fields. The addition of an articulated camera arm will allow for operations monitoring and will be equipped with contingency tools to aid in calcine removal. Possible challenges (calcine bridging and rat-holing) associated with calcine retrieval and transport, including potential solutions for bin pressurization, calcine fluidization and waste confinement, are also addressed. The Calcine Disposition Project has the responsibility to retrieve, treat, and package HLW calcine. The calcine retrieval system has been designed to incorporate the functions and technical characteristics as established by the retrieval system functional analysis. By adequately implementing the highest ranking technical characteristics into the design of the retrieval system, the system will be able to satisfy the functional requirements. The retrieval system conceptual design provides the means for removing bulk calcine from the bins of the CSSF vaults. Top-down vacuum retrieval coupled with an articulating camera arm will allow for a robust, contained process capable of evacuating bulk calcine from bins and transporting it to the processing facility. The system is designed to fluidize, vacuum, transport and direct the calcine from its current location to the CSSF roof-top transport lines. An articulating camera arm, deployed through an adjacent access riser, will work in conjunction with the

Fluidized bed system for calcination of high level radioactive waste

Energy Technology Data Exchange (ETDEWEB)

Pande, D P; Prasad, T L; Yadgiri, N K; Theyyunni, T K [Process Engineering and Systems Development Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

During the operation of nuclear facilities significant quantities of radiochemical liquid effluents of different concentrations and varying chemical compositions are generated. These effluents contain activated radionuclides, corrosion products and fission products. The advantage of feeding the waste in solid form into the vitrifying equipment are multifold. Efforts are therefore made in many countries to calcine the high level waste, and obtain waste in the oxide form before the same is mixed with glass forming additives and fed into the melter unit. An experimental rig for fluidized bed calcination is constructed for carrying out the detailed investigation of this process, in order to adopt the same for plant scale application. To achieve better gas-solid contact and avoid raining down of solids, a distributor of bubble cap type was designed. A review of existing experience at various laboratories and design of new experimental facility for development of calciners are given. (author). 11 refs., 5 figs.

Waste Calcining Facility remote inspection report

International Nuclear Information System (INIS)

Patterson, M.W.; Ison, W.M.

1994-08-01

The purpose of the Waste Calcining Facility (WCF) remote inspections was to evaluate areas in the facility which are difficult to access due to high radiation fields. The areas inspected were the ventilation exhaust duct, waste hold cell, adsorber manifold cell, off-gas cell, calciner cell and calciner vessel. The WCF solidified acidic, high-level mixed waste generated during nuclear fuel reprocessing. Solidification was accomplished through high temperature oxidation and evaporation. Since its shutdown in 1981, the WCFs vessels, piping systems, pumps, off-gas blowers and process cells have remained contaminated. Access to the below-grade areas is limited due to contamination and high radiation fields. Each inspection technique was tested with a mock-up in a radiologically clean area before the equipment was taken to the WCF for the actual inspection. During the inspections, essential information was obtained regarding the cleanliness, structural integrity, in-leakage of ground water, indications of process leaks, indications of corrosion, radiation levels and the general condition of the cells and equipment. In general, the cells contain a great deal of dust and debris, as well as hand tools, piping and miscellaneous equipment. Although the building appears to be structurally sound, the paint is peeling to some degree in all of the cells. Cracking and spalling of the concrete walls is evident in every cell, although the east wall of the off-gas cell is the worst. The results of the completed inspections and lessons learned will be used to plan future activities for stabilization and deactivation of the facility. Remote clean-up of loose piping, hand tools, and miscellaneous debris can start immediately while information from the inspections is factored into the conceptual design for deactivating the facility

Determination of the Rate of Formation of Hydroceramic Waste Forms made with INEEL Calcined Wastes

Energy Technology Data Exchange (ETDEWEB)

Barry Scheetz; Johnson Olanrewaju

2001-10-15

The formulation, synthesis, characterization and hydration kinetics of hydroceramic waste forms designed as potential hosts for existing INEEL calcine high-level wastes have been established as functions of temperature and processing time. Initial experimentations were conducted with several aluminosilicate pozzolanic materials, ranging from fly ash obtained from various power generating coal and other combustion industries to reactive alumina, natural clays and ground bottled glass powders. The final selection criteria were based on the ease of processing, excellent physical properties and chemical durability (low-leaching) determined from the PCT test produced in hydroceramic. The formulation contains vermiculite, Sr(NO32), CsC1, NaOH, thermally altered (calcined natural clay) and INEEL simulated calcine high-level nuclear wastes and 30 weight percent of fluorinel blend calcine and zirconia calcine. Syntheses were carried out at 75-200 degree C at autogeneous water pressure (100% relative humidity) at various time intervals. The resulting monolithic compact products were hard and resisted breaking when dropped from a 5 ft height. Hydroceramic host mixed with fluorinel blend calcine and processed at 75 degree C crumbled into rice hull-side grains or developed scaly flakes. However, the samples equally possessed the same chemical durability as their unbroken counterparts. Phase identification by XRD revealed that hydroceramic host crystallized type zeolite at 75-150 degree C and NaP1 at 175-200 degree C in addition to the presence of quartz phase originating from the clay reactant. Hydroceramic host mixed with either fluorinel blend calcine or zirconia calcine crystallized type A zeolite at 75-95 degree C, formed a mixture of type A zeolite and hydroxysodalite at 125-150 degree C and hydroxysodalite at 175-200 degree C. Quartz, calcium fluoride and zirconia phases from the clay reactant and the two calcine wastes were also detected. The PCT test solution

Volatilities of ruthenium, iodine, and technetium on calcining fission product nitrate wastes

International Nuclear Information System (INIS)

Rimshaw, S.J.; Case, F.N.

1980-01-01

Various high-level nitrate wastes were subjected to formic acid denitration. Formic acid reacts with the nitrate anion to yield noncondensable, inert gases according to the following equation: 4 HCOOH + 2 HNO 3 → N 2 O + 4 CO 2 + 5 H 2 O. These gases can be scrubbed free of 106 Ru, 131 I, and 99 Tc radioactivities prior to elimination from the plant by passage through HEPA filters. The formation of deleterious NO/sub x/ is avoided. Moreover, formic acid reduces ruthenium to a lower valence state with a sharp reduction in RuO 4 volatility during subsequent calcination of the pretreated waste. It is shown that a minimum of 3% of RuO 4 in an off-gas stream reacts with Davison silica gel (Grade 40) to give a fine RuO 2 aerosol having a particle size of 0.5 μ. This RuO 2 aerosol passes through water or weak acid scrub solutions but is trapped by a caustic scrub solution. Iodine volatilizes almost completely on calcining an acidic waste, and the iodine volatility increases with increasing calcination temperature. On calcining an alkaline sodium nitrate waste the iodine volatility is about an order of magnitude lower, with a relatively low iodine volatility of 0.39% at a calcination temperature of 250 0 C and a moderate volatility of 9.5% at 600 0 C. Volatilities of 99 Tc were generally 0 C. Data are presented to indicate that 99 Tc concentrates in the alkaline sodium nitrate supernatant waste, with approx. 10 mg 99 Tc being associated with each curie of 137 Cs present in the waste. It is shown that lutidine (2,4 dimethyl-pyridine) extracts Tc(VII) quantitatively from alkaline supernatant wastes. The distribution coefficient (K/sub D/) for Tc(VII) going into the organic phase in the above system is 102 for a simulated West Valley waste and 191 for a simulated Savannah River Plant (SRP) waste

Proposed Atomic Energy of Canada Ltd. 99Mo waste calcination process

International Nuclear Information System (INIS)

Ramey, D.W.; Haas, P.A.; Malkemus, D.W.; McGinnis, C.P.; Meyers, E.S.; Patton, B.D.; Birdwell, J.F.; Jubin, R.T.; Coltharp, K.A.

1994-10-01

Atomic Energy of Canada Limited (AECL), at its Chalk River Laboratory, generates from 3000 to 5000 L/year of high-level fissile waste solution from the production of 99 Mo. In this Mo process, highly enriched uranium (93 wt % 235 U, total uranium basis) contained in uranium-aluminum alloy target rods is irradiated to produce the 99 Mo product. The targets are removed from the reactor and dissolved in a mercury nitrate-catalyzed reaction with nitric acid. The 99 Mo product is then recovered by passing the solution through an alumina (Al 2 O 3 ) column. During discussions with personnel from the Oak Ridge National Laboratory (ORNL) on September 10, 1992, the ORNL-developed technology formerly applied to the solidification of aqueous uranium waste (Consolidated Edison Uranium Solidification Program or CEUSP) was judged potentially applicable to the AECL 99 Mo waste. Under a Work-for-Others contract (no. ERD-92-1132), which began May 24, 1993, ORNL was tasked to determine the feasibility of applying the CEUSP (or a similar) calcination process to solidify AECL's 99 Mo waste for > 30 years of safe dry storage. This study was to provide sufficient detailed information on the applicability of a CEUSP-type waste solidification process to allow AECL to select the process which best suited its needs. As with the CEUSP process, evaporation of the waste and a simultaneously partial destruction of acid by reaction with formaldehyde followed by in situ waste can thermal denitration waste was chosen as the best means of solidification. Unlike the CEUSP material, the 99 Mo waste has a considerable number of problem volatile and semivolatile constituents which must be recovered in the off-gas treatment system. Mercury removal before calcination was seen as the best option

Pyrochemical separation of radioactive components from inert materials in ICPP high-level calcined waste

International Nuclear Information System (INIS)

Del Debbio, J.A.; Nelson, L.O.; Todd, T.A.

1995-05-01

Since 1963, calcination of aqueous wastes from reprocessing of DOE-owned spent nuclear fuels has resulted in the accumulation of approximately 3800 m 3 of high-level waste (HLW) at the Idaho Chemical Processing Plant (ICPP). The waste is in the form of a granular solid called calcine and is stored on site in stainless steel bins which are encased in concrete. Due to the leachability of 137 Cs and 90 Sr and possibly other radioactive components, the calcine is not suitable for final disposal. Hence, a process to immobilize calcine in glass is being developed. Since radioactive components represent less than 1 wt % of the calcine, separation of actinides and fission products from inert components is being considered to reduce the volume of HLW requiring final disposal. Current estimates indicate that compared to direct vitrification, a volume reduction factor of 10 could result in significant cost savings. Aqueous processes, which involve calcine dissolution in nitric acid followed by separation of actinide and fission products by solvent extraction and ion exchange methods, are being developed. Pyrochemical separation methods, which generate small volumes of aqueous wastes and do not require calcine dissolution, have been evaluated as alternatives to aqueous processes. This report describes three proposed pyrochemical flowsheets and presents the results of experimental studies conducted to evaluate their feasibility. The information presented is a consolidation of three reports, which should be consulted for experimental details

Determination of the Rate of Formation of Hydroceramic Waste Forms made with INEEL Calcined Wastes; FINAL

International Nuclear Information System (INIS)

Barry Scheetz; Johnson Olanrewaju

2001-01-01

The formulation, synthesis, characterization and hydration kinetics of hydroceramic waste forms designed as potential hosts for existing INEEL calcine high-level wastes have been established as functions of temperature and processing time. Initial experimentations were conducted with several aluminosilicate pozzolanic materials, ranging from fly ash obtained from various power generating coal and other combustion industries to reactive alumina, natural clays and ground bottled glass powders. The final selection criteria were based on the ease of processing, excellent physical properties and chemical durability (low-leaching) determined from the PCT test produced in hydroceramic. The formulation contains vermiculite, Sr(NO32), CsC1, NaOH, thermally altered (calcined natural clay) and INEEL simulated calcine high-level nuclear wastes and 30 weight percent of fluorinel blend calcine and zirconia calcine. Syntheses were carried out at 75-200 degree C at autogeneous water pressure (100% relative humidity) at various time intervals. The resulting monolithic compact products were hard and resisted breaking when dropped from a 5 ft height. Hydroceramic host mixed with fluorinel blend calcine and processed at 75 degree C crumbled into rice hull-side grains or developed scaly flakes. However, the samples equally possessed the same chemical durability as their unbroken counterparts. Phase identification by XRD revealed that hydroceramic host crystallized type zeolite at 75-150 degree C and NaP1 at 175-200 degree C in addition to the presence of quartz phase originating from the clay reactant. Hydroceramic host mixed with either fluorinel blend calcine or zirconia calcine crystallized type A zeolite at 75-95 degree C, formed a mixture of type A zeolite and hydroxysodalite at 125-150 degree C and hydroxysodalite at 175-200 degree C. Quartz, calcium fluoride and zirconia phases from the clay reactant and the two calcine wastes were also detected. The PCT test solution

Ninth Processing Campaign in the Waste Calcining Facility

International Nuclear Information System (INIS)

Childs, K.F.; Donovan, R.I.; Swenson, M.C.

1982-04-01

This report discusses the Ninth (and final) Processing Campaign at the Waste Calcining Facility. Several processing interruptions were experienced during this campaign and the emphasis of this report is on process and equipment performance with operating problems and corrective actions discussed in detail

Optimization of Calcine Blending During Retrieval from Binsets

International Nuclear Information System (INIS)

Nelson, Lee Orville; Mohr, Charles Milton; Taylor, Dean Dalton

2000-01-01

This report documents a study performed during advanced feasibility studies for the INTEC Technology Development Facility (ITDF). The study was commissioned to provide information about functional requirements for the ITDF related to development of equipment and procedures for retrieving radioactive calcine from binset storage at the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). Calcine will be retrieved prior to treating it for permanent disposal in a national repository for high level waste. The objective this study was to estimate the degree of homogenization of the calcine that might be achieved through optimized retrieval and subsequent blending. Such homogenization has the potential of reducing the costs for treatment of the calcine and for qualifying of the final waste forms for acceptance at the repository. Results from the study indicate that optimized retrieval and blending can reduce the peak concentration variations of key components (Al, Zr, F) in blended batches of retrieved calcine. During un-optimized retrieval these variations are likely to be 81-138% while optimized retrieval can reduce them to the 5-10% range

Design and performance of a full-scale spray calciner for nonradioactive high-level-waste-vitrification studies

International Nuclear Information System (INIS)

Miller, F.A.

1981-06-01

In the spray calcination process, liquid waste is spray-dried in a heated-wall spray dryer (termed a spray calciner), and then it may be combined in solid form with a glass-forming frit. This mixture is then melted in a continuous ceramic melter or in an in-can melter. Several sizes of spray calciners have been tested at PNL- laboratory scale, pilot scale and full scale. Summarized here is the experience gained during the operation of PNL's full-scale spray calciner, which has solidified approx. 38,000 L of simulated acid wastes and approx. 352,000 L of simulated neutralized wastes in 1830 h of processing time. Operating principles, operating experience, design aspects, and system descriptions of a full-scale spray calciner are discussed. Individual test run summaries are given in Appendix A. Appendices B and C are studies made by Bechtel Inc., under contract by PNL. These studies concern, respectively, feed systems for the spray calciner process and a spray calciner vibration analysis. Appendix D is a detailed structural analysis made at PNL of the spray calciner. These appendices are included in the report to provide a complete description of the spray calciner and to include all major studies made concerning PNL's full-scale spray calciner

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...

Design criteria for the new waste calcining facility at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Anderson, F.H.; Bingham, G.E.; Buckham, J.A.; Dickey, B.R.; Slansky, C.M.; Wheeler, B.R.

1976-01-01

The New Waste Calcining Facility (NWCF) at the Idaho Chemical Processing Plant (ICPP) is being built to replace the existing fluidized-bed, high-level waste calcining facility (WCF). Performance of the WCF is reviewed, equipment failures in WCF operation are examined, and pilot-plant studies on calciner improvements are given in relation to NWCF design. Design features of the NWCF are given with emphasis on process and equipment improvements. A major feature of the NWCF is the use of remote maintenance facilities for equipment with high maintenance requirements, thereby reducing personnel exposures during maintenance and reducing downtime resulting from plant decontamination. The NWCF will have a design net processing rate of 11.36 m 3 of high-level waste per day, and will incorporate in-bed combustion of kerosene for heating the fluidized bed calciner. The off-gas cleaning system will be similar to that for the WCF

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.

Biodiesel production from waste cooking oil using calcined scallop shell as catalyst

International Nuclear Information System (INIS)

Sirisomboonchai, Suchada; Abuduwayiti, Maidinamu; Guan, Guoqing; Samart, Chanatip; Abliz, Shawket; Hao, Xiaogang; Kusakabe, Katsuki; Abudula, Abuliti

2015-01-01

Highlights: • Calcined scallop shell was used as low-cost and effective catalyst for biodiesel production. • BDF yield from waste cooking oil reached 86% at 65 °C with a catalyst loading amount of 5 wt%. • Calcined scallop shell showed good reusability. • Calcium glyceroxide played an important role on the reusability of calcined scallop shell. • Water in the waste cooking oil had negative effect on the catalytic activity of calcined scallop shell. - Abstract: Transesterification of waste cooking oil (WCO) and methanol by using calcined scallop shell (CSS) as catalyst was carried out in a closed system for biodiesel fuel (BDF) production. It is found that the optimum calcination temperature for the preparation of CSS was 1000 °C. The effects of transesterification temperature, reaction time, methanol/oil molar ratio and catalyst loading amount on the BDF yield were investigated. Compared with the commercial CaO, CSS showed higher catalytic activity and the BDF yield reached 86% at 65 °C with a catalyst loading amount of 5 wt% (WCO basis) and a reaction time of 2 h. The catalyst was reused for 5 cycles whilst the BDF yield decreased 23%. It is found that CaO in CSS was transferred to calcium glyceroxide after the transesterification reaction, and calcium glyceroxide also showed good catalytic activity and reusability. Furthermore, Water content in WCO had negative effect on BDF yield. It is found that BDF yield reduced 15% due to the occurring of saponification when the water content was increased from 0.64% to 2.48%. It is expected that CCS can be used as an alternative and cheap catalyst for the biodiesel production

Calcined solids storage facility closure study

International Nuclear Information System (INIS)

Dahlmeir, M.M.; Tuott, L.C.; Spaulding, B.C.

1998-02-01

The disposal of radioactive wastes now stored at the Idaho National Engineering and Environmental Laboratory is currently mandated under a open-quotes Settlement Agreementclose quotes (or open-quotes Batt Agreementclose quotes) between the Department of Energy and the State of Idaho. Under this agreement, all high-level waste must be treated as necessary to meet the disposal criteria and disposed of or made road ready to ship from the INEEL by 2035. In order to comply with this agreement, all calcined waste produced in the New Waste Calcining Facility and stored in the Calcined Solids Facility must be treated and disposed of by 2035. Several treatment options for the calcined waste have been studied in support of the High-Level Waste Environmental Impact Statement. Two treatment methods studied, referred to as the TRU Waste Separations Options, involve the separation of the high-level waste (calcine) into TRU waste and low-level waste (Class A or Class C). Following treatment, the TRU waste would be sent to the Waste Isolation Pilot Plant (WIPP) for final storage. It has been proposed that the low-level waste be disposed of in the Tank Farm Facility and/or the Calcined Solids Storage Facility following Resource Conservation and Recovery Act closure. In order to use the seven Bin Sets making up the Calcined Solids Storage Facility as a low-level waste landfill, the facility must first be closed to Resource Conservation and Recovery Act (RCRA) standards. This study identifies and discusses two basic methods available to close the Calcined Solids Storage Facility under the RCRA - Risk-Based Clean Closure and Closure to Landfill Standards. In addition to the closure methods, the regulatory requirements and issues associated with turning the Calcined Solids Storage Facility into an NRC low-level waste landfill or filling the bin voids with clean grout are discussed

Calcined solids storage facility closure study

Energy Technology Data Exchange (ETDEWEB)

Dahlmeir, M.M.; Tuott, L.C.; Spaulding, B.C. [and others

1998-02-01

The disposal of radioactive wastes now stored at the Idaho National Engineering and Environmental Laboratory is currently mandated under a {open_quotes}Settlement Agreement{close_quotes} (or {open_quotes}Batt Agreement{close_quotes}) between the Department of Energy and the State of Idaho. Under this agreement, all high-level waste must be treated as necessary to meet the disposal criteria and disposed of or made road ready to ship from the INEEL by 2035. In order to comply with this agreement, all calcined waste produced in the New Waste Calcining Facility and stored in the Calcined Solids Facility must be treated and disposed of by 2035. Several treatment options for the calcined waste have been studied in support of the High-Level Waste Environmental Impact Statement. Two treatment methods studied, referred to as the TRU Waste Separations Options, involve the separation of the high-level waste (calcine) into TRU waste and low-level waste (Class A or Class C). Following treatment, the TRU waste would be sent to the Waste Isolation Pilot Plant (WIPP) for final storage. It has been proposed that the low-level waste be disposed of in the Tank Farm Facility and/or the Calcined Solids Storage Facility following Resource Conservation and Recovery Act closure. In order to use the seven Bin Sets making up the Calcined Solids Storage Facility as a low-level waste landfill, the facility must first be closed to Resource Conservation and Recovery Act (RCRA) standards. This study identifies and discusses two basic methods available to close the Calcined Solids Storage Facility under the RCRA - Risk-Based Clean Closure and Closure to Landfill Standards. In addition to the closure methods, the regulatory requirements and issues associated with turning the Calcined Solids Storage Facility into an NRC low-level waste landfill or filling the bin voids with clean grout are discussed.

Optimization of Calcine Blending During Retrieval From Binsets

International Nuclear Information System (INIS)

Taylor, D.D.; Mohr, C.M.; Nelson, L.O.

2000-01-01

This report documents a study performed during advanced feasibility studies for the INTEC Technology Development Facility (ITDF). The study was commissioned to provide information about functional requirements for the ITDF related to development of equipment and procedures for retrieving radioactive calcine from binset storage at the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). Calcine will be retrieved prior to treating it for permanent disposal in a national repository for high level waste. The objective this study was to estimate the degree of homogenization of the calcine that might be achieved through optimized retrieval and subsequent blending. Such homogenization has the potential of reducing the costs for treatment of the calcine and for qualifying of the final waste forms for acceptance at the repository. Results from the study indicate that optimized retrieval and blending can reduce the peak c oncentration variations of key components (Al, Zr, F) in blended batches of retrieved calcine. During un-optimized retrieval these variations are likely to be 81-138% while optimized retrieval can reduce them to the 5-10% range

Radiant-heat spray-calcination process for the solid fixation of radioactive waste. Part 1, Non-radioactive pilot unit

Energy Technology Data Exchange (ETDEWEB)

Allemann, R.T.; Johnson, B.M. Jr.

1960-11-14

The fixation of radioactive waste in a stable solid media by means of calcination of these aqueous solutions has been the subject of considerable-effort throughout the U. S. Atomic Energy Commission and by atomic energy organizations in other countries. Several methods of doing this on a continuous or semi-continuous basis have been devised, and a fev have been demonstrated to be feasible for the handling of non-radioactive, or low-activity, simulated wastes. Notable among methods currently under development are: (a) batch-operated pot calcination of waste generated from reprocessing stainless steel clad fuel elements (Darex process) and Purex waste, (b) combination rotary kiln and ball mill calcination of aluminum nitrate (TBP-25 and Redox process), and (c) fluidized bed calcination of TBP-25 and Purex wastes. Although a considerable amount of engineering experience has been obtained on the calcination of dissolved salts in a fluidized bed, and the other methods have been the subjects of a great deal of study, none of them have been developed to-the extent which would rule out the desirability of further investigation of other possible methods of calcination.

Effect of aluminum and silicon reactants and process parameters on glass-ceramic waste form characteristics for immobilization of high-level fluorinel-sodium calcined waste

International Nuclear Information System (INIS)

Vinjamuri, K.

1993-06-01

In this report, the effects of aluminum and silicon reactants, process soak time and the initial calcine particle size on glass-ceramic waste form characteristics for immobilization of the high-level fluorinel-sodium calcined waste stored at the Idaho Chemical Processing Plant (ICPP) are investigated. The waste form characteristics include density, total and normalized elemental leach rates, and microstructure. Glass-ceramic waste forms were prepared by hot isostatically pressing (HIPing) a pre-compacted mixture of pilot plant fluorinel-sodium calcine, Al, and Si metal powders at 1050 degrees C, 20,000 psi for 4 hours. One of the formulations with 2 wt % Al was HIPed for 4, 8, 16 and 24 hours at the same temperature and pressure. The calcine particle size range include as calcined particle size smaller than 600 μm (finer than -30 mesh, or 215 μm Mass Median Diameter, MMD) and 180 μm (finer than 80 mesh, or 49 μm MMD)

Mobile calcination and cementation unit for solidification of concentrated radioactive wastes

International Nuclear Information System (INIS)

Napravnik, J.; Sazavsky, P.; Skaba, V.; Skvarenina, R.; Ditl, P.

1985-01-01

Mobile experimental unit MESA-1 was developed and manufactured for processing radioactive concentrates by direct cementation. The unit is mainly designed for processing low-level liquid wastes from nuclear power plants and other nuclear installations, in which the level of radioactivity does not exceed 10 10 Bq/m 3 , the salt content of liquid solutions does not exceed 500 kg/m 3 and the maximum amount of boric acid is 130 kg/m 3 . The equipment is built into three modules which may be assembled and dismantled in a short time and transported separately. The unit without the calciner module was tested in non-radioactive mode and in operation with actual radioactive wastes from the V-1 nuclear power plant. The course and results of the tests are described in detail. All project design values were achieved, a total of 18 dm 3 model solutions were processed and 1 m 3 of actual wastes with a salt content of 450 kg/m 3 . The test showed that with regard to the radiation level reached it will be necessary in the process of calcination to increase the shielding of certain exposed points. The calciner module is being assembled for completion. (Z.M.)

Optimization of Calcine Blending During Retrieval From Binsets; TOPICAL

International Nuclear Information System (INIS)

Taylor, D.D.; Mohr, C.M.; Nelson, L.O.

2000-01-01

This report documents a study performed during advanced feasibility studies for the INTEC Technology Development Facility (ITDF). The study was commissioned to provide information about functional requirements for the ITDF related to development of equipment and procedures for retrieving radioactive calcine from binset storage at the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). Calcine will be retrieved prior to treating it for permanent disposal in a national repository for high level waste. The objective this study was to estimate the degree of homogenization of the calcine that might be achieved through optimized retrieval and subsequent blending. Such homogenization has the potential of reducing the costs for treatment of the calcine and for qualifying of the final waste forms for acceptance at the repository. Results from the study indicate that optimized retrieval and blending can reduce the peak c oncentration variations of key components (Al, Zr, F) in blended batches of retrieved calcine. During un-optimized retrieval these variations are likely to be 81-138% while optimized retrieval can reduce them to the 5-10% range

Bin Set 1 Calcine Retrieval Feasibility Study

International Nuclear Information System (INIS)

Adams, R.D.; Berry, S.M.; Galloway, K.J.; Langenwalter, T.A.; Lopez, D.A.; Noakes, C.M.; Peterson, H.K.; Pope, M.I.; Turk, R.J.

1999-01-01

At the Department of Energy's Idaho Nuclear Technology and Engineering Center, as an interim waste management measure, both mixed high-level liquid waste and sodium bearing waste have been solidified by a calculation process and are stored in the Calcine Solids Storage Facilities. This calcined product will eventually be treated to allow final disposal in a national geologic repository. The Calcine Solids Storage Facilities comprise seven ''bit sets.'' Bin Set 1, the first to be constructed, was completed in 1959, and has been in service since 1963. It is the only bin set that does not meet current safe-shutdown earthquake seismic criteria. In addition, it is the only bin set that lacks built-in features to aid in calcine retrieval. One option to alleviate the seismic compliance issue is to transport the calcine from Bin Set 1 to another bin set which has the required capacity and which is seismically qualified. This report studies the feasibility of retrieving the calcine from Bi n Set 1 and transporting it into Bin Set 6 which is located approximately 650 feet away. Because Bin Set 1 was not designed for calcine retrieval, and because of the high radiation levels and potential contamination spread from the calcined material, this is a challenging engineering task. This report presents preconceptual design studies for remotely-operated, low-density, pneumatic vacuum retrieval and transport systems and equipment that are based on past work performed by the Raytheon Engineers and Constructors architectural engineering firm. The designs presented are considered feasible; however, future development work will be needed in several areas during the subsequent conceptual design phase

Bin Set 1 Calcine Retrieval Feasibility Study

Energy Technology Data Exchange (ETDEWEB)

R. D. Adams; S. M. Berry; K. J. Galloway; T. A. Langenwalter; D. A. Lopez; C. M. Noakes; H. K. Peterson; M. I. Pope; R. J. Turk

1999-10-01

At the Department of Energy's Idaho Nuclear Technology and Engineering Center, as an interim waste management measure, both mixed high-level liquid waste and sodium bearing waste have been solidified by a calculation process and are stored in the Calcine Solids Storage Facilities. This calcined product will eventually be treated to allow final disposal in a national geologic repository. The Calcine Solids Storage Facilities comprise seven ''bit sets.'' Bin Set 1, the first to be constructed, was completed in 1959, and has been in service since 1963. It is the only bin set that does not meet current safe-shutdown earthquake seismic criteria. In addition, it is the only bin set that lacks built-in features to aid in calcine retrieval. One option to alleviate the seismic compliance issue is to transport the calcine from Bin Set 1 to another bin set which has the required capacity and which is seismically qualified. This report studies the feasibility of retrieving the calcine from Bi n Set 1 and transporting it into Bin Set 6 which is located approximately 650 feet away. Because Bin Set 1 was not designed for calcine retrieval, and because of the high radiation levels and potential contamination spread from the calcined material, this is a challenging engineering task. This report presents preconceptual design studies for remotely-operated, low-density, pneumatic vacuum retrieval and transport systems and equipment that are based on past work performed by the Raytheon Engineers and Constructors architectural engineering firm. The designs presented are considered feasible; however, future development work will be needed in several areas during the subsequent conceptual design phase.

High Temperature Calcination - MACT Upgrade Equipment Pilot Plant Test

Energy Technology Data Exchange (ETDEWEB)

Richard D. Boardman; B. H. O& #39; Brien; N. R. Soelberg; S. O. Bates; R. A. Wood; C. St. Michel

2004-02-01

About one million gallons of acidic, hazardous, and radioactive sodium-bearing waste are stored in stainless steel tanks at the Idaho Nuclear Technology and Engineering Center (INTEC), which is a major operating facility of the Idaho National Engineering and Environmental Laboratory. Calcination at high-temperature conditions (600 C, with alumina nitrate and calcium nitrate chemical addition to the feed) is one of four options currently being considered by the Department of Energy for treatment of the remaining tank wastes. If calcination is selected for future processing of the sodium-bearing waste, it will be necessary to install new off-gas control equipment in the New Waste Calcining Facility (NWCF) to comply with the Maximum Achievable Control Technology (MACT) standards for hazardous waste combustors and incinerators. This will require, as a minimum, installing a carbon bed to reduce mercury emissions from their current level of up to 7,500 to <45 {micro}g/dscm, and a staged combustor to reduce unburned kerosene fuel in the off-gas discharge to <100 ppm CO and <10 ppm hydrocarbons. The staged combustor will also reduce NOx concentrations of about 35,000 ppm by 90-95%. A pilot-plant calcination test was completed in a newly constructed 15-cm diameter calciner vessel. The pilot-plant facility was equipped with a prototype MACT off-gas control system, including a highly efficient cyclone separator and off-gas quench/venturi scrubber for particulate removal, a staged combustor for unburned hydrocarbon and NOx destruction, and a packed activated carbon bed for mercury removal and residual chloride capture. Pilot-plant testing was performed during a 50-hour system operability test January 14-16, followed by a 100-hour high-temperature calcination pilot-plant calcination run January 19-23. Two flowsheet blends were tested: a 50-hour test with an aluminum-to-alkali metal molar ratio (AAR) of 2.25, and a 50-hour test with an AAR of 1.75. Results of the testing

Processing of concentrated radioactive wastes into cement and bitumens following calcination

International Nuclear Information System (INIS)

Napravnik, J.; Sazavsky, P.; Ditl, P.; Prikryl, P.

1985-01-01

A brief characteristic is presented of the most frequently used processes of solidification of liquid radioactive wastes, viz., bituminization, cementation and their combination with calcination. The effect of individual parameters is assessed on the choice of the type of solidification process as is their importance in the actual process, in temporary storage, during transportation and under conditions of long-term storage. It has been found that a combination of the procedures could lead to a modular system of methods and equipment. This would allow to approach optimal solidification of wastes in the present period and to establish a research reserve for the development of more modern, economically advantageous and safer procedures. A rough estimate is made of the costs of the solidification of 1 m 3 of radioactive concentrate from the V-1 power plant at a production of 380 m 3 /year, this for the cementation-calcination and bituminization-calcination procedures. The said rough economic analysis only serves to identify the major operating components which have the greatest effect on the economic evaluation of the solidification procedures. (Z.M.)

Alternative calcination development status report

International Nuclear Information System (INIS)

Boardman, R.D.

1997-12-01

The Programmatic Spent Nuclear Fuel and (INEEL) Environmental Restoration and Waste Management Programs Environmental Impact Statement Record of Decision, dated June 1, 1995, specifies that high-level waste stored in the underground tanks at the ICPP continue to be calcined while other options to treat the waste are studied. Therefore, the High-Level Waste Program has funded a program to develop new flowsheets to increase the liquid waste processing rate. Simultaneously, a radionuclide separation process, as well as other options, are also being developed, which will be compared to the calcination treatment option. Two alternatives emerged as viable candidates; (1) elevated temperature calcination (also referred to as high temperature calcination), and (2) sugar-additive calcination. Both alternatives were determined to be viable through testing performed in a lab-scale calcination mockup. Subsequently, 10-cm Calciner Pilot Plant scoping tests were successfully completed for both flowsheets. The results were compared to the standard 500 C, high-ANN flow sheet (baseline flowsheet). The product and effluent streams were characterized to help elucidate the process chemistry and to investigate potential environmental permitting issues. Several supplementary tests were conducted to gain a better understanding of fine-particles generation, calcine hydration, scrub foaming, feed makeup procedures, sugar/organic elimination, and safety-related issues. Many of the experiments are only considered to be scoping tests, and follow-up experiments will be required to establish a more definitive understanding of the flowsheets. However, the combined results support the general conclusion that flowsheet improvements for the NWCF are technically viable

Idaho National Engineering and Environmental Laboratory, Old Waste Calcining Facility, Scoville vicinity, Butte County, Idaho -- Photographs, written historical and descriptive data. Historical American engineering record

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

This report describes the history of the Old Waste Calcining Facility. It begins with introductory material on the Idaho National Engineering and Environmental Laboratory, the Materials Testing Reactor fuel cycle, and the Idaho Chemical Processing Plant. The report then describes management of the wastes from the processing plant in the following chapters: Converting liquid to solid wastes; Fluidized bed waste calcining process and the Waste Calcining Facility; Waste calcining campaigns; WCF gets a new source of heat; New Waste Calcining Facility; Last campaign; Deactivation and the RCRA cap; Significance/context of the old WCF. Appendices contain a photo key map for HAER photos, a vicinity map and neighborhood of the WCF, detailed description of the calcining process, and chronology of WCF campaigns.

Idaho National Engineering and Environmental Laboratory, Old Waste Calcining Facility, Scoville vicinity, Butte County, Idaho -- Photographs, written historical and descriptive data. Historical American engineering record

International Nuclear Information System (INIS)

1997-01-01

This report describes the history of the Old Waste Calcining Facility. It begins with introductory material on the Idaho National Engineering and Environmental Laboratory, the Materials Testing Reactor fuel cycle, and the Idaho Chemical Processing Plant. The report then describes management of the wastes from the processing plant in the following chapters: Converting liquid to solid wastes; Fluidized bed waste calcining process and the Waste Calcining Facility; Waste calcining campaigns; WCF gets a new source of heat; New Waste Calcining Facility; Last campaign; Deactivation and the RCRA cap; Significance/context of the old WCF. Appendices contain a photo key map for HAER photos, a vicinity map and neighborhood of the WCF, detailed description of the calcining process, and chronology of WCF campaigns

Pecularities of carrying out radioactive wastes vitrification process without preliminary calcination of wastes

International Nuclear Information System (INIS)

Konstantinovich, A.A.; Kulichenko, V.V.; Bel'tyukov, V.A.; Nikiforov, A.S.; Nikipelov, B.V.; Stepanov, S.E.; Baskov, L.I.; Kulakov, S.I.

1978-01-01

Vitrification technology is considered for liquid radioactive wastes by means of electric furnace where heating of glass-paste is done by electric current passing through the melt. Continious process of gehydration, calcination and vitrification is going on in one apparatus. Testing if the method has been performed by use of a model solution, containing sodium and aluminium nitrates. To obtain phosphoric acid has been added into the solution. Lay-out of the device and its description as well as technical parameters of the electric furnace are given. The results are stated for determination of the optimum operation conditions for the device. To reduce entrainment of solid components, molasses has been added in the solution. Parameters are given for the process of the solution containing 80 g/l molasses processing. It has been shown that edding molasses to the solution permitted to reduse power consumption of the process due to the heat generation during oxidation-reduction reaction on the melt surface. The results are given for investigations of the nitrogen oxides catching in scrubbers. These results have shown that introduction of molasses reduces nitrigen oxides concentration. The results of the experimental works have shown the possibility of the continious process of dehydration, calcination and vitrification in single device with application of remote control and monitoring by means of automatics. (I.T.) [ru

Calcination of Fluorinel-sodium waste blends using sugar as a feed additive (formerly WINCO-11879)

International Nuclear Information System (INIS)

Newby, B.J.; Thomson, T.D.; O'Brien, B.H.

1992-06-01

Methods were studied for using sugar as a feed additive for converting the sodium-bearing wastes stored at the Idaho Chemical Processing Plant into granular, free flowing solids by fluidized-bed calcination at 500 degrees C. All methods studied blended sodium-bearing wastes with Fluorinel wastes but differed in the types of sugar (sucrose or dextrose) that were added to the blend. The most promising sugar additive was determined to be sucrose, since it is converted more completely to inorganic carbon than is dextrose. The effect of the feed aluminum-to-alkali metal mole ratio on calcination of these blends with sugar was also investigated. Increasing the aluminum-to-alkali metal ratio from 0.6 to 1.0 decreased the calcine product-to-fines ratio from 3.0 to 1.0 and the attrition index from 80 to 15%. Further increasing the ratio to 1.25 had no effect

Chemistry of application of calcination/dissolution to the Hanford tank waste inventory

International Nuclear Information System (INIS)

Delegard, C.H.; Elcan, T.D.; Hey, B.E.

1994-05-01

Approximately 330,000 metric tons of sodium-rich radioactive waste originating from separation of plutonium from irradiated uranium fuel are stored in underground tanks at the Hanford Site in Washington State. Fractionation of the waste into low-level waste (LLW) and high-level waste (HLW) streams is envisioned via partial water dissolution and limited radionuclide extraction operations. Under optimum conditions, LLW would contain most of the chemical bulk while HLW would contain virtually all of the transuranic and fission product activity. Calcination at around 850 C, followed by water dissolution, has been proposed as an alternative initial treatment of Hanford Site waste to improve waste dissolution and the envisioned LLW/HLW split. Results of literature and laboratory studies are reported on the application of calcination/dissolution (C/D) to the fractionation of the Hanford Site tank waste inventory. Both simulated and genuine Hanford Site waste materials were used in the lab tests. To evaluation confirmed that C/D processing reduced the amount of several components from the waste. The C/D dissolutions of aluminum and chromium allow redistribution of these waste components from the HLW to the LLW fraction. Comparisons of simple water-washing with C/D processing of genuine Hanford Site waste are also reported based on material (radionuclide and chemical) distributions to solution and solid residue phases. The lab results show that C/D processing yielded superior dissolution of aluminum and chromium sludges compared to simple water dissolution. 57 refs., 26 figs., 18 tabs

Chemistry of application of calcination/dissolution to the Hanford tank waste inventory

Energy Technology Data Exchange (ETDEWEB)

Delegard, C.H.; Elcan, T.D.; Hey, B.E.

1994-05-01

Approximately 330,000 metric tons of sodium-rich radioactive waste originating from separation of plutonium from irradiated uranium fuel are stored in underground tanks at the Hanford Site in Washington State. Fractionation of the waste into low-level waste (LLW) and high-level waste (HLW) streams is envisioned via partial water dissolution and limited radionuclide extraction operations. Under optimum conditions, LLW would contain most of the chemical bulk while HLW would contain virtually all of the transuranic and fission product activity. Calcination at around 850 C, followed by water dissolution, has been proposed as an alternative initial treatment of Hanford Site waste to improve waste dissolution and the envisioned LLW/HLW split. Results of literature and laboratory studies are reported on the application of calcination/dissolution (C/D) to the fractionation of the Hanford Site tank waste inventory. Both simulated and genuine Hanford Site waste materials were used in the lab tests. To evaluation confirmed that C/D processing reduced the amount of several components from the waste. The C/D dissolutions of aluminum and chromium allow redistribution of these waste components from the HLW to the LLW fraction. Comparisons of simple water-washing with C/D processing of genuine Hanford Site waste are also reported based on material (radionuclide and chemical) distributions to solution and solid residue phases. The lab results show that C/D processing yielded superior dissolution of aluminum and chromium sludges compared to simple water dissolution. 57 refs., 26 figs., 18 tabs.

Design and performance of atomizing nozzles for spray calcination of high-level wastes

International Nuclear Information System (INIS)

Miller, F.A.; Stout, L.A.

1981-05-01

A key aspect of high-level liquid-waste spray calcination is waste-feed atomization by using air atomizing nozzles. Atomization substantially increases the heat transfer area of the waste solution, which enhances rapid drying. Experience from the spray-calciner operations has demonstrated that nozzle flow conditions that produce 70-μ median-volume-diameter or smaller spray droplets are required for small-scale spray calciners (drying capacity less than 80 L/h). For large-scale calciners (drying capacity greater than 300 L/h), nozzle flow conditions that produce 100-μ median-volume-diameter or smaller spray droplets are required. Mass flow ratios of 0.2 to 0.4, depending on nozzle size, are required for proper operation of internal-mix atomizing nozzles. Both internal-mix and external-mix nozzles have been tested at PNL. Due to the lower airflow requirements and fewer large droplets produced, the internal-mix nozzle has been chosen for primary development in the spray calciner program at PNL. Several nozzle air-cap materials for internal-mix nozzles have been tested for wear resistance. Results show that nozzle air caps of stainless steel and Cer-vit (a machineable glass ceramic) are suceptible to rapid wear by abrasive slurries, whereas air caps of alumina and reaction-bonded silicon nitride show only slow wear. Longer-term testing is necessary to determine more accurately the actual frequency of nozzle replacement. Atomizing nozzle air caps of alumina are subject to fracture from thermal shock, whereas air caps of silicon nitride and Cer-vit are not. Fractured nozzles are held in place by the air-cap retaining ring and continue to atomize satisfactorily. Therefore, fractures caused by thermal shocking do not necessarily result in nozzle failure

Redox calcination study of Synroc D powder containing simulated SRL waste

International Nuclear Information System (INIS)

Chen, C.

1982-01-01

According to Ringwood [A.E. Ringwood, W. Sinclair, and G.M. McLaughlin, Nuclear Waste Immobilization, Lawrence Livermore Laboratory, Livermore, Rept. UCRL-15147 (1979)], the iron oxidation state is important in controlling, the spinel mineralogy and composition if the amount of titania (TiO 2 ) consumed in spinel formation is to be minimized in favor of the formation of the Synroc phases, zirconolite, perovskite, and nepheline. In our redox calcination studies we observed that the iron oxidation state of FeO/Fe 2 O 3 can be controlled by the redoxcalcining atmosphere. In a CO atmosphere, the oxidation state was reduced to less than 7 wt % Fe 2 O 3 . With appropriate CO 2 /CO gas mixtures the resultant iron oxidation states were in the range of 45 to 59 wt % Fe 2 O 3 . Direct rotary redox calcination of spray dried powder at 600 0 C, without prior air calcination, showed increased redox efficiency when compared to powder that had been previously air calcined at 650 0 C. We believe this is caused by a reduction in particle size. Rotary calcination at 800 0 C in argon has no measurable reduction affect on the iron oxidation state of Synroc D powder

Physical, Chemical and Structural Evolution of Zeolite-Containing Waste Forms Produced from Metakaolinite and Calcined HLW

International Nuclear Information System (INIS)

Grutzeck, Michael; Jantzen, Carol M.

1999-01-01

fashion could be a viable alternative for fixation of low activity waste (LAW) calcines. Our objective is to adapt this technology for use in site remediation and clean up of caustic waste solutions now in storage in tanks at Hanford and the Savannah River sites. The proposed work is meant to develop a clearer understanding of the advantages and limitations of producing a zeolite-containing waste form (hydroceramic) from calcined radioactive waste, i.e. the effect of processing variables, reaction kinetics, crystal and phase chemistry, and microstructure on their performance

Calcination under negative atmosphere for SYNROC preparation

International Nuclear Information System (INIS)

Ambashta, R.D.; Wattal, P.K.; Govindankutty, K.V.

2006-01-01

SYNROC-C is a ceramic waste formulation designed to immobilise reprocessing waste from fast breeder reactor. This formulation is capable of incorporating noble metals, other fission products, corrosion products and activation products in its multiphase assemblage. Calcination is an important step of SYNROC preparation for decomposition of nitrates of the radioactive waste and conversion to oxide precursors. This paper presents a comparison between properties of calcine prepared under different calcination procedures to obtain product suitable for compaction

Evaluation of Calcine Disposition Path Forward

International Nuclear Information System (INIS)

Birrer, S.A.; Heiser, M.B.

2003-01-01

This document describes an evaluation of the baseline and two alternative disposition paths for the final disposition of the calcine wastes stored at the Idaho Nuclear Technology and Engineering Center at the Idaho National Engineering and Environmental Laboratory. The pathways are evaluated against a prescribed set of criteria and a recommendation is made for the path forward

Attrition, elutriation, and growth of particles produced in fluidized-bed waste calciners

International Nuclear Information System (INIS)

McDonald, F.N.

1982-09-01

The Idaho Chemical Processing Plant reduces the volume of high-level liquid radioactive wastes in a fluidized bed to produce a granular calcine product. In the past, difficulties have been experienced in controlling the product's particle size when processing certain blends of sodium-bearing waste. Therefore, experiments in attrition, elutriation, and particle growth were done to characterize how best to control these three parameters. 15 figures, 16 tables

Letter report: Evaluation of dryer/calciner technologies for testing

International Nuclear Information System (INIS)

Sevigny, G.

1996-02-01

This letter report describes some past experiences on the drying and calcination of radioactive materials or corresponding simulants; and the information needed from testing. The report also includes an assessment of informational needs including possible impacts to a full-scale plant. This includes reliability, maintenance, and overall size versus throughput. Much of the material was previously compiled and reported by Mike Elliott of PNL open-quotes Melter Performance Assessmentclose quotes and Larry Eisenstatt of SEG on contract to WHC in a letter to Rod Powell. Also, an annotated bibliography was prepared by Reagan Seymour of WHC. Descriptions of the drying and calciner technologies, development status, advantages and disadvantages of using a WFE or calciner, and recommendations for future testing are discussed in this report

Remote process connectors for the new waste calcining facility

International Nuclear Information System (INIS)

Jacobs, R.T.; Carter, J.A.; Hohback, A.C.

1978-01-01

The remote process connectors developed, used, and tested at the Remote Maintenance Development Facility are described. These connectors, including the three-bolt kinematic-graphite flange and watertight electrical connectors, are assembled on master jigs (holding-welding fixture) to form interchangeable pump and valve loop assemblies. These assemblies, with their guide-in platforms, make possible a method of performing remote maintenance at the New Waste Calcining Facility which is a departure from methods that until now have been the standard of the industry

Comparison of the rotary calciner-metallic melter and the slurry-fed ceramic melter technologies for vitrifying West Valley high-level wastes

International Nuclear Information System (INIS)

Chapman, C.C.

1983-01-01

Two processes which are believed applicable and available for vitrification of West Valley's high-level (HLW) wastes were technically evaluated and compared. The rotary calciner-metallic melter (AVH) and the slurry-fed ceramic melter (SFCM) were evaluated under the following general categories: process flow sheet, remote operability, safety and environmental considerations, and estimated cost and schedules

Vitrification of radioactive high-level waste by spray calcination and in-can melting

Science.gov (United States)

Hanson, M. S.; Bjorklund, W. J.

1980-07-01

After several nonradioactive test runs, radioactive waste from the processing of 1.5 t of spent, light water reactor fuel was successfully concentrated, dried and converted to a vitreous product. A total of 97 L of waste glass (in two stainless steel canisters) was produced. The spray calcination process coupled to the in-can melting process, as developed at Pacific Northwest Labortory, was used to vitrify the waste. An effluent system consisting of a variety of condensation of scrubbing steps more than adequately decontaminated the process off gas before it was released to the atmosphere.

Evaluation of blends bauxite-calcination-method red mud with other industrial wastes as a cementitious material: Properties and hydration characteristics

International Nuclear Information System (INIS)

Zhang Na; Liu Xiaoming; Sun Henghu; Li Longtu

2011-01-01

Red mud is generated from alumina production, and its disposal is currently a worldwide problem. In China, large quantities of red mud derived from bauxite calcination method are being discharged annually, and its utilization has been an urgent topic. This experimental research was to evaluate the feasibility of blends red mud derived from bauxite calcination method with other industrial wastes for use as a cementitious material. The developed cementitious material containing 30% of the bauxite-calcination-method red mud possessed compressive strength properties at a level similar to normal Portland cement, in the range of 45.3-49.5 MPa. Best compressive strength values were demonstrated by the specimen RSFC2 containing 30% bauxite-calcination-method red mud, 21% blast-furnace slag, 10% fly ash, 30% clinker, 8% gypsum and 1% compound agent. The mechanical and physical properties confirm the usefulness of RSFC2. The hydration characteristics of RSFC2 were characterized by XRD, FTIR, 27 Al MAS-NMR and SEM. As predominant hydration products, ettringite and amorphous C-S-H gel are principally responsible for the strength development of RSFC2. Comparing with the traditional production for ordinary Portland cement, this green technology is easier to be implemented and energy saving. This paper provides a key solution to effectively utilize bauxite-calcination-method red mud.

Innovative technologies for the treatment of hazardous and mixed wastes

International Nuclear Information System (INIS)

Eyman, L.D.; Anderson, T.D.

1988-01-01

The treatment, storage, and disposal of hazardous and mixed wastes incur significant costs for Department of Energy (DOE) installations. These wastes must be managed under strict environmental controls and regulations to prevent the possibility of migration of hazardous materials to the biosphere. Through the Hazardous Waste Remedial Actions Program, the DOE is seeking to develop innovative ways of improving current treatment technologies to eliminate the hazardous components of wastes, reduce waste management costs, and minimize the volume requiring disposal as hazardous or mixed waste. Sponsored projects progress from research and development to field demonstration. Among the innovative technologies under development are supercritical water oxidation of hazardous chemicals, microwave-assisted destruction of chlorinated hydrocarbons, paramagnetic separation of metals from waste, detoxification and reclamation of waste acid, nitrate destruction through calcination, treatment/disposal of reactive metals, and methodologies for encapsulation. Technologies at a demonstration phase include detoxification of mixed waste sludge, microbial degradation of polychlorinated biphenyls in soil, and the remediation process for a hydrocarbon spill. 14 refs

An Assessment of Using Vibrational Compaction of Calcined HLW and LLW in DWPF Canisters

International Nuclear Information System (INIS)

Yi, Yun-Bo; Amme, Robert C.; Shayer, Zeev

2008-01-01

Since 1963, the INEL has calcined almost 8 million gallons of liquid mixed waste and liquid high-level waste, converting it to some 1.1 million gallons of dry calcine (about 4275.0 m3), which consists of alumina-and zirconia-based calcine and zirconia-sodium blend calcine. In addition, if all existing and projected future liquid wastes are solidified, approximately 2,000 m3 of additional calcine will be produced primarily from sodium-bearing waste. Calcine is a more desirable material to store than liquid radioactive waste because it reduces volume, is much less corrosive, less chemically reactive, less mobile under most conditions, easier to monitor and more protective of human health and the environment. This paper describes the technical issue involved in the development of a feasible solution for further volume reduction of calcined nuclear waste for transportation and long term storage, using a standard DWPF canister. This will be accomplished by developing a process wherein the canisters are transported into a vibrational machine, for further volume reduction by about 35%. The random compaction experiments show that this volume reduction is achievable. The main goal of this paper is to demonstrate through computer modeling that it is feasible to use volume reduction vibrational machine without developing stress/strain forces that will weaken the canister integrity. Specifically, the paper presents preliminary results of the stress/strain analysis of the DWPF canister as a function of granular calcined height during the compaction and verifying that the integrity of the canister is not compromised. This preliminary study will lead to the development of better technology for safe compactions of nuclear waste that will have significant economical impact on nuclear waste storage and treatment. The preliminary results will guide us to find better solutions to the following questions: 1) What are the optimum locations and directions (vertical versus horizontal or

Cost Comparison for the Transfer of Select Calcined Waste Canisters to the Monitored Geologic Repository at Yucca Mountain, NV

International Nuclear Information System (INIS)

Michael B. Heiser; Clark B. Millet

2005-01-01

This report performs a life-cycle cost comparison of three proposed canister designs for the shipment and disposition of Idaho National Laboratory high-level calcined waste currently in storage at the Idaho Nuclear Technology and Engineering Center to the proposed national monitored geologic repository at Yucca Mountain, Nevada. Concept A (2 x 10-ft) and Concept B (2 x 15-ft) canisters are comparable in design, but they differ in size and waste loading options and vary proportionally in weight. The Concept C (5.5 x 17.5-ft) canister (also called the ''super canister''), while similar in design to the other canisters, is considerably larger and heavier than Concept A and B canisters and has a greater wall thickness. This report includes estimating the unique life-cycle costs for the three canister designs. Unique life-cycle costs include elements such as canister purchase and filling at the Idaho Nuclear Technology and Engineering Center, cask preparation and roundtrip consignment costs, final disposition in the monitored geologic repository (including canister off-loading and placement in the final waste disposal package for disposition), and cask purchase. Packaging of the calcine ''as-is'' would save $2.9 to $3.9 billion over direct vitrification disposal in the proposed national monitored geologic repository at Yucca Mountain, Nevada. Using the larger Concept C canisters would use 0.75 mi less of tunnel space, cost $1.3 billion less than 10-ft canisters of Concept A, and would be complete in 6.2 years

Separation technologies for the treatment of Idaho National Engineering Laboratory wastes

Energy Technology Data Exchange (ETDEWEB)

Todd, T.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

1997-10-01

Currently about 6.8 million L of acidic, radioactive liquid waste that is not amenable to calcination, and about 3800 m{sup 3} of calcine exist at the ICPP. Legal drivers (court orders) and agreements between the state of Idaho, the U.S. Navy, and DOE exist that obligate INEL to develop, demonstrate, and implement technologies for treatment and interim storage of the radioactive liquid and calcine wastes. Per these agreements, all tank waste must be removed from the underground liquid storage tanks by the year 2012, and high-level radioactive waste must be treated and removed from INEL by 2035. Separation of the radionuclides from the wastes, followed by immobilization of the high-activity and low-activity fractions in glass and grout, respectively, is the approach preferred by INEL. Technologies to remove actinides (U, Np, Pu, and Am), Cs, Sr, and possibly Tc from highly acidic solutions are required to process INEL wastes. Decontamination of the wastes to NRC Class A low-level waste (LLW) is planned. Separation and isolation of Resource Conservation and Recovery Act (RCRA) metals (Hg, Pb, Cd, and Cr) from the highly radioactive waste streams may also be required. Remediation efforts will begin in FY 1997 to remove volatile organic compounds (VOCs) and radionuclides (Cs and Sr) from groundwater located at the Test Area North facility at INEL. A plume of VOCs and radionuclides has spread from the former TSF-05 injection well, and a Comprehensive Environmental Response, Conservation, and Liability Act (CERCLA) remediation action is under way. A Record of Decision was signed in August 1995 that commits INEL to remediate the plume from TSF-05. Removal of Sr and Cs from the groundwater using commercially available ion-exchange resins has been unsuccessful at meeting maximum contaminant levels, which are 119 pCi/L and 8 pCi/L for Cs and Sr, respectively. Cesium and Sr are the major contaminants that must be removed from the groundwater.

Sodium-bearing Waste Treatment Technology Evaluation Report

Energy Technology Data Exchange (ETDEWEB)

Charles M. Barnes; Arlin L. Olson; Dean D. Taylor

2004-05-01

Sodium-bearing waste (SBW) disposition is one of the U.S. Department of Energy (DOE) Idaho Operation Office’s (NE-ID) and State of Idaho’s top priorities at the Idaho National Engineering and Environmental Laboratory (INEEL). The INEEL has been working over the past several years to identify a treatment technology that meets NE-ID and regulatory treatment requirements, including consideration of stakeholder input. Many studies, including the High-Level Waste and Facilities Disposition Environmental Impact Statement (EIS), have resulted in the identification of five treatment alternatives that form a short list of perhaps the most appropriate technologies for the DOE to select from. The alternatives are (a) calcination with maximum achievable control technology (MACT) upgrade, (b) steam reforming, (c) cesium ion exchange (CsIX) with immobilization, (d) direct evaporation, and (e) vitrification. Each alternative has undergone some degree of applied technical development and preliminary process design over the past four years. This report presents a summary of the applied technology and process design activities performed through February 2004. The SBW issue and the five alternatives are described in Sections 2 and 3, respectively. Details of preliminary process design activities for three of the alternatives (steam reforming, CsIX, and direct evaporation) are presented in three appendices. A recent feasibility study provides the details for calcination. There have been no recent activities performed with regard to vitrification; that section summarizes and references previous work.

Practical results of the MESA 1 line calcinator trial operation

International Nuclear Information System (INIS)

Napravnik, J.; Sazavsky, P.; Skaba, V.; Zahalka, F.; Vild, J.; Kulovany, J.

1987-01-01

Mobile calcination and cementation unit MESA 1 was designed and built by UJV Rez in cooperation with many enterprises, mainly with the Kralovopolske Strojirny Brno. This facility for direct fixation of liquid radioactive wastes was experimentally tested using model non-radioactive solutions and model and actual wastes from the Jaslovske Bohunice nuclear power plant. The calciner was run in trial operation at the Kralovopolske SAtrojirny Brno. A total of 1.3 m 3 of model solutions was processed into 180 kg of calcinate. The fixation of the calcinate in cement, the times of solidification and of hardening and the moisture content of concrete blocks were studied. The application was also tested of the calciner in drying ion exchangers from WWER-440 prior to their bituminization. Following the despatch of the cementation module to the Chernobyl nuclear power plant, the direct calcination module was tested at Dukovany together with an auxiliary module which makes possible self-contained calciner operation. Model non-radioactive solutions from the Dukovany nuclear power plant were treated containing H 3 BO 3 and NaNO 3 as main components. The usability in actual conditions of the mobile calcination and cementation unit for radioactive wastes was tested in a total of about 70 operating hours. (E.S.). 2 figs., 2 refs

Hot isostatically-pressed aluminosilicate glass-ceramic with natural crystalline analogues for immobilizing the calcined high-level nuclear waste at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Raman, S.

1993-12-01

The additives Si, Al, MgO, P 2 O 5 were mechanically blended with fluorinelsodium calcine in varying proportions. The batches were vacuum sealed in stainless steel canisters and hot isostatically pressed at 20,000 PSI and 1000 C for 4 hours. The resulting suite of glass-ceramic waste forms parallels the natural rocks in microstructural and compositional heterogeneity. Several crystalline phases ar analogous in composition and structure to naturally occurring minerals. Additional crystalline phases are zirconia and Ca-Mg borate. The glasses are enriched in silica and alumina. Approximately 7% calcine elements occur dissolved in this glass and the total glass content in the waste forms averages 20 wt%. The remainder of the calcine elements are partitioned into crystalline phases at 75 wt% calcine waste loading. The waste forms were tested for chemical durability in accordance with the MCC1-test procedure. The leach rates are a function of the relative proportions of additives and calcine, which in turn influence the composition and abundances of the glass and crystalline phases. The DOE leach rate criterion of less than 1 g/m 2 -day is met by all the elements B, Cs and Na are increased by lowering the melt viscosity. This is related to increased crystallization or devitrification with increases in MgO addition. This exploratory work has shown that the increases in waste loading occur by preferred partitioning of the calcine components among crystalline and glass phases. The determination of optimum processing parameters in the form of additive concentration levels, homogeneous blending among the components, and pressure-temperature stabilities of phases must be continued to eliminate undesirable effects of chemical composition, microstructure and glass devitrification

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

Flexible waste management to increase the effectiveness of minor actinide PT technology

Energy Technology Data Exchange (ETDEWEB)

Fukasawa, T. [Hitachi-GE Nuclear Energy, Ltd., 3-1-1 Saiwai, Hitachi 317-0073 (Japan); Inagaki, Y.; Arima, T. [Kyshu University, 744 Motooka, Nishi, Fukuoka 819-0395 (Japan); Sato, S. [Fukushima National College of Technology, 30 Aza-Nagao, Tairakamiarakawa, Iwaki 970-8034 (Japan)

2016-07-01

Partitioning and transmutation (PT) technologies have been developed for minor actinides (MA) to reduce the high level waste (HLW) volume and long-term radiotoxicity. Although the MA PT can reduce the potential radiotoxicity effectively by 1-3 orders of magnitude, the actual operation of PT requires several tens of years for developing elemental technologies of nuclide separation, MA containing fuel fabrication, transmutation and their practical systematization. The high level liquid waste (HLLW) containing MA is presently vitrified immediately after spent fuel reprocessing, stored about 50 years at surface facility and will be disposed of at deep geological repository. Vitrified HLW form works as an excellent artificial barrier against nuclides release during storage and disposal. On the other hand, it is difficult to recover MA from the form. So the present waste management scheme has an issue of MA PT technology application until its deployment, which will produce much amount of vitrified HLW including long-lived MA without PT application. Thus the authors proposed the flexible waste management method to increase the effectiveness of the MA PT. The system adopts the HLLW calcination instead of the vitrification to produce granule for its dry storage of about 50 years until the MA PT technology will be applicable. The granule should be easily dissolved by the nitric acid solution to apply the typical aqueous MA partitioning technologies to be developed. This paper reports the purpose of the study, the feasibility evaluation results for the calcined granule storage and the evaluation results for the environmental burden reduction effect. (authors)

Potential dispositioning flowsheets for ICPP SNF and wastes

Energy Technology Data Exchange (ETDEWEB)

Olson, A.L. [ed.; Anderson, P.A.; Bendixsen, C.L. [and others

1995-11-01

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation`s radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995.

Potential dispositioning flowsheets for ICPP SNF and wastes

International Nuclear Information System (INIS)

Olson, A.L.; Anderson, P.A.; Bendixsen, C.L.

1995-11-01

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation's radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995

Mercury removal at Idaho National Engineering and Environmental Laboratory's New Waste Calciner Facility

International Nuclear Information System (INIS)

Ashworth, S.C.

2000-01-01

Technologies were investigated to determine viable processes for removing mercury from the calciner (NWCF) offgas system at the Idaho National Engineering and Environmental Laboratory. Technologies for gas phase and aqueous phase treatment were evaluated. The technologies determined are intended to meet EPA Maximum Achievable Control Technology (MACT) requirements under the Clean Air Act and Resource Conservation and Recovery Act (RCRA). Currently, mercury accumulation in the calciner off-gas scrubbing system is transferred to the tank farm. These transfers lead to accumulation in the liquid heels of the tanks. The principal objective for aqueous phase mercury removal is heel mercury reduction. The system presents a challenge to traditional methods because of the presence of nitrogen oxides in the gas phase and high nitric acid in the aqueous scrubbing solution. Many old and new technologies were evaluated including sorbents and absorption in the gas phase and ion exchange, membranes/sorption, galvanic methods, and UV reduction in the aqueous phase. Process modifications and feed pre-treatment were also evaluated. Various properties of mercury and its compounds were summarized and speciation was predicted based on thermodynamics. Three systems (process modification, NOxidizer combustor, and electrochemical aqueous phase treatment) and additional technology testing were recommended

Technology Evaluations Related to Mercury, Technetium, and Chloride in Treatment of Wastes at the Idaho Nuclear Technology and Engineering Center of the Idaho National Engineering and Environmental Laboratory

Energy Technology Data Exchange (ETDEWEB)

C. M. Barnes; D. D. Taylor; S. C. Ashworth; J. B. Bosley; D. R. Haefner

1999-10-01

The Idaho High-Level Waste and Facility Disposition Environmental Impact Statement defines alternative for treating and disposing of wastes stored at the Idaho Nuclear Technology and Engineering Center. Development is required for several technologies under consideration for treatment of these wastes. This report contains evaluations of whether specific treatment is needed and if so, by what methods, to remove mercury, technetium, and chlorides in proposed Environmental Impact Statement treatment processes. The evaluations of mercury include a review of regulatory requirements that would apply to mercury wastes in separations processes, an evaluation of the sensitivity of mercury flowrates and concentrations to changes in separations processing schemes and conditions, test results from laboratory-scale experiments of precipitation of mercury by sulfide precipitation agents from the TRUEX carbonate wash effluent, and evaluations of methods to remove mercury from New Waste Calcining Facility liquid and gaseous streams. The evaluation of technetium relates to the need for technetium removal and alternative methods to remove technetium from streams in separations processes. The need for removal of chlorides from New Waste Calcining Facility scrub solution is also evaluated.

Technology Evaluations Related to Mercury, Technetium, and Chloride in Treatment of Wastes at the Idaho Nuclear Technology and Engineering Center of the Idaho National Engineering and Environmental Laboratory

International Nuclear Information System (INIS)

Barnes, C.M.; Taylor, D.D.; Ashworth, S.C.; Bosley, J.B.; Haefner, D.R.

1999-01-01

The Idaho High-Level Waste and Facility Disposition Environmental Impact Statement defines alternative for treating and disposing of wastes stored at the Idaho Nuclear Technology and Engineering Center. Development is required for several technologies under consideration for treatment of these wastes. This report contains evaluations of whether specific treatment is needed and if so, by what methods, to remove mercury, technetium, and chlorides in proposed Environmental Impact Statement treatment processes. The evaluations of mercury include a review of regulatory requirements that would apply to mercury wastes in separations processes, an evaluation of the sensitivity of mercury flowrates and concentrations to changes in separations processing schemes and conditions, test results from laboratory-scale experiments of precipitation of mercury by sulfide precipitation agents from the TRUEX carbonate wash effluent, and evaluations of methods to remove mercury from New Waste Calcining Facility liquid and gaseous streams. The evaluation of technetium relates to the need for technetium removal and alternative methods to remove technetium from streams in separations processes. The need for removal of chlorides from New Waste Calcining Facility scrub solution is also evaluated

PHYSICAL, CHEMICAL, AND STRUCTURAL EVOLUTION OF ZEOLITE-CONTAINING WASTE FORMS PRODUCED FROM METAKAOLINITE AND CALCINED HLW

International Nuclear Information System (INIS)

Pareizs, J. M.; Jantzenm, C.M.

2000-01-01

Natural and synthetic zeolites are extremely versatile materials. They can adsorb a variety of liquids and gases, and also take part in cation exchange reactions. Zeolites have the ability to sequester ions in lattice positions or within their networks of channels and voids. The zeolites can host alkali, alkaline earth and a variety of higher valance cations. As such they may be a viable alternative for immobilization of low activity waste (LAW) salts and calcines. The process for synthesizing zeolites is well documented for pure starting materials. A reactive aluminosilicate is reacted with an alkaline hydroxide at low temperature (<300 C) to form a zeolite. Processing time and temperature and specific reactants determine the type of zeolite formed. Zeolites are easy to make, and can be synthesized from a wide variety of natural and man made materials. However, relatively little is known about the process if one of the starting materials is a poorly characterized complex mixture of oxides (waste) containing nearly every element in the periodic table. The purpose of this work is to develop a clearer understanding of the advantages and limitations of producing a zeolite waste form from radioactive waste. Dr. M. W. Grutzeck at the Pennsylvania State University is investigating the production of a zeolite waste form using nonradioactive simulants. Dr. C. M. Jantzen and J. M. Pareizs at the Savannah River Technology Center will use the results from simulant work as a starting point for producing a zeolite waste form from an actual Savannah River Site radioactive waste stream

Physical, chemical, and structural evolution of zeolite-containing waste forms produced from metakaolinite and calcined HLW

International Nuclear Information System (INIS)

Pareizs, J.M.

2000-01-01

Natural and synthetic zeolites are extremely versatile materials. They can adsorb a variety of liquids and gases, and also take part in cation exchange reactions. Zeolites have the ability to sequester ions in lattice positions or within their networks of channels and voids. The zeolites can host alkali, alkaline earth and a variety of higher valence cations. As such they may be a viable alternative for immobilization of low activity waste (LAW) salts and calcines. The process for synthesizing zeolites is well documented for pure starting materials. A reactive aluminosilicate is reacted with an alkaline hydroxide at low temperature to form a zeolite. Processing time and temperature and specific reactants determine the type of zeolite formed. Zeolites are easy to make, and can be synthesized from a wide variety of natural and man made materials. However, relatively little is known about the process if one of the starting materials is a poorly characterized complex mixture of oxides (waste) containing nearly every element in the periodic table. The purpose of this work is to develop a clearer understanding of the advantages and limitations of producing a zeolite waste form from radioactive waste. Dr. M. W. Grutzeck at the Pennsylvania State University is investigating the production of a zeolite waste form using non-radioactive simulants. Dr. C. M. Jantzen and J. M. Pareizs at the Savannah River Technology Center will use the results from simulant work as a starting point for producing a zeolite waste form from an actual Savannah River Site radioactive waste stream

Generalized Test Plan for the Vitrification of Simulated High-Level -Waste Calcine in the Idaho National Laboratory's Bench -Scale Cold Crucible Induction Melter

International Nuclear Information System (INIS)

Maio, Vince

2011-01-01

This Preliminary Idaho National Laboratory (INL) Test Plan outlines the chronological steps required to initially evaluate the validity of vitrifying INL surrogate (cold) High-Level-Waste (HLW) solid particulate calcine in INL's Cold Crucible Induction Melter (CCIM). Its documentation and publication satisfies interim milestone WP-413-INL-01 of the DOE-EM (via the Office of River Protection) sponsored work package, WP 4.1.3, entitled 'Improved Vitrification' The primary goal of the proposed CCIM testing is to initiate efforts to identify an efficient and effective back-up and risk adverse technology for treating the actual HLW calcine stored at the INL. The calcine's treatment must be completed by 2035 as dictated by a State of Idaho Consent Order. A final report on this surrogate/calcine test in the CCIM will be issued in May 2012-pending next fiscal year funding In particular the plan provides; (1) distinct test objectives, (2) a description of the purpose and scope of planned university contracted pre-screening tests required to optimize the CCIM glass/surrogate calcine formulation, (3) a listing of necessary CCIM equipment modifications and corresponding work control document changes necessary to feed a solid particulate to the CCIM, (4) a description of the class of calcine that will be represented by the surrogate, and (5) a tentative tabulation of the anticipated CCIM testing conditions, testing parameters, sampling requirements and analytical tests. Key FY -11 milestones associated with this CCIM testing effort are also provided. The CCIM test run is scheduled to be conducted in February of 2012 and will involve testing with a surrogate HLW calcine representative of only 13% of the 4,000 m3 of 'hot' calcine residing in 6 INL Bin Sets. The remaining classes of calcine will have to be eventually tested in the CCIM if an operational scale CCIM is to be a feasible option for the actual INL HLW calcine. This remaining calcine's make-up is HLW containing

Spent Fuel and Waste Management Technology Development Program. Annual progress report

Energy Technology Data Exchange (ETDEWEB)

Bryant, J.W.

1994-01-01

This report provides information on the progress of activities during fiscal year 1993 in the Spent Fuel and Waste Management Technology Development Program (SF&WMTDP) at the Idaho Chemical Processing Plant (ICPP). As a new program, efforts are just getting underway toward addressing major issues related to the fuel and waste stored at the ICPP. The SF&WMTDP has the following principal objectives: Investigate direct dispositioning of spent fuel, striving for one acceptable waste form; determine the best treatment process(es) for liquid and calcine wastes to minimize the volume of high level radioactive waste (HLW) and low level waste (LLW); demonstrate the integrated operability and maintainability of selected treatment and immobilization processes; and assure that implementation of the selected waste treatment process is environmentally acceptable, ensures public and worker safety, and is economically feasible.

Manufacture of barium hexaferrite (BaO3.98Fe2O3) from iron oxide waste of grinding process by using calcination process

Science.gov (United States)

Idayanti, N.; Dedi; Kristiantoro, T.; Mulyadi, D.; Sudrajat, N.; Alam, G. F. N.

2018-03-01

The utilization of iron oxide waste of grinding process as raw materials for making barium hexaferrite has been completed by powder metallurgy method. The iron oxide waste was purified by roasting at 800 °C temperature for 3 hours. The method used varying calcination temperature at 1000, 1100, 1200, and 1250 °C for 3 hours. The starting iron oxide waste (Fe2O3) and barium carbonate (BaCO3) were prepared by mol ratio of Fe2O3:BaCO3 from the formula BaO3.98Fe2O3. Some additives such as calcium oxide (CaO), silicon dioxide (SiO2), and polyvinyl alcohol (PVA) were added after calcination process. The samples were formed at the pressure of 2 ton/cm2 and sintered at the temperature of 1250 °C for 1 hour. The formation of barium hexaferrite compounds after calcination is determined by X-Ray diffraction. The magnetic properties were observed by Permagraph-Magnet Physik with the optimum characteristic at calcination temperature of 1250 °C with the induction of remanence (Br) = 1.38 kG, coercivity (HcJ) = 4.533 kOe, product energy maximum (BHmax) = 1.086 MGOe, and density = 4.33 g/cm3.

Product removal and solids transport from fluidized-bed calciners

International Nuclear Information System (INIS)

Grimmett, E.S.; Munger, D.H.

1978-09-01

Methods of removing the solid product from pilot-plant and production fluidized-bed calciners, and transporting product to underground storage vaults are reported here. Testing of dense-phase solids transport systems in test loops during development of a 15-cm-diam. and 30-cm-diam. calciner are described. A lean-phase solid transport system is used with the Waste Calcining Facility. The results of some recent tests done in a lean-phase transport system connected to the 30-cm-diam. calciner are included in this report

Fluidized-bed calcination of LWR fuel-reprocessing HLLW: requirements and potential for off-gas cleanup

International Nuclear Information System (INIS)

Schindler, R.E.

1979-01-01

Fluidized-bed solidification (calcination) was developed on a pilot scale for a variety of simulated LWR high-level liquid-waste (HLLW) and blended high-level and intermediate-level liquid-waste (ILLW) compositions. It has also been demonstrated with ICPP fuel-reprocessing waste since 1963 in the Waste Calcining Facility (WCF) at gross feed rates of 5 to 12 m 3 /day. A fluidized-bed calciner produces a relatively large volume of off-gas. A calciner solidifying 6 m 3 /day of liquid waste would generate about 13 standard m 3 /min of off-gas containing 10 to 20 g of entrained solids per standard m 3 of off-gas. Use of an off-gas system similar to that of the WCF could provide an overall process decontamination factor for particulates of about 2 x 10 10 . A potential advantage of fluidized-bed calcination over other solidification methods is the ability to control ruthenium volatilization from the calciner at less than 0.01% by calcining at 500 0 C or above. Use of an off-gas system similar to that of the WCF would provide an overall process decontamination factor for volatile ruthenium of greater than 1.6 x 10 7

Fluidized bed calciner

International Nuclear Information System (INIS)

Sheely, W.F.

1986-01-01

A unique way to convert radioactive scrap into useful nuclear fuel products was developed for the Department of Energy at Hanford. An advanced, fluidized bed calciner is used to convert metallic nitrate scrap or waste solutions into benign, solid and gaseous products. There are broad potential applications of this concept beyond those in the nuclear industry

Volume reduction of low- and medium-level waste by incineration/calcination

International Nuclear Information System (INIS)

Buzonniere, A. de; Gauthey, J.C.

1993-01-01

Nuclear installations generate large quantities of low- and medium-level radwaste. This waste comes from various installations in the fuel cycle, reactor operation, research institute, hospitals, nuclear plate dismantling, etc.. TECHNICATOME did the project development work for the incineration plant of PIERRELATE (France) on behalf of COGEMA (Compagnie Generale des d'Etudes Technique). This plant has been in active service since November 1987. In addition, TECHNICATOME was in charge of the incinerator by a turnkey contract. This incinerator was commissioned in 1992. For a number of years, TECHNICATOME has been examining, developing and producing incineration and drying/calcination installations. They are used for precessing low- and medium-level radwaste

Remotely replaceable fuel and feed nozzles for the NWCF calciner vessel

International Nuclear Information System (INIS)

Fletcher, R.D.; Carter, J.A.; May, K.W.

1978-01-01

The development and testing of remotely replaceable fuel and feed nozzles for calcination of liquid radioactive wastes in the calciner vessel of the New Waste Calcining Facility (NWCF) being built at the Idaho National Engineering Laboratory are described. A complete fuel nozzle assembly was fabricated and tested at the Remote Maintenance Development Facility to evolve design refinements, identify required support equipment, and develop handling techniques. The design also provided for remote replacement of the nozzle support carriage and adjacent feed and fuel pipe loops using two pairs of master-slave manipulators

Evaluation and selection of aqueous-based technology for partitioning radionuclides from ICPP calcine

International Nuclear Information System (INIS)

Olson, A.L.; Schulz, W.W.; Burchfield, L.A.; Carlson, C.D.; Swanson, J.L.; Thompson, M.C.

1993-02-01

Early in 1993 Westinghouse Idaho Nuclear Company (WINCO) chartered a Panel of Nuclear Separations Experts. The purpose of this Panel was to assist WINCO scientists and engineers in selecting, evaluating, and ranking candidate aqueous-based processes and technologies for potential use in partitioning selected radionuclides from nitric acid solutions of retrieved Idaho Chemical Processing Plant (ICPP) calcine. Radionuclides of interest are all transuranium elements, 90 Sr, 99 Tc, 129 I, and 137 Cs. The six man Panel met for 4 days (February 16--19, 1993) on the campus of the Idaho State University in Pocatello, Idaho. Principal topics addressed included: Available radionuclide removal technology; applicability of separations technology and processes to ICPP calcine; and potential integrated radionuclide partitioning schemes. This report, prepared from contributions from all Panel members, presents a comprehensive account of the proceedings and significant findings of the February, 1993 meeting in Pocatello

Fall 2010 Semiannual (III.H. and I.U.) Report for the HWMA/RCRA Post Closure Permit for the INTEC Waste Calcining Facility and the CPP 601/627/640 Facility at the INL Site

Energy Technology Data Exchange (ETDEWEB)

Boehmer, Ann

2010-11-01

The Waste Calcining Facility is located at the Idaho Nuclear Technology and Engineering Center. In 1999, the Waste Calcining Facility was closed under an approved Hazardous Waste Management Act/Resource Conservation and Recovery Act (HWMA/RCRA) Closure Plan. Vessels and spaces were grouted and then covered with a concrete cap. The Idaho Department of Environmental Quality issued a final HWMA/RCRA post-closure permit on September 15, 2003, with an effective date of October 16, 2003. This permit sets forth procedural requirements for groundwater characterization and monitoring, maintenance, and inspections of the Waste Calcining Facility to ensure continued protection of human health and the environment. The post closure permit also includes semiannual reporting requirements under Permit Conditions III.H. and I.U. These reporting requirements have been combined into this single semiannual report, as agreed between the Idaho Cleanup Project and Idaho Department of Environmental Quality. The Permit Condition III.H. portion of this report includes a description and the results of field methods associated with groundwater monitoring of the Waste Calcining Facility. Analytical results from groundwater sampling, results of inspections and maintenance of monitoring wells in the Waste Calcining Facility groundwater monitoring network, and results of inspections of the concrete cap are summarized. The Permit Condition I.U. portion of this report includes noncompliances not otherwise required to be reported under Permit Condition I.R. (advance notice of planned changes to facility activity which may result in a noncompliance) or Permit Condition I.T. (reporting of noncompliances which may endanger human health or the environment). This report also provides groundwater sampling results for wells that were installed and monitored as part of the Phase 1 post-closure period of the landfill closure components in accordance with HWMA/RCRA Landfill Closure Plan for the CPP-601 Deep

Dissolution of two NWCF calcines: Extent of dissolution and characterization of undissolved solids

International Nuclear Information System (INIS)

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

1995-01-01

A study was undertaken to determine the dissolution characteristics of two NWCF calcine types. A two-way blended calcine made from 4 parts nonradioactive aluminum nitrate and one part WM-102 was studied to determine the extent of dissolution for aluminum-type calcines. A two-way blend of 3.5 parts fluorinel waste from WM-187 and 1 part sodium waste from WM-185 was used to determine the extent of dissolution for zirconium-type calcines. This study was necessary to develop suitable aqueous separation flowsheets for the partitioning of actinides and fission products from ICPP calcines and to determine the disposition of the resulting undissolved solids (UDS). The dissolution flowsheet developed by Herbst was used to dissolve these two NWCF calcine types. Results show that greater than 95 wt% of aluminum and zirconium calcine types were dissolved after a single batch contact with 5 M HNO 3 . A characterization of the UDS indicates that the weight percent of TRU elements in the UDS resulting from both calcine type dissolutions increases by approximately an order of magnitude from their concentrations prior to dissolution. Substantial activities of cesium and strontium are also present in the UDS resulting from the dissolution of both calcine types. Multiple TRU, Cs, and Sr analyses of both UDS types show that these solids are relatively homogeneous. From this study, it is estimated that between 63.5 and 635 cubic meters of UDS will be generated from the dissolution of 3800 M 3 of calcine. The significant actinide and fission product activities in these UDS will preclude their disposal as low-level waste. If the actinide and fission activity resulting from the UDS is the only considered source in the dissolved calcine solutions, an estimated 99.9 to 99.99 percent of the solids must be removed from this solution for it to meet non-TRU Class A low-level waste

Behavior of radioactive iodine and technetium in the spray calcination of high-level waste

Science.gov (United States)

Knox, C. A.; Farnsworth, R. K.

1981-08-01

The Remote Laboratory-Scale Waste Treatment Facility (RLSWTF) was designed and built as a part of the High-Level Waste Immobilization Program (now the High-Level Waste Process Development Program) at the Pacific Northwest Laboratory. In facility, installed in a radiochemical cell, is described in which installed in a radiochemical cell is described in which small volumes of radioactive liquid wastes can be solidified, the process off gas can be analyzed, and the methods for decontaminating this off gas can be tested. During the spray calcination of commercial high-level liquid waste spiked with Tc-99 and I-131 and 31 wt% loss of I-131 past the sintered-metal filters. These filters and venturi scrubber were very efficient in removing particulates and Tc-99 from the the off-gas stream. Liquid scrubbers were not efficient in removing I-131 as 25% of the total lost went to the building off-gas system. Therefore, solid adsorbents are needed to remove iodine. For all future operations where iodine is present, a silver zeolite adsorber is to be used.

Physical, Chemical and Structural Evolution of Zeolite - Containing Waste Forms Produced from Metakaolinite and Calcined HLW

International Nuclear Information System (INIS)

Grutzeck, Michael

2005-01-01

During the seventh year of the current grant (DE-FG02-05ER63966) we completed an exhaustive study of cold calcination and began work on the development of tank fill materials to fill empty tanks and control residuals. Cold calcination of low and high NOx low activity waste (LAW) SRS Tank 44 and Hanford AN-107 simulants, respectively with metallic Al + Si powders was evaluated. It was found that a combination of Al and Si powders could be used as reducing agents to reduce the nitrate and nitrite content of both low and high NOx LAW to low enough levels to allow the LAW to be solidified directly by mixing it with metakaolin and allowing it to cure at 90 C. During room temperature reactions, NOx was reduced and nitrogen was emitted as N2 or NH3. This was an important finding because now one can pretreat LAW at ambient temperatures which provides a low-temperature alternative to thermal calcination. The significant advantage of using Al and Si metals for denitration/denitrition of the LAW is the fact that the supernate could potentially be treated in situ in the waste tanks themselves. Tank fill materials based upon a hydroceramic binder have been formulated from mixtures of metakaolinite, Class F fly ash and Class C flue gas desulphurization (FGD) ash mixed with various concentrations of NaOH solution. These harden over a period of hours or days depending on composition. A systematic study of properties of the tank fill materials (leachability) and ability to adsorb and hold residuals is under way

Removal of acid blue 062 on aqueous solution using calcinated colemanite ore waste

Energy Technology Data Exchange (ETDEWEB)

Atar, Necip [Department of Chemistry, Faculty of Arts and Science, University of Dumlupinar, Kuetahya (Turkey); Olgun, Asim [Department of Chemistry, Faculty of Arts and Science, University of Dumlupinar, Kuetahya (Turkey)]. E-mail: aolgun@dumlupinar.edu.tr

2007-07-19

Colemanite ore waste (CW) has been employed as adsorbent for the removal of acid blue 062 anionic dye (AB 062) from aqueous solution. The adsorption of AB 062 onto CW was examined with respect to contact time, calcination temperature, particle size, pH, adsorbent dosage and temperature. The physical and chemical properties of the CW, such as particle sizes and calcinations temperature, play important roles in dye adsorption. The dye adsorption largely depends on the initial pH of the solution with maximum uptake occurring at pH 1.Three simplified kinetics models, namely, pseudo-first order, pseudo-second order, and intraparticle diffusion models were tested to investigate the adsorption mechanisms. The kinetic adsorption of AB 062 on CW follows a pseudo-second order equation. The adsorption data have been analyzed using Langmuir and Freundlich isotherms. The results indicate that the Langmuir model provides the best correlation of the experimental data. Isotherms have also been used to obtain the thermodynamic parameters such as free energy, enthalpy and entropy of the adsorption of dye onto CW.

Removal of acid blue 062 on aqueous solution using calcinated colemanite ore waste

International Nuclear Information System (INIS)

Atar, Necip; Olgun, Asim

2007-01-01

Colemanite ore waste (CW) has been employed as adsorbent for the removal of acid blue 062 anionic dye (AB 062) from aqueous solution. The adsorption of AB 062 onto CW was examined with respect to contact time, calcination temperature, particle size, pH, adsorbent dosage and temperature. The physical and chemical properties of the CW, such as particle sizes and calcinations temperature, play important roles in dye adsorption. The dye adsorption largely depends on the initial pH of the solution with maximum uptake occurring at pH 1.Three simplified kinetics models, namely, pseudo-first order, pseudo-second order, and intraparticle diffusion models were tested to investigate the adsorption mechanisms. The kinetic adsorption of AB 062 on CW follows a pseudo-second order equation. The adsorption data have been analyzed using Langmuir and Freundlich isotherms. The results indicate that the Langmuir model provides the best correlation of the experimental data. Isotherms have also been used to obtain the thermodynamic parameters such as free energy, enthalpy and entropy of the adsorption of dye onto CW

Mercury removal at Idaho National Engineering and Environmental Laboratory's New Waste Calcining Facility

Energy Technology Data Exchange (ETDEWEB)

S. C. Ashworth

2000-02-27

Technologies were investigated to determine viable processes for removing mercury from the calciner (NWCF) offgas system at the Idaho National Engineering and Environmental Laboratory. Technologies for gas phase and aqueous phase treatment were evaluated. The technologies determined are intended to meet EPA Maximum Achievable Control Technology (MACT) requirements under the Clean Air Act and Resource Conservation and Recovery Act (RCRA). Currently, mercury accumulation in the calciner off-gas scrubbing system is transferred to the tank farm. These transfers lead to accumulation in the liquid heels of the tanks. The principal objective for aqueous phase mercury removal is heel mercury reduction. The system presents a challenge to traditional methods because of the presence of nitrogen oxides in the gas phase and high nitric acid in the aqueous scrubbing solution. Many old and new technologies were evaluated including sorbents and absorption in the gas phase and ion exchange, membranes/sorption, galvanic methods, and UV reduction in the aqueous phase. Process modifications and feed pre-treatment were also evaluated. Various properties of mercury and its compounds were summarized and speciation was predicted based on thermodynamics. Three systems (process modification, NOxidizer combustor, and electrochemical aqueous phase treatment) and additional technology testing were recommended.

Mercury Removal at Idaho National Engineering and Environmental Laboratory's New Waste Calcining Facility

Energy Technology Data Exchange (ETDEWEB)

Ashworth, Samuel Clay; Wood, R. A.; Taylor, D. D.; Sieme, D. D.

2000-03-01

Technologies were investigated to determine viable processes for removing mercury from the calciner (NWCF) offgas system at the Idaho National Engineering and Environmental Laboratory. Technologies for gas phase and aqueous phase treatment were evaluated. The technologies determined are intended to meet EPA Maximum Achievable Control Technology (MACT) requirements under the Clean Air Act and Resource Conservation and Recovery Act (RCRA). Currently, mercury accumulation in the calciner off-gas scrubbing system is transferred to the tank farm. These transfers lead to accumulation in the liquid heels of the tanks. The principal objective for aqueous phase mercury removal is heel mercury reduction. The system presents a challenge to traditional methods because of the presence of nitrogen oxides in the gas phase and high nitric acid in the aqueous scrubbing solution. Many old and new technologies were evaluated including sorbents and absorption in the gas phase and ion exchange, membranes/sorption, galvanic methods, and UV reduction in the aqueous phase. Process modifications and feed pre-treatment were also evaluated. Various properties of mercury and its compounds were summarized and speciation was predicted based on thermodynamics. Three systems (process modification, NOxidizer combustor, and electrochemical aqueous phase treatment) and additional technology testing were recommended.

Calcined Eggshell Waste for Mitigating Soil Antibiotic-Resistant Bacteria/Antibiotic Resistance Gene Dissemination and Accumulation in Bell Pepper.

Science.gov (United States)

Ye, Mao; Sun, Mingming; Feng, Yanfang; Li, Xu; Schwab, Arthur P; Wan, Jinzhong; Liu, Manqiang; Tian, Da; Liu, Kuan; Wu, Jun; Jiang, Xin

2016-07-13

The combined accumulation of antibiotics, heavy metals, antibiotic-resistant bacteria (ARB)/antibiotic resistance genes (ARGs) in vegetables has become a new threat to human health. This is the first study to investigate the feasibility of calcined eggshells modified by aluminum sulfate as novel agricultural wastes to impede mixed contaminants from transferring to bell pepper (Capsicum annuum L.). In this work, calcined eggshell amendment mitigated mixed pollutant accumulation in bell pepper significantly, enhanced the dissipation of soil tetracycline, sulfadiazine, roxithromycin, and chloramphenicol, decreased the water-soluble fractions of antibiotics, and declined the diversity of ARB/ARGs inside the vegetable. Moreover, quantitative polymerase chain reaction analysis detected that ARG levels in the bell pepper fruits significantly decreased to 10(-10) copies/16S copies, indicating limited risk of ARGs transferring along the food chain. Furthermore, the restoration of soil microbial biological function suggests that calcined eggshell is an environmentally friendly amendment to control the dissemination of soil ARB/ARGs in the soil-vegetable system.

Microwave calcination for plutonium immobilization and residue stabilization

Energy Technology Data Exchange (ETDEWEB)

Harris, M.J.; Rising, T.L.; Roushey, W.J.; Sprenger, G.S. [Kaiser-Hill Co., Golden, CO (United States)

1995-12-01

In the late 1980`s development was begun on a process using microwave energy to vitrify low level mixed waste sludge and transuranic mixed waste sludge generated in Building 374 at Rocky Flats. This process was shown to produce a dense, highly durable waste form. With the cessation of weapons production at Rocky Flats, the emphasis has changed from treatment of low level and TRU wastes to stabilizaiton of plutonium oxide and residues. This equipment is versatile and can be used as a heat source to calcine, react or vitrify many types of residues and oxides. It has natural economies in that it heats only the material to be treated, significantly reducing cycle times over conventional furnaces. It is inexpensive to operate in that most of the working components remain outside of any necessary contamination enclosure and therefore can easily be maintained. Limited testing has been successfully performed on cerium oxide (as a surrogate for plutonium oxide), surrogate electrorefining salts, surrogate residue sludge and residue ash. Future plans also include tests on ion exchange resins. In an attempt to further the usefullness of this technology, a mobile, self-contained microwave melting system is currently under development and expected to be operational at Rocky Flats Enviromental Technology Site by the 4th quarter of FY96.

Microwave calcination for plutonium immobilization and residue stabilization

International Nuclear Information System (INIS)

Harris, M.J.; Rising, T.L.; Roushey, W.J.; Sprenger, G.S.

1995-01-01

In the late 1980's development was begun on a process using microwave energy to vitrify low level mixed waste sludge and transuranic mixed waste sludge generated in Building 374 at Rocky Flats. This process was shown to produce a dense, highly durable waste form. With the cessation of weapons production at Rocky Flats, the emphasis has changed from treatment of low level and TRU wastes to stabilizaiton of plutonium oxide and residues. This equipment is versatile and can be used as a heat source to calcine, react or vitrify many types of residues and oxides. It has natural economies in that it heats only the material to be treated, significantly reducing cycle times over conventional furnaces. It is inexpensive to operate in that most of the working components remain outside of any necessary contamination enclosure and therefore can easily be maintained. Limited testing has been successfully performed on cerium oxide (as a surrogate for plutonium oxide), surrogate electrorefining salts, surrogate residue sludge and residue ash. Future plans also include tests on ion exchange resins. In an attempt to further the usefullness of this technology, a mobile, self-contained microwave melting system is currently under development and expected to be operational at Rocky Flats Enviromental Technology Site by the 4th quarter of FY96

Retrofit design of remotely removable decontamination spray nozzles for the new waste calcining facility at the Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Gay, J.A.

1988-01-01

High level radioactive liquid waste is converted to a solid form at the Idaho Chemical Processing Plant (ICPP). The conversion is done by a fluidized bed combustion process in the calciner vessel. The interior decontamination system for the calciner vessel uses a common header bolted to four decontamination nozzles around the upper head. The retrofit was required to eliminate hands-on maintenance and difficulty in nozzle removal because of nozzle plugging. The retrofit design for this project demonstrates the solution of problems associated with thermal phenomena, structural supports, seismic requirements, remote handling and installations into extremely restricted spaces

High-Level Waste Vitrification Facility Feasibility Study

Energy Technology Data Exchange (ETDEWEB)

D. A. Lopez

1999-08-01

A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035.

High-Level Waste Vitrification Facility Feasibility Study

International Nuclear Information System (INIS)

D. A. Lopez

1999-01-01

A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035

Leaching properties and chemical compositions of calcines produced at the Idaho Chemical Processing Plant

International Nuclear Information System (INIS)

Staples, B.A.; Paige, B.E.; Rhodes, D.W.; Wilding, M.W.

1980-01-01

No significant chemical differences were determined between retrieved and fresh calcine based on chemical and spectrochemical analyses. Little can be derived from the amounts of the radioisotopes present in the retrieved calcine samples other than the ratios of strontium-90 to cesium-137 are typical of aged fission product. The variations in concentrations of radionuclides within the composite samples of each bin also reflect the differences in compositions of waste solutions calcined. In general the leaching characteristics of both calcines by distilled water are similar. In both materials the radionuclides of cesium and strontium were selectively leached at significant rates, although cesium leached much more completely from the alumina calcine than from the zirconia calcine. Cesium and strontium are probably contained in both calcines as nitrate salts and also as fluoride salts in zirconia calcine, all of which are at least slightly soluble in water. Radionuclides of cerium, ruthenium, and plutonium in both calcines were highly resistant to leaching and leached at rates similar to or less than those of the matrix elements. These elements exist as polyvalent metal ions in the waste solutions before calcination and they probably form insoluble oxides and fluorides in the calcine. The relatively slow leaching of nitrate ion from zirconia calcine and radiocesium from both calcines suggests that the calcine matrix in some manner prevents complete or immediate contact of the soluble ions with water. Whether radiostrontium forms slightly fluoride salts or forms nitrate salts which are protected in the same manner as radiocesium is unknown. Nevertheless, selective leaching of cesium and strontim is retarded in some manner by the calcine matrix

Solidification of highly active wastes

International Nuclear Information System (INIS)

Morris, J.B.

1984-11-01

Final reports are presented on work on the following topics: glass technology; enhancement of off-gas aerosol collection; formation and trapping of volatile ruthenium; volatilisation of caesium, technetium and tellurium in high-level waste vitrification; deposition of ruthenium; and calcination of high-level waste liquors. (author)

Behavior of radioactive iodine and technetium in the spray calcination of high-level waste

International Nuclear Information System (INIS)

Knox, C.A.; Farnsworth, R.K.

1981-08-01

The Remote Laboratory-Scale Waste Treatment Facility (RLSWTF) was designed and built as a part of the High-Level Waste Immobilization Program (now the High-Level Waste Process Development Program) at the Pacific Northwest Laboratory. In this facility, which is installed in a radiochemical cell, small volumes of radioactive liquid wastes can be solidified, the process off gas can be analyzed, and the methods for decontaminating this off gas can be tested. Initial operations were completed with nonradioactive, simulated waste solutions (Knox, Siemens and Berger 1981). The first radioactive operations in this facility were performed with a simulated, commercial waste composition containing tracer levels of 99 Tc and 131 I. This report describes the facility and test operations and presents the results of the behavior of 131 I and 99 Tc during solidification of radioactive liquid wastes. During the spray calcination of commercial high-level liquid waste spiked with 99 Tc and 131 I, there was a 0.3 wt% loss of particulates, a 0.15 wt% loss of 99 Tc and a 31 wt% loss of 131 I past the sintered-metal filters. These filters and a venturi scrubber were very efficient in removing particulates and 99 Tc from the off-gas stream. Liquid scrubbers were not efficient in removing 131 I, as 25% of the total lost went to the building off-gas system. Therefore, solid adsorbents will be needed to remove iodine. For all future RLSWTF operations where iodine is present, a silver zeolite adsorber will be used

Foreign travel report: Visits to UK, Belgium, Germany, and France to benchmark European spent fuel and waste management technology

International Nuclear Information System (INIS)

Ermold, L.F.; Knecht, D.A.

1993-08-01

The ICPP WINCO Spent Fuel and Waste Management Development Program recently was funded by DOE-EM to develop new technologies for immobilizing ICPP spent fuels, sodium-bearing liquid waste, and calcine to a form suitable for disposal. European organizations are heavily involved, in some cases on an industrial scale in areas of waste management, including spent fuel disposal and HLW vitrification. The purpose of this trip was to acquire first-hand European efforts in handling of spent reactor fuel and nuclear waste management, including their processing and technical capabilities as well as their future planning. Even though some differences exist in European and U.S. DOE waste compositions and regulations, many aspects of the European technologies may be applicable to the U.S. efforts, and several areas offer potential for technical collaboration

Foreign travel report: Visits to UK, Belgium, Germany, and France to benchmark European spent fuel and waste management technology

Energy Technology Data Exchange (ETDEWEB)

Ermold, L.F.; Knecht, D.A.

1993-08-01

The ICPP WINCO Spent Fuel and Waste Management Development Program recently was funded by DOE-EM to develop new technologies for immobilizing ICPP spent fuels, sodium-bearing liquid waste, and calcine to a form suitable for disposal. European organizations are heavily involved, in some cases on an industrial scale in areas of waste management, including spent fuel disposal and HLW vitrification. The purpose of this trip was to acquire first-hand European efforts in handling of spent reactor fuel and nuclear waste management, including their processing and technical capabilities as well as their future planning. Even though some differences exist in European and U.S. DOE waste compositions and regulations, many aspects of the European technologies may be applicable to the U.S. efforts, and several areas offer potential for technical collaboration.

Research about the pozzolanic activity of waste materials from calcined clay

Directory of Open Access Journals (Sweden)

Sánchez de Rojas, M. I.

2001-03-01

Full Text Available To recycle and reutilise waste materials and find definite applications for their use, it is necessary to have a deep knowledge of them. The aim of this study is to study the possibility of using waste materials from calcined clay, actually ceramic tile, once crushed and grounded, as pozzolanic material. For this purpose, different tests are carried out in order to establish the pozzolanic activity of this material. At the same time, these results are compared to those of other industrial by-products, fly ash and silica fume, which are pozzolanic materials usually employed to elaborate mortars and concretes.

Para llevar a cabo labores encaminadas al reciclado y revalorización de residuos es necesario un conocimiento profundo de los mismos, de forma que se busquen aplicaciones concretas de uso. El objetivo de este estudio es investigar la posibilidad de utilizar materiales de desecho procedentes de arcilla cocida, concretamente teja cerámica, una vez triturada y molida, como puzolana. Para ello, se efectúan diferentes ensayos dirigidos a establecer la actividad puzolanica del material. A su vez, estos resultados son comparados con otros residuos industriales, ceniza volante y humo de sílice, habituales en la elaboración de morteros y hormigones.

Spray solidification of nuclear waste

International Nuclear Information System (INIS)

Bonner, W.F.; Blair, H.T.; Romero, L.S.

1976-08-01

The spray calciner is a relatively simple machine. Operation is simple and is easily automated. Startup and shutdown can be performed in less than an hour. A wide variety of waste compositions and concentrations can be calcined under easily maintainable conditions. Spray calcination of high-level and mixed high- and intermediate-level liquid wastes has been demonstrated. Waste concentrations of from near infinite dilution to less than 225 liters per tonne of fuel are calcinable. Wastes have been calcined containing over 2M sodium. Feed concentration, composition, and flowrate can vary rapidly by over a factor of two without requiring operator action. Wastes containing mainly sodium cations can be spray calcined by addition of finely divided silica to the feedstock. A remotely replaceable atomizing nozzle has been developed for use in plant-scale equipment. Calciner capacity of over 75 l/h has been demonstrated in pilot-scale equipment. Sintered stainless steel filters are effective in deentraining over 99.9 percent of the solids that result from calcining the feedstock. The volume of recycle required from the effluent treatment system is very small. Vibrator action maintains the calcine holdup in the calciner at less than 1 kg. Successful remote operation and maintenance of a heated-wall spray calciner have been demonstrated while processing high-level waste. Radionuclide volatilization was acceptably low

Solidification of highly active wastes

International Nuclear Information System (INIS)

Morris, J.B.

1986-07-01

This document contains the annual reports for the contracts: (A) Glass Technology; (B) Calcination of Highly Active Waste Liquors; (C) Formation and Trapping of Volatile Ruthenium; (D) Deposition of Ruthenium; (E) Enhancement of Off-Gas Aerosol Collection; (F) Volatilisation of Cs, Tc and Te in High Level Waste Vitrification. (author)

Environmental assessment: Closure of the Waste Calcining Facility (CPP-633), Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

1996-07-01

The U.S. Department of Energy (DOE) proposes to close the Waste Calcining Facility (WCF). The WCF is a surplus DOE facility located at the Idaho Chemical Processing Plant (ICPP) on the Idaho National Engineering Laboratory (INEL). Six facility components in the WCF have been identified as Resource Conservation and Recovery Ace (RCRA)-units in the INEL RCRA Part A application. The WCF is an interim status facility. Consequently, the proposed WCF closure must comply with Idaho Rules and Standards for Hazardous Waste contained in the Idaho Administrative Procedures Act (IDAPA) Section 16.01.05. These state regulations, in addition to prescribing other requirements, incorporate by reference the federal regulations, found at 40 CFR Part 265, that prescribe the requirements for facilities granted interim status pursuant to the RCRA. The purpose of the proposed action is to reduce the risk of radioactive exposure and release of hazardous constituents and eliminate the need for extensive long-term surveillance and maintenance. DOE has determined that the closure is needed to reduce potential risks to human health and the environment, and to comply with the Idaho Hazardous Waste Management Act (HWMA) requirements

Dissolution studies with pilot plant and actual INTEC calcines

International Nuclear Information System (INIS)

Herbst, R.S.; Garn, T.G.

1999-01-01

The dissolution of Idaho Nuclear Technology and Engineering Center (INTEC) pilot plant calcines was examined to determine solubility of calcine matrix components in acidic media. Two representatives pilot plant calcine types were studied: Zirconia calcine and Zirconia/Sodium calcine. Dissolution of these calcines was evaluated using lower initial concentrations of nitric acid than used in previous tests to decrease the [H+] concentration in the final solutions. Lower [H+] concentrations contribute to more favorable TRUEX/SREX solvent extraction flowsheet performance. Dissolution and analytical results were also obtained for radioactive calcines produced using high sodium feeds blended with non-radioactive Al(NO 3 ) 3 solutions to dilute the sodium concentration and prevent bed agglomeration during the calcination process. Dissolution tests indicated >95 wt.% of the initial calcine mass can be dissolved using the baseline dissolution procedure, with the exception that higher initial nitric acid concentrations are required. The higher initial acid concentration is required for stoichiometric dissolution of the oxides, primarily aluminum oxide. Statistically designed experiments using pilot plant calcine were performed to determine the effect of mixing rate on dissolution efficiency. Mixing rate was determined to provide minimal effects on wt.% dissolution. The acid/calcine ratio and temperature were the predominate variables affecting the wt.% dissolution, a result consistent with previous studies using other similar types of pilot plant calcines

Idaho Chemical Processing Plant Spent Fuel and Waste Management Technology Development Program Plan

International Nuclear Information System (INIS)

1993-09-01

The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage and reprocessing since 1953. Reprocessing of SNF has resulted in an existing inventory of 1.5 million gallons of radioactive sodium-bearing liquid waste and 3800 cubic meters (m 3 ) of calcine, in addition to the 768 metric tons (MT) of SNF and various other fuel materials in inventory. To date, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, recent changes in world events have diminished the demand to recover and recycle this material. As a result, DOE has discontinued reprocessing SNF for uranium recovery, making the need to properly manage and dispose of these and future materials a high priority. In accordance with the Nuclear Waste Policy Act (NWPA) of 1982, as amended, disposal of SNF and high-level waste (HLW) is planned for a geological repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP Spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will properly stored and prepared for final disposal. Program elements in support of acceptable interim storage and waste minimization include: developing and implementing improved radioactive waste treatment technologies; identifying and implementing enhanced decontamination and decommissioning techniques; developing radioactive scrap metal (RSM) recycle capabilities; and developing and implementing improved technologies for the interim storage of SNF

Review of metal-matrix encapsulation of solidified radioactive high-level waste

International Nuclear Information System (INIS)

Jardine, L.J.; Steindler, M.J.

1978-05-01

Literature describing previous and current work on the encapsulation of solidified high-level waste forms in a metal matrix was reviewed. Encapsulation of either stabilized calcine pellets or glass beads in alloys by casting techniques was concluded to be the most developed and direct approach to fabricating solid metal-matrix waste forms. Further characterizations of the physical and chemical properties of metal-matrix waste forms are still needed to assess the net attributes of metal-encapsulation alternatives. Steady-state heat transfer properties of waste canisters in air and water environments were calculated for four reference waste forms: (1) calcine, (2) glass monoliths, (3) metal-encapsulated calcine, and (4) metal-encapsulated glass beads. A set of criteria for the maximum allowable canister centerline and surface temperatures and heat generation rates per canister at the time of shipment to a Federal repository was assumed, and comparisons were made between canisters of these reference waste forms of the shortest time after reactor discharge that canisters could be filled and the subsequent ''interim'' storage times prior to shipment to a Federal repository for various canister diameters and waste ages. A reference conceptual flowsheet based on existing or developing technology for encapsulation of stabilized calcine pellets is discussed. Conclusions and recommendations are presented

Radioactive waste management

International Nuclear Information System (INIS)

Slansky, C.M.

1975-01-01

High-level radioactive waste is produced at Idaho Chemical Processing Plant (ICPP) during the recovery of spent highly enriched nuclear fuels. Liquid waste is stored safely in doubly contained tanks made of steel. The liquid waste is calcined to a solid and stored safely in a retrievable form in doubly contained underground bins. The calcine can be treated further or left untreated in anticipation of ultimate storage. Fluidized bed calcination has been applied to many kinds of high-level waste. The environmental impact of high-level waste management at the ICcP has been negligible and should continue to be negligible. 13 refs

Final safety-analysis report for the Fifth Calcined Solids Storage Facility

International Nuclear Information System (INIS)

1982-01-01

Radioactive aqueous wastes generated by the solvent extraction of uranium from expended fuels at ICPP will be calcined in the New Waste Calcining Facility (NWCF). The calcined solids are pneumatically transferred to stainless steel bins enclosed in concrete vaults for interim storage of up to 500 years. The Fifth Calcined Solids Storage Facility (CSSF) provides 1000 m 3 of storage and consists of seven annular stainless steel bins inside a reinforced concrete vault set on bedrock. Storage of calcined solids is essentially a passive operation with very little opportunity for release of radionuclides and with no potential for criticality. There will be no potential for fire or explosion. Shielding has been designed to assure that the radiation levels at the vault exterior surfaces will be limited to less than 0.5 mRem/h. A sump in the vault floor will collect any in-leakage that may occur. Any water that collects in the sump will be sampled then removed with the sump jet. There will be an extremely small chance of release of radioactive particulates into the atmosphere as a result of a bin leak. The Design Basis Accident (DBA) postulates the spill of solids from an eroded fill line into the vault coupled with a failure of the vault cooling air radiation monitor. For the DBA, the maximum calculated radiation dose to an exposed individual near the site boundary is less than 1.2 μRem to the bone and lung

Kinetic Study of Calcination of Jakura Limestone Using Power Rate ...

African Journals Online (AJOL)

National Research Institute for Chemical Technology, P.M. B 1052, Zaria, ... calcination of Jakura limestone was also found to be first order reaction with respect to CaCO3 ... Keywords: Jakura, limestone, calcination, kinetics, power law model.

Inductive classification of operating data from a fluidized bed calciner

International Nuclear Information System (INIS)

O'Brien, B.H.

1990-01-01

A process flowsheet expert system for a fluidized bed calciner which solidifies high-level radioactive liquid waste was developed from pilot-plant data using a commercial, inductive classification program. After initial classification of the data, the resulting rules were inspected and adjusted to match existing knowledge of process chemistry. The final expert system predicts performance of process flowsheets based upon the chemical composition of the calciner feed and has been successfully used to identify potential operational problems prior to calciner pilot-plant testing of new flowsheets and to provide starting parameters for pilot-plant tests. By using inductive classification techniques to develop the initial rules from the calciner pilot-plant data and using existing process knowledge to verify the accuracy of these rules, an effective expert system was developed with a minimum amount of effort. This method may be applied for developing expert systems for other processes where numerous operating data are available and only general process chemistry effects are known

Effect of the heat curing on strength development of self-compacting mortars containing calcined silt of dams and Ground Brick Waste

Directory of Open Access Journals (Sweden)

B. Safi

2013-01-01

Full Text Available The strength development of self-compacting mortars (SCM containing calcined silt (CS and ground brick waste (GWB was investigated. The variables are the nature of addition (CS and GWB in the binder and the heat curing at different temperatures (20 ºC and 60 ºC at 7 and 14 days of curing. Two temperatures 20 and 60 ºC were applied to samples with intermediate levels (depending on the drying method applied to precast for 18 hours in total. In this study, a Portland cement (CEMII, Calcined silt (750 ºC for 5 hours, ground waste brick, were used in the binders of SCM. The results show that the compressive strength to 14 days of mortars, increases with annealing (60 ºC compared to that measured at 20 ºC. Also, values of compressive strength of mortars at 14 days that are close to those obtained without 28 days curing treatment. Indeed, a strength gain of about 20.5% and 27.3% was obtained respectively for the SCM with GWB and the SCM with CS. However, a small change in mass recorded for both types of mortars.

Development of multibarrier nuclear waste forms

International Nuclear Information System (INIS)

1979-03-01

The multibarrier concept aims to separate the radionuclide-containing inner core material and the environment by the use of coatings and matrices. Two options were developed for the inner core of the multibarrier concept: supercalcine pellets and glass marbles. Supercalcine is a crystalline assemblage of mutually compatible, refractory, and leach-resistant solid solution phases incorporating high-level liquid waste ions. Supercalcine powder is produced by spray calcining the liquid waste stream to which Al 2 O 3 , CaO, SiO 2 , and SrO have been added. Supercalcine pellets are produced by disc pelletizing. The amorphous supercalcine crystallizes into solid solution phases after subsequent heat treatment. Based on the multibarrier processes described, several conclusions can be made: gravity sintering and vacuum casting are both applicable methods for metal matrix encapsulation. The multibarrier concept of glass marbles encapsulated in a vacuum-cast lead alloy provides enhanced inertness at a minimum increase in technological complexity. If it were desirable to develop a crystalline multibarrier waste form, uncoated sintered supercalcine pellets would offer enhanced inertness at a much lower level of technological complexity than glaze- or CVD-coated supercalcine. The 16-inch diameter pelletizer unit has enough capacity to handle the output of a large PNL spray calciner (52.5 kg of calcine/hr) and it can form spray-calcined material into pellets with diameters of 2 mm to 20 mm having strength enough to withstand handling without significant breakage.Chemical vapor deposition coating of supercalcine should be pursued only if a very high level of inertness is required

Waste disposal options report. Volume 2

International Nuclear Information System (INIS)

Russell, N.E.; McDonald, T.G.; Banaee, J.; Barnes, C.M.; Fish, L.W.; Losinski, S.J.; Peterson, H.K.; Sterbentz, J.W.; Wenzel, D.R.

1998-02-01

Volume 2 contains the following topical sections: estimates of feed and waste volumes, compositions, and properties; evaluation of radionuclide inventory for Zr calcine; evaluation of radionuclide inventory for Al calcine; determination of k eff for high level waste canisters in various configurations; review of ceramic silicone foam for radioactive waste disposal; epoxides for low-level radioactive waste disposal; evaluation of several neutralization cases in processing calcine and sodium-bearing waste; background information for EFEs, dose rates, watts/canister, and PE-curies; waste disposal options assumptions; update of radiation field definition and thermal generation rates for calcine process packages of various geometries-HKP-26-97; and standard criteria of candidate repositories and environmental regulations for the treatment and disposal of ICPP radioactive mixed wastes

Waste disposal options report. Volume 2

Energy Technology Data Exchange (ETDEWEB)

Russell, N.E.; McDonald, T.G.; Banaee, J.; Barnes, C.M.; Fish, L.W.; Losinski, S.J.; Peterson, H.K.; Sterbentz, J.W.; Wenzel, D.R.

1998-02-01

Volume 2 contains the following topical sections: estimates of feed and waste volumes, compositions, and properties; evaluation of radionuclide inventory for Zr calcine; evaluation of radionuclide inventory for Al calcine; determination of k{sub eff} for high level waste canisters in various configurations; review of ceramic silicone foam for radioactive waste disposal; epoxides for low-level radioactive waste disposal; evaluation of several neutralization cases in processing calcine and sodium-bearing waste; background information for EFEs, dose rates, watts/canister, and PE-curies; waste disposal options assumptions; update of radiation field definition and thermal generation rates for calcine process packages of various geometries-HKP-26-97; and standard criteria of candidate repositories and environmental regulations for the treatment and disposal of ICPP radioactive mixed wastes.

Conceptual design for vitrification of HLW at West Valley using a rotary calciner/metallic melter

International Nuclear Information System (INIS)

Giraud, J.P.; Conord, J.P.; Saverot, P.M.

1984-01-01

The CEA has had an extensive research program in the field of vitrification technology for over 24 years, and several testing facilities were used throughout all phases of development and engineering: The Vulcain facility comprises a vitrification hot cell and four auxiliary hot cells. Vulcain allows the production of 2-kg samples of active glass. The off-gas treatment system allows testing the DF of each equipment. The auxiliary cells are equipped with leach-rate tests, diffusion tests, and irradiation tests on the glass samples. The Atlas facility is a reproduction of AVM calcination and vitrification furnaces at 1/2 scale enclosed in a glove box. This facility is used for testing ruthenium volatility and containment in the vitrification process. The full-scale AVM inactive pilot facility is used for testing calcination and vitrification of new compositions of high-level waste and for developing new types of vitrification furnaces. The inactive test loop is for testing air cooling of glass containers. The full-scale AVH inactive pilot facility is used for testing AVH technology and has been in operation since late 1981

Radioactive waste management

International Nuclear Information System (INIS)

Kizawa, Hideo

1982-01-01

A system of combining a calciner for concentrated radioactive liquid waste and an incinerator for miscellaneous radioactive solid waste is being developed. Both the calciner and the incinerator are operated by fluidized bed method. The system features the following points: (1) Inflammable miscellaneous solids and concentrated liquid can be treated in combination to reduce the volume. (2) Used ion-exchange resin can be incinerated. (3) The system is applicable even if any final waste disposal method is adopted; calcinated and incinerated solids obtained as intermediate products are easy to handle and store. (4) The system is readily compatible with other waste treatment systems to form optimal total system. The following matters are described: the principle of fluidized-bed furnaces, the objects of treatment, system constitution, the features of the calciner and incinerator, and the current status of development. (J.P.N.)

Nuclear waste solidification

Science.gov (United States)

Bjorklund, William J.

1977-01-01

High level liquid waste solidification is achieved on a continuous basis by atomizing the liquid waste and introducing the atomized liquid waste into a reaction chamber including a fluidized, heated inert bed to effect calcination of the atomized waste and removal of the calcined waste by overflow removal and by attrition and elutriation from the reaction chamber, and feeding additional inert bed particles to the fluidized bed to maintain the inert bed composition.

Nuclear waste solidification

International Nuclear Information System (INIS)

Bjorklund, W.J.

1977-01-01

High level liquid waste solidification is achieved on a continuous basis by atomizing the liquid waste and introducing the atomized liquid waste into a reaction chamber including a fluidized, heated inert bed to effect calcination of the atomized waste and removal of the calcined waste by overflow removal and by attrition and elutriation from the reaction chamber, and feeding additional inert bed particles to the fluidized bed to maintain the inert bed composition

Physical and chemical characterization of synthetic calcined sludge

International Nuclear Information System (INIS)

Slates, R.V.; Mosley, W.C. Jr.; Tiffany, B.; Stone, J.A.

1982-03-01

Calcined synthetic sludge was chemically characterized in support of engineering studies to design a processing plant to solidify highly radioactive waste at the Savannah River Plant. An analytical technique is described which provides quantitative data by mass spectrometric analysis of gases evolved during thermogravimetric analysis without measurements of gas flow rates or mass spectrometer sensitivities. Scanning electron microprobe analysis, Mossbauer spectroscopy, and several other common analytical methods were also used. Calcined sludge consists primarily of amorphous particles of hydrous oxides with iron, manganese, nickel, and calcium distributed fairly uniformly throughout the powder. Iron, manganese, nickel, and calcium exist in forms that are highly insoluble in water, but aluminum, sulfate, nitrate, and sodium exhibit relative water solubilities that increase in the given order from 60% to 94%. Evolved gas analysis in a helium atmosphere showed that calcined sludge is completely dehydrated by heating to 400 0 C, carbon dioxide is evolved between 100 to 700 0 C with maximum evolution at 500 0 C, and oxygen is evolved between 400 and 1000 0 C. Evolved gas analyses are also reported for uncalcined sludge. A spinel-type oxide similar to NiFe 2 O 4 was detected by x-ray diffraction analysis at very low-level in calcined sludge

Technology for Treatment of Liquid Radioactive Waste Generated during Uranium and Plutonium Chemical and Metallurgical Manufacturing in FSUE PO Mayak - 13616

Energy Technology Data Exchange (ETDEWEB)

Adamovich, D. [SUE MosSIA Radon, 2/14 7th Rostovsky lane, Moscow, 119121 (Russian Federation); Batorshin, G.; Logunov, M.; Musalnikov, A. [FSUE ' PO Mayak' , 31 av. Lenin, Ozyorsk, Chelyabinsk region, 456780 (Russian Federation)

2013-07-01

Created technological scheme for treatment of liquid radioactive waste generated while uranium and plutonium chemical and metallurgical manufacturing consists of: - Liquid radioactive waste (LRW) purification from radionuclides and its transfer into category of manufacturing waste; - Concentration of suspensions containing alpha-nuclides and their further conversion to safe dry state (calcinate) and moving to long controlled storage. The following technologies are implemented in LRW treatment complex: - Settling and filtering technology for treatment of liquid intermediate-level waste (ILW) with volume about 1500m{sup 3}/year and alpha-activity from 10{sup 6} to 10{sup 8} Bq/dm{sup 3} - Membrane and sorption technology for processing of low-level waste (LLW) of radioactive drain waters with volume about 150 000 m{sup 3}/year and alpha-activity from 10{sup 3} to 10{sup 4} Bq/dm{sup 3}. Settling and filtering technology includes two stages of ILW immobilization accompanied with primary settling of radionuclides on transition metal hydroxides with the following flushing and drying of the pulp generated; secondary deep after settling of radionuclides on transition metal hydroxides with the following solid phase concentration by the method of tangential flow ultrafiltration. Besides, the installation capacity on permeate is not less than 3 m{sup 3}/h. Concentrates generated are sent to calcination on microwave drying (MW drying) unit. Membrane and sorption technology includes processing of averaged sewage flux by the method of tangential flow ultrafiltration with total capacity of installations on permeate not less than 18 m{sup 3}/h and sorption extraction of uranium from permeate on anionite. According to radionuclide contamination level purified solution refers to general industrial waste. Concentrates generated during suspension filtering are evaporated in rotary film evaporator (RFE) in order to remove excess water, thereafter they are dried on infrared heating

Comparative waste forms study

International Nuclear Information System (INIS)

Wald, J.W.; Lokken, R.O.; Shade, J.W.; Rusin, J.M.

1980-12-01

A number of alternative process and waste form options exist for the immobilization of nuclear wastes. Although data exists on the characterization of these alternative waste forms, a straightforward comparison of product properties is difficult, due to the lack of standardized testing procedures. The characterization study described in this report involved the application of the same volatility, mechanical strength and leach tests to ten alternative waste forms, to assess product durability. Bulk property, phase analysis and microstructural examination of the simulated products, whose waste loading varied from 5% to 100% was also conducted. The specific waste forms investigated were as follows: Cold Pressed and Sintered PW-9 Calcine; Hot Pressed PW-9 Calcine; Hot Isostatic Pressed PW-9 Calcine; Cold Pressed and Sintered SPC-5B Supercalcine; Hot Isostatic pressed SPC-5B Supercalcine; Sintered PW-9 and 50% Glass Frit; Glass 76-68; Celsian Glass Ceramic; Type II Portland Cement and 10% PW-9 Calcine; and Type II Portland Cement and 10% SPC-5B Supercalcine. Bulk property data were used to calculate and compare the relative quantities of waste form volume produced at a spent fuel processing rate of 5 metric ton uranium/day. This quantity ranged from 3173 L/day (5280 Kg/day) for 10% SPC-5B supercalcine in cement to 83 L/day (294 Kg/day) for 100% calcine. Mechanical strength, volatility, and leach resistance tests provide data related to waste form durability. Glass, glass-ceramic and supercalcine ranked high in waste form durability where as the 100% PW-9 calcine ranked low. All other materials ranked between these two groupings

Management of radioactive wastes from the nuclear fuel cycle

International Nuclear Information System (INIS)

1976-01-01

The increased emphasis in many countries on the development and utilization of nuclear power is leading to an expansion of all sectors of the nuclear fuel cycle, giving rise to important policy issues and radioactive-waste management requirements. Consequently, the IAEA and the Nuclear Energy Agency of OECD felt that it would be timely to review latest technology for the management of the radioactive wastes arising from nuclear fuel cycle facilities, to identify where important advances have been made, and to indicate those areas where further technological development is needed. Beginning in 1959, the IAEA, either by itself or jointly with OECD/NEA has held seven international symposia on the management of radioactive wastes. The last symposium, on the management of radioactive wastes from fuel reprocessing, was held jointly by the IAEA and OECD/NEA in Paris in November 1972. An objective of the 1976 symposium was to update the information presented at the previous symposia with the latest technological developments and thinking regarding the management and disposal of all categories of radioactive wastes. Consequently, although the scope of the symposium was rather broad, attention was focussed on operational experience and progress in unresolved areas of radioactive waste management. The programme dealt primarily with the solidification of liquid radioactive wastes and disposal of the products, especially the high-level fission products and actinide-containing waste from fuel reprocessing. Other topics covered policy and planning, treatment of hulls and solvent, management of plutonium-contaminated waste, and removal of gaseous radionuclides. The major topic of interest was the current state of the technology for the reduction and incorporation of the high-level radioactive liquid from fuel reprocessing into solid forms, such as calcines, glasses or ceramics, for safe interim storage and eventual disposal. The approaches to vitrification ranged from two stage

Design of a hot pilot plant facility for demonstration of the pot calcination process

Energy Technology Data Exchange (ETDEWEB)

Buckham, J A

1962-01-01

A facility was designed for demonstration of the pot calcination process with wastes from processing aluminum alloy fuels, Darex or electrolytic processing of stainless-steel fuels, and Purex processes. This facility will also permit determination of procedures required for economical production of low-porosity, relatively nonleachable materials by addition of suitable reagents to the wastes fed to the calciner. The process consists of concentration by evaporation and thermal decomposition in situ in pots which also serve as the final disposal containers. This unit permits determination of pot loading and density, leachability, melting point, volatile material content, heat release, and thermal conductivity of the calcine. Also to be determined are transient calcine temperature distributions, fission product behavior during calcination, deentrainment obtained in the various parts of the system, decontamination achieved on all liquid and gaseous effluent streams, need for venting of stored pots, optimum means of remotely sealing the pots, and methods required for production of a minimum volume of noncondensable off-gas. This facility will employ nominal full-scale pots 8 and 12 in. in diameter and 8 ft long. A unique evaporator design was evolved to permit operation either with close-coupled continuous feed preparation or with bath feed preparation. Provisions were made to circumvent possible explosions due to organic material in feed solutions and other suspected hazards.

Status, direction, and critical issues of waste treatment technology

International Nuclear Information System (INIS)

Knowles, D.E.; Bonner, W.F.

1983-01-01

Nuclear power production and related activities generate radioactive wastes that must be safely managed to protect workers and the general public. The liquid-fed ceramic melting (LFCM) vitrification process is the reference process for vitrifying high-level nuclear waste in the U.S. as well as in Japan and India. The French are currently using a rotary kiln calciner/metallic melter system at their reprocessing facility. Compaction or controlled-air incineration are the currently preferred options for low-level waste solids, followed by immobilization in an appropriate matrix. The Nuclear Waste Policy Act of 1982 is a significant step in proceeding with waste treatment and disposal. Programs can now build on past work to assure that public safety and regulations are met in a cost-effective manner

Perspective of metal encapsulation of waste

International Nuclear Information System (INIS)

Jardine, L.J.; Steindler, M.J.

1978-01-01

A conceptual flow sheet is presented for encapsulating solid, stabilized calcine (e.g., supercalcine) in a solid lead alloy, using existing or developing technologies. Unresolved and potential problem areas of the flow sheet are outlined and suggestions are made as how metal encapsulation might be applied to other solid wastes from the fuel cycle. It is concluded that metal encapsulation is a technique applicable to many forms of solid wastes and is likely to meet future waste isolation criteria and regulations

Nuclear fuel cycle waste recycling technology deverlopment - Radioactive metal waste recycling technology development

International Nuclear Information System (INIS)

Oh, Won Zin; Moon, Jei Kwon; Jung, Chong Hun; Park, Sang Yoon

1998-08-01

With relation to recycling of the radioactive metal wastes which are generated during operation and decommissioning of nuclear facilities, the following were described in this report. 1. Analysis of the state of the art on the radioactive metal waste recycling technologies. 2. Economical assessment on the radioactive metal waste recycling. 3. Process development for radioactive metal waste recycling, A. Decontamination technologies for radioactive metal waste recycling. B. Decontamination waste treatment technologies, C. Residual radioactivity evaluation technologies. (author). 238 refs., 60 tabs., 79 figs

Status, direction, and critical issues of waste treatment technology

International Nuclear Information System (INIS)

Knowlton, D.E.; Bonner, W.F.

1983-06-01

Nuclear power production and related activities generate radioactive wastes that must be safely managed to protect workers and the general public. The liquid-fed ceramic melting (LFCM) vitrification process is the reference process for vitrifying high-level nuclear waste in the US as well as in Japan and India. The French are currently using a rotary kiln calciner/metallic melter system at their reprocessing facility. Compaction or controlled-air incineration are the currently preferred options for low-level waste solids, followed by immobilization in an appropriate matrix. The Nuclear Waste Policy Act of 1982 is a significant step in proceeding with waste treatment and disposal. Programs can now build on past work to assure that public safety and regulations atre met in a cost-effective manner. 7 references, 2 figures, 3 tables

Volatilization and trapping of ruthenium during calcination of nitric acid solutions

International Nuclear Information System (INIS)

Klein, M.; Weyers, C.; Goossens, W.R.A.; Smet, M. de; Trine, J.

1983-01-01

Solid radioactive aerosols and semi-volatile fission products e.g. Ru, Cs, Sb are generated during high level liquid waste calcination and vitrification processes. The retention of Ruthenium was studied because of its strong tendency to form volatile compounds in oxidative media. Since RuO 4 was the suspected form for high temperature processes, the study was carried out on the behaviour of RuO 4 and its retention on adsorbants and catalysts for various gas compositions. The behaviour of volatilized Ru species obtained by calcination of nitrosyl Ru compounds was then compared with the RuO 4 case

Development and Testing of the Advanced CHP System Utilizing the Off-Gas from the Innovative Green Coke Calcining Process in Fluidized Bed

Energy Technology Data Exchange (ETDEWEB)

Chudnovsky, Yaroslav [Gas Technology Inst., Des Plaines, IL (United States); Kozlov, Aleksandr [Gas Technology Inst., Des Plaines, IL (United States)

2013-08-15

Green petroleum coke (GPC) is an oil refining byproduct that can be used directly as a solid fuel or as a feedstock for the production of calcined petroleum coke. GPC contains a high amount of volatiles and sulfur. During the calcination process, the GPC is heated to remove the volatiles and sulfur to produce purified calcined coke, which is used in the production of graphite, electrodes, metal carburizers, and other carbon products. Currently, more than 80% of calcined coke is produced in rotary kilns or rotary hearth furnaces. These technologies provide partial heat utilization of the calcined coke to increase efficiency of the calcination process, but they also share some operating disadvantages. However, coke calcination in an electrothermal fluidized bed (EFB) opens up a number of potential benefits for the production enhancement, while reducing the capital and operating costs. The increased usage of heavy crude oil in recent years has resulted in higher sulfur content in green coke produced by oil refinery process, which requires a significant increase in the calcinations temperature and in residence time. The calorific value of the process off-gas is quite substantial and can be effectively utilized as an “opportunity fuel” for combined heat and power (CHP) production to complement the energy demand. Heat recovered from the product cooling can also contribute to the overall economics of the calcination process. Preliminary estimates indicated the decrease in energy consumption by 35-50% as well as a proportional decrease in greenhouse gas emissions. As such, the efficiency improvement of the coke calcinations systems is attracting close attention of the researchers and engineers throughout the world. The developed technology is intended to accomplish the following objectives: - Reduce the energy and carbon intensity of the calcined coke production process. - Increase utilization of opportunity fuels such as industrial waste off-gas from the novel

Connecting section and associated systems concept for the spray calciner/in-can melter process

International Nuclear Information System (INIS)

Petkus, L.L.; Gorton, P.S.; Blair, H.T.

1981-06-01

For a number of years, researchers at the Pacific Northwest Laboratory have been developing processes and equipment for converting high-level liquid wastes to solid forms. One of these processes is the Spray Calciner/In-Can Melter system. To immobilize high-level liquid wastes, this system must be operated remotely, and the calcine must be reliably conveyed from the calciner to the melting furnace. A concept for such a remote conveyance system was developed at the Pacific Northwest Laboratory, and equipment was tested under full-scale, nonradioactive conditions. This concept and the design of demonstration equipment are described, and the results of equipment operation during experimental runs of 7 d are presented. The design includes a connecting section and its associated systems - a canister sypport and alignment concept and a weight-monitoring system for the melting furnace. Overall, the runs demonstrated that the concept design is an acceptable method of connecting the two pieces of process equipment together. Although the connecting section has not been optimized in all areas of concern, it provides a first-generation design of a production-oriented system

Experimental and modeling study of flash calcination of kaolinite rich clay particles in a gas suspension calciner

DEFF Research Database (Denmark)

Gebremariam, Abraham Teklay; Yin, Chungen; Rosendahl, Lasse

2015-01-01

gas suspension calciner, with the aim to derive useful guidelines on smart calcination for obtaining products of the best pozzolanic properties. Calcination tests are performed in the calciner under six different operation conditions. The raw feed and the calcined clay samples are all characterized...

Characterization of norm sources in petroleum coke calcining processes - 16314

International Nuclear Information System (INIS)

Hamilton, Ian S.; Halter, Donald A.; Fruchtnicht, Erich H.; Arno, Matthew G.; Haumann, Donald F

2009-01-01

Petroleum coke, or 'petcoke', is a waste by-product of the oil refining industry. The majority of petcoke consumption is in energy applications; catalyst coke is used as refinery fuel, anode coke for electricity conduction, and marketable coke for heating cement kilns. Roskill has predicted that long-term growth in petroleum coke production will be maintained, and may continue to increase slightly through 2012. Petcoke must first be calcined to drive off any undesirable petroleum by-products that would shorten the coke product life cycle. As an example, the calcining process can take place in large, rotary kilns heated to maximum temperatures as high as approximately 1400-1540 deg. C. The kilns and combustion/settling chambers, as well as some cooler units, are insulated with refractory bricks and other, interstitial materials, e.g., castable refractory materials, to improve the efficiency of the calcining process. The bricks are typically made of 70-85-percent bauxite, and are slowly worn away by the calcining process; bricks used to line the combustion chambers wear away, as well, but at a slower rate. It has been recognized that the refractory materials contain slight amounts of naturally occurring radioactive materials (NORM) from the uranium- and thorium-decay series. Similarly, low levels of NORM could be present in the petcoke feed stock given the nature of its origin. Neither the petcoke nor the refractory bricks represent appreciable sources of radiation or radioactive waste. However, some of the demolished bricks that have been removed from service because of the aforementioned wearing process have caused portal alarms to activate at municipal disposal facilities. This has lead to the current investigation into whether there is a NORM concentrating mechanism facilitated by the presence of the slightly radioactive feed stock in the presence of the slightly radioactive refractory materials, at calcining-zone temperatures. Research conducted to date has been

Arc plasma incineration of surrogate radioactive wastes

International Nuclear Information System (INIS)

Girold, C.; Cartier, R.; Taupiac, J.P.; Vandensteendam, C.; Baronnet, J.M.

1995-01-01

The aim of this presentation is to demonstrate the feasibility to substitute a single plasma reactor, where the arc is transferred on a melt glass bath, for several steps in an existing nuclear technological wastes incinerator. The incineration of wastes, the produced gas treatment and the vitrification of ashes issued from waste incineration are the three simultaneous functions of this new kind of reactor. The three steps of the work are described: first, post-combustion in an oxygen plasma of gases generated from the waste pyrolysis, then, vitrification of ashes from the calcination of wastes in the transferred plasma furnace and finally, incineration/vitrification of wastes in the same furnace

Corrosion monitoring of storage bins for radioactive calcines

International Nuclear Information System (INIS)

Hoffman, T.L.

1975-01-01

Highly radioactive liquid waste produced at the Idaho Chemical Processing Plant is calcined to a granular solid for long term storage in stainless steel bins. Corrosion evaluation of coupons withdrawn from these bins indicates excellent performance for the materials of construction of the bins. At exposure periods of up to six years the average penetration rates are 0.01 and 0.05 mils per year for Types 304 and 405 stainless steels, respectively. (auth)

Idaho Chemical Processing Plant spent fuel and waste management technology development program plan: 1994 Update

International Nuclear Information System (INIS)

1994-09-01

The Department of Energy has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage since 1951 and reprocessing since 1953. Until April 1992, the major activity of the ICPP was the reprocessing of SNF to recover fissile uranium and the management of the resulting high-level wastes (HLW). In 1992, DOE chose to discontinue reprocessing SNF for uranium recovery and shifted its focus toward the continued safe management and disposition of SNF and radioactive wastes accumulated through reprocessing activities. Currently, 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3,800 cubic meters of calcine waste, and 289 metric tons heavy metal of SNF are in inventory at the ICPP. Disposal of SNF and high-level waste (HLW) is planned for a repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP spent Fuel and Waste Management Technology Development Program (SF ampersand WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will be properly stored and prepared for final disposal in accordance with regulatory drivers. This Plan presents a brief summary of each of the major elements of the SF ampersand WMTDP; identifies key program assumptions and their bases; and outlines the key activities and decisions that must be completed to identify, develop, demonstrate, and implement a process(es) that will properly prepare the SNF and radioactive wastes stored at the ICPP for safe and efficient interim storage and final disposal

Separation of non-hazardous, non-radioactive components from ICPP calcine via chlorination

International Nuclear Information System (INIS)

Nelson, L.O.

1995-05-01

A pyrochemical treatment method for separating non-radioactive from radioactive components in solid granular waste accumulated at the Idaho Chemical Processing Plant was investigated. The goal of this study was to obtain kinetic and chemical separation data on the reaction products of the chlorination of the solid waste, known as calcine. Thermodynamic equilibrium calculations were completed to verify that a separation of radioactive and non-radioactive calcine components was possible. Bench-scale chlorination experiments were completed subsequently in a variety of reactor configurations including: a fixed-bed reactor (reactive gases flowed around and not through the particle bed), a packed/fluidized-bed reactor, and a packed-bed reactor (reactive gases flowed through the particle bed). Chemical analysis of the reaction products generated during the chlorination experiments verified the predictions made by the equilibrium calculations. An empirical first-order kinetic rate expression was developed for each of the reactor configurations. 20 refs., 16 figs., 21 tabs

High level waste facilities - Continuing operation or orderly shutdown

International Nuclear Information System (INIS)

Decker, L.A.

1998-04-01

Two options for Environmental Impact Statement No action alternatives describe operation of the radioactive liquid waste facilities at the Idaho Chemical Processing Plant at the Idaho National Engineering and Environmental Laboratory. The first alternative describes continued operation of all facilities as planned and budgeted through 2020. Institutional control for 100 years would follow shutdown of operational facilities. Alternatively, the facilities would be shut down in an orderly fashion without completing planned activities. The facilities and associated operations are described. Remaining sodium bearing liquid waste will be converted to solid calcine in the New Waste Calcining Facility (NWCF) or will be left in the waste tanks. The calcine solids will be stored in the existing Calcine Solids Storage Facilities (CSSF). Regulatory and cost impacts are discussed

Method and apparatus for evaporating radioactive liquid and calcinating the residue

International Nuclear Information System (INIS)

Sridhar, T.S.

1984-01-01

This invention provides an apparatus and a process for evaporating liquid wastes and calcining the residue. The liquid is sprayed against a hollow, rotating heated cylinder within a casing. The dried residue is scraped from the rotating cylinder and released through a valve at the bottom of the casing, while the effluent gas is filtered

Immobilization in ceramic waste forms of the residues from treatment of mixed wastes

International Nuclear Information System (INIS)

Oversby, V.M.; van Konynenburg, R.A.; Glassley, W.E.; Curtis, P.G.

1993-11-01

The Environmental Restoration and Waste Management Applied Technology Program at LLNL is developing a Mixed Waste Management Facility to demonstrate treatment technologies that provide an alternative to incineration. As part of that program, we are developing final waste forms using ceramic processing methods for the immobilization of the treatment process residues. The ceramic phase assemblages are based on using Synroc D as a starting point and varying the phase assemblage to accommodate the differences in chemistry between the treatment process residues and the defense waste for which Synroc D was developed. Two basic formulations are used, one for low ash residues resulting from treatment of organic materials contaminated with RCRA metals, and one for high ash residues generated from the treatment of plastics and paper products. Treatment process residues are mixed with ceramic precursor materials, dried, calcined, formed into pellets at room temperature, and sintered at 1150 to 1200 degrees C to produce the final waste form. This paper discusses the chemical composition of the waste streams and waste forms, the phase assemblages that serve as hosts for inorganic waste elements, and the changes in waste form characteristics as a function of variation in process parameters

Flash calcination of kaolinite rich clay and impact of process conditions on the quality of the calcines

DEFF Research Database (Denmark)

Gebremariam, Abraham Teklay; Yin, Chungen; Rosendahl, Lasse

2016-01-01

Use of properly calcined kaolinite rich clay (i.e., metakaolin) to offset part of CO2-intensive clinkers not only reduces CO2 footprint from cement industry but also improves the performance of concrete. However, calcination under inappropriately high temperatures or long retention times may...... suspension calciner. The model is validated by the experimental data (e.g., the degree of dehydroxylation and the density of the calcines). Based on the model, the impacts of process conditions and feed properties on the quality of the calcination products are thoroughly examined....

Wasting the Future: The Technological Sublime, Communications Technologies, and E-waste

Directory of Open Access Journals (Sweden)

Sebine Label

2012-08-01

Full Text Available Literally speaking, e-waste is the future of communications. E-waste is the fastest growing waste stream in the world, much of it communications technologies from cell phones to laptops, televisions to peripherals. As a result of policies of planned obsolescence working computers, cell phones, and tablets are routinely trashed. One of the most powerful and enduring discourses associated with emerging technologies is the technological sublime, in which technology is seen as intellectually, emotionally, or spiritually transcendent. It comprises a contradictory impulse that elevates technology with an almost religious fervor, while simultaneously overlooking some of the consequences of industrialism, as well as ignoring the necessity of social, economic, and governmental infrastructures necessary to the implementation and development of new technologies. The idea that a new technology will not pollute or harm the environment is a persistent, though often quickly passed over, theme in the technological sublime, echoed in discourses about emerging technologies such as the silicon chip, the internet, and other ICTs. In this paper, I make connections between the discourse of newness, the practice of planned obsolescence, and the mountains of trashed components and devices globally. Considering the global context demonstrates the realities of the penetration of ICTs and their enduring pollution and negative implications for the health of humans and nonhumans, including plants, animals, waterways, soil, air and so on. I use the discourse of the technological sublime to open up and consider the future of communications, to argue that this discourse not only stays with us but also contains within it two important and related components, the promise of ecological harmony and a future orientation. I argue that these lingering elements keep us from considering the real future of communications – e-waste – and that, as communications scholars, we must also

Vitrified waste option study report

International Nuclear Information System (INIS)

Lopez, D.A.; Kimmitt, R.R.

1998-02-01

A open-quotes Settlement Agreementclose quotes between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) will be treated so that it is ready to be moved out of Idaho for disposal by a target date of 2035. This report investigates vitrification treatment of all ICPP calcine, including the existing and future HLW calcine resulting from calcining liquid Sodium-Bearing Waste (SBW). Currently, the SBW is stored in the tank farm at the ICPP. Vitrification of these wastes is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the calcined waste and casting the vitrified mass into stainless steel canisters that will be ready to be moved out of the Idaho for disposal by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory (INEEL) until they are sent to a HLW national repository. The operating period for vitrification treatment will be from 2013 through 2032; all HLW will be treated and in storage by the end of 2032

Vitrified waste option study report

Energy Technology Data Exchange (ETDEWEB)

Lopez, D.A.; Kimmitt, R.R.

1998-02-01

A {open_quotes}Settlement Agreement{close_quotes} between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) will be treated so that it is ready to be moved out of Idaho for disposal by a target date of 2035. This report investigates vitrification treatment of all ICPP calcine, including the existing and future HLW calcine resulting from calcining liquid Sodium-Bearing Waste (SBW). Currently, the SBW is stored in the tank farm at the ICPP. Vitrification of these wastes is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the calcined waste and casting the vitrified mass into stainless steel canisters that will be ready to be moved out of the Idaho for disposal by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory (INEEL) until they are sent to a HLW national repository. The operating period for vitrification treatment will be from 2013 through 2032; all HLW will be treated and in storage by the end of 2032.

Uranium dioxide calcining apparatus

International Nuclear Information System (INIS)

Cole, E.A.; Peterson, R.S.

1978-01-01

This invention relates to an improved continuous calcining apparatus for consistently and controllably producing from calcinable reactive solid compounds of uranium, such as ammonium diuranate, uranium dioxide (UO 2 ) having an oxygen to uranium ratio of less than 2.2. The apparatus comprises means at the outlet end of a calciner kiln for receiving hot UO 2 , means for cooling the UO 2 to a temperature of below 100 deg C and conveying the cooled UO 2 to storage or to subsequent UO 2 processing apparatus where it finally comes into contact with air, the means for receiving cooling and conveying being sealed to the outlet end of the calciner and being maintained full of UO 2 and so operable as to exclude atmospheric oxygen from coming into contact with any UO 2 which is at elevated temperatures where it would readily oxidize, without the use of extra hydrogen gas in said means. (author)

Properties of radioactive calcine retrieved from the second calcined solids storage facility at ICPP

International Nuclear Information System (INIS)

Staples, B.A.; Pomiak, G.S.; Wade, E.L.

1979-03-01

The chemical and physical properties of radioactive alumina and zirconia calcine samples retrieved from the storage bins at ICPP were measured. Chemical properties measured include chemical composition, crystalline structure, and radiochemical composition. The physical properties measured and reported include density, size distribution, relative attrition, solubility in 8 M HNO 3 , thermal stability, and flow characteristics. The chemical and physical properties of the retrieved calcine after the 10 to 12 years of storage are very similar to freshly prepared simulated calcine

Regeneration of Waste Edible Oil by the Use of Virgin and Calcined Magnesium Hydroxide as Adsorbents.

Science.gov (United States)

Ogata, Fumihiko; Kawasaki, Naohito

2016-01-01

In this study, we prepared virgin (S, L) and calcined (S-380, S-1000, L-380, L-1000) magnesium hydroxide for regeneration of waste edible oil. Deterioration of soybean oil, rapeseed oil, and olive oil was achieved by heat and aeration treatment. The properties of the different adsorbents were investigated using specific surface area measurements, scanning electron microscopy, X-ray diffraction analysis, thermogravimetric-differential thermal analysis, and surface pH measurement. Moreover, the relationship between the changes in acid value (AV) and carbonyl value (CV) and the adsorbent properties were evaluated. The specific surface areas of S-380 and L-380 were greater than that of other adsorbents. In addition, the XRD results show that S-380 and L-380 contain both magnesium hydroxide and magnesium oxide structures. The decreases in AV and CV using S-380 and L-380 were greater than achieved using other adsorbents. The correlation coefficients between the decrease in AV and CV and specific surface area were 0.947 for soybean oil, 0.649 for rapeseed oil, and 0.773 for olive oil, respectively. The results obtained in this study suggest that a physical property of the adsorbent, namely specific surface area, was primarily responsible for the observed decreases in AV and CV. Overall, the results suggest that S-380 and L-380 are useful for the regeneration of waste edible oil.

Direct cementitious waste option study report

International Nuclear Information System (INIS)

Dafoe, R.E.; Losinski, S.J.

1998-02-01

A settlement agreement between the Department of Energy (DOE) and the State of Idaho mandates that all high-level radioactive waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) will be treated so that it is ready to be moved out of Idaho for disposal by a target data of 2035. This study investigates the direct grouting of all ICPP calcine (including the HLW dry calcine and those resulting from calcining sodium-bearing liquid waste currently residing in the ICPP storage tanks) as the treatment method to comply with the settlement agreement. This method involves grouting the calcined waste and casting the resulting hydroceramic grout into stainless steel canisters. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory (INEEL) until they are sent to a national geologic repository. The operating period for grouting treatment will be from 2013 through 2032, and all the HLW will be treated and in interim storage by the end of 2032

Direct cementitious waste option study report

Energy Technology Data Exchange (ETDEWEB)

Dafoe, R.E.; Losinski, S.J.

1998-02-01

A settlement agreement between the Department of Energy (DOE) and the State of Idaho mandates that all high-level radioactive waste (HLW) now stored at the Idaho Chemical Processing Plant (ICPP) will be treated so that it is ready to be moved out of Idaho for disposal by a target data of 2035. This study investigates the direct grouting of all ICPP calcine (including the HLW dry calcine and those resulting from calcining sodium-bearing liquid waste currently residing in the ICPP storage tanks) as the treatment method to comply with the settlement agreement. This method involves grouting the calcined waste and casting the resulting hydroceramic grout into stainless steel canisters. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory (INEEL) until they are sent to a national geologic repository. The operating period for grouting treatment will be from 2013 through 2032, and all the HLW will be treated and in interim storage by the end of 2032.

Comparative evaluation of glasses reprocessing and reversible conditioning of calcinates

International Nuclear Information System (INIS)

Boen, R.

2000-01-01

Fission products and minor actinides separated during the spent fuel reprocessing treatment are industrially vitrified on-line and thus confined inside a glass matrix with admittedly durability properties. In the framework of the feasibility of a reversible conditioning, this document examines first the possible alternative ways of conditioning and storage of calcinates before vitrification, which may simplify the reversibility aspect. Such a conditioning must be compatible with the storage process, with a possible extraction of actinides and long-lived fission products, and with the vitrification process if no extraction is performed. Calcinates are pulverulent and comprise an important soluble fraction, a proportion of nitrates of about 30%, and release a high thermal power (17 kW/m 3 ) combined to a low thermal conductivity (0.1 to 0.15 W.m -1 k -1 ). Among the different foreseeable solutions (denitration, mixing with another material, with or without compacting, dissolution inside another material..), the dissolution inside a borate seems to be the most acceptable with respect to the safety, feasibility and vitrification aspects. The thermal aspect of the storage remains complex as a specific container is necessary. In a second part, this report analyzes the possibility to re-extract back the long-lived radionuclides from vitrified wastes. The different possible ways to destroy the glass structure and to transfer the fission products and minor actinides in an aqueous solution compatible with an hydrometallurgical separation process are explored. Two processes are foreseeable: a low temperature dissolution process which requires a preliminary crushing and the handling of huge amounts of acids, and a both high and low temperature process which comprises the following steps: melting, fractionation by water tempering, addition of Na 2 O or sodium tetraborate to make it sensible to hot leaching, separation of fission products and minor actinides, recycling of

Characterisation of sugar cane straw waste as pozzolanic material for construction: Calcining temperature and kinetic parameters

International Nuclear Information System (INIS)

Frias, Moises; Villar-Cocina, E.; Valencia-Morales, E.

2007-01-01

This paper reports on the influence of calcining temperature (800 and 1000 deg. C) on the pozzolanic activation of sugar cane straw (SCS). The reaction kinetics of SCS ash-lime mixtures were inferred from physicochemical characteristics (X-ray diffraction patterns and thermogravimetry analysis. The fitting of a kinetic-diffusive model to the experimental data (fixed lime versus time) allowed the computing of the kinetic parameters (reaction rate constant) of the pozzolanic reaction. Results obtained confirm that the sugar cane straw ash (SCSA) calcined at 800 and 1000 deg. C have properties indicative of very high pozzolanic activity. No influence of calcining temperature on the pozzolanic activity was observed. Also, no crystalline compounds during the pozzolanic reaction were identified up to 90 days of reaction. Environmental durability and strength of the consequential mortars remain to be assessed

Physical and chemical characteristics of fluorinel/sodium calcine generated during 30 cm Pilot-Plant Run 17

International Nuclear Information System (INIS)

Brewer, K.N.; Kessinger, G.F.; Littleton, L.L.; Olson, A.L.

1993-07-01

The 30 centimeter (cm) pilot plant calciner Run 17, of March 9, 1987, was performed to study the calcination of fluroinel-sodium blended waste blended at the ratio 3.5:1 fluorinel to sodium, respectively. The product of the run was analyzed by a variety of analytical techniques that included X-ray powder diffraction (XRD), inductively coupled plasma spectroscopy (ICP), scanning electron microscopy (SEM), and X-ray photoelectron spectroscopy (XPS) to deduce physical and chemical characteristics. The analytical data, as well as data analyses and conclusions drawn from the data, are presented

Selection of a glass-ceramic formulation to immobilize fluorinel- sodium calcine

International Nuclear Information System (INIS)

Staples, B.A.; Wood, H.C.

1994-12-01

One option for immobilizing calcined high level wastes produced by nuclear fuel reprocessing activities at the Idaho Chemical Processing Plant (ICPP) is conversion to a glass-ceramic form through hot isostatic pressing. Calcines exist in several different chemical compositions, and thus candidate formulations have been developed for converting each to glass-ceramic forms which are potentially resistant to aqueous corrosion and stable enough to qualify for repository storage. Fluorinel/Na, a chemically complex calcine type, is one of the types being stored at ICPP, and development efforts have identified three formulations with potential for immobilizing it. These are a glass forming additive that uses aluminum metal to enhance reactivity, a second glass forming additive that uses titanium metal to enhance reactivity and a third that uses not only a combination of silicon and titanium metals but enough phosphorous pentoxide to form a calcium phosphate host phase in the glass-ceramic product. Glass-ceramics of each formulation performed well in restricted characterization tests. However, none of the three was subjected to rigorous testing that would provide information on whether each was processable, that is able to retain favorable characteristics over a practical range of processing conditions

Volatility of ruthenium-106, technetium-99, and iodine-129, and the evolution of nitrogen oxide compounds during the calcination of high-level, radioactive nitric acid waste

International Nuclear Information System (INIS)

Rimshaw, S.J.; Case, F.N.; Tompkins, J.A.

1980-02-01

The nitrate anion is the predominant constituent in all high-level nuclear wastes. Formic acid reacts with the nitrate anion to yield noncondensable, inert gases (N 2 or N 2 O), which can be scrubbed free of 106 Ru, 129 I, and 99 Tc radioactivities prior to elimination from the plant by passing through HEPA filters. Treatment of a high-level authentic radioactive waste with two moles of formic acid per mole of nitrate anion leads to a low RuO 4 volatility of about 0.1%, which can be reduced to an even lower level of 0.007% on adding a 15% excess of formic acid. Without pretreatment of the nitrate waste with formic acid, a high RuO 4 volatility of approx. 35% is observed on calcining a 4.0 N HNO 3 solution in quartz equipment at 350 0 C. The RuO 4 volatility falls to approx. 1.0% on decreasing the initial HNO 3 concentration to 1.0 N or lower. It is postulated that thermal denitration of a highly nitrated ruthenium complex leads to the formation of volatile RuO 4 , while decarboxylation of a ruthenium-formate complex leads to the formation of nonvolatile RuO 2 . Wet scrubbing with water is used to remove RuO 4 from the off-gas stream. In all glass equipment, small amounts of particulate RuO 2 are formed in the gas phase by decomposition of RuO 4 . The 99 Tc volatility was found to vary from 0.2 to 1.4% on calcining HNO 3 and HCOOH (formic acid) solutions over the temperature range of 250 to 600 0 C. These unexpectedly low volatilities of 99 Tc are correlated to the high thermal stability limits of various metal pertechnetates and technetates. Iodine volatilities were high, varying from a low of 30% at 350 0 C to a high of 97% at 650 0 C. It is concluded that with a proper selection of pretreatment and operating conditions the 106 Ru and 99 Tc activities can be retained in the calcined solid with recycle of the wet scrubbing solution

Hanford Waste Vitrification Plant Technology Plan

International Nuclear Information System (INIS)

Sexton, R.A.

1988-06-01

The reference Hanford plan for disposal of defense high-level waste is based on waste immobilization in glass by the vitrification process and temporary vitrified waste storage at the Hanford Site until final disposal in a geologic repository. A companion document to the Hanford Waste Management Plan (HWMP) is the Draft, Interim Hanford Waste Management Technology Plan (HWMTP), which provides a description of the technology that must be developed to meet the reference waste management plan. One of the issues in the HWMTP is DST-6, Immobilization (Glass). The HWMTP includes all expense funding needed to complete the Hanford Waste Vitrification Plant (HWVP) project. A preliminary HWVP Technology Plan was prepared in 1985 as a supporting document to the HWMTP to provide a more detailed description of the technology needed to construct and operate a vitrification facility. The plan was updated and issued in 1986, and revised in 1987. This document is an annual update of the plan. The HWVP Technology Plan is limited in scope to technology that requires development or confirmation testing. Other expense-funded activities are not included. The relationship between the HWVP Technology Plan and other waste management issues addressed in the HWMTP is described in section 1.6 of this plan. 6 refs., 4 figs., 34 tabs

Phosphorous removal from aqueous solution can be enhanced through the calcination of lime sludge.

Science.gov (United States)

Bal Krishna, K C; Niaz, Mohamed R; Sarker, Dipok C; Jansen, Troy

2017-09-15

Water treatment plants generate an enormous amount of the sludge which is normally treated as waste. In the recent past, many investigations have been focused on developing an economical adsorbent using water treatment sludge to remove phosphorous (P) from aqueous solutions. However, the great extents of the studies have been limited in the use of alum- and iron-based sludges. This study, therefore, investigated the P removal performance of the calcined lime sludge. Calcined lime sludge at 700 °C significantly enhanced the P removal efficiency whereas marginal improvement was noted when the sludge calcined at 400 °C was tested. With increase P removal efficiency, final pH values of the solution also significantly increased. P removal efficiency of the calcined sludge decreased with increasing the initial P concentrations. However, the removal efficiency could be improved by increasing the weight of the sludge. Further analysis demonstrated that P removal trend followed both pseudo-second order and diffusion-chemisorption kinetics signifying the P removal is potentially due to a multi-mechanistic reaction in which, the process is controlled by intra-particle diffusion followed by chemisorptions. Copyright © 2017 Elsevier Ltd. All rights reserved.

Design and performance of a 100-kg/h, direct calcine-fed electric-melter system for nuclear-waste vitrification

International Nuclear Information System (INIS)

Dierks, R.D.

1980-11-01

This report describes the physical characteristics of a ceramic-lined, joule-heated glass melter that is directly connected to the discharge of a spray calciner and is currently being used to study the vitrification of simulated nuclear-waste slurries. Melter performance characteristics and subsequent design improvements are described. The melter contains 0.24 m 3 of glass with a glass surface area of 0.76 m 2 , and is heated by the flow of an alternating current (ranging from 600 to 1200 amps) between two Inconel-690 slab-type electrodes immersed in the glass at either end of the melter tank. The melter was maintained at operating temperature (900 to 1260 0 C) for 15 months, and produced 62,000 kg of glass. The maximum sustained operating period was 122 h, during which glass was produced at the rate of 70 kg/h

[Infrared spectral analysis for calcined borax].

Science.gov (United States)

Zhao, Cui; Ren, Li-Li; Wang, Dong; Zhou, Ping; Zhang, Qian; Wang, Bo-Tao

2011-08-01

To valuate the quality of calcined borax which is sold in the market, 18 samples of calcined borax were studied using the Fourier transform infrared, and samples with different water content were selected and analyzed. Then, the results of analysis were used to evaluate the quality of calcined borax. Results show that the infrared spectra of calcined borax include OH vibration, BO3(-3) vibration and BO4(5-) vibration absorption bands. The position and width of OH vibration absorption band depend on the level of water content, and the more the water content, the wider the absorption band. The number of BO3(3-) vibration and BO4(5-) vibration bands also depend on the level of water content, and the more the water content, and the stronger the hydrogen bond and the lower the symmetry of B atoms, the more the number of infrared absorption peaks. It was concluded that because the quality of calcined borax has direct correlation with water content, the infrared spectroscopy is an express and objective approach to quality analysis and evaluation of calcined borax.

Process description and plant design for preparing ceramic high-level waste forms

International Nuclear Information System (INIS)

Grantham, L.F.; McKisson, R.L.; Guon, J.; Flintoff, J.F.; McKenzie, D.E.

1983-01-01

The ceramics process flow diagram has been simplified and upgraded to utilize only two major processing steps - fluid-bed calcination and hot isostatic press consolidating. Full-scale fluid-bed calcination has been used at INEL to calcine high-level waste for 18 y; and a second-generation calciner, a fully remotely operated and maintained calciner that meets ALARA guidelines, started calcining high-level waste in 1982. Full-scale hot isostatic consolidation has been used by DOE and commercial enterprises to consolidate radioactive components and to encapsulate spent fuel elements for several years. With further development aimed at process integration and parametric optimization, the operating knowledge of full-scale demonstration of the key process steps should be rapidly adaptable to scale-up of the ceramic process to full plant size. Process flowsheets used to prepare ceramic and glass waste forms from defense and commercial high-level liquid waste are described. Preliminary layouts of process flow diagrams in a high-level processing canyon were prepared and used to estimate the preliminary cost of the plant to fabricate both waste forms. The estimated costs for using both options were compared for total waste management costs of SRP high-level liquid waste. Using our design, for both the ceramic and glass plant, capital and operating costs are essentially the same for both defense and commercial wastes, but total waste management costs are calculated to be significantly less for defense wastes using the ceramic option. It is concluded from this and other studies that the ceramic form may offer important advantages over glass in leach resistance, waste loading, density, and process flexibility. Preliminary economic calculations indicate that ceramics must be considered a leading candidate for the form to immobilize high-level wastes

SYNROC production using a fluid bed calciner

International Nuclear Information System (INIS)

Ackerman, F.J.; Grens, J.Z.; Ryerson, F.J.; Hoenig, C.L.; Bazan, F.; Campbell, J.H.

1982-01-01

SYNROC is a titanate-based ceramic developed for immobilization of high-level nuclear reactor wastes in solid form. Fluid-bed SYNROC production permits slurry drying, calcining and redox to be carried out in a single unit. We present results of studies from two fluid beds; the Idaho Exxon internally-heated unit and the externally-heated unit constructed at Lawrence Livermore National laboratory. Bed operation over a range of temperature, feed rate, fluidizing rate and redox conditions indicate that high density, uniform particle-size SYNROC powders are produced which facilitate the densification step and give HUP parts with dense, well-developed phases and good leaching characteristics. 3 figures, 3 tables

Engineering-scale vitrification of commercial high-level waste

International Nuclear Information System (INIS)

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

1980-04-01

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

Interim Hanford Waste Management Technology Plan

International Nuclear Information System (INIS)

1985-09-01

The Interim Hanford Waste Management Technology Plan (HWMTP) is a companion document to the Interim Hanford Waste Management Plan (HWMP). A reference plan for management and disposal of all existing and certain projected future radioactive Hanford Site Defense Wastes (HSDW) is described and discussed in the HWMP. Implementation of the reference plan requires that various open technical issues be satisfactorily resolved. The principal purpose of the HWMTP is to present detailed descriptions of the technology which must be developed to close each of the technical issues associated with the reference plan identified in the HWMP. If alternative plans are followed, however, technology development efforts including costs and schedules must be changed accordingly. Technical issues addressed in the HWMTP and HWMP are those which relate to disposal of single-shell tank wastes, contaminated soil sites, solid waste burial sites, double-shell tank wastes, encapsulated 137 CsCl and 90 SrF 2 , stored and new solid transuranic (TRU) wastes, and miscellaneous wastes such as contaminated sodium metal. Among the high priority issues to be resolved are characterization of various wastes including early determination of the TRU content of future cladding removal wastes; completion of development of vitrification (Hanford Waste Vitrification Plant) and grout technology; control of subsidence in buried waste sites; and development of criteria and standards including performance assessments of systems proposed for disposal of HSDW. Estimates of the technology costs shown in this report are made on the basis that all identified tasks for all issues associated with the reference disposal plan must be performed. Elimination of, consolidation of, or reduction in the scope of individual tasks will, of course, be reflected in corresponding reduction of overall technology costs

Alternative processes for managing existing commercial high-level radioactive wastes

International Nuclear Information System (INIS)

1976-04-01

A number of alternatives are discussed for managing high-level radioactive waste presently stored at the West Valley, New York, plant owned by Nuclear Fuel Services, Inc. These alternatives (liquid storage, conversion to cement, shale fracturing, shale cement, calcination, aqueous silicate, conversion to glass, and salt cake) are limited to concepts presently under active investigation by ERDA. Each waste management option is described and examined regarding the status of the technology; its applications to managing NFS waste; its advantages and disadvantages; the research and development needed to implement the option; safety considerations; and estimated costs and time to implement the process

Plasma technology for treatment of waste

Energy Technology Data Exchange (ETDEWEB)

Cohn, D [Massachusetts Inst. of Technology, Cambridge, MA (United States). Plasma Fusion Center

1997-12-31

Meeting goals for waste cleanup will require new technology with improved environmental attractiveness and reduced cost. Plasma technology appears promising because of the high degree of controllability; capability to process waste without the adverse effects of combustion; and a very wide temperature range of operation. At the Plasma Fusion Center at the Massachusetts Institute of Technology, a range of plasma technologies are being investigated. `Hot` thermal plasmas produced by DC arc technology are being examined for treatment of solid waste. In conjunction with this activity, new diagnostics are being developed for monitoring arc furnace operating parameters and gaseous emissions. Electron-beam generated plasma technology is being investigated as a means of producing non-thermal `cold` plasmas for selective processing of dilute concentrations of gaseous waste. (author). 4 figs., 5 refs.

Mixed Waste Integrated Program: A technology assessment for mercury-containing mixed wastes

International Nuclear Information System (INIS)

Perona, J.J.; Brown, C.H.

1993-03-01

The treatment of mixed wastes must meet US Environmental Protection Agency (EPA) standards for chemically hazardous species and also must provide adequate control of the radioactive species. The US Department of Energy (DOE) Office of Technology Development established the Mixed Waste Integrated Program (MWIP) to develop mixed-waste treatment technology in support of the Mixed Low-Level Waste Program. Many DOE mixed-waste streams contain mercury. This report is an assessment of current state-of-the-art technologies for mercury separations from solids, liquids, and gases. A total of 19 technologies were assessed. This project is funded through the Chemical-Physical Technology Support Group of the MWIP

Experimental results: Pilot plant calcine dissolution and liquid feed stability

International Nuclear Information System (INIS)

Herbst, R.S.; Fryer, D.S.; Brewer, K.N.; Johnson, C.K.; Todd, T.A.

1995-02-01

The dissolution of simulated Idaho Chemical Processing Plant pilot plant calcines, containing none of the radioactive actinides, lanthanides or fission products, was examined to evaluate the solubility of calcine matrix materials in acidic media. This study was a necessary precursor to dissolution and optimization experiments with actual radionuclide-containing calcines. The importance of temperature, nitric acid concentration, ratio of acid volume to calcine mass, and time on the amount, as a weight percentage of calcine dissolved, was evaluated. These parameters were studied for several representative pilot plant calcine types: (1) Run No. 74 Zirconia calcine; (2) Run No. 17 Zirconia/Sodium calcine; (3) Run No. 64 Zirconia/Sodium calcine; (3) Run No. 1027 Alumina calcine; and (4) Run No. 20 Alumina/Zirconia/Sodium calcine. Statistically designed experiments with the different pilot plant calcines indicated the effect of the studied process variables on the amount of calcine dissolved decreases in the order: Acid/Calcine Ratio > Temperature > HNO 3 Concentration > Dissolution Time. The following conditions are suitable to achieve greater than 90 wt. % dissolution of most Zr, Al, or Na blend calcines: (1) Maximum nitric acid concentration of 5M; (2) Minimum acid/calcine ratio of 10 mL acid/1 gram calcine; (3) Minimum dissolution temperature of 90 degrees C; and (4) Minimum dissolution time of 30 minutes. The formation of calcium sulphate (CaSO 4 ) precipitates was observed in certain dissolved calcine solutions during the dissolution experiments. Consequently, a study was initiated to evaluate if and under what conditions the resulting dissolved calcine solutions would be unstable with regards to precipitate formation. The results indicate that precipitate formation in the calcine solutions prepared under the above proposed dissolution conditions are not anticipated

Technologies to remediate hazardous waste sites

International Nuclear Information System (INIS)

Falco, J.W.

1990-03-01

Technologies to remediate hazardous wastes must be matched with the properties of the hazardous materials to be treated, the environment in which the wastes are imbedded, and the desired extent of remediation. Many promising technologies are being developed, including biological treatment, immobilization techniques, and in situ methods. Many of these new technologies are being applied to remediate sites. The management and disposal of hazardous wastes is changing because of federal and state legislation as well as public concern. Future waste management systems will emphasize the substitution of alternatives for the use of hazardous materials and process waste recycling. Onsite treatment will also become more frequently adopted. 5 refs., 7 figs

FY-95 technology catalog. Technology development for buried waste remediation

International Nuclear Information System (INIS)

1995-01-01

The US Department of Energy's (DOE) Buried Waste Integrated Demonstration (BWID) program, which is now part of the Landfill Stabilization Focus Area (LSFA), supports applied research, development, demonstration, and evaluation of a multitude of advanced technologies dealing with underground radioactive and hazardous waste remediation. These innovative technologies are being developed as part of integrated comprehensive remediation systems for the effective and efficient remediation of buried waste sites throughout the DOE complex. These efforts are identified and coordinated in support of Environmental Restoration (EM-40) and Waste Management (EM-30) needs and objectives. Sponsored by the DOE Office of Technology Development (EM-50), BWID and LSFA work with universities and private industry to develop technologies that are being transferred to the private sector for use nationally and internationally. This report contains the details of the purpose, logic, and methodology used to develop and demonstrate DOE buried waste remediation technologies. It also provides a catalog of technologies and capabilities with development status for potential users. Past FY-92 through FY-94 technology testing, field trials, and demonstrations are summarized. Continuing and new FY-95 technology demonstrations also are described

FY-95 technology catalog. Technology development for buried waste remediation

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-10-01

The US Department of Energy`s (DOE) Buried Waste Integrated Demonstration (BWID) program, which is now part of the Landfill Stabilization Focus Area (LSFA), supports applied research, development, demonstration, and evaluation of a multitude of advanced technologies dealing with underground radioactive and hazardous waste remediation. These innovative technologies are being developed as part of integrated comprehensive remediation systems for the effective and efficient remediation of buried waste sites throughout the DOE complex. These efforts are identified and coordinated in support of Environmental Restoration (EM-40) and Waste Management (EM-30) needs and objectives. Sponsored by the DOE Office of Technology Development (EM-50), BWID and LSFA work with universities and private industry to develop technologies that are being transferred to the private sector for use nationally and internationally. This report contains the details of the purpose, logic, and methodology used to develop and demonstrate DOE buried waste remediation technologies. It also provides a catalog of technologies and capabilities with development status for potential users. Past FY-92 through FY-94 technology testing, field trials, and demonstrations are summarized. Continuing and new FY-95 technology demonstrations also are described.

Toward integrated design of waste management technologies

International Nuclear Information System (INIS)

Carnes, S.A.; Wolfe, A.K.

1994-01-01

Implementation of waste management technologies has been hindered by the intervention of diverse interests. Relying on a perceived history of inadequate and improper management, operations, and technological design, critics have stymied the implementation of scientifically and governmentally approved technologies and facilities, leading to a critical shortage of hazardous, mixed, and radioactive waste management capacity. The research and development (R ampersand D) required to identify technologies that are simultaneously (1) scientifically valid, (2) economically sound, and (3) publicly acceptable must necessarily address, in an integrated and interdisciplinary manner, these three criteria and how best to achieve the integration of stakeholders early in the technology implementation process (i.e., R ampersand D, demonstration, and commercialization). The goal of this paper is to initiate an identification of factors likely to render radioactive and hazardous waste management technologies publicly acceptable and to provide guidance on how technological R ampersand D might be revised to enhance the acceptability of alternative waste management technologies. Principal among these factors are the equitable distribution of costs, risks, and benefits of waste management policies and technologies, the equitable distribution of authority for making waste management policy and selecting technologies for implementation, and the equitable distribution of responsibility for resolving waste management problems. Stakeholder participation in assessing the likely distribution of these factors and mitigative mechanisms to enhance their equitable distribution, together with stakeholder participation in policy and technology R ampersand D, as informed by stakeholder assessments, should enhance the identification of acceptable policies and technologies

Alternatives for conversion to solid interim waste forms of the radioactive liquid high-level wastes stored at the Western New York Nuclear Service Center

International Nuclear Information System (INIS)

Vogler, S.; Trevorrow, L.E.; Ziegler, A.A.; Steindler, M.J.

1981-08-01

Techniques for isolating and solidifying the nuclear wastes in the storage tanks at the Western New York Nuclear Service Center plant have been examined. One technique involves evaporating the water and forming a molten salt containing the precipitated sludge. The salt is allowed to solidify and is stored in canisters until processing into a final waste form is to be done. Other techniques involve calcining the waste material, then agglomerating the calcine with sodium silicate to reduce its dispersibility. This option can also involve a prior separation and decontamination of the supernatant salt. The sludge and all resins containing fission-product activity are then calcined together. The technique of removing the water and solidifying the salt may be the simplest method for removing the waste from the West Valley Plant

The KS-KT-100 plant for two-stage vitrification of radioactive waste: results of tests with simulators

International Nuclear Information System (INIS)

Davydov, V.I.; Dobrygin, P.G.; Dolgov, V.V.; Sergeev, G.A.

1976-01-01

The Soviet Union has developed a two-stage process for phosphate vitrification of liquid radioactive waste involving the use, at the initial stage, of calcination in the pseudo-liquefied layer, followed by melting of the calcinate in a ceramic crucible (second stage). On the basis of the laboratory studies and bench tests using experimental equipment, the authors have developed and tried out an enlarged plant - the KS-KT-100. The plant includes units for preparing the solution, evaporation, calcination, melting and gas purification. The initial solution containing 240 g/litre of aluminium nitrate, 125 g/litre of sodium nitrate, 120 to 130 g/litre of orthophosphoric acid, and 90 to 150 g/litre of industrial molasses simulated fluxed nitrate waste. The tests have shown that the various units operate satisfactorily. The authors have determined the technological parameters for evaporation, calcination of the solution and melting of the calcinate. The presence of molasses in the solution (150 g/litre) makes it possible to decompose and distil 40% of the nitrate ion during evaporation. The calcination temperature is 350 to 400 0 C, and the fluidization rate 1.5 m/s. The capacity of the plant for the initial solution is 100 litres/h, for the evaporated solution 65 litres/h, and for the glass 20 kg/h. The efficiency of the gas purification system ranges between 10 7 and 10 9 . The test results show the feasibility of the two-stage method of vitrification in actual practice. (author)

Low-risk alternative waste forms for problematic high-level and long-lived nuclear wastes

International Nuclear Information System (INIS)

Stewart, M.W.A.; Begg, B.D.; Moricca, S.; Day, R.A.

2006-01-01

Full text: The highest cost component the nuclear waste clean up challenge centres on high-level waste (HLW) and consequently the greatest opportunity for cost and schedule savings lies with optimising the approach to HLW cleanup. The waste form is the key component of the immobilisation process. To achieve maximum cost savings and optimum performance the selection of the waste form should be driven by the characteristics of the specific nuclear waste to be immobilised, rather than adopting a single baseline approach. This is particularly true for problematic nuclear wastes that are often not amenable to a single baseline approach. The use of tailored, high-performance, alternative waste forms that include ceramics and glass-ceramics, coupled with mature process technologies offer significant performance improvements and efficiency savings for a nuclear waste cleanup program. It is the waste form that determines how well the waste is locked up (chemical durability), and the number of repository disposal canisters required (waste loading efficiency). The use of alternative waste forms for problematic wastes also lowers the overall risk by providing high performance HLW treatment alternatives. The benefits tailored alternative waste forms bring to the HLW cleanup program will be briefly reviewed with reference to work carried out on the following: The HLW calcines at the Idaho National Laboratory; SYNROC ANSTO has developed a process utilising a glass-ceramic combined with mature hot-isostatic pressing (HIP) technology and has demonstrated this at a waste loading of 80 % and at a 30 kg HIP scale. The use of this technology has recently been estimated to result in a 70 % reduction in waste canisters, compared to the baseline borosilicate glass technology; Actinide-rich waste streams, particularly the work being done by SYNROC ANSTO with Nexia Solutions on the Plutonium-residues wastes at Sellafield in the UK, which if implemented is forecast to result in substantial

XANES analysis of dried and calcined bones

International Nuclear Information System (INIS)

Rajendran, Jayapradhi; Gialanella, Stefano; Aswath, Pranesh B.

2013-01-01

The structure of dried and calcined bones from chicken, bovine, deer, pig, sheep and chamois was examined using X-ray Absorption Near Edge Structure (XANES) spectroscopy. The oxygen K-edge absorption edge indicates that the surface of dried bone has a larger proportion of carbonate than the interior that is made up of phosphates. The phosphorus L and K edge clearly indicate that pyrophosphates, α-tricalcium phosphate (α-TCP) and hydrogen phosphates of Ca do not exist in either the dried bone or calcined bone and phosphorus exists as either β-tricalcium phosphate (β-TCP) or hydroxyapatite, both in the dried and calcined conditions. The Ca K-edge analysis indicates that β-TCP is the likely form of phosphate in both the dried and calcined conditions. - Highlights: • For the first time bones of five different species of vertebrates have been compared in both the dried and calcined states. • O, P and Ca edges detail the local coordination of these atoms in dried and calcined bone. • O K-edge shows that the surface of bone has more CO 3 while the interior has more PO 4 . • P and Ca edges eliminate the presence of pyrophosphates and confirmed the presence of HA and β-TCP. • The stability of these phosphates on calcination has been examined using XANES

XANES analysis of dried and calcined bones

Energy Technology Data Exchange (ETDEWEB)

Rajendran, Jayapradhi [Materials Science and Engineering Department, University of Texas at Arlington (United States); Gialanella, Stefano [Materials Science and Industrial Technology Department, University of Trento (Italy); Aswath, Pranesh B., E-mail: aswath@uta.edu [Materials Science and Engineering Department, University of Texas at Arlington (United States)

2013-10-15

The structure of dried and calcined bones from chicken, bovine, deer, pig, sheep and chamois was examined using X-ray Absorption Near Edge Structure (XANES) spectroscopy. The oxygen K-edge absorption edge indicates that the surface of dried bone has a larger proportion of carbonate than the interior that is made up of phosphates. The phosphorus L and K edge clearly indicate that pyrophosphates, α-tricalcium phosphate (α-TCP) and hydrogen phosphates of Ca do not exist in either the dried bone or calcined bone and phosphorus exists as either β-tricalcium phosphate (β-TCP) or hydroxyapatite, both in the dried and calcined conditions. The Ca K-edge analysis indicates that β-TCP is the likely form of phosphate in both the dried and calcined conditions. - Highlights: • For the first time bones of five different species of vertebrates have been compared in both the dried and calcined states. • O, P and Ca edges detail the local coordination of these atoms in dried and calcined bone. • O K-edge shows that the surface of bone has more CO{sub 3} while the interior has more PO{sub 4}. • P and Ca edges eliminate the presence of pyrophosphates and confirmed the presence of HA and β-TCP. • The stability of these phosphates on calcination has been examined using XANES.

Programme and french realizations concerning alpha wastes

International Nuclear Information System (INIS)

Sousselier, Y.

1978-01-01

Water reactors and breeder spent fuels are reprocessed to recover plutonium, minimise wastes and decrease irradiation risks. Alloys formation, glass addition and vitrification or metallic matrix are studied to treat cladding hulls. Plutonium content is controlled by alpha spectrometry or prompt neutrons measurements or neutrons activation. Wastes are calcinated or crushed at low temperature to recover transuranium elements by solvent extraction or precipitation or ionic exchange or ultrafiltration. Wastes are calcinated or crushed at low temperature to recover transuranium elements by solvent extraction or precipitation or ionic exchange or ultrafiltration. Wastes are embedded into bitumen or thermosetting resins and long term storage in geologic formation is studied [fr

Development of thermal conditioning technology for Alpha-containment wastes: Alpha-contaminated waste incineration technology

Energy Technology Data Exchange (ETDEWEB)

Kim, Joon Hyung; Kim, Jeong Guk; Yang, Hee Chul; Choi, Byung Seon; Jeong, Myeong Soo

1999-03-01

As the first step of a 3-year project named 'development of alpha-contaminated waste incineration technology', the basic information and data were reviewed, while focusing on establishment of R and D direction to develop the final goal, self-supporting treatment of {alpha}- wastes that would be generated from domestic nuclear industries. The status on {alpha} waste incineration technology of advanced states was reviewed. A conceptual design for {alpha} waste incineration process was suggested. Besides, removal characteristics of volatile metals and radionuclides in a low-temperature dry off-gas system were investigated. Radiation dose assessments and some modification for the Demonstration-scale Incineration Plant (DSIP) at Korea Atomic Energy Research Institute (KAERI) were also done.

Development of thermal conditioning technology for Alpha-containment wastes: Alpha-contaminated waste incineration technology

International Nuclear Information System (INIS)

Kim, Joon Hyung; Kim, Jeong Guk; Yang, Hee Chul; Choi, Byung Seon; Jeong, Myeong Soo

1999-03-01

As the first step of a 3-year project named 'development of alpha-contaminated waste incineration technology', the basic information and data were reviewed, while focusing on establishment of R and D direction to develop the final goal, self-supporting treatment of α- wastes that would be generated from domestic nuclear industries. The status on α waste incineration technology of advanced states was reviewed. A conceptual design for α waste incineration process was suggested. Besides, removal characteristics of volatile metals and radionuclides in a low-temperature dry off-gas system were investigated. Radiation dose assessments and some modification for the Demonstration-scale Incineration Plant (DSIP) at Korea Atomic Energy Research Institute (KAERI) were also done

Plasma technology for waste treatment

International Nuclear Information System (INIS)

Cohn, D.R.

1995-01-01

Improved environmental cleanup technology is needed to meet demanding goals for remediation and treatment of future waste streams. Plasma technology has unique features which could provide advantages of reduced secondary waste, lower cost, and onsite treatment for a wide variety of applications. Plasma technology can provide highly controllable processing without the need for combustion heating. It can be used to provide high temperature processing (∼10,000 degrees C). Plasma technology can also be employed for low temperature processing (down to room temperature range) through selective plasma chemistry. A graphite electrode arc plasma furnace at MIT has been used to investigate high temperature processing of simulated solid waste for Department of Energy environmental cleanup applications. Stable, non-leachable glass has been produced. To ensure reliable operation and to meet environmental objectives, new process diagnostics have been developed to measure furnace temperature and to determine metals emissions in the gaseous effluent. Selective plasma destruction of dilute concentrations of hazardous compounds in gaseous waste streams has been investigated using electron beam generated plasmas. Selective destruction makes it possible to treat the gas steam at relatively low temperatures in the 30-300 degrees C range. On-line infrared measurements have been used in feedback operation to maximize efficiency and ensure desired performance. Plasma technology and associated process diagnostics will be used in future studies of a wide range of waste streams

DOE low-level waste long term technology development

International Nuclear Information System (INIS)

Barainca, M.J.

1982-01-01

The objective of the Department of Energy's Low-Level Waste Management Program is to provide a low-level waste management system by 1986. Areas of concentration are defined as: (1) Waste Generation Reduction Technology, (2) Process and Handling Technology, (3) Environmental Technology, (4) Low-Level Waste Disposal Technology. A program overview is provided with specific examples of technical development. 2 figures

Proceedings of emerging technologies for hazardous waste management

International Nuclear Information System (INIS)

Tedder, D.W.

1992-01-01

This book contains proceedings of emerging technologies for hazardous waste management. Topics covered include: Low-temperature oxidation of organic chemical wastes; Advanced waste minimization strategies; Treatment of manufactured gas plant (MGP) and similar wastes; Bioremediation of soils and sediments; Advances in radioactive waste treatment; Computer aides approaches to hazardous waste management; Advances in soil remediation; Low-temperature oxidation of organic chemical waste; Boremediation: Micro, meso, and macro-scale processes; In situ remediation techniques; Treatment of hazardous organics with radiation or solar energy; Technologies for management of municipal waste combustion residues; Environmental restoration and waste management; and Advanced separation and stabilization technologies

Mixed Waste Integrated Program -- Problem-oriented technology development

International Nuclear Information System (INIS)

Hart, P.W.; Wolf, S.W.; Berry, J.B.

1994-01-01

The Mixed Waste Integrated Program (MWIP) is responding to the need for DOE mixed waste treatment technologies that meet these dual regulatory requirements. MWIP is developing emerging and innovative treatment technologies to determine process feasibility. Technology demonstrations will be used to determine whether processes are superior to existing technologies in reducing risk, minimizing life-cycle cost, and improving process performance. Technology development is ongoing in technical areas required to process mixed waste: materials handling, chemical/physical treatment, waste destruction, off-gas treatment, final forms, and process monitoring/control. MWIP is currently developing a suite of technologies to process heterogeneous waste. One robust process is the fixed-hearth plasma-arc process that is being developed to treat a wide variety of contaminated materials with minimal characterization. Additional processes encompass steam reforming, including treatment of waste under the debris rule. Advanced off-gas systems are also being developed. Vitrification technologies are being demonstrated for the treatment of homogeneous wastes such as incinerator ash and sludge. An alternative to conventional evaporation for liquid removal--freeze crystallization--is being investigated. Since mercury is present in numerous waste streams, mercury removal technologies are being developed

Technology for commercial radioactive waste management

International Nuclear Information System (INIS)

1979-05-01

Conceptual processes and facilities for treating gaseous and various transuranium (TRU) wastes produced during the past fission portion of the light water reactor fuel cycle are described in volume 2. The goal of the treatment process for TRU wastes and for long-lived radionuclides removed from the gaseous waste streams is to convert these wastes to stable products suitable for placement in geologic isolation repositories. The treatment concepts are based on available technology. They do not necessarily represent an optimum design but are representative of what could be achieved with current technology. In actual applications it is reasonable to expect that there could be some improvement over these concepts that might be reflected in either lower costs or lower environmental impacts or both. These conceptual descriptions do provide a reasonable basis for cost analysis and for development of estimates of environmental impacts. The waste treatment technologies considered here include: high-level waste solidification, packaging of fuel residue, failed equipment and noncombustible waste treatment, general trash and combustible waste treatment, degraded solvent treatment, dilute aqueous waste pretreatment, immobilization of wet and solid wastes, off-gas particle removal systems, fuel reprocessing plant dissolver off-gas treatment, process off-gas treatment, and fuel reprocessing plant atmospheric protection system

1st International Conference on Calcined Clays for Sustainable Concrete

CERN Document Server

Favier, Aurélie

2015-01-01

This volume focuses on research and practical issues linked to Calcined Clays for Sustainable Concrete. The main subjects are geology of clays, hydration and performance of blended systems with calcined clays, alkali activated binders, economic and environmental impacts of the use of calcined clays in cement based materials. Topics addressed in this book include the influence of processing on reactivity of calcined clays, influence of clay mineralogy on reactivity, geology of clay deposits, Portland-calcined clay systems, hydration, durability, performance, Portland-calcined clay-limestone systems, hydration, durability, performance, calcined clay-alkali systems, life cycle analysis, economics and environmental impact of use of calcined clays in cement and concrete, and field applications. This book compiles the different contributions of the 1st International Conference on Calcined Clays for Sustainable Concrete, which took place in Lausanne, Switzerland, June, 23-25, 2015.The papers present the latest  res...

Regulatory barriers to hazardous waste technology innovation

International Nuclear Information System (INIS)

Kuusinen, T.L.; Siegel, M.R.

1991-02-01

The primary federal regulatory programs that influence the development of new technology for hazardous waste are the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation and Liability Act (CERCLA, also commonly known as Superfund). Two important aspects of RCRA that can create barriers to hazardous waste technology innovation are technology-based waste pre-treatment standards and a cumbersome permitting program. By choosing a technology-based approach to the RCRA land disposal restrictions program, the US Environmental Protection Agency (EPA) has simultaneously created tremendous demand for the technologies specified in its regulations, while at the same time significantly reduced incentives for technology innovation that might have otherwise existed. Also, the RCRA hazardous waste permitting process can take years and cost hundreds of thousands of dollars. The natural tendency of permit writers to be cautious of unproven (i.e., innovative) technology also can create a barrier to deployment of new technologies. EPA has created several permitting innovations, however, to attempt to mitigate this latter barrier. Understanding the constraints of these permitting innovations can be important to the success of hazardous waste technology development programs. 3 refs

Oxyfuel carbonation/calcination cycle for low cost CO2 capture in existing power plants

International Nuclear Information System (INIS)

Romeo, Luis M.; Abanades, J. Carlos; Escosa, Jesus M.; Pano, Jara; Gimenez, Antonio; Sanchez-Biezma, Andres; Ballesteros, Juan C.

2008-01-01

Postcombustion CO 2 capture is the best suitable capture technology for existing coal power plants. This paper focuses on an emerging technology that involves the separation of CO 2 using the reversible carbonation reaction of CaO to capture CO 2 from the flue gas, and the calcination of CaCO 3 to regenerate the sorbent and produce concentrated CO 2 for storage. We describe the application to this concept to an existing (with today's technology) power plant. The added capture system incorporates a new supercritical steam cycle to take advantage of the large amount of heat coming out from the high temperature capture process (oxyfired combustion of coal is needed in the CaCO 3 calciner). In these conditions, the capture system is able to generate additional power (26.7% efficiency respect to LHV coal input to the calciner after accounting for all the penalties in the overall system), without disturbing the steam cycle of the reference plant (that retains its 44.9 efficiency). A preliminary cost study of the overall system, using well established analogues in the open literature for the main components, yields capture cost around 16 Euro /ton CO 2 avoided and incremental cost of electricity of just over 1 Euro /MW h e

Mixed Waste Integrated Program emerging technology development

International Nuclear Information System (INIS)

Berry, J.B.; Hart, P.W.

1994-01-01

The US Department of Energy (DOE) is responsible for the management and treatment of its mixed low-level wastes (MLLW). MLLW are regulated under both the Resource Conservation and Recovery Act and various DOE orders. Over the next 5 years, DOE will manage over 1.2 m 3 of MLLW and mixed transuranic (MTRU) wastes. In order to successfully manage and treat these mixed wastes, DOE must adapt and develop characterization, treatment, and disposal technologies which will meet performance criteria, regulatory approvals, and public acceptance. Although technology to treat MLLW is not currently available without modification, DOE is committed to developing such treatment technologies and demonstrating them at the field scale by FY 1997. The Office of Research and Development's Mixed Waste Integrated Program (MWIP) within the DOE Office of Environmental Management (EM), OfFice of Technology Development, is responsible for the development and demonstration of such technologies for MLLW and MTRU wastes. MWIP advocates and sponsors expedited technology development and demonstrations for the treatment of MLLW

Design and installation of a laboratory-scale system for radioactive waste treatment

International Nuclear Information System (INIS)

Berger, D.N.; Knox, C.A.; Siemens, D.H.

1980-05-01

Described are the mechanical design features and remote installation of a laboratory-scale radiochemical immobilization system which is to provide a means at Pacific Northwest Laboratory of studying effluents generated during solidification of high-level liquid radioactive waste. Detailed are the hot cell, instrumentation, two 4-in. and 12-in. service racks, the immobilization system modules - waste feed, spray calciner unit, and effluent - and a gamma emission monitor system for viewing calcine powder buildup in the spray calciner/in-can melter

Uranium dioxide calcining apparatus and method

International Nuclear Information System (INIS)

Cole, E.A.; Peterson, R.S.

1978-01-01

This invention relates to an improved continuous calcining apparatus for consistently and controllably producing from calcinable reactive solid compounds of uranium, such as ammonium diuranate, uranium dioxide (UO 2 ) having an oxygen to uranium ratio of less than 2.2. The apparatus comprises means at the outlet end of a calciner kiln for receiving hot UO 2 , means for cooling the UO 2 to a temperature of below 100 0 C and conveying the cooled UO 2 to storage or to subsequent UO 2 processing apparatus where it finally comes into contact with air, the means for receiving, cooling and conveying being sealed to the outlet end of the calciner and being maintained full of UO 2 and so operable as to exclude atmospheric oxygen from coming into contact with any UO 2 which is at elevated temperatures where it would readily oxidize, without the use of extra hydrogen gas in said means

Calcination of the cerium concentrate to be cerium oxide

International Nuclear Information System (INIS)

Suyanti; MV Purwani

2016-01-01

Calcination of the cerium concentrate to be cerium oxide has done. The cerium concentrate were obtained from the Ce making process wear KBrO_3 and without using KBrO_3. The calcination were done with a variation of time 1, 2, 3 and 4 hours with the temperature variations of 700, 800 and 900°C. The easiest calcination of Ce concentrates to be CeO_2 containing majority of Ce(OH)_4 and contains least impurities as Th(OH)_4, (NH_4)_2Y(NO_3), H_4N_5O_1_2La, H_1_2N_3NdO_1_5 and N_3O_9Sm. On the calcination of Ce concentrates process results without using KBrO_3 1, the calcination temperature 900°C was obtained CeO_2 content of 73.53% for calcination time of 4 hours, has little difference when compared with the predictions and calculation result of complete calcination was equal 73.84%. (author)

Development of a SREX flowsheet for the separation of strontium from dissolved INEEL zirconium calcine

International Nuclear Information System (INIS)

Law, J.D.; Wood, D.J.; Todd, T.A.

1999-01-01

Laboratory experimentation has indicated that the SREX process is effective for partitioning 90 Sr from acidic radioactive waste solutions located at the Idaho Nuclear Technology and Engineering Center. These laboratory results were used to develop a flowsheet for countercurrent testing of the SREX process with dissolved pilot plant calcine. Testing was performed using 24 stages of 2-cm diameter centrifugal contactors which are installed in the Remote Analytical Laboratory hot cell. Dissolved Run No.64 pilot plant calcine spiked with 85 Sr was used as feed solution for the testing. The flowsheet tested consisted of an extraction section (0.15 M 4prime,4prime(5prime)-di-(tert-butylcyclohexo)-18-crown-6 and 1.5 M TBP in Isopar-L.), a 1.0 M NaNO 3 scrub section to remove extracted K from the SREX solvent, a 0.01 M HNO 3 strip section for the removal of Sr from the SREX solvent, a 0.25 M Na2CO 3 wash section to remove degradation products from the solvent, and a 0.1 M HNO 3 rinse section. The behavior of 85 Sr, Na, K, Al, B, Ca, Cr, Fe, Ni, and Zr was evaluated. The described flowsheet successfully extracted 85 Sr from the dissolved pilot plant calcine with a removal efficiency of 99.6%. Distribution coefficients for 85 Sr ranged from 3.6 to 4.5 in the extraction section. With these distribution coefficients a removal efficiency of approximately >99.99% was expected. It was determined that the lower than expected removal efficiency can be attributed to a stage efficiency of only 60% in the extraction section. Extracted K was effectively scrubbed from the SREX solvent with the 1.0 M NaNO 3 resulting in only 6.4% of the K in the HLW strip product. Sodium was not extracted from the dissolved calcine by the SREX solvent; however, the use of a 1.0 M NaNO 3 scrub solution resulted in a Na concentration of 70 mg/L (12.3% of the feed concentration) in the HLW strip product. Al, B, Ca, Cr, Fe, Ni, and Zr were determined to be essentially inextractable

Hazardous waste market and technology trends

International Nuclear Information System (INIS)

Anon.

1993-01-01

What forces are currently driving the growth of the hazardous waste remediation market? Which factors will control the development of cleanup technologies during the next decade? At what types of sites are various technologies being applied? In an effort to answer these questions, EPA has produced an overview of trends in the demand for remedial technologies at CERCLA, RCRA corrective action, underground storage tank (UST), and other cleanup sites across the United States. The 160-page document, entitled Cleaning Up the Nation's Waste Sites: Markets and Technology Trends, was developed by EPA's Office of Solid Waste and Emergency Response. Highlights from the report are presented below. 1 ref., 2 figs., 1 tab

Demonstration of the TRUEX process for the treatment of actual high activity tank waste at the INEEL using centrifugal contactors

International Nuclear Information System (INIS)

Law, J.D.; Brewer, K.N.; Todd, T.A.; Olson, L.G.

1997-01-01

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering and Environmental Laboratory (INEEL), formerly reprocessed spent nuclear fuel to recover fissionable uranium. The radioactive raffinates from the solvent extraction uranium recovery processes were converted to granular solids (calcine) in a high temperature fluidized bed. A secondary liquid waste stream was generated during the course of reprocessing, primarily from equipment decontamination between campaigns and solvent wash activities. This acidic tank waste cannot be directly calcined due to the high sodium content and has historically been blended with reprocessing raffinates or non-radioactive aluminum nitrate prior to calcination. Fuel reprocessing activities are no longer being performed at the ICPP, thereby eliminating the option of waste blending to deplete the waste inventory. Currently, approximately 5.7 million liters of high-activity waste are temporarily stored at the ICPP in large underground stainless-steel tanks. The United States Environmental Protection Agency and the Idaho Department of Health and Welfare filed a Notice of Noncompliance in 1992 contending some of the underground waste storage tanks do not meet secondary containment. As part of a 1995 agreement between the State of Idaho, the Department of Energy, and the Department of Navy, the waste must be removed from the tanks by 2012. Treatment of the tank waste inventories by partitioning the radionuclides and immobilizing the resulting high-activity and low-activity waste streams is currently under evaluation. A recent peer review identified the most promising radionuclide separation technologies for evaluation. The Transuranic Extraction-(TRUEX) process was identified as a primary candidate for separation of the actinides from ICPP tank waste

Mixed Waste Integrated Program emerging technology development

Energy Technology Data Exchange (ETDEWEB)

Berry, J.B. [Oak Ridge National Lab., TN (United States); Hart, P.W. [USDOE, Washington, DC (United States)

1994-06-01

The US Department of Energy (DOE) is responsible for the management and treatment of its mixed low-level wastes (MLLW). MLLW are regulated under both the Resource Conservation and Recovery Act and various DOE orders. Over the next 5 years, DOE will manage over 1.2 m{sup 3} of MLLW and mixed transuranic (MTRU) wastes. In order to successfully manage and treat these mixed wastes, DOE must adapt and develop characterization, treatment, and disposal technologies which will meet performance criteria, regulatory approvals, and public acceptance. Although technology to treat MLLW is not currently available without modification, DOE is committed to developing such treatment technologies and demonstrating them at the field scale by FY 1997. The Office of Research and Development`s Mixed Waste Integrated Program (MWIP) within the DOE Office of Environmental Management (EM), OfFice of Technology Development, is responsible for the development and demonstration of such technologies for MLLW and MTRU wastes. MWIP advocates and sponsors expedited technology development and demonstrations for the treatment of MLLW.

Review: Waste-Pretreatment Technologies for Remediation of Legacy Defense Nuclear Wastes

Energy Technology Data Exchange (ETDEWEB)

Wilmarth, William R.; Lumetta, Gregg J.; Johnson, Michael E.; Poirier, Micheal R.; Thompson, Major C.; Suggs, Patricia C.; Machara, N.

2011-01-13

The U.S. Department of Energy (DOE) is responsible for retrieving, immobilizing, and disposing of radioactive waste that has been generated during the production of nuclear weapons in the United States. The vast bulk of this waste material is stored in underground tanks at the Savannah River Site in South Carolina and the Hanford Site in Washington State. The general strategy for treating the radioactive tank waste consists of first separating the waste into high-level and low-activity fractions. This initial partitioning of the waste is referred to as pretreatment. Following pretreatment, the high-level fraction will be immobilized in a glass form suitable for disposal in a geologic repository. The low-activity waste will be immobilized in a waste form suitable for disposal at the respective site. This paper provides a review of recent developments in the application of pretreatment technologies to the processing of the Hanford and Savannah River radioactive tank wastes. Included in the review are discussions of 1) solid/liquid separations methods, 2) cesium separation technologies, and 3) other separations critical to the success of the DOE tank waste remediation effort. Also included is a brief discussion of the different requirements and circumstances at the two DOE sites that have in some cases led to different choices in pretreatment technologies.

Proceedings of emerging technologies for hazardous waste management

International Nuclear Information System (INIS)

Tedder, D.W.

1992-01-01

This paper contains the proceedings of emergin technologies for hazardous waste management. Topics covered include: advanced transuranic waste managements; remediation of soil/water systems contaminated with nonaqueous pollutants; advances in molten salt oxidation; air treatment and protection; advanced waste minimization strategies; removal of hazardous materials from soils or groundwater; bioremediation of soils and sediment; innovation, monitoring, and asbestos; high-level liquid waste chemistry in the Hanford tanks; biological contributions to soil and groundwater remediation; soil treatment technologies; pollution prevention; incineration and vitrification; current technology; systematic design approaches to hazardous waste management; waste management and environmental restoration at Savannah River; soil washing and flushing for remediation of hazardous wastes

Activity and structure of calcined coal gangue

Energy Technology Data Exchange (ETDEWEB)

Gong Chenchen; Li Dongxu; Wang Xiaojun; Li Zongjin [Nanjing University of Technology, Nanjing (China). College of Materials Science and Engineering

2007-12-15

Coal gangue was activated by means of calcination in seven temperature ranges. Systematic research was made about activation mechanism and structural evolution. The glycerin-ethanol method, SEM, MIP and XRD were used to determine the variation of structure and activation of coal gangue during calcination. The experimental results show that because of heat treatment in the range of calcination temperatures, mineral composition and microstructure of coal gangue are changed. In addition, its activity is improved. The amount of lime absorbed by the sample calcined at 700{sup o}C is 2-4 times that by uncalcined coal gangue in the course of hydration. When NaOH is added to coal gangue-lime system, the hydration reaction rate of the system is increased and the microstructure of hydrating samples of coal gangue is improved.

Technologies 1995: environment and wastes treatment

International Nuclear Information System (INIS)

Anon.

1995-03-01

From new technical or scientific developments, new products launching, and markets evolutions, this catalog gives informations selection on research and development projects, new fabrication processes, activities and plants strategies, licences or technology transfers opportunities. The covered fields are: atmospheric pollution controls, water and liquid wastes treatment, polluted soils treatments, noise and odors treatments, municipal and industrial wastes treatments (metal, plastic, paper, glass), clean materials and technologies, radioactive wastes, and european cooperation programs. (A.B.)

Low-level waste volume reduction--physicochemical systems

International Nuclear Information System (INIS)

Ferrigno, D.P.

1980-01-01

In some cases, volume reduction (VR) equipment may be called upon to reduce noncombustible liquid wastes to essentially dry salts and/or oxides. In other cases, it may be called upon to reduce combustible solids and liquids to ashes and innocuous gases. In brand terms, four kinds of processes are available to further reduce the volume of waste generated at nuclear facilities. These include high-solids evaporation, alternative evaporative designs, extruders/mixers, and calciner/incinerators. This paper discusses the following VR processes for radioactive wastes at nuclear facilities: evaporator/crystallizer; fluid bed dryer/incinerator; fluid bed calciner/incinerator; inert carrier radwaste processor; and molten glass incinerator

Technology Summary Advancing Tank Waste Retrieval And Processing

International Nuclear Information System (INIS)

Sams, T.L.; Mendoza, R.E.

2010-01-01

This technology overview provides a high-level summary of technologies being investigated and developed by Washington River Protection Solutions (WRPS) to advance Hanford Site tank waste retrieval and processing. Technology solutions are outlined, along with processes and priorities for selecting and developing them. This technology overview provides a high-level summary of technologies being investigated, developed, and deployed by WRPS to advance Hanford Site tank waste retrieval and processing. Transformational technologies are needed to complete Hanford tank waste retrieval and treatment by 12/31/2047. Hanford's underground waste storage tanks hold approximately 57 million gallons of radiochemical waste from nuclear defense production - more tank waste than any other site in the United States. In addition, the waste is uniquely complicated because it contains constituents from at least six major radiochemical processes and several lesser processes. It is intermixed and complexed more than any other waste collection known to exist in the world. The multi-faceted nature of Hanford's tank waste means that legally binding agreements in the Federal Facility Agreement and Consent Order (known as the Tri-Party Agreement) and between the Department of Energy (DOE) and its contractors may not be met using current vitrification schedules, plans, and methods. WRPS and the DOE are developing, testing, and deploying technologies to meet the necessary commitments and complete the DOE's River Protection Project (RPP) mission within environmentally acceptable requirements. Technology solutions are outlined, along with processes and priorities for selecting and developing them. DOE's Office of Environmental Management (EM) identifies the environmental management technology needs and the activities necessary to address them. The U.S. Congress then funds these activities through EM or the DOE field offices. Finally, an array of entities that include DOE site prime contractors and

Heat transfer in high-level waste management

International Nuclear Information System (INIS)

Dickey, B.R.; Hogg, G.W.

1979-01-01

Heat transfer in the storage of high-level liquid wastes, calcining of radioactive wastes, and storage of solidified wastes are discussed. Processing and storage experience at the Idaho Chemical Processing Plant are summarized for defense high-level wastes; heat transfer in power reactor high-level waste processing and storage is also discussed

Durability, mechanical, and thermal properties of experimental glass-ceramic forms for immobilizing ICPP high level waste

International Nuclear Information System (INIS)

Vinjamuri, K.

1990-01-01

The high-level liquid waste generated at the Idaho Chemical Processing Plant (ICPP) is routinely solidified into granular calcined high-level waste (HLW) and stored onsite. Research is being conducted at the ICPP on methods of immobilizing the HLW, including developing a durable glass-ceramic form which has the potential to significantly reduce the final waste volume by up to 60% compared to a glass form. Simulated, pilot plant, non-radioactive, calcines similar to the composition of the calcined HLW and glass forming additives are used to produce experimental glass-ceramic forms. The objective of the research reported in this paper is to study the impact of ground calcine particle size on durability and mechanical and thermal properties of experimental glass-ceramic forms

Research and development of improved type radioactive waste volume reduction system

International Nuclear Information System (INIS)

Okamoto, Masahiro; Watanabe, Yoshifumi; Yamaoka, Katsuaki; Masaki, Tetsuo; Akagawa, Yoshihiro; Murakami, Tadashi; Miyake, Takashi.

1985-01-01

Development and research had been conducted since 1978 on an improved type radioactive waste volume reduction system incorporating calcining and incinerating fluidized bed type furnaces. This system can dispose of concentrated liquid wastes, combustible solid wastes, spent ion exchange resins and so forth by calcination or incineration to turn them into reduced-volume products. Recently a pilot test facility has constructed and tests has been conducted to demonstrate actual performance. Representative results of pilot tests are reported in this paper. (author)

Waste-to-energy technologies and project implementation

CERN Document Server

Rogoff, Marc J

2011-01-01

This book covers in detail programs and technologies for converting traditionally landfilled solid wastes into energy through waste-to-energy projects. Modern Waste-to-Energy plants are being built around the world to reduce the levels of solid waste going into landfill sites and contribute to renewable energy and carbon reduction targets. The latest technologies have also reduced the pollution levels seen from early waste incineration plants by over 99 per cent. With case studies from around the world, Rogoff and Screve provide an insight into the different approaches taken to the planning and implementation of WTE. The second edition includes coverage of the latest technologies and practical engineering challenges as well as an exploration of the economic and regulatory context for the development of WTE.

Fossil energy waste management. Technology status report

Energy Technology Data Exchange (ETDEWEB)

Bossart, S.J.; Newman, D.A.

1995-02-01

This report describes the current status and recent accomplishments of the Fossil Energy Waste Management (FE WM) projects sponsored by the Morgantown Energy Technology Center (METC) of the US Department of Energy (DOE). The primary goal of the Waste Management Program is to identify and develop optimal strategies to manage solid by-products from advanced coal technologies for the purpose of ensuring the competitiveness of advanced coal technologies as a future energy source. The projects in the Fossil Energy Waste Management Program are divided into three types of activities: Waste Characterization, Disposal Technologies, and Utilization Technologies. This technology status report includes a discussion on barriers to increased use of coal by-products. Also, the major technical and nontechnical challenges currently being addressed by the FE WM program are discussed. A bibliography of 96 citations and a list of project contacts is included if the reader is interested in obtaining additional information about the FE WM program.

Waste management issues, a set of technologies

International Nuclear Information System (INIS)

Gautrot, J.J.

2000-01-01

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

Technologies for environmental cleanup: Toxic and hazardous waste management

International Nuclear Information System (INIS)

Ragaini, R.C.

1993-12-01

This is the second in a series of EUROCOURSES conducted under the title, ''Technologies for Environmental Cleanup.'' To date, the series consist of the following courses: 1992, soils and groundwater; 1993, Toxic and Hazardous Waste Management. The 1993 course focuses on recent technological developments in the United States and Europe in the areas of waste management policies and regulations, characterization and monitoring of waste, waste minimization and recycling strategies, thermal treatment technologies, photolytic degradation processes, bioremediation processes, medical waste treatment, waste stabilization processes, catalytic organic destruction technologies, risk analyses, and data bases and information networks. It is intended that this course ill serve as a resource of state-of-the-art technologies and methodologies for the environmental protection manager involved in decisions concerning the management of toxic and hazardous waste

Development of the destruction technology for radioactive organic solid wastes

Energy Technology Data Exchange (ETDEWEB)

Oh, Won Zin; Park, H.S.; Lee, K.W. [and others

1999-04-01

The followings were studied through the project entitled 'Technology development for nuclear fuel cycle waste treatment'. 1. Organic waste decomposition technology development A. Destruction technology for organic wastes using Ag(2)-mediated electrochemical oxidation B. Recovery and regeneration technology for the spent chemicals used in the MEO process 2. Radioactive metal waste recycling technology A. Surface decontamination processes B. Decontamination waste treatment technology 3. Volume reduction technology nuclear fuel cycle (NFC) technology A. Estimation of the amount of radwastes and the optimum volume reduction methodology of domestic NFC B. Pretreatment of spent fuel cladding by electrochemical decontamination C. Hot cell process technology for the treatment of NFC wastes 4. Design and fabrication of the test equipment of volume reduction and reuse of alpha contaminated wastes 5. Evaluation on environmental compatibility of NFC A. Development of evaluation methodology on environmental friendliness of NFC B. Residual activity assessment of recycling wastes. (author). 321 refs., 54 tabs., 183 figs.

Development of the destruction technology for radioactive organic solid wastes

International Nuclear Information System (INIS)

Oh, Won Zin; Park, H.S.; Lee, K.W.

1999-04-01

The followings were studied through the project entitled 'Technology development for nuclear fuel cycle waste treatment'. 1. Organic waste decomposition technology development A. Destruction technology for organic wastes using Ag(2)-mediated electrochemical oxidation B. Recovery and regeneration technology for the spent chemicals used in the MEO process 2. Radioactive metal waste recycling technology A. Surface decontamination processes B. Decontamination waste treatment technology 3. Volume reduction technology nuclear fuel cycle (NFC) technology A. Estimation of the amount of radwastes and the optimum volume reduction methodology of domestic NFC B. Pretreatment of spent fuel cladding by electrochemical decontamination C. Hot cell process technology for the treatment of NFC wastes 4. Design and fabrication of the test equipment of volume reduction and reuse of alpha contaminated wastes 5. Evaluation on environmental compatibility of NFC A. Development of evaluation methodology on environmental friendliness of NFC B. Residual activity assessment of recycling wastes. (author). 321 refs., 54 tabs., 183 figs

Technology for commercial radioactive waste management

International Nuclear Information System (INIS)

1979-05-01

The scope of this report is limited to technology for management of past-fission wastes produced in the commercial nuclear power light water reactor fuel cycle. Management of spent fuel (as a waste), high-level and other transuranic wastes, and gaseous wastes are characterized. Non-transuranic wastes are described, but management of these wastes, except for gaseous wastes, is excluded from the scope of this report. Volume 1 contains the summary and the bases and background information

Analysis of capital and operating costs associated with high level waste solidification processes

International Nuclear Information System (INIS)

Heckman, R.A.; Kniazewycz, B.G.

1978-03-01

An analysis was performed to evaluate the sensitivity of annual operating costs and capital costs of waste solidification processes to various parameters defined by the requirements of a proposed Federal waste repository. Five process methods and waste forms examined were: salt cake, spray calcine, fluidized bed calcine, borosilicate glass, and supercalcine multibarrier. Differential cost estimates of the annual operating and maintenance costs and the capital costs for the five HLW solidification alternates were developed

Development and evaluation of alumina calcination

International Nuclear Information System (INIS)

Bennett, I.J.

2000-01-01

This thesis focuses on a number of aspects governing the transformation of gibbsite, via intermediate phases, to α-alumina. These aspects include the size and morphology of the gibbsite grains, the influence of additions of foreign elements, the effect of a mechanical treatment of the gibbsite prior to calcination, and combinations of these factors. The materials were characterised by scanning electron microscopy, X-ray diffraction and surface area measurements. For some of the calcined materials an attempt was made to sinter the powders to a dense body to investigate if any of the treatments during calcination had an effect on this process. The literature review covers the current state of understanding of the production of bulk alumina powder by the Bayer process and the phase changes seen on calcination of precursors to the stable α-alumina phase. A detailed description of the phase changes is given and the various routes and conditions necessary for the transformations to occur are considered. The transformations are examined in relation to the morphology of the crystals and the variables controlling the phase transformation route are discussed. Calcination in air showed that the size of the gibbsite grain governs the calcination route taken to reach oc-alumina. The standard gibbsites used in this work show a mixed calcination sequence transforming both via the boehmite phase, followed by the γ, δ and θ phases, and via the χ and κ phases. The formation of boehmite is attributed to retention of water vapour within the grain. Differences in morphology of the starting materials showed that for the range of materials seen, the morphology of the grain is less important than its size. The super fine material confirmed that a small grain size transforms via the non-boehmite route only, with the other gibbsites taking intermediate routes as for the standard gibbsites. Of the additions made prior to calcination, aluminium fluoride was found to reduce the

Microwave heating application in calcination and SYNROC formation

International Nuclear Information System (INIS)

Ambashta, R.D.; Wattal, P.K.; Malav, R.K.; Mallik, G.K.

2006-01-01

Microwave for calcination of titanate based ceramic wasteform (SYNROC) is being reported for the first time in this paper. Although major constituents in SYNROC were non microwave active, the combination with microwave active constituents rendered the mixture calcinable. Calcine was sintered at 1150 degC under hot uniaxial conditions at an applied pressure of ∼30 MPa. XRD shows presence of major phases of SYNROC in the compacted sample. (author)

Mathematical modeling of an in-line low-NOx calciner

DEFF Research Database (Denmark)

Iliuta, Ion; Dam-Johansen, Kim; Jensen, Lars Skaarup

2002-01-01

The reduction of the NOx content in in-line-calciner-type kiln systems can be made by optimization of the primary filing in the rotary kiln and of the secondary firing in the calciner. Because the optimization of calciner offers greater opportunities the mathematical modeling of this reactor...

An overview of in situ waste treatment technologies

International Nuclear Information System (INIS)

Walker, S.; Hyde, R.A.; Piper, R.B.; Roy, M.W.

1992-01-01

In situ technologies are becoming an attractive remedial alternative for eliminating environmental problems. In situ treatments typically reduce risks and costs associated with retrieving, packaging, and storing or disposing-waste and are generally preferred over ex situ treatments. Each in situ technology has specific applications, and, in order to provide the most economical and practical solution to a waste problem, these applications must be understood. This paper presents an overview of thirty different in situ remedial technologies for buried wastes or contaminated soil areas. The objective of this paper is to familiarize those involved in waste remediation activities with available and emerging in situ technologies so that they may consider these options in the remediation of hazardous and/or radioactive waste sites. Several types of in situ technologies are discussed, including biological treatments, containment technologies, physical/chemical treatments, solidification/stabilization technologies, and thermal treatments. Each category of in situ technology is briefly examined in this paper. Specific treatments belonging to these categories are also reviewed. Much of the information on in situ treatment technologies in this paper was obtained directly from vendors and universities and this information has not been verified

Long-term high-level waste technology. Composite quarterly technical report: April-June 1981

International Nuclear Information System (INIS)

Cornman, W.R.

1981-12-01

This series of reports summarizes research and development studies on the immobilization of high-level wastes from the chemical reprocessing of nuclear reactor fuels. The reports are grouped under the following tasks: (1) program management and support; (2) waste preparation; (3) waste fixation; and (4) final handling. Some of the highlights are: leaching properties were obtained for titanate and tailored ceramic materials being developed at ICPP to immobilize zirconia calcine; comparative leach tests, hot-cell tests, and process evaluations were conducted of waste form alternatives to borosilicate glass for the immobilization of SRP high-level wastes, experiments were run at ANL to qualify neutron activation analysis and radioactive tracers for measuring leach rates from simulated waste glasses; comparative leach test samples of SYNROC D were prepared, characterized, and tested at LLNL; encapsulation of glass marbles with lead or lead alloys was demonstrated on an engineering scale at PNL; a canister for reference Commercial HLW was designed at PNL; a study of the optimization of salt-crete was completed at SRL; a risk assessment showed that an investment for tornado dampers in the interim storage building of the DWPF is unjustified

Technology Roadmapping for Waste Management

International Nuclear Information System (INIS)

Bray, O.

2003-01-01

Technology roadmapping can be an effective strategic technology planning tool. This paper describes a process for customizing a generic technology roadmapping process. Starting with a generic process reduces the learning curve and speeds up the roadmap development. Similarly, starting with a generic domain model provides leverage across multiple applications or situations within the domain. A process that combines these two approaches facilitates identifying technology gaps and determining common core technologies that can be reused for multiple applications or situations within the domain. This paper describes both of these processes and how they can be integrated. A core team and a number of technology working groups develop the technology roadmap, which includes critical system requirements and targets, technology areas and metrics for each area, and identifies and evaluates possible technology alternatives to recommend the most appropriate ones to pursue. A generalized waste management model, generated by considering multiple situations or applications in terms of a generic waste management model, provides the domain requirements for the technology roadmapping process. Finally, the paper discusses lessons learns from a number of roadmapping projects

Processing of palm oil mill wastes based on zero waste technology

Science.gov (United States)

Irvan

2018-02-01

Indonesia is currently the main producer of palm oil in the world with a total production reached 33.5 million tons per year. In the processing of fresh fruit bunches (FFB) besides producing palm oil and kernel oil, palm oil mills also produce liquid and solid wastes. The increase of palm oil production will be followed by an increase in the production of waste generated. It will give rise to major environmental issues especially the discharge of liquid waste to the rivers, the emission of methane from digestion pond and the incineration of empty fruit bunches (EFB). This paper describes a zero waste technology in processing palm oil mill waste after the milling process. The technology involves fermentation of palm oil mill effluent (POME) to biogas by using continuous stirred tank reactor (CSTR) in the presence of thermophilic microbes, producing activated liquid organic fertilizer (ALOF) from discharge of treated waste effluent from biogas digester, composting EFB by spraying ALOF on the EFB in the composter, and producing pellet or biochar from EFB by pyrolysis process. This concept can be considered as a promising technology for palm oil mills with the main objective of eliminating the effluent from their mills.

Municipal solid waste disposal by using metallurgical technologies and equipments

Energy Technology Data Exchange (ETDEWEB)

Cai, Jiuju; Sun, Wenqiang [State Environmental Protection Key Laboratory of Eco-industry, Institute of Thermal and Environmental Engineering, Northeastern University, Shenyang 110819 (China)

2012-07-01

Pyrolysis of municipal solid waste can take full advantage of energy and resource and avoid producing hazardous material during this period. In combination with mature metallurgical technologies of coking by coke oven, regenerative flame furnace technology and melting by electric arc furnace, technologies of regenerative fixed bed pyrolysis technology for household waste, co-coking technology for waste plastic and blend coal, and incineration ash melting technology by electric arc technology for medical waste were respectively developed to improve current unsatisfied sorting status of waste. The investigation results of laboratory experiments, semi-industrial experiments and industrial experiments as well as their economic benefits and environmental benefits for related technologies were separately presented.

A big picture prospective for wet waste processing management

International Nuclear Information System (INIS)

Gibson, J.D.

1996-01-01

This paper provides an overview of general observations made relative to the technical and economical considerations being evaluated by many commercial nuclear power plants involving their decision making process for implementation of several new wet waste management technologies. The waste management processes reviewed include the use of, Reverse Osmosis, Non-Precoat Filters, Resin Stripping ampersand Recycling, Evaporation ampersand Calcination (RVR trademark, ROVER trademark ampersand Thermax trademark), Compression Dewatering (PressPak trademark), Incineration (Resin Express trademark), Survey ampersand Free Release (Green Is Clean) and Quantum Catalytic Extraction Processing (QCEP trademark). These waste management processes are reviewed relative to their general advantages and disadvantages associated with the processing of various wet waste streams including: reactor make-up water, floor drain sludges and other liquid waste streams such as boric acid concentrates and steam generator cleaning solutions. A summary of the conclusions generally being derived by most utilities associated with the use of these waste management processes is also provided

Assessment of selected furnace technologies for RWMC waste

International Nuclear Information System (INIS)

Batdorf, J.; Gillins, R.; Anderson, G.L.

1992-03-01

This report provides a description and initial evaluation of five selected thermal treatment (furnace) technologies, in support of earlier thermal technologies scoping work for application to the Idaho National Engineering Laboratory Radioactive Waste Management Complex (RWMC) buried wastes. The cyclone furnace, molten salt processor, microwave melter, ausmelt (fuel fired lance) furnace, and molten metal processor technologies are evaluated. A system description and brief development history are provided. The state of development of each technology is assessed, relative to treatment of RWMC buried waste

Overview: Defense high-level waste technology program

International Nuclear Information System (INIS)

Shupe, M.W.; Turner, D.A.

1987-01-01

Defense high-level waste generated by atomic energy defense activities is stored on an interim basis at three U.S. Department of Energy (DOE) operating locations; the Savannah River Plant in South Carolina, the Hanford Site in Washington, and the Idaho National Engineering Laboratory in Idaho. Responsibility for the permanent disposal of this waste resides with DOE's Office of Defense Waste and Transportation Management. The objective of the Defense High-Level Wast Technology Program is to develop the technology for ending interim storage and achieving permanent disposal of all U.S. defense high-level waste. New and readily retrievable high-level waste are immobilized for disposal in a geologic repository. Other high-level waste will be stabilized in-place if, after completion of the National Environmental Policy Act (NEPA) process, it is determined, on a site-specific basis, that this option is safe, cost effective and environmentally sound. The immediate program focus is on implementing the waste disposal strategy selected in compliance with the NEPA process at Savannah River, while continuing progress toward development of final waste disposal strategies at Hanford and Idaho. This paper presents an overview of the technology development program which supports these waste management activities and an assessment of the impact that recent and anticipated legal and institutional developments are expected to have on the program

Technological and organizational aspects of radioactive waste management

International Nuclear Information System (INIS)

2005-01-01

This document comprises collected lecture on radioactive waste management which were given by specialists of the Radioactive Waste Management Section of the IAEA, scientific-industrial enterprise 'Radon' (Moscow, RF) and A.A. Bochvar's GNTs RF VNIINM (Moscow, RF) on various courses, seminars and conferences. These lectures include the following topics: basic principles and national systems of radioactive waste management; radioactive waste sources and their classification; collection, sorting and initial characterization of radioactive wastes; choice of technologies of radioactive waste processing and minimization of wastes; processing and immobilization of organic radioactive wastes; thermal technologies of radioactive waste processing; immobilization of radioactive wastes in cements, asphalts, glass and polymers; management of worked out closed radioactive sources; storage of radioactive wastes; deactivation methods; quality control and assurance in radioactive waste management

Application and research of special waste plasma disposal technology

International Nuclear Information System (INIS)

Lan Wei

2007-12-01

The basic concept of plasma and the principle of waste hot plasma disposal technology are simply introduced. Several sides of application and research of solid waste plasma disposal technology are sumed up. Compared to the common technology, the advantages of waste hot plasma disposal technology manifest further. It becomes one of the most prospective and the most attended high tech disposal technology in particular kind of waste disposal field. The article also simply introduces some experiment results in Southwest Institute of Physics and some work on the side of importation, absorption, digestion, development of foreign plasma torch technology and researching new power sources for plasma torch. (authors)

Incorporating regulatory considerations into waste treatment technology development

International Nuclear Information System (INIS)

Siegel, M.R.; Powell, J.A.; Williams, T.A.; Kuusinen, T.L.; Lesperance, A.M.

1991-02-01

It is generally recognized that the development of new and innovative waste treatment technologies can significantly benefit the US Department of Energy's (DOE) environmental restoration and waste management program. DOE has established a research, development, demonstration, testing, and evaluation (RDDT ampersand E) program, managed by its Office of Technology Development, to encourage and direct the development of new waste treatment and management technologies. The treatment, storage, and disposal of hazardous and radioactive waste is heavily regulated both at the federal and state levels. In order to achieve the goals of applying the best new technologies in the fastest and most cost-effective manner possible, it is essential that regulatory factors be considered early and often during the development process. This paper presents a number of regulatory issues that are relevant to any program intended to encourage the development of new waste treatment and management technologies. It will also address how the use of these basic regulatory considerations can help ensure that technologies that are developed are acceptable to regulators and can therefore be deployed in the field. 2 refs

Vectra GSI, Inc. low-level waste melter testing Phase 1 test report

Energy Technology Data Exchange (ETDEWEB)

Stegen, G.E.; Wilson, C.N.

1996-02-21

A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Vectra GSI, Inc. was one of seven vendors selected for Phase 1 of the melter demonstration tests using simulated LLW that were completed during fiscal year 1995. The attached report prepared by Vectra GSI, Inc. describes results of melter testing using slurry feed and dried feeds. Results of feed drying and prereaction tests using a fluid bed calciner and rotary dryer also are described.

Vectra GSI, Inc. low-level waste melter testing Phase 1 test report

International Nuclear Information System (INIS)

Stegen, G.E.; Wilson, C.N.

1996-01-01

A multiphase program was initiated in 1994 to test commercially available melter technologies for the vitrification of the low-level waste (LLW) stream from defense wastes stored in underground tanks at the Hanford Site in southeastern Washington State. Vectra GSI, Inc. was one of seven vendors selected for Phase 1 of the melter demonstration tests using simulated LLW that were completed during fiscal year 1995. The attached report prepared by Vectra GSI, Inc. describes results of melter testing using slurry feed and dried feeds. Results of feed drying and prereaction tests using a fluid bed calciner and rotary dryer also are described

The new technologies in city waste management

International Nuclear Information System (INIS)

Marti, C.

2016-01-01

The new EU objectives included in its Circular Economy Package and the Spanish 2016-2022 Waste Plan define a new scenario of transformation of municipal solid waste management. They also define the hierarchization of waste treatment: reduction, reuse, recycling, energy valorization and, as a last resort, landfill. The use of new technologies is contributing to this transformation, including both separation at source and collection and treatment. Improved traceability of wastes via the use of sensors, technological innovation in management and the emergence of a fifth bin for selective collection of organic wastes are only some of the new elements that are increasingly common in Spanish cities. (Author)

Technology Summary Advancing Tank Waste Retreival And Processing

International Nuclear Information System (INIS)

Sams, T.L.

2010-01-01

This technology overview provides a high-level summary of technologies being investigated and developed by Washington River Protection Solutions (WRPS) to advance Hanford Site tank waste retrieval and processing. Technology solutions are outlined, along with processes and priorities for selecting and developing them. Hanford's underground waste storage tanks hold approximately 57 million gallons of radiochemical waste from nuclear defense production - more tank waste than any other site in the United States. In addition, the waste is uniquely complicated since it contains constituents from at least six major radiochemical processes and several lesser processes. It is intermixed and complexed more than any other waste collection known to exist in the world. The multi-faceted nature of Hanford's tank waste means that legally binding agreements in the Federal Facility Agreement and Consent Order (known as the Tri-Party Agreement) and between the Department of Energy (DOE) and its contractors may not be met using current vitrification schedules, plans and methods. WRPS and the DOE are therefore developing, testing, and deploying technologies to ensure that they can meet the necessary commitments and complete the DOE's River Protection Project (RPP) mission within environmentally acceptable requirements. Technology solutions are outlined, along with processes and priorities for selecting and developing them.

High-level waste management technology program plan

Energy Technology Data Exchange (ETDEWEB)

Harmon, H.D.

1995-01-01

The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs.

High-level waste management technology program plan

International Nuclear Information System (INIS)

Harmon, H.D.

1995-01-01

The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs

Reactivation properties of carbide slag as a CO{sub 2} sorbent during calcination/carbonation cycles

Energy Technology Data Exchange (ETDEWEB)

Li, Yingjie; Sun, Rongyue; Liu, Hongling; Lu, Chunmei [Shandong Univ., Jinan (China). School of Energy and Power Engineering

2013-07-01

The carbide slag from polyvinyl chloride production as industry hazardous wastes was proposed as CO{sub 2} sorbent at high temperature in calcium looping cycle. The cyclic CO{sub 2} capture behavior and the microstructure characteristics of the carbide slag as one of the typical calcium-based industrial wastes during the multiple calcination/carbonation cycles. Also, the comparisons between the carbide slag and the natural limestone in cyclic CO{sub 2} capture behavior were made. XRD analysis demonstrates that the predominating constituent of the carbide slag is Ca(OH){sub 2}. The carbonation temperature ranging from 650 to 700 C is favourable to cyclic carbonation of the carbide slag. The cyclic carbonation conversions of the carbide slag is lower than that of the limestone before a certain time, but the situation is converse after that time in a thermogravimetric analyzer. The carbide slag has better cyclic CO{sub 2} capture capacity. The carbonation conversion of the carbide slag retains 0.28 after 100 calcination/carbonation cycles, while the two limestones achieve 0.08 and 0.14 respectively at the same reaction conditions in a dual fixed-bed reactor. The microstructure of the carbide slag by SEM reveals the reason why it possesses better CO{sub 2} capture capacity.

Calcination/dissolution chemistry development Fiscal year 1995

International Nuclear Information System (INIS)

Delegard, C.H.

1995-09-01

The task open-quotes IPC Liaison and Chemistry of Thermal Reconstitutionclose quotes is a $300,000 program that was conducted in Fiscal Year (FY) 1995 with U.S. Department of Energy (DOE) Office of Research and Development (EM-53) Efficient Separations and Processing Crosscutting Program supported under technical task plan (TTP) RL4-3-20-04. The principal investigator was Cal Delegard of the Westinghouse Hanford Company (WHC). The task encompassed the following two subtasks related to the chemistry of alkaline Hanford Site tank waste: (1) Technical Liaison with the Institute of Physical Chemistry of the Russian Academy of Science (IPC/RAS) and its research into the chemistry of transuranic elements (TRU) and technetium (Tc) in alkaline media. (2) Laboratory investigation of the chemistry of calcination/dissolution (C/D) (or thermal reconstitution) as an alternative to the present reference Hanford Site tank waste pretreatment flowsheet, Enhanced Sludge Washing (ESW). This report fulfills the milestone for the C/D subtask to open-quotes Provide End-of-Year Report on C/D Laboratory Test Resultsclose quotes due 30 September 1995. A companion report, fulfilling the milestone to provide an end-of-year report on the IPC/RAS liaison, also has been prepared

Status of technology for nuclear waste management

International Nuclear Information System (INIS)

Lieberman, J.A.

1984-01-01

In the area of low- and intermediate-level radioactive wastes the successful development and application of specific management technologies have been demonstrated over the years. The major area in which technology remains to be effectively implemented is in the management of high-level wastes from the nuclear fuel cycle. Research and development specifically directed at the management of high-level radioactive wastes in the USA and other countries is briefly reviewed in the article introduced

Hot isostatic press waste option study report

International Nuclear Information System (INIS)

Russell, N.E.; Taylor, D.D.

1998-02-01

A Settlement Agreement between the Department of Energy and the State of Idaho mandates that all high-level radioactive waste now stored at the Idaho Chemical Processing Plant be treated so that it is ready to move out of Idaho for disposal by the target date of 2035. This study investigates the immobilization of all Idaho Chemical Processing Plant calcine, including calcined sodium bearing waste, via the process known as hot isostatic press, which produces compact solid waste forms by means of high temperature and pressure (1,050 C and 20,000 psi), as the treatment method for complying with the settlement agreement. The final waste product would be contained in stainless-steel canisters, the same type used at the Savannah River Site for vitrified waste, and stored at the Idaho National Engineering and Environmental Laboratory until a national geological repository becomes available for its disposal. The waste processing period is from 2013 through 2032, and disposal at the High Level Waste repository will probably begin sometime after 2065

Hot isostatic press waste option study report

Energy Technology Data Exchange (ETDEWEB)

Russell, N.E.; Taylor, D.D.

1998-02-01

A Settlement Agreement between the Department of Energy and the State of Idaho mandates that all high-level radioactive waste now stored at the Idaho Chemical Processing Plant be treated so that it is ready to move out of Idaho for disposal by the target date of 2035. This study investigates the immobilization of all Idaho Chemical Processing Plant calcine, including calcined sodium bearing waste, via the process known as hot isostatic press, which produces compact solid waste forms by means of high temperature and pressure (1,050 C and 20,000 psi), as the treatment method for complying with the settlement agreement. The final waste product would be contained in stainless-steel canisters, the same type used at the Savannah River Site for vitrified waste, and stored at the Idaho National Engineering and Environmental Laboratory until a national geological repository becomes available for its disposal. The waste processing period is from 2013 through 2032, and disposal at the High Level Waste repository will probably begin sometime after 2065.

Assessment of soda ash calcination treatment of Turkish trona ore

Directory of Open Access Journals (Sweden)

Gezer Sibel

2016-01-01

Full Text Available Trona is relatively rare, non-metallic mineral, Na2CO3 · NaHCO3 · 2H2O. The pure material contains 70.3% sodium carbonate and by calcination the excess CO2 and water can be driven off, yielding natural soda ash. The terms soda ash and sodium carbonate are used interchangeably. Trona calcining is a key process step in production of soda ash (sodium carbonate anhydrate from the relatively cheap trona ore. The calcination reaction may proceeds in a sequence of steps. Depending on the conditions, it may result in formation of either sodium carbonate monohydrate (Na2CO3 · H2O, sodium sesquicarbonate or weigschederite (Na2CO3 · 3NaHCO3. The Beypazarı Turkish trona deposit is the second largest deposit in the world with the content of 84% trona. The decomposition of trona appeared to be a single stage process across the temperature range studied (150-200 °C with the representative samples of different size fractions in the draught up metallurgical furnace. The optimum particle size and calcination time were −6.35 mm and 30 minutes, respectively, at calcination temperature of 175 °C in a metallurgical furnace. Microwave-induced dry calcination of trona was possible and 5 minutes of calcination time at a power level of 900 was sufficient for complete calcination of −6.35 mm feed. This includes short time calcinations with the goal of improving economics and simplifying the thermal process.

Technology applications for radioactive waste minimization

International Nuclear Information System (INIS)

Devgun, J.S.

1994-01-01

The nuclear power industry has achieved one of the most successful examples of waste minimization. The annual volume of low-level radioactive waste shipped for disposal per reactor has decreased to approximately one-fifth the volume about a decade ago. In addition, the curie content of the total waste shipped for disposal has decreased. This paper will discuss the regulatory drivers and economic factors for waste minimization and describe the application of technologies for achieving waste minimization for low-level radioactive waste with examples from the nuclear power industry

New Technological Options to Manage High Level Waste

International Nuclear Information System (INIS)

Gonzalez Romero, E. M.

2007-01-01

Nuclear energy renaissance and its expansion in time and space has renewed the need for minimization technologies applicable to nuclear wastes. The minimization technologies include new power reactor concepts, Generation IV, and dedicated technologies like Partitioning and Transmutation of the actinides contained in the spent fuel. These technologies apply the principle of classification and recycling to the spent fuel to transform what at present is an environmental hazard into an energy source. the waste minimization technologies are also relevant for countries planning the reduction or phase-out of nuclear energy, as they will allow minimizing the size and number of the final waste repositories. Present estimations indicate that reductions as large as a factor 100 in the amount (radiotoxicity) of long lived nuclear waste are feasibly, with a modest increase on the final electricity cost. (Author)

Radioactive waste treatment technology at Czech nuclear power plants

International Nuclear Information System (INIS)

Kulovany, J.

2001-01-01

This presentation describes the main technologies for the treatment and conditioning of radioactive wastes at Czech nuclear power plants. The main technologies are bituminisation for liquid radioactive wastes and supercompaction for solid radioactive wastes. (author)

Feasibility Study for Vitrification of Sodium-Bearing Waste

International Nuclear Information System (INIS)

Quigley, J.J.; Raivo, B.D.; Bates, S.O.; Berry, S.M.; Nishioka, D.N.; Bunnell, P.J.

2000-01-01

Treatment of sodium-bearing waste (SBW) at the Idaho Nuclear Technology and Engineering Center (INTEC) within the Idaho National Engineering and Environmental Laboratory is mandated under a Settlement Agreement between the Department of Energy and the State of Idaho. One of the requirements of the Settlement Agreement is the complete calcination (i.e., treatment) of all SBW by December 31, 2012. One of the proposed options for treatment of SBW is vitrification. This study will examine the viability of SBW vitrification. This study describes the process and facilities to treat the SBW, from beginning waste input from INTEC Tank Farm to the final waste forms. Schedules and cost estimates for construction and operation of a Vitrification Facility are included. The study includes a facility layout with drawings, process description and flow diagrams, and preliminary equipment requirements and layouts

Feasibility Study for Vitrification of Sodium-Bearing Waste

Energy Technology Data Exchange (ETDEWEB)

J. J. Quigley; B. D. Raivo; S. O. Bates; S. M. Berry; D. N. Nishioka; P. J. Bunnell

2000-09-01

Treatment of sodium-bearing waste (SBW) at the Idaho Nuclear Technology and Engineering Center (INTEC) within the Idaho National Engineering and Environmental Laboratory is mandated under a Settlement Agreement between the Department of Energy and the State of Idaho. One of the requirements of the Settlement Agreement is the complete calcination (i.e., treatment) of all SBW by December 31, 2012. One of the proposed options for treatment of SBW is vitrification. This study will examine the viability of SBW vitrification. This study describes the process and facilities to treat the SBW, from beginning waste input from INTEC Tank Farm to the final waste forms. Schedules and cost estimates for construction and operation of a Vitrification Facility are included. The study includes a facility layout with drawings, process description and flow diagrams, and preliminary equipment requirements and layouts.

Influence of calcined mud on the mechanical properties and shrinkage of self-compacting concrete

Directory of Open Access Journals (Sweden)

Fatima Taieb

2018-01-01

Full Text Available The use of SCC has a particular interest in terms of sustainable development. Indeed, their specific formulation leads to a greater volume of dough than for common concretes, thus, a larger quantity of cement. However, for economical, ecological and technical reasons, it is sought to limit their cement content [1]. It is therefore necessary to almost always use mineral additions as a partial replacement for cement because the technology of self-compacting concretes can consume large quantities of fines, in this case calcinated mud issued from dams dredging sediments that can give and/or ameliorate characteristics and performances of this type of concretes. Four SCCs had been formulated from the same composition where the only percentage of calcinated mud of Chorfa (west of Algeria dam changed (0%, 10%, 20% and 30%. The effect of calcinated mud on characteristics at fresh state of SCC according to AFGC was quantified. Mechanical strengths and shrinkage deformation (total, autogenous, drying were evaluated. The results show the possibility to make SCCs with different dosages of calcinated mud having strengths that can defy those of the control SCC. The analysis of free deformations indicates the beneficial impact of the mud by contributing to decrease the amplitudes of the shrinkage compared to those of the control SCC.

Remote ignitability analysis of high-level radioactive waste

International Nuclear Information System (INIS)

Lundholm, C.W.; Morgan, J.M.; Shurtliff, R.M.; Trejo, L.E.

1992-09-01

The Idaho Chemical Processing Plant (ICPP), was used to reprocess nuclear fuel from government owned reactors to recover the unused uranium-235. These processes generated highly radioactive liquid wastes which are stored in large underground tanks prior to being calcined into a granular solid. The Resource Conservation and Recovery Act (RCRA) and state/federal clean air statutes require waste characterization of these high level radioactive wastes for regulatory permitting and waste treatment purposes. The determination of the characteristic of ignitability is part of the required analyses prior to calcination and waste treatment. To perform this analysis in a radiologically safe manner, a remoted instrument was needed. The remote ignitability Method and Instrument will meet the 60 deg. C. requirement as prescribed for the ignitability in method 1020 of SW-846. The method for remote use will be equivalent to method 1020 of SW-846

TMI-2: Unique waste management technology

International Nuclear Information System (INIS)

Bixby, W.W.; Young, W.R.; Grant, P.J.

1987-01-01

The 1979 accident at TMI-2 severely damaged the reactor core and contaminated more than a million gallons of water. Subsequent activities created another million gallons of water. The damaged reactor core represented a new waste form and cleanup of the contaminated water and system components created other new waste forms requiring creative approaches to waste management. This paper focuses on technologies that were developed specific to fuel waste management, core debris shipping, processing accident generated water, and disposal of the resultant waste forms

Hanford Waste Vitrification Plant applied technology plan

International Nuclear Information System (INIS)

Kruger, O.L.

1990-09-01

This Applied Technology Plan describes the process development, verification testing, equipment adaptation, and waste form qualification technical issues and plans for resolution to support the design, permitting, and operation of the Hanford Waste Vitrification Plant. The scope of this Plan includes work to be performed by the research and development contractor, Pacific Northwest Laboratory, other organizations within Westinghouse Hanford Company, universities and companies with glass technology expertise, and other US Department of Energy sites. All work described in this Plan is funded by the Hanford Waste Vitrification Plant Project and the relationship of this Plan to other waste management documents and issues is provided for background information. Work to performed under this Plan is divided into major areas that establish a reference process, develop an acceptable glass composition envelope, and demonstrate feed processing and glass production for the range of Hanford Waste Vitrification Plant feeds. Included in this work is the evaluation and verification testing of equipment and technology obtained from the Defense Waste Processing Facility, the West Valley Demonstration Project, foreign countries, and the Hanford Site. Development and verification of product and process models and other data needed for waste form qualification documentation are also included in this Plan. 21 refs., 4 figs., 33 tabs

Modeling of calcination of single kaolinitic clay particle

DEFF Research Database (Denmark)

Gebremariam, Abraham Teklay; Yin, Chungen; Rosendahl, Lasse

The present work aims at modeling of the calcination (dehydroxylation) process of clay particles, specifically kaolinite, and its thermal transformation. For such purpose, 1D single particle calcination model was developed based on the concept of shrinking core model to assess the dehydroxylation...

Biodiesel production via the transesterification of soybean oil using waste starfish (Asterina pectinifera).

Science.gov (United States)

Jo, Yong Beom; Park, Sung Hoon; Jeon, Jong-Ki; Ko, Chang Hyun; Ryu, Changkook; Park, Young-Kwon

2013-07-01

Calcined waste starfish was used as a base catalyst for the production of biodiesel from soybean oil for the first time. A batch reactor was used for the transesterification reaction. The thermal characteristics and crystal structures of the waste starfish were investigated by thermo-gravimetric analysis and X-ray diffraction. The biodiesel yield was determined by measuring the content of fatty acid methyl esters (FAME). The calcination temperature appeared to be a very important parameter affecting the catalytic activity. The starfish-derived catalyst calcined at 750 °C or higher exhibited high activity for the transesterification reaction. The FAME content increased with increasing catalyst dose and methanol-over-oil ratio.

Calcination of kaolinite clay particles for cement production

DEFF Research Database (Denmark)

Gebremariam, Abraham Teklay; Yin, Chungen; Rosendahl, Lasse

2014-01-01

Kaolinite rich clay particles calcined under certain conditions can attain favorable pozzolanic properties and can be used to substitute part of the CO2 intensive clinker in cement production. To better guide calcination of a clay material, a transient one-dimensional single particle model...

Coating of waste containing ceramic granules

International Nuclear Information System (INIS)

Neumann, W.; Kofler, O.

1979-01-01

Simulated high-level waste granules produced by fluidized-bed calcination were overcoated by chemical vapor deposition (CVD) with pyrocarbon and nickel in laboratory-scale experiments. Successful development enables pyrocrbon deposition at temperatures of 600 to 800 0 K. The coated granules have excellent properties for long-term waste storage

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Current high-level waste solidification technology

International Nuclear Information System (INIS)

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

1976-01-01

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

Effects of rapid calcination on properties of calcium-based sorbents

International Nuclear Information System (INIS)

Yan, Chang-Feng; Grace, John R.; Lim, C. Jim

2010-01-01

The calcination process may influence subsequent fragmentation, sintering and swelling when CaO derived from limestone acts as a CO 2 or SO 2 -sorbent in combustion, gasification and reforming. Sorbent properties are affected by CO 2 partial pressure, total pressure, temperature, heating rate, impurities and sample size. In this study, the effect of calcination heating rate was investigated based on an electrically heated platinum foil. The effects of heating rate (up to 800 C/s), calcination temperature (700-950 C), particle size (90-180 μm) and sweep gas velocity were investigated. Higher initial heating rates led to lower extents of limestone calcination, but the extents of carbonation of the resulting CaO were similar to each other. Calcium utilization declined markedly during carbonation or sulphation of CaO after calcination by rapid heating. Experimental results show that carbonation and calcium utilization were most effective for carbonation temperatures between 503 and 607 C. Increasing the extent of calcination is not the best way to improve overall calcium utilization due to the vast increase in energy consumption. (author)

Study on technology for radioactive waste treatment and management from uranium production

International Nuclear Information System (INIS)

Vu Hung Trieu; Vu Thanh Quang; Nguyen Duc Thanh; Trinh Giang Huong; Tran Van Hoa; Hoang Minh Chau; Ngo Van Tuyen; Nguyen Hoang Lan; Vuong Huu Anh

2007-01-01

There is some solid and liquid radioactive waste created during producing Uranium that needs being treated and managed to keep our environment safe. This radioactive waste contains Uranium (U-238), Thorium (Th-232), Radium (Ra-226) and some heavy metals and mainly is low radioactive waste. Our project has researched and built up appropriate technology for treating and managing the radioactive waste. After researching and experimenting, we have built up four technology processes as follows: Technology for separating Radium from liquid waste; Technology for treating and managing solid waste containing Ra; Technology for separating Thorium from liquid waste after recovering radium; Technology for stabilizing solid waste from Uranium production. (author)

Pilot plant experience on high-level waste solidification and design of the engineering prototype VERA

Energy Technology Data Exchange (ETDEWEB)

Guber, W; Diefenbacher, W; Hild, W; Krause, H; Schneider, E; Schubert, G

1972-11-01

In the present paper the solidification process for highly active waste solutions as developed in the Karlsruhe Nuclear Research Center is presented. Its principal steps are: denitration, calcination in a spray calciner operated with superheated steam, melting of the calcine with appropriate additives to borosilicate glass in an induction-heated melting furnace. The operational experiences gained so far in the inactive 1:1 pilot plant are reported. Furthermore, a description is given of the projected multi-purpose experimental facility VERA 2 which is provided for processing the highly active waste solutions from the first German reprocessing plant WAK.

CSER 99-001: PFP LAB Dentirating calciner

International Nuclear Information System (INIS)

MILLER, E.M.; DOBBIN, K.D.

1999-01-01

A criticality safety evaluation report was prepared for the Plutonium Finishing Plant (PFP) laboratory denigrating calciner, located in Glovebox 188-1, that converts Pu(NO 3 ) 4 solutions to the high fired stable oxide PuO 2 . Fissile mass limits and volume limits are set for the glovebox for testing operations and training operators using only nitric acid feed to a plutonium oxide bed in the calciner

Treatment of waste

International Nuclear Information System (INIS)

1981-01-01

A method of treating radioactive waste to substantially reduce the volume and which is especially useful in the treatment of material which includes radioactive halogens such as 131 I, is described. A fluidised bed incinerator and calciner are used to reduce all the liquid and combustible solid waste to anhydrous granular solids, all of which is carried by fluidizing gases into an off-gas system designed for their collection. (U.K.)

Development of thermal conditioning technology for alpha-contaminated wastes

International Nuclear Information System (INIS)

Kim, Joon Hyung; Kim, H. Y.; Kim, J. G.

2001-04-01

To develop a thermal conditioning technology for alpha-contaminated wastes, which are presumed to generate from pyrochemical processing of spent fuel, research on the three different fields have been performed; incineration, off-gas treatment, and vitrification/cementation technology. Through the assessment on the amount of alpha-contaminated waste and incineration characterises, an oxygen-enriched incineration process, which can greatly reduce the off-gas volume, was developed by our own technology. Trial burn test with paper waste resulted in a reduction of off-gas volume by 3.5. A study on the behavior and adsorption of nuclides/heavy metals at high-temperature was performed to develop an efficient removal technology. Off-gas treatment technologies for radioiodine at high-temperature and 14 CO 2 , acidic gases, and radioactive gaseous wastes such as Xe/Kr at room temperature were established. As a part of development of high-level waste solidification technology, manufacture of high-frequency induction melter, fabrication and characterization of base-glass media fabricated with spent HEPA filter medium, and development of titanate ceramic material as a precursor of SYNROC by a self-combustion method were performed. To develop alpha-contaminated waste solidification technology, a process to convert periodontal in the cement matrix to calcite with SuperCritical Carbon Dioxide (SCCD) was manufactured. The SCCD treatment enhanced the physicochemical properties of cement matrices, which increase the long-term integrity of cement waste forms during transportation and storage

Mixed and low-level waste treatment facility project. Volume 3, Waste treatment technologies (Draft)

Energy Technology Data Exchange (ETDEWEB)

1992-04-01

The technology information provided in this report is only the first step toward the identification and selection of process systems that may be recommended for a proposed mixed and low-level waste treatment facility. More specific information on each technology will be required to conduct the system and equipment tradeoff studies that will follow these preengineering studies. For example, capacity, maintainability, reliability, cost, applicability to specific waste streams, and technology availability must be further defined. This report does not currently contain all needed information; however, all major technologies considered to be potentially applicable to the treatment of mixed and low-level waste are identified and described herein. Future reports will seek to improve the depth of information on technologies.

Embedding methods of solidified waste in metal matrices

International Nuclear Information System (INIS)

Neumann, W.

1979-01-01

The embedding of simulated waste calcines by three different methods (vacuum-pressure casting, centrifugal casting, and metal stirred with the calcines) was investigated. The experimental performance is described and advantages and disadvantages noted. The feasibility of embedding fines by stirring in metal was shown. In addition, an estimation of the influence of porosity on the properties of composites was carried out

Waste processing building with incineration technology

Science.gov (United States)

Wasilah, Wasilah; Zaldi Suradin, Muh.

2017-12-01

In Indonesia, waste problem is one of major problem of the society in the city as part of their life dynamics. Based on Regional Medium Term Development Plan of South Sulawesi Province in 2013-2018, total volume and waste production from Makassar City, Maros, Gowa, and Takalar Regency estimates the garbage dump level 9,076.949 m3/person/day. Additionally, aim of this design is to present a recommendation on waste processing facility design that would accommodate waste processing process activity by incineration technology and supported by supporting activity such as place of education and research on waste, and the administration activity on waste processing facility. Implementation of incineration technology would reduce waste volume up to 90% followed by relative negative impact possibility. The result planning is in form of landscape layout that inspired from the observation analysis of satellite image line pattern of planning site and then created as a building site pattern. Consideration of building orientation conducted by wind analysis process and sun path by auto desk project Vasari software. The footprint designed by separate circulation system between waste management facility interest and the social visiting activity in order to minimize the croos and thus bring convenient to the building user. Building mass designed by inseparable connection series system, from the main building that located in the Northward, then connected to a centre visitor area lengthways, and walked to the waste processing area into the residue area in the Southward area.

Properties of Formula 127 glass prepared with radioactive zirconia calcine

International Nuclear Information System (INIS)

Staples, B.A.; Pavlica, D.A.; Cole, H.S.

1982-09-01

Formula 127 glass has been developed to immobilize ICPP zirconia calcine. This glass has been prepared remotely on a laboratory scale basis with actual radioactive zirconia calcine retrieved after ten years of storage from Bin Set 2. The aqueous leachability of the glass produced was investigated and compared through application of the MCC-1, MCC-2 and Soxhlet leach tests with that of Formula 127 glass prepared with simulated calcine. The solid state properties of the glasses prepared with actual and simulated calcines were also measured by electron spectroscopy for chemical analysis (ESCA) and scanning electron microscopy energy dispersive x-ray (SEM-EDX). Based on the application of these leaching tests and analysis techniques the properties measured in this study are similar for 127 glass prepared with either simulated or radioactive calcine. 13 figures, 16 tables

Innovative technologies for managing oil field waste

International Nuclear Information System (INIS)

Veil, J.A.

2003-01-01

Each year, the oil industry generates millions of barrels of wastes that need to be properly managed. For many years, most oil field wastes were disposed of at a significant cost. However, over the past decade, the industry has developed many processes and technologies to minimize the generation of wastes and to more safely and economically dispose of the waste that is generated. Many companies follow a three-tiered waste management approach. First, companies try to minimize waste generation when possible. Next, they try to find ways to reuse or recycle the wastes that are generated. Finally, the wastes that cannot be reused or recycled must be disposed of. Argonne National Laboratory (Argonne) has evaluated the feasibility of various oil field waste management technologies for the U.S. Department of Energy. This paper describes four of the technologies Argonne has reviewed. In the area of waste minimization, the industry has developed synthetic-based drilling muds (SBMs) that have the desired drilling properties of oil-based muds without the accompanying adverse environmental impacts. Use of SBMs avoids significant air pollution from work boats hauling offshore cuttings to shore for disposal and provides more efficient drilling than can be achieved with water-based muds. Downhole oil/water separators have been developed to separate produced water from oil at the bottom of wells. The produced water is directly injected to an underground formation without ever being lifted to the surface, thereby avoiding potential for groundwater or soil contamination. In the area of reuse/recycle, Argonne has worked with Southeastern Louisiana University and industry to develop a process to use treated drill cuttings to restore wetlands in coastal Louisiana. Finally, in an example of treatment and disposal, Argonne has conducted a series of four baseline studies to characterize the use of salt caverns for safe and economic disposal of oil field wastes.

MIxed Waste Integrated Program (MWIP): Technology summary

International Nuclear Information System (INIS)

1994-02-01

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

Implementation of industrial waste ferrochrome slag in conventional and low cement castables: Effect of calcined alumina

Directory of Open Access Journals (Sweden)

Pattem Hemanth Kumar

2014-12-01

Full Text Available A new class of conventional and low-cement ferrochrome slag-based castables were prepared from 40 wt.% ferrochrome slag and 45 wt.% calcined bauxite. Rest fraction varied between high alumina cement (HAC acting as hydraulic binder and calcined alumina as pore filling additive. Standard ASTM size briquettes were prepared for crushing and bending strengths evaluation, and the samples were then subjected to firing at 800, 1100 and 1300 °C for a soaking period of 3 h. The microstructure and refractory properties of the prepared castables have been investigated using X-ray diffraction (XRD, scanning electron microscopy (SEM, cold crushing strength, modulus of rupture and permanent linear changes (PLCs test. Castables show good volume stability (linear change <0.7% at 1300 °C. The outcomes of these investigations were efficacious and in accordance with previously reported data of similar compositions. High thermo-mechanical and physico-chemical properties were attained pointing out an outstanding potential to increase the refractory lining working life of non-recovery coke oven and reheating furnaces.

RFID technology for hazardous waste management and tracking.

Science.gov (United States)

Namen, Anderson Amendoeira; Brasil, Felipe da Costa; Abrunhosa, Jorge José Gouveia; Abrunhosa, Glaucia Gomes Silva; Tarré, Ricardo Martinez; Marques, Flávio José Garcia

2014-09-01

The illegal dumping of hazardous waste is one of the most concerning occurrences related to illegal waste activities. The waste management process is quite vulnerable, especially when it comes to assuring the right destination for the delivery of the hazardous waste. The purpose of this paper is to present a new system design and prototype for applying the RFID technology so as to guarantee the correct destination for the hazardous waste delivery. The aim of this innovative approach, compared with other studies that employ the same technology to the waste disposal process, is to focus on the certification that the hazardous waste will be delivered to the right destination site and that no inappropriate disposal will occur in the transportation stage. These studies were carried out based on data collected during visits to two hazardous waste producer companies in Brazil, where the material transportation and delivery to a company in charge of the waste disposal were closely monitored. © The Author(s) 2014.

Chemistry and technology of radioactive waste management - the IAEA perspective

International Nuclear Information System (INIS)

Efremenkov, V.M.; )

2003-01-01

The paper refers the consideration of chemical composition of radioactive waste in selection of particular method and technology for waste treatment and conditioning, importance of physico-chemical parameters of waste processing techniques for optimisation of waste processing to produce waste form of appropriate quality. Consideration of waste chemistry is illustrated by several IAEA activities on radioactive waste management and by outlining the scope of some selected technical reports on different waste management subjects. Different components of the IAEA activities on radioactive waste management and on technology transfer are presented and discussed. (author)

Technology transfer and the management of radioactive waste

International Nuclear Information System (INIS)

Bonne, A.; Chan-Sands, C.

1998-01-01

One of the IAEA's fundamental roles is to act as a centre for the transfer of nuclear technologies, including those for managing radioactive wastes. In the area of waste management technology, the Agency is actively working to improve and develop new and efficient means to fulfill that responsibility. Recognizing its responsibilities and challenges, IAEA efforts related to radioactive waste management technologies into the next century are framed around three major areas: the development and implementation of mechanisms for better technology transfer and information exchange; the promotion of sustainable and safer processes and procedures; and the provision of peer reviews and direct technical assistance that help facilitate bilateral and multinational efforts. To illustrate some specific elements of the overall programme, this article reviews selected technology-transfer activities that have been initiated in the field

Waste disposal technologies: designs and evaluations

International Nuclear Information System (INIS)

Shaw, R.A.

1987-01-01

Many states and compacts are presently in the throes of considering what technology to select for their low level waste disposal site. Both the technical and economic aspects of disposal technology are important considerations in these decisions. It is also important that they be considered in the context of the entire system. In the case of a nuclear power plant, that system encompasses the various individual waste streams that contain radioactivity, the processing equipment which reduces the volume and/or alters the form in which the radioisotopes are contained, the packaging of the processed wastes in shipment, and finally its disposal. One further part of this is the monitoring that takes place in all stages of this operation. This paper discusses the results of some research that has been sponsored by EPRI with the principal contractor being Rogers and Associates Engineering Corporation. Included is a description of the distinguishing features found in disposal technologies developed in a generic framework, designs for a selected set of these disposal technologies and the costs which have been derived from these designs. In addition, a description of the early efforts towards defining the performance of these various disposal technologies is described. 5 figures, 1 table

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-03-01

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

A perspective of hazardous waste and mixed waste treatment technology at the Savannah River Site

International Nuclear Information System (INIS)

England, J.L.; Venkatesh, S.; Bailey, L.L.; Langton, C.A.; Hay, M.S.; Stevens, C.B.; Carroll, S.J.

1991-01-01

Treatment technologies for the preparation and treatment of heavy metal mixed wastes, contaminated soils, and mixed mercury wastes are being considered at the Savannah River Site (SRS), a DOE nuclear material processing facility operated by Westinghouse Savannah River Company (WSRC). The proposed treatment technologies to be included at the Hazardous Waste/Mixed Waste Treatment Building at SRS are based on the regulatory requirements, projected waste volumes, existing technology, cost effectiveness, and project schedule. Waste sorting and size reduction are the initial step in the treatment process. After sorting/size reduction the wastes would go to the next applicable treatment module. For solid heavy metal mixed wastes the proposed treatment is macroencapsulation using a thermoplastic polymer. This process reduces the leachability of hazardous constituents from the waste and allows easy verification of the coating integrity. Stabilization and solidification in a cement matrix will treat a wide variety of wastes (i.e. soils, decontamination water). Some pretreatments may be required (i.e. Ph adjustment) before stabilization. Other pretreatments such as soil washing can reduce the amount of waste to be stabilized. Radioactive contaminated mercury waste at the SRS comes in numerous forms (i.e. process equipment, soils, and lab waste) with the required treatment of high mercury wastes being roasting/retorting and recovery. Any unrecyclable radioactive contaminated elemental mercury would be amalgamated, utilizing a batch system, before disposal

Mixed waste focus area alternative technologies workshop

International Nuclear Information System (INIS)

Borduin, L.C.; Palmer, B.A.; Pendergrass, J.A.

1995-01-01

This report documents the Mixed Waste Focus Area (MWFA)-sponsored Alternative Technology Workshop held in Salt Lake City, Utah, from January 24--27, 1995. The primary workshop goal was identifying potential applications for emerging technologies within the Options Analysis Team (OAT) ''wise'' configuration. Consistent with the scope of the OAT analysis, the review was limited to the Mixed Low-Level Waste (MLLW) fraction of DOE's mixed waste inventory. The Los Alamos team prepared workshop materials (databases and compilations) to be used as bases for participant review and recommendations. These materials derived from the Mixed Waste Inventory Report (MWIR) data base (May 1994), the Draft Site Treatment Plan (DSTP) data base, and the OAT treatment facility configuration of December 7, 1994. In reviewing workshop results, the reader should note several caveats regarding data limitations. Link-up of the MWIR and DSTP data bases, while representing the most comprehensive array of mixed waste information available at the time of the workshop, requires additional data to completely characterize all waste streams. A number of changes in waste identification (new and redefined streams) occurred during the interval from compilation of the data base to compilation of the DSTP data base with the end result that precise identification of radiological and contaminant characteristics was not possible for these streams. To a degree, these shortcomings compromise the workshop results; however, the preponderance of waste data was linked adequately, and therefore, these analyses should provide useful insight into potential applications of alternative technologies to DOE MLLW treatment facilities

US Department of Energy interim mixed waste inventory report: Waste streams, treatment capacities and technologies

International Nuclear Information System (INIS)

1993-04-01

The United States Department of Energy (DOE) has prepared this report to provide an inventory of its mixed wastes and treatment capacities and technologies in response to section 3021(a) of the Resource Conservation and Recovery Act (RCRA), as amended by section 105(a) of the Federal Facility Compliance Act (FFCA) of 1992 (Pub. L. No. 102-386). DOE has prepared this report for submission to EPA and the States in which DOE stores, generates, or treats mixed wastes. As required by the FFCA, this report contains: a national inventory of all mixed wastes in the DOE system that are currently stored or will be generated over the next five years, including waste stream name, description, EPA waste codes, basis for characterization (i.e., sampling and analysis or process knowledge), effect of radionuclides on treatment, quantity stored that is subject to the Land Disposal Restrictions (LDRs) storage prohibition, quantity stored that is not subject to the LDRS, expected generation over the next five years, Best Demonstrated Available Technology (BDAT) used for developing the LDR requirements, and waste minimization activities; and a national inventory of mixed waste treatment capacities and technologies, including information such as the descriptions, capacities, and locations of all existing and proposed treatment facilities, explanations for not including certain existing facilities in capacity evaluations, information to support decisions on unavailability of treatment technologies for certain mixed wastes, and the planned technology development activities

Stabilization of As-, Pb-, and Cu-contaminated soil using calcined oyster shells and steel slag.

Science.gov (United States)

Moon, Deok Hyun; Wazne, Mahmoud; Cheong, Kyung Hoon; Chang, Yoon-Young; Baek, Kitae; Ok, Yong Sik; Park, Jeong-Hun

2015-07-01

In this study, As-, Pb-, and Cu-contaminated soil was stabilized using calcined oyster shells (COS) and steel slag (SS). The As-contaminated soil was obtained from a timber mill site where chromate copper arsenate (CCA) was used as a preservative. On the other hand, Pb- and Cu-contaminated soil was obtained from a firing range. These two soils were thoroughly mixed to represent As-, Pb-, and Cu-contaminated soil. Calcined oyster shells were obtained by treating waste oyster shells at a high temperature using the calcination process. The effectiveness of stabilization was evaluated by 1-N HCl extraction for As and 0.1-N HCl extraction for Pb and Cu. The treatment results showed that As, Pb, and Cu leachability were significantly reduced upon the combination treatment of COS and SS. The sole treatment of SS (10 wt%) did not show effective stabilization. However, the combination treatment of COS and SS showed a significant reduction in As, Pb, and Cu leachability. The best stabilization results were obtained from the combination treatment of 15 wt% COS and 10 wt% SS. The SEM-EDX results suggested that the effective stabilization of As was most probably achieved by the formation of Ca-As and Fe-As precipitates. In the case of Pb and Cu, stabilization was most probably associated with the formation of pozzolanic reaction products such as CSHs and CAHs.

Social Technology Apply to National Policy on Solid Waste: Solid Waste Management Integrated in the Countryside

Directory of Open Access Journals (Sweden)

Greice Kelly Lourenco Porfirio de Oliveira

2016-06-01

Full Text Available This article aims to study the environmentally friendly social technologies through appropriate techniques to the treatment of solid waste disposed of improperly. After exposure of concepts, a reflection on the use of social technologies as a mechanism for realization of integrated management objectives of waste set by the National Solid Waste Policy will be made – 12.305/10 . Finally, data from the Social Technologies Bank of Brazil Foundation will be displayed showing the results of the use of technology to promote the integrated management of solid waste in rural communities Crateús/CE , through a provision aimed at PNRS, selective collection

WASTE-FREE PRODUCTION TECHNOLOGY OF DRY MASHED POTATOES

Directory of Open Access Journals (Sweden)

G. V. Kalashnikov

2015-01-01

Full Text Available Summary. According to data on norms of consumption of vegetable production of scientific research institute of Food of the Russian Academy of Medical Science, potatoes win first place with norm of 120 kg a year on the person. In this regard much attention is paid to processing of potatoes that allows to prolong the term of its validity, and also to reduce the capacity of storages and to reduce transport transportations as 1 kg of a dry potatoes produсt is equivalent 7-8 kg of fresh potatoes. Thus industrial processing of potatoes on dry mashed potatoes allows to reduce losses of potatoes at storage and transportation, there is a possibility of enrichment of products vitamins and other useful components, its nutrition value remains better, conditions for complex processing of raw materials with full recycling and creations of stocks of products from potatoes on a crop failure case are created. Dry mashed potatoes are a product of long storage. On the basis of studying of the production technology of mashed potatoes the analysis of technological processes as sources of creation of waste, and the directions of recovery of secondary raw materials for complex waste-free technology of processing of potatoes are defined is provided. The waste-free technological scheme of processing of potatoes and production of dry instant mashed potatoes on the basis of dehydration and moisture thermal treatment a component providing recovery of secondary carbohydrate content raw materials in the form of waste of the main production is developed. The main stages of production of dry instant mashed potatoes are described. It is offered the technological scheme of a production line of mashed potatoes on the basis of waste-free technology. Advantages of the offered waste-free production technology of dry instant mashed potatoes with processing of secondary starch-containing raw materials are given.

Process for solidifying high-level nuclear waste

Science.gov (United States)

Ross, Wayne A.

1978-01-01

The addition of a small amount of reducing agent to a mixture of a high-level radioactive waste calcine and glass frit before the mixture is melted will produce a more homogeneous glass which is leach-resistant and suitable for long-term storage of high-level radioactive waste products.

Application of thermal technologies for processing of radioactive waste

International Nuclear Information System (INIS)

2006-12-01

The primary objective of this publication is to provide an overview of the various thermal technologies for processing various solid, liquid, organic and inorganic radioactive waste streams. The advantages, limitations and operating experience of various thermal technologies are explained. This publication also goes beyond previous work on thermal processes by addressing the applicability of each technology to national or regional nuclear programmes of specific relative size (major advanced programmes, small to medium programmes, and emerging programmes with other nuclear applications). The most commonly used thermal processing technologies are reviewed, and the key factors influencing the selection of thermal technologies as part of a national waste management strategy are discussed. Accordingly, the structure and content of this publication is intended to assist decision-makers, regulators, and those charged with developing such strategies to identify and compare thermal technologies for possible inclusion in the mix of available, country-specific waste management processes. This publication can be used most effectively as an initial cutting tool to identify whether any given technology will best serve the local waste management strategy in terms of the waste generated, technical complexity, available economic resources, environmental impact considerations, and end product (output) of the technology. If multiple thermal technologies are being actively considered, this publication should be instrumental in comparing the technologies and assisting the user to reach an informed decision based on local needs, economics and priorities. A detailed set of conclusions is provided in Section 7

Alternative oxidation technologies for organic mixed waste

International Nuclear Information System (INIS)

Borduin, L.C.; Fewell, T.

1998-01-01

The Mixed Waste Focus Area (MWFA) is currently supporting the development and demonstration of several alternative oxidation technology (AOT) processes for treatment of combustible mixed low-level wastes. AOTs have been defined as technologies that destroy organic material without using open-flame reactions. AOTs include both thermal and nonthermal processes that oxidize organic wastes but operate under significantly different physical and chemical conditions than incinerators. Nonthermal processes currently being studied include Delphi DETOX and acid digestion at the Savannah River Site (SRS), and direct chemical oxidation at Lawrence Livermore National Laboratory (LLNL). All three technologies are at advanced stages of development or are entering the demonstration phase. Nonflame thermal processes include catalytic chemical oxidation, which is being developed and deployed at Lawrence Berkeley National Laboratory (LBNL), and steam reforming, a commercial process being supported by the Department of Energy (DOE). Although testing is complete on some AOT technologies, most require additional support to complete some or all of the identified development objectives. Brief descriptions, status, and planned paths forward for each of the technologies are presented

Calcinating petroleum coke in a furnace with a rotating hearth

Energy Technology Data Exchange (ETDEWEB)

Akhmetov, M M; Ezhov, B M; Galeeva, Z G; Goriunov, V S; Karpinskaia, N N; Zaitseva, S A

1980-01-01

A scheme is described for an industrial device with a bottom furnace for calcinating coke from slow coking. The consumption and operational indicators of the process during the calcination of standard and needle cokes are given, together with data on the quality of dry and calcinated cokes under different conditions. The basic drawbacks in the operation of the device are described, and measures are proposed for increasing its operational effectiveness.

Evaluation of Technologies for Retrieval of Waste from Leaking Tanks

International Nuclear Information System (INIS)

Bamberger, Judith A.; Hatchell, Brian K.; Lewis, Benjamin E.; Randolph, John D.; Killough, Stephen M.

2000-01-01

The US Department of Energy Environmental and Waste Management Tanks Focus Area selected as a strategic initiative the need to identify and develop technologies for remediation of tanks that are known or are suspected to leak. This investigation identified and evaluated technical options for single-shell tank waste retrieval applicable to retrieve waste from potentially leaking tanks. Technologies that minimize leakage use minimal water, and dry retrieval technologies were evaluated. Safety, cost, authorization basis, and schedule risks were identified for each technology to provide River Protection Program with information to evaluate technical and programmatic risk. A workshop was held to identify technology needs and solutions. These approaches grouped into five categories: those related to waste dislodging, waste conveyance, both waste dislodging and conveyance, the deployment platform, and technologies related to leak detection, monitoring, and mitigation. Based on the ranking, six technologies were selected as potential candidates for further evaluation. These items were prioritized into four technologies to recommend for further evaluation (1) Air assisted TORE(R). The TORE(R) produces a processing vortex core with the ability to convey solids at pre-determined slurry concentrations over great distances. The dry TORE(R) concept uses air to develop the vortex to fluidize dry solids. The TORE(R)the solids in a slurry transport line. (2) Sonication for waste dislodging utilizes ultrasonic energy to fracture and dislodge hard waste types such as salt cake and sludge. (3) Novel long-reach manipulators concept is to investigate novel cost effective approaches for long-reach manipulator technology. (4) Next generation crawler technology envisions a non-umbilical dislodger, possibly radio controlled and powered remotely to provide a deployment platform not affected by path, or the need to retrace steps

Model for acquiring innovative waste immobilization technologies

International Nuclear Information System (INIS)

Dole, L.R.; Singh, S.P.N.

1988-01-01

The US Department of Energy's (DOE's) Oak Ridge Operations (ORO) has established the Waste Management Technology Center (WMTC) at Oak Ridge National Laboratory to assist in meeting the environmental requirements for federal facilities as stated in the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). The WMTC will bring innovative mixed chemical and radioactive waste treatment and site closure technologies to bear on the many mixed chemical and radioactive waste problems at the DOE-ORO facilities located in Tennessee, Kentucky, and Ohio. The WMTC seeks innovative technologies through a phased procurement cycle that encourages the teaming of emerging technologies with experienced contractors in order to comply with on-site requirements of DOE orders concerning protection of the environment. This three-phase procurement cycle includes: (1) a feasibility study and implementation plan, (2) an on-site pilot demonstration, and (3) full-scale implementation. This paper describes the statements of work for some related demonstrations and remedial actions

Environmental assessment of solid waste systems and technologies: EASEWASTE

DEFF Research Database (Denmark)

Kirkeby, Janus Torsten; Birgisdottir, Harpa; Hansen, Trine Lund

2006-01-01

A new model has been developed for evaluating the overall resource consumption and environmental impacts of municipal solid waste management systems by the use of life cycle assessment. The model is named EASEWASTE (Environmental Assessment of Solid Waste Systems and Technologies) and is able...... may not always be the most environmentally friendly. The EASEWASTE model can identify the most environmentally sustainable solution, which may differ among waste materials and regions and can add valuable information about environmental achievements from each process in a solid waste management system....... to compare different waste management strategies, waste treatment methods and waste process technologies. The potential environmental impacts can be traced back to the most important processes and waste fractions that contribute to the relevant impacts. A model like EASEWASTE can be used by waste planners...

Sodium waste technology: A summary report

International Nuclear Information System (INIS)

Abrams, C.S.; Witbeck, L.C.

1987-01-01

The Sodium Waste Technology (SWT) Program was established to resolve long-standing issues regarding disposal of sodium-bearing waste and equipment. Comprehensive SWT research programs investigated a variety of approaches for either removing sodium from sodium-bearing items, or disposal of items containing sodium residuals. The most successful of these programs was the design, test, and the production operation of the Sodium Process Demonstration Facility at ANL-W. The technology used was a series of melt-drain-evaporate operations to remove nonradioactive sodium from sodium-bearing items and then converting the sodium to storable compounds

Technical Division quarterly progress report, October 1--December 31, 1977. [Fuel cycle research and development; special materials production

Energy Technology Data Exchange (ETDEWEB)

Slansky, C.M. (ed.)

1978-02-01

Results are presented on the fluidized-bed calcination of simulated radioactive waste from the reprocessing of spent commercial nuclear fuel and on the removal of actinide elements from the waste prior to calcination. Other programs include the development of storage technology for /sup 85/Kr waste; and the behavior of volatile radionuclides during the combustion of HTGR graphite-based fuel. The long-term management of defense waste from the ICPP covers post-calcination treatment of ICPP calcined waste; the removal of actinide elements from first-cycle raffinate; the retrieval and handling of calcined waste from ICPP storage vaults; and the preparation of the Programmatic Environmental Impact Statement on ICPP waste. Process improvements are reported on the Fluorinel headend process for Zircaloy-clad fuels and on uranium accountability measurements. Other development results cover the process for recovering spent Rover fuel, buried pipeline transfer systems, support to the Waste Management Program, Waste Calcining Facility support, New Waste Calcining Facility support, and effluent monitoring methods evaluation and development. In this category are studies on nuclear materials security, application of a liquid-solid fluidized-bed heat exchanger to the recovery of geothermal heat, inplant reactor source term measurements, burnup methods for fast breeder reactor fuels, absolute thermal fission yield measurements, analytical support to light water breeder reactor development, research on analytical methods, and the behavior of environmental species of iodine.

Future directions of defense programs high-level waste technology programs

International Nuclear Information System (INIS)

Chee, T.C.; Shupe, M.W.; Turner, D.A.; Campbell, M.H.

1987-01-01

The Department of Energy has been managing high-level waste from the production of nuclear materials for defense activities over the last forty years. An objective for the Defense Waste and Transportation Management program is to develop technology which ensures the safe, permanent disposal of all defense radioactive wastes. Technology programs are underway to address the long-term strategy for permanent disposal of high-level waste generated at each Department of Energy site. Technology is being developed for assessing the hazards, environmental impacts, and costs of each long-term disposal alternative for selection and implementation. This paper addresses key technology development areas, and consideration of recent regulatory requirements associated with the long-term management of defense radioactive high-level waste

DC plasma arc melter technology for waste vitrification

International Nuclear Information System (INIS)

Hamilton, R.A.; Wittle, J.K.; Trescot, J.

1995-01-01

This paper describes the features and benefits of a breakthrough DC Arc Melter for the permanent treatment of all types of solid wastes including nonhazardous, hazardous and radioactive. This DC Arc Furnace system, now commercially available, is the low cost permanent solution for solid waste pollution prevention and remediation. Concern over the effective disposal of wastes generated by the industrial society, worldwide, has prompted development of technologies to address the problem. For the most part these technologies have resulted in niche solutions with limited application. The only solution that has the ability to process almost all wastes, and to recover/recycle metallic and inorganic matter, is the group of technologies known as melters. Melters have distinct advantages over traditional technologies such as incineration because melters operate at higher temperatures, are relatively unaffected by changes in the waste stream, produce a vitrified stable product, and have the capability to recover/recycle slag, metals and gas. The system, DC Plasma Arc Melter, has the lowest capital, maintenance and operating cost of any melter technology because of its patented DC Plasma Arc with graphite electrode. DC Plasma Arc Melter systems are commercially available in sizes from 50 kg/batch or 250--3,000 kg/hr on a continuous feed basis. This paper examines the design and operating benefits of a DC Plasma Arc Melter System

Zero Waste and Conversion Efficiencies of Various Technologies for Disposal of Municipal Solid Waste

Institute of Scientific and Technical Information of China (English)

Zhang Wenyang

2005-01-01

Zero waste is a philosophy and a design principle of dealing with our waste stream for the 21st century. After reviewing the available information, the goal of zero waste from landfill is considered to be unachievable by using known and proven methods and approaches. The comparison of various technologies shows that the conversion efficiencies depend upon the type of system chosen for processing residual waste, and the best overall diversion rate of waste management system that can be achieved is about 71%. The maximum achievable overall diversion rate can be increased to approximate 92% if current environmental regulations to permit the routine use of the bottom ash or char for advanced thermal technologies.

Remote technologies for buried waste retrieval

International Nuclear Information System (INIS)

Smith, A.M.; Rice, P.

1995-01-01

The DOE is evaluating what should be done with this buried waste. Although the radioactive waste is not particularly mobile unless airborne, some of it was buried with volatile organics and/or other substances that tend to spread easily to surrounding soil or water tables. Volatile organics are hazardous materials (such as trichloroethylene) and require clean-up at certain levels in drinking water. There is concern that the buried volatile organics will spread into the water table and contaminate drinking water. Because of this, the DOE is considering options for handling this buried waste and reducing the risks of spreading or exposure. There are two primary options: containment and stabilization, or retrieval. Containment and stabilization systems would include systems that would leave the waste where it is, but contain and stabilize it so that the radioactive and hazardous materials would not spread to the surrounding soil, water, or air. For example, an in situ vitrification system could be used to melt the waste into a composite glass-like material that would not leach into the surrounding soil, water, or air. Retrieval systems are those that would remove the waste from its burial location for treatment and/or repackaging for long term storage. The objective of this project was to develop and demonstrate remote technologies that would minimize dust generation and the spread of airborne contaminants during buried waste retrieval. Remote technologies are essential for the retrieval of buried waste because they remove workers from the hazardous environment and provide greater automation, reducing the chances of human error. Minimizing dust generation is also essential to increased safety for the workers and the environment during buried waste retrieval. The main contaminants within the waste are micron-sized particles of plutonium and americium oxides, chlorides, and hydroxides, which are easily suspended in air and spread if disturbed

Impact of partitioning and transmutation on high-level waste disposal for the fast breeder reactor fuel cycle

International Nuclear Information System (INIS)

Nishihara, Kenji; Oigawa, Hiroyuki; Nakayama, Shinichi; Ono, Kiyoshi; Shiotani, Hiroki

2010-01-01

The impact of partitioning and/or transmutation (PT) technology on high-level waste management was investigated for the equilibrium state of several potential fast breeder reactor (FBR) fuel cycles. Three different fuel cycle scenarios involving PT technology were analyzed: 1) partitioning process only (separation of some fission products), 2) transmutation process only (separation and transmutation of minor actinides), and 3) both partitioning and transmutation processes. The conventional light water reactor (LWR) fuel cycle without PT technology, on which the current repository design is based, was also included for comparison. We focused on the thermal constraints in a geological repository and determined the necessary predisposal storage quantities and time periods (by defining a storage capacity index) for several predefined emplacement configurations through transient thermal analysis. The relation between this storage capacity index and the required repository emplacement area was obtained. We found that the introduction of the FBR fuel cycle without PT can yield a 35% smaller repository per unit electricity generation than the LWR fuel cycle, although the predisposal storage period is prolonged from 50 years for the LWR fuel cycle to 65 years for the FBR fuel cycle without PT. The introduction of the partitioning-only process does not result in a significant reduction of the repository emplacement area from that for the FBR fuel cycle without PT, but the introduction of the transmutation-only process can reduce the emplacement area by a factor of 5 when the storage period is extended from 65 to 95 years. When a coupled partitioning and transmutation system is introduced, the repository emplacement area can be reduced by up to two orders of magnitude by assuming a predisposal storage of 60 years for glass waste and 295 years for calcined waste containing the Sr and Cs fraction. The storage period of 295 years for the calcined waste does not require a large

Cost of organic waste technologies: A case study for New Jersey

Directory of Open Access Journals (Sweden)

Gal Hochman

2015-09-01

Full Text Available This paper evaluates the benefits of converting food waste and manure to biogas and/or fertilizer, while focusing on four available waste treatment technologies: direct combustion, landfilling, composting, and anaerobic digestion. These four alternative technologies were simulated using municipal-level data on food waste and manure in New Jersey. The criteria used to assess the four technologies include technological productivity, economic benefits, and impact on land scarcity. Anaerobic digestion with gas collection has the highest technological productivity; using anaerobic digesters would supply electricity to nearly ten thousand families in New Jersey. In terms of economic benefits, the landfill to gas method is the least costly method of treating waste. In comparison, direct combustion is by far the most costly method of all four waste-to-energy technologies.

Design and operation of high level waste vitrification and storage facilities

International Nuclear Information System (INIS)

1992-01-01

The conversion of high level wastes (HLW) into solids has been studied for the past 30 years, primarily in those countries engaged in the reprocessing of nuclear fuels. Production and demonstration calcination and solidification plants have been operated by using waste solutions from fuels irradiated at various burnup rates, depending on the reactor type. Construction of more advanced solidification processes is now in progress in several countries to permit the handling of high burnup power reactor fuel wastes. The object of this report is to provide detailed information and references for those vitrification systems in advanced stages of implementation. Some less detailed information will be provided for previously developed immobilization systems. The report will examine the HLLW arising from the various locations, the features of each process as well as the stage of development, scale-up potential and flexibility of the processes. Since the publication of IAEA Technical Reports Series No. 176, Techniques for the Solidification of High-Level Wastes great progress on this subject has been made. The AVM in France has been operated successfully for 11 years and France has completed construction at La Hague of two vitrification plants that are based on the AVM rotary calciner/metallic melter process. A similar plant is under construction at Sellafield. The ceramic melter process has been chosen by several countries. Germany has successfully operated the PAMELA vitrification plant. Since 1986, Belgoprocess has continued to operate this facility. The former USSR operated the EP-500 plant from 1986 to 1988. In addition, two ceramic melter vitrification plants are nearing completion in the USA at Savannah River and West Valley and plans are being made to use this technology at Hanford as well as in Japan, Germany and India. This major progress attests to the maturity of these technologies for vitrifying HLLW to make a borosilicate glass for disposal of the waste. 67

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

Microwave waste processing technology overview

Energy Technology Data Exchange (ETDEWEB)

Petersen, R.D.

1993-02-01

Applications using microwave energy in the chemical processing industry have increased within the last ten years. Recently, interest in waste treatment applications process development, especially solidification, has grown. Microwave waste processing offers many advantages over conventional waste treatment technologies. These advantages include a high density, leach resistant, robust waste form, volume and toxicity reduction, favorable economics, in-container treatment, good public acceptance, isolated equipment, and instantaneous energy control. The results from the {open_quotes}cold{close_quotes} demonstration scale testing at the Rocky Flats nuclear weapons facility are described. Preliminary results for a transuranic (TRU) precipitation sludge indicate that volume reductions of over 80% are achievable over the current immobilization process. An economic evaluation performed demonstrated cost savings of $11.68 per pound compared to the immobilization process currently in use on wet sludge.

Microwave waste processing technology overview

International Nuclear Information System (INIS)

Petersen, R.D.

1993-02-01

Applications using microwave energy in the chemical processing industry have increased within the last ten years. Recently, interest in waste treatment applications process development, especially solidification, has grown. Microwave waste processing offers many advantages over conventional waste treatment technologies. These advantages include a high density, leach resistant, robust waste form, volume and toxicity reduction, favorable economics, in-container treatment, good public acceptance, isolated equipment, and instantaneous energy control. The results from the open-quotes coldclose quotes demonstration scale testing at the Rocky Flats nuclear weapons facility are described. Preliminary results for a transuranic (TRU) precipitation sludge indicate that volume reductions of over 80% are achievable over the current immobilization process. An economic evaluation performed demonstrated cost savings of $11.68 per pound compared to the immobilization process currently in use on wet sludge

Plasma technologies: applications to waste processing

International Nuclear Information System (INIS)

Fauchais, P.

2007-01-01

Since the 1990's, plasma technologies have found applications in the processing of toxic wastes of military and industrial origin, like the treatment of contaminated solids and low level radioactive wastes, the decontamination of soils etc.. Since the years 2000, this development is becoming exponential, in particular for the processing of municipal wastes and the recovery of their synthesis gas. The advantage of thermal plasmas with respect to conventional combustion techniques are: a high temperature (more than 6000 K), a pyrolysis capability (CO formation instead of CO 2 ), about 90% of available energy above 1500 K (with respect to 23% with flames), a greater energy density, lower gas flow rates, and plasma start-up and shut-down times of only few tenth of seconds. This article presents: 1 - the present day situation of thermal plasmas development; 2 - some general considerations about plasma waste processing; 3 - the plasma processes: liquid toxic wastes, solid wastes (contaminated soils and low level radioactive wastes, military wastes, vitrification of incinerators fly ash, municipal wastes processing, treatment of asbestos fibers, treatment of chlorinated industrial wastes), metallurgy wastes (dusts, aluminium slags), medical and ship wastes, perspectives; 4 -conclusion. (J.S.)

Environmental restoration/waste management-applied technology semiannual report, January--June 1992

International Nuclear Information System (INIS)

Adamson, M.; Kline-Simon, K.

1992-01-01

This is the first issue from the Lawrence Livermore National Laboratory of The Environmental Restoration/Waste Management-Applied Technology (ER/WM-AT) Semiannual Report, a continuation of the Advanced Processing Technology (APT) Semiannual Report. The name change reflects the consolidation of the APT Program with the Environmental Restoration and Waste Management Program to form the Environmental Restoration/Waste Management-Applied Technology (ER/WM-AT) Program. The Livermore site mirrors, on a small scale, many of the environmental and waste management problems of the DOE Complex. The six articles in this issue cover incineration- alternative technologies, process development for waste minimization, the proposed Mixed Waste Management Facility, dynamic underground stripping, electrical resistance tomography, and Raman spectroscopy for remote characterization of underground tanks

Development of high-level waste solidification technology 1

Energy Technology Data Exchange (ETDEWEB)

Kim, Joon Hyung; Kim, Hwan Young; Kim, In Tae [and others

1999-02-01

Spent nuclear fuel contains useful nuclides as valuable resource materials for energy, heat and catalyst. High-level wastes (HLW) are expected to be generated from the R and D activities and reuse processes. It is necessary to develop vitrification or advanced solidification technologies for the safe long-term management of high level wastes. As a first step to establish HLW vitrification technology, characterization of HLWs that would arise at KAERI site, glass melting experiments with a lab-scale high frequency induction melter, and fabrication and property evaluation of base-glass made of used HEPA filter media and additives were performed. Basic study on the fabrication and characterization of candidate ceramic waste form (Synroc) was also carried out. These HLW solidification technologies would be directly useful for carrying out the R and Ds on the nuclear fuel cycle and waste management. (author). 70 refs., 29 tabs., 35 figs.

French industrial plant AVM for continuous vitrification of high level radioactive wastes

International Nuclear Information System (INIS)

Bonniaud, Roger; Sombret, Claude; Barbe, Alain

1975-01-01

The A.V.M. plant is a continuous process plant now under construction at Marcoule and intended for vitrifying the whole of fission product solutions from the C.E.A. (Commissariat a l'Energie Atomique) - Marcoule reprocessing plant. The outset of the construction took place in the second 1974 half year; the first radioactive run is scheduled in July 1977. The two steps of the process are shown: first a continuous calcination then a continuous glass making from the calcined product and suitable additives. The plant consists in two parts: vitrification and storage facilities. Some wastes will be continuously produced day after day due to gas clean up and worn out materials. Characteristics of the solutions processed, calcined products, glass composition, and expected liquid wastes are given in tables [fr

Synthesis of aluminum oxide by the polymer precursor method (Pechini) in 4: 1 ratio of citric acid: metal cation: calcination temperature effect

International Nuclear Information System (INIS)

Silva, M.C.; Lira, H.L.; Ribeiro, P.C.; Freitas, N.L.

2014-01-01

The technology field is nanopowders prominent in science since these materials fall in various sectors regarding their applications. This work aims at the synthesis of aluminum oxide by polymeric precursors in 4:1 ratio of citric acid:metal cation and evaluate the influence of calcination temperature on their structural and morphological characteristics. The samples after reaction were characterized by XRD and thermal analysis. After calcination 500-1200°C the samples were characterized by XRD, SEM and particle size distribution. The results showed that the variation of the calcination temperature is sufficient to achieve a same material with different structural and morphological characteristics. The most stable phase aluminum oxide arose only after calcination at 1100°C, below 900°C, the amorphous material appeared. As regards the morphology, the change was not as significant as compared to the structure. (author)

FY94 Office of Technology Development Mixed Waste Operations Robotics Demonstration

International Nuclear Information System (INIS)

Kriikku, E.M.

1994-01-01

The Department of Energy (DOE) Office of Technology Development (OTD) develops technologies to help solve waste management and environmental problems at DOE sites. The OTD includes the Robotics Technology Development Program (RTDP) and the Mixed Waste Integrated Program (MWIP). Together these programs will provide technologies for DOE mixed waste cleanup projects. Mixed waste contains both radioactive and hazardous constituents. DOE sites currently store over 240,000 cubic meters of low level mixed waste and cleanup activities will generate several hundred thousand more cubic meters. Federal and state regulations require that this waste must be processed before final disposal. The OTD RTDP Mixed Waste Operations (MWO) team held several robotic demonstrations at the Savannah River Site (SRS) during November of 1993. Over 330 representatives from DOE, Government Contractors, industry, and universities attended. The MWO team includes: Fernald Environmental Management Project (FEMP), Idaho National Engineering Laboratory (INEL), Lawrence Livermore National Laboratory (LLNL), Oak Ridge National Engineering Laboratory (ORNL), Sandia National Laboratory (SNL), and Savannah River Technology Center (SRTC). SRTC is the lead site for MWO and provides the technical coordinator. The primary demonstration objective was to show that robotic technologies can make DOE waste facilities run better, faster, more cost effective, and safer. To meet the primary objective, the demonstrations successfully showed the following remote waste drum processing activities: non-destructive drum examination, drum transportation, drum opening, removing waste from a drum, characterize and sort waste items, scarify metal waste, and inspect stored drums. To further meet the primary objective, the demonstrations successfully showed the following remote waste box processing activities: swing free crane control, workcell modeling, and torch standoff control

Preparation and leaching of radioactive INEL waste forms

International Nuclear Information System (INIS)

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

1982-01-01

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

Review of high-level waste form properties

International Nuclear Information System (INIS)

Rusin, J.M.

1980-12-01

This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison

Criticality safety analysis of a calciner exit chute

International Nuclear Information System (INIS)

Haught, C.F.; Basoglu, B.; Brewer, R.W.; Hollenback, D.F.; Wilkinson, A.D.; Dodds, H.L.

1994-01-01

Calcination of uranyl nitrate into uranium oxide is part of normal operations of some enrichment plants. Typically, a calciner discharges uranium oxide powder (U 3 O 8 ) into an exit chute that directs the powder into a receiving can located in a glove box. One possible scenario for a criticality accident is the exit chute becoming blocked with powder near its discharge. The blockage restricts the flow of powder causing the exit chute to become filled with the powder. If blockage does occur, the height of the powder could reach a level that would not be safe from a criticality point of view. In this analysis, the subcritical height limit is examined for 98% enriched U 3 O 8 in the exit chute with full water reflection and optimal water moderation. The height limit for ensuring criticality safety during such an accumulation is 28.2 cm above the top of the discharge pipe at the bottom of the chute. Chute design variations are also evaluated with full water reflection and optimal water moderation. Subcritical configurations for the exit chute variation are developed, but the configurations are not safe when combined with the calciner. To ensure criticality safety, modifications must be made to the calciner tube or safety measures must be implemented if these designs are to be utilized with 98% enriched material. A geometrically safe configuration for the exit chute is developed for a blockage of 20% enriched powder with full water reflection and optimal water moderation, and this configuration is safe when combined with the existing calciner

Long-lived legacy: Managing high-level and transuranic waste at the DOE Nuclear Weapons Complex. Background paper

International Nuclear Information System (INIS)

1991-05-01

The document focuses on high-level and transuranic waste at the DOE nuclear weapons complex. Reviews some of the critical areas and aspects of the DOE waste problem in order to provide data and further analysis of important issues. Partial contents, High-Level Waste Management at the DOE Weapons Complex, are as follows: High-Level Waste Management: Present and Planned; Amount and Distribution; Current and Potential Problems; Vitrification; Calcination; Alternative Waste Forms for the Idaho National Engineering Laboratory; Technologies for Pretreatment of High-Level Waste; Waste Minimization; Regulatory Framework; Definition of High-Level Waste; Repository Delays and Contingency Planning; Urgency of High-Level Tank Waste Treatment; Technologies for High-Level Waste Treatment; Rethinking the Waste Form and Package; Waste Form for the Idaho National Engineering Laboratory; Releases to the Atmosphere; Future of the PUREX Plant at Hanford; Waste Minimization; Tritium Production; International Cooperation; Scenarios for Future HLW Production. Partial contents of Chapter 2, Managing Transuranic Waste at the DOE Nuclear Weapons Complex, are as follows: Transuranic Waste at Department of Energy Sites; Amount and Distribution; Waste Management: Present and Planned; Current and Potential Problems; Three Technologies for Treating Retrievably Stored Transuranic Waste; In Situ Vitrification; The Applied Research, Development, Demonstration, Testing, and Evaluation Plan (RDDT ampersand E); Actinide Conversion (Transmutation); Waste Minimization; The Regulatory Framework; Definition of, and Standards for, Disposal of Transuranic Waste; Repository Delays; Alternative Storage and Disposal Strategies; Remediation of Buried Waste; The Waste Isolation Pilot Plant; Waste Minimization; Scenarios for Future Transuranic Waste Production; Conditions of No-Migration Determination

Vitrification technology for treating low-level waste from nuclear facilities

International Nuclear Information System (INIS)

Oniki, Toshiro; Nabemoto, Toyonobu; Fukui, Toshiki

2016-01-01

The development of technologies for treating nuclear waste generated by nuclear power plants and reprocessing plants during their operation or decommissioning is underway both in Japan and abroad. Of the many types of treatment technologies that have been developed, vitrification technology is attracting attention as being the most promising technology for converting such waste into a stable state. As a brief review of technical developments aimed at reducing nuclear waste and finding a solution to the final disposal issue, this paper describes approaches to completing the development of vitrification technology in Japan, including IHI's activities. (author)

Technology Successes in Hanford Tank Waste Storage and Retrieval

International Nuclear Information System (INIS)

Cruz, E. J.

2002-01-01

The U. S. Department of Energy (DOE), Office of River Protection (ORP) is leading the River Protection Project (RPP), which is responsible for dispositioning approximately 204,000 cubic meters (54 million gallons) of high-level radioactive waste that has accumulated in 177 large underground tanks at the Hanford Site since 1944. The RPP is comprised of five major elements: storage of the waste, retrieval of the waste from the tanks, treatment of the waste, disposal of treated waste, and closure of the tank facilities. Approximately 3785 cubic meters (1 million gallons) of waste have leaked from the older ''single-shell tanks.'' Sixty-seven of the 147 single shell tanks are known or assumed ''leakers.'' These leaks have resulted in contaminant plumes that extend from the tank to the groundwater in a number of tank farms. Retrieval and closure of the leaking tanks complicates the ORP technical challenge because cleanup decisions must consider the impacts of past leaks along with a strategy for retrieving the waste in the tanks. Completing the RPP mission as currently planned and with currently available technologies will take several decades and tens of billions of dollars. RPP continue to pursue the benefits from deploying technologies that reduce risk to human health and the environment, as well as, the cost of cleanup. This paper discusses some of the recent technology partnering activities with the DOE Office of Science and Technology activities in tank waste retrieval and storage

Prediction of combustible waste generation and estimate of potential energy by applying waste to energy technologies in Korea

International Nuclear Information System (INIS)

Lee, Jang-Soo; Cho, Sung-Jin; Jung, Hae-Young; Lee, Ki-Bae; Seo, Yong-Chil

2010-01-01

In 2007 total waste generation rate in Korea was 318,670 ton,day. In general waste generation rate shows rising trend since 2000. Wastes are composed of municipal waste 14.9 % industrial waste 34.1 % and construction waste 51.0 %. Treatment of wastes by recycling was 81.1 % landfill 11.1 % incineration 5.3 % and ocean dumping 2.4 %. National waste energy policies have been influenced by various factors such as environmental problem economy technology level (could be made energy), and so on. Korea has the worlds third dense population density environmental pollution load per unit land area is the highest in OECD countries caused due to the fast development in economy, industrialization and urbanization in recent. Also, land area per person is just 2,072 m 2 . Landfill capacity reaches the upper limit, industrial waste generation is increasing. Searching new-renewable energy is vital to substitute fossil fuel considering its increasing price. Korea is the world's 10th biggest energy consuming country and 97% of energy depends on importing. Korea aims to increases supply of new-renewable energy by 5% until the 2011. In this study, we computed the amount of combustible waste from municipality generated by the multiple regression analysis. The existing technologies for converting waste to energy were surveyed and the technologies under development or utilizing in future were also investigated. Based on the technology utilization, the amount of energy using waste to energy technology could be estimated in future. (author)

Investigation of corrosion experienced in a spray calciner/ceramic melter vitrification system

International Nuclear Information System (INIS)

Dierks, R.D.; Mellinger, G.B.; Miller, F.A.; Nelson, T.A.; Bjorklund, W.J.

1980-08-01

After periodic testing of a large-scale spray calciner/ceramic melter vitrification system over a 2-yr period, sufficient corrosion was noted on various parts of the vitrification system to warrant its disassembly and inspection. A majority of the 316 SS sintered metal filters on the spray calciner were damaged by chemical corrosion and/or high temperature oxidation. Inconel-601 portions of the melter lid were attacked by chlorides and sulfates which volatilized from the molten glass. The refractory blocks, making up the walls of the melter, were attacked by the waste glass. This attack was occurring when operating temperatures were >1200 0 C. The melter floor was protected by a sludge layer and showed no corrosion. Corrosion to the Inconel-690 electrodes was minimal, and no corrosion was noted in the offgas treatment system downstream of the sintered metal filters. It is believed that most of the melter corrosion occurred during one specific operating period when the melter was operated at high temperatures in an attempt to overcome glass foaming behavior. These high temperatures resulted in a significant release of volatile elements from the molten glass, and also created a situation where the glass was very fluid and convective, which increased the corrosion rate of the refractories. Specific corrosion to the calciner components cannot be proven to have occurred during a specific time period, but the mechanisms of attack were all accelerated under the high-temperature conditions that were experienced with the melter. A review of the materials of construction has been made, and it is concluded that with controlled operating conditions and better protection of some materials of construction corrosion of these systems will not cause problems. Other melter systems operating under similar strenuous conditions have shown a service life of 3 yr

Low-level waste management program and interim waste operations technologies

International Nuclear Information System (INIS)

Mezga, L.J.

1983-01-01

The Department of Energy currently supports an integrated technology development and transfer program aimed at ensuring that the technology necessary for the safe management and disposal of LLW by the commercial and defense sectors is available. The program focuses on five technical areas: (1) corrective measures technology, (2) improved shallow land burial technology, (3) greater confinement disposal technology, (4) model development and validation, and (5) treatment methods for problem wastes. The results of activities in these areas are reported in the open literature and the Proceedings of the LLWMP Annual Participants Information Meeting

Risk assessments of innovative technologies for treatment of mixed waste

International Nuclear Information System (INIS)

Ragaini, R.C.; Aycock, M.T.; Russell, J.E.

1993-01-01

The mission of the US Department of Energy's (DOE'S) Mixed Waste Integrated Program (MWIP) is to develop complete and appropriate technologies for the treatment of DOE mixed low-level waste and transuranic wastes in order to ensure that all affected DOE installations and projects can come into compliance with environmental law and meet DOE's 30-yr cleanup and operational goals. The MWIP will achieve its goal by developing technologies that are in compliance with regulatory requirements, are socially and politically viable, and are cost beneficial and effective in disposed waste source term and volume reduction. The project management plan for MWIP requires that technologies be evaluated in accordance with criteria that rank technologies with regard to performance, risk, and cost-effectiveness. This paper addresses the methodology used to rank alternative mixed-water treatment technologies with regard to relative risk

Process considerations for hot pressing ceramic nuclear waste forms

International Nuclear Information System (INIS)

Wilson, C.N.; Brite, D.W.

1981-01-01

Spray calcined simulated ceramic nuclear waste powders were hot pressed in graphite, nickel-lined graphite and ZrO 2 -lined Al 2 O 3 dies. Densification, initial off-gas, waste element retention and pellet-die interactions were evaluated. Indicated process considerations and limitations are discussed. 15 figures

The state of the art on the radioactive metal waste recycling technologies

International Nuclear Information System (INIS)

Oh, Won Jin; Moon, Jei Kwon; Jung, Chong Hun; Park, Sang Yoon

1997-09-01

As the best strategy to manage the radioactive metal wastes which are generated during operation and decommissioning of nuclear facilities, the following recycling technologies are investigated. 1. decontamination technologies for radioactive metal waste recycling 2. decontamination waste treatment technologies. 3. residual radioactivity evaluation technologies. (author). 260 refs., 26 tabs., 31 figs

Subject bibliography of radioactive waste management publications at Pacific Northwest Laboratory, 1975-1978

International Nuclear Information System (INIS)

Powell, J.A.

1981-10-01

This bibliography contains publications from 1975 to 1978 written by PNL staff. PNL translations are also announced in this document. The following areas are covered: actinides; airborne wastes; alternative waste forms; calcination; characterization; containers; decontamination; disposal; high-level wastes; liquid wastes; radionuclide migration; safety; separation processes; soils; solidification; storage; transport; transuranic waste; and vitrification

Barriers to development and deployment of innovative waste minimization technologies

International Nuclear Information System (INIS)

Flores, E.A.; Donaghue, J.F.

1994-08-01

Increasing regulation and scrutiny is driving waste generators towards reducing the use of scarce natural resources and reducing or eliminating was streams. There is increasing emphasis on developing and deploying technologies that meet industry needs for recovering valuable materials in a cost-effective manner. At the Department of Energy's (DOE) Hanford Site, Battelle operates Pacific Northwest Laboratory (PNL). PNL's mission is to develop technologies to clean up the environment, and to assist industry in being competitive on a global scale. One such technology developed by PNL is the Waste Acid Detoxification and Reclamation (WADR) process. This technology recovers acids from metal-bearing spent solutions, separating out the metals (which are a valuable byproduct of the acid recycling operation) from the acids. WADR uses selective precipitation and distillation together in an innovative waste recycling technology. Selective precipitation removes the heavy metals, and vacuum distillation recovers clean acid. However, WADR and other innovative waste reduction technologies face numerous barriers to successful development and deployment in the field

Microstructural changes in porous hematite nanoparticles upon calcination

DEFF Research Database (Denmark)

Johnsen, Rune; Knudsen, Kenneth D.; Molenbroek, Alfons M.

2011-01-01

This combined study using small-angle neutron scattering (SANS), X-ray powder diffraction (XRPD), transmission electron microscopy (TEM) and adsorption isotherm techniques demonstrates radical changes in the microstructure of porous hematite (-Fe2O3) nanoparticles upon calcination in air. TEM....... The change in microstructure also causes a reduction in the surface area as calculated by gaseous adsorption. The XRPD and SANS data show that the crystallite and SANS particle sizes are virtually unchanged by calcination at 623 K. Calcination at 973 K induces a significant alteration of the sample. The XRPD...... data reveal that the crystallite size increases significantly, and the SANS and adsorption isotherm studies suggest that the specific surface area decreases by a factor of 5–6. The TEM images show that the particles are sintered into larger agglomerates, but they also show that parts of the porous...

Technology development activities supporting tank waste remediation

International Nuclear Information System (INIS)

Bonner, W.F.; Beeman, G.H.

1994-06-01

This document summarizes work being conducted under the U.S. Department of Energy's Office of Technology Development (EM-50) in support of the Tank Waste Remediation System (TWRS) Program. The specific work activities are organized by the following categories: safety, characterization, retrieval, barriers, pretreatment, low-level waste, and high-level waste. In most cases, the activities presented here were identified as supporting tank remediation by EM-50 integrated program or integrated demonstration lead staff and the selections were further refined by contractor staff. Data sheets were prepared from DOE-HQ guidance to the field issued in September 1993. Activities were included if a significant portion of the work described provides technology potentially needed by TWRS; consequently, not all parts of each description necessarily support tank remediation

Sodium-Bearing Waste Treatment, Applied Technology Plan

International Nuclear Information System (INIS)

Lance Lauerhass; Vince C. Maio; S. Kenneth Merrill; Arlin L. Olson; Keith J. Perry

2003-01-01

Settlement Agreement between the Department of Energy and the State of Idaho mandates treatment of sodium-bearing waste at the Idaho Nuclear Technology and Engineering Center within the Idaho National Engineering and Environmental Laboratory. One of the requirements of the Settlement Agreement is to complete treatment of sodium-bearing waste by December 31, 2012. Applied technology activities are required to provide the data necessary to complete conceptual design of four identified alternative processes and to select the preferred alternative. To provide a technically defensible path forward for the selection of a treatment process and for the collection of needed data, an applied technology plan is required. This document presents that plan, identifying key elements of the decision process and the steps necessary to obtain the required data in support of both the decision and the conceptual design. The Sodium-Bearing Waste Treatment Applied Technology Plan has been prepared to provide a description/roadmap of the treatment alternative selection process. The plan details the results of risk analyzes and the resulting prioritized uncertainties. It presents a high-level flow diagram governing the technology decision process, as well as detailed roadmaps for each technology. The roadmaps describe the technical steps necessary in obtaining data to quantify and reduce the technical uncertainties associated with each alternative treatment process. This plan also describes the final products that will be delivered to the Department of Energy Idaho Operations Office in support of the office's selection of the final treatment technology

Sodium-Bearing Waste Treatment, Applied Technology Plan

Energy Technology Data Exchange (ETDEWEB)

Lance Lauerhass; Vince C. Maio; S. Kenneth Merrill; Arlin L. Olson; Keith J. Perry

2003-06-01

Settlement Agreement between the Department of Energy and the State of Idaho mandates treatment of sodium-bearing waste at the Idaho Nuclear Technology and Engineering Center within the Idaho National Engineering and Environmental Laboratory. One of the requirements of the Settlement Agreement is to complete treatment of sodium-bearing waste by December 31, 2012. Applied technology activities are required to provide the data necessary to complete conceptual design of four identified alternative processes and to select the preferred alternative. To provide a technically defensible path forward for the selection of a treatment process and for the collection of needed data, an applied technology plan is required. This document presents that plan, identifying key elements of the decision process and the steps necessary to obtain the required data in support of both the decision and the conceptual design. The Sodium-Bearing Waste Treatment Applied Technology Plan has been prepared to provide a description/roadmap of the treatment alternative selection process. The plan details the results of risk analyzes and the resulting prioritized uncertainties. It presents a high-level flow diagram governing the technology decision process, as well as detailed roadmaps for each technology. The roadmaps describe the technical steps necessary in obtaining data to quantify and reduce the technical uncertainties associated with each alternative treatment process. This plan also describes the final products that will be delivered to the Department of Energy Idaho Operations Office in support of the office's selection of the final treatment technology.

Vitrification technology for Hanford Site tank waste

International Nuclear Information System (INIS)

Weber, E.T.; Calmus, R.B.; Wilson, C.N.

1995-04-01

The US Department of Energy's (DOE) Hanford Site has an inventory of 217,000 m 3 of nuclear waste stored in 177 underground tanks. The DOE, the US Environmental Protection Agency, and the Washington State Department of Ecology have agreed that most of the Hanford Site tank waste will be immobilized by vitrification before final disposal. This will be accomplished by separating the tank waste into high- and low-level fractions. Capabilities for high-capacity vitrification are being assessed and developed for each waste fraction. This paper provides an overview of the program for selecting preferred high-level waste melter and feed processing technologies for use in Hanford Site tank waste processing

Overview of non-thermal mixed waste treatment technologies: Treatment of mixed waste (ex situ); Technologies and short descriptions

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-07-01

This compendium contains brief summaries of new and developing non- thermal treatment technologies that are candidates for treating hazardous or mixed (hazardous plus low-level radioactive) wastes. It is written to be all-encompassing, sometimes including concepts that presently constitute little more than informed ``ideas``. It bounds the universe of existing technologies being thought about or considered for application on the treatment of such wastes. This compendium is intended to be the very first step in a winnowing process to identify non-thermal treatment systems that can be fashioned into complete ``cradle-to-grave`` systems for study. The purpose of the subsequent systems paper studies is to investigate the cost and likely performance of such systems treating a representative sample of U.S. Department of Energy (DOE) mixed low level wastes (MLLW). The studies are called Integrated Non-thermal Treatment Systems (INTS) Studies and are being conducted by the Office of Science and Technology (OST) of the Environmental Management (EM) of the US Department of Energy. Similar studies on Integrated Thermal Treatment Systems have recently been published. These are not designed nor intended to be a ``downselection`` of such technologies; rather, they are simply a systems evaluation of the likely costs and performance of various non- thermal technologies that have been arranged into systems to treat sludges, organics, metals, soils, and debris prevalent in MLLW.

Overview of non-thermal mixed waste treatment technologies: Treatment of mixed waste (ex situ); Technologies and short descriptions

International Nuclear Information System (INIS)

1995-07-01

This compendium contains brief summaries of new and developing non- thermal treatment technologies that are candidates for treating hazardous or mixed (hazardous plus low-level radioactive) wastes. It is written to be all-encompassing, sometimes including concepts that presently constitute little more than informed ''ideas''. It bounds the universe of existing technologies being thought about or considered for application on the treatment of such wastes. This compendium is intended to be the very first step in a winnowing process to identify non-thermal treatment systems that can be fashioned into complete ''cradle-to-grave'' systems for study. The purpose of the subsequent systems paper studies is to investigate the cost and likely performance of such systems treating a representative sample of U.S. Department of Energy (DOE) mixed low level wastes (MLLW). The studies are called Integrated Non-thermal Treatment Systems (INTS) Studies and are being conducted by the Office of Science and Technology (OST) of the Environmental Management (EM) of the US Department of Energy. Similar studies on Integrated Thermal Treatment Systems have recently been published. These are not designed nor intended to be a ''downselection'' of such technologies; rather, they are simply a systems evaluation of the likely costs and performance of various non- thermal technologies that have been arranged into systems to treat sludges, organics, metals, soils, and debris prevalent in MLLW

Incineration technology for alpha-bearing radioactive waste in Germany

International Nuclear Information System (INIS)

Dirks, Friedlich; Pfeiffer, Reinhard

1997-01-01

Since 1971 the Karlsruhe Research Center has developed and operated plants for the incineration of radioactive waste. Three incineration plants for pure β/γ solid, α-bearing solid and radioactive liquid waste have been successfully utilized during last two decades. Recently more than 20 year-old β/γ plant was shut down with the economic point of view, mainly due to the recently reduced volume of burnable β/γ waste. Burnable β/γ solid waste is now being treated with α-bearing waste in a α solid incineration plant. The status of incineration technology for α-bearing waste and other radioactive waste treatment technologies, which are now utilized in Karlsruhe Research Center, such as conditioning of incineration ash, supercompaction, scrapping, and decontamination of solid radioactive waste, etc. are introduced in this presentation. Additionally, operational results of the recently installed new dioxin adsorber and fluidized-bed drier for scrubber liquid in α incineration plant are also described in this presentation. (author) 1 tab., 13 figs

Future-proof radioactive waste treatment technologies for nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Buettner, Klaus; Braehler, Georg [NUKEM Technologies Engineering Services GmbH, Alzenau (Germany)

2014-08-15

In order to select the optimal treatment method for radioactive waste three options can be considered. First, to treat the radioactive waste only to allow long term interim storage until the waste acceptance criteria are defined and the disposal sites are operable. Second, to select treatment methods just in compliance with the current state of discussion with the regard to the above. Third, taking also the future development in the field of waste acceptance criteria and disposal into account. When developing waste treatment systems for Nuclear Power Plants NUKEM Technologies follows the following targets, minimisation of the amount of radioactive waste, maximisation of free release material, volume reduction, avoidance of unwanted materials in the waste package, as well as efficient waste treatment solutions (low investment, high volume reduction). With its technologies produced waste packages fulfil the most stringent waste acceptance criteria.

DC graphite plasma arc melter technology for waste vitrification

International Nuclear Information System (INIS)

Hamilton, R.A.; Wittle, J.K.; Trescot, J.; Wilver, P.

1995-01-01

This paper describes the features and benefits of a DC Arc Melter for the permanent treatment of all types of solid wastes including nonhazardous, hazardous and radioactive. This DC Arc Melter system is the low cost permanent solution for solid waste pollution prevention and remediation. Concern over the effective disposal of wastes generated by our industrial society, worldwide, has prompted development of technologies to address the problem. The only solution that has the ability to process almost all wastes, and to recover/recycle metallic and inorganic matter, is the group of technologies known as melters. Melters have distinct advantages over traditional technologies such as incineration because melters; operate at higher temperatures, are relatively unaffected by changes in the waste stream, produce a vitrified stable product, reduce gaseous emissions, and have the capability to recover/recycle slag, metals and gas. The system, DC Plasma Arc Melter, has the lowest capital, maintenance and operating cost of any melter technology because of its patented DC Plasma Arc with graphite electrode. DC Plasma Arc Melter systems are available in sizes from 50 kg/batch or 250-3,000 kg/hr on a continuous basis

Calcination of Rod-like Hydroxyapatite Nanocrystals with an Anti-sintering Agent Surrounding the Crystals

International Nuclear Information System (INIS)

Okada, M.; Furuzono, T.

2007-01-01

Sintering-free nanocrystals of calcined hydroxyapatite (HAp) having a rod-like morphology were fabricated by calcination at 800 deg. C for 1 h with an anti-sintering agent surrounding original HAp particles and the agent was subsequently removed after calcination. The original HAp particles having a rod-like morphology with a size ranging from 30 to 80 nm (short axis) and 300 to 500 nm (long axis) were prepared by wet chemical process, and poly(acrylic acid, calcium salt) (PAA-Ca) was used as the anti-sintering agent. In the case of calcination without additives, the mean size of HAp crystals dispersed in an ethanol medium increased by about 4 times and the specific surface area of the crystals exhibited a 25% decrease compared to those of the original HAp particles because of calcination-induced sintering among the crystals. On the other hand, the HAp crystals calcined with the anti-sintering agent, PAA-Ca, could be dispersed in an ethanol medium at the same size as the original particles, and they preserved the specific surface area after calcination. These results indicate that PAA-Ca and/or its thermally decomposed product, CaO, surrounded the HAp particles and protected them against calcination-induced sintering during calcination. The HAp crystals calcined with PAA-Ca showed high crystallinity, and no other calcium phosphate phases could be detected after washing with water

Application of new technologies for characterization of Hanford Site high-level waste

International Nuclear Information System (INIS)

Winters, W.I.

1998-01-01

To support remediation of Hanford Site high-level radioactive waste tanks, new chemical and physical measurement technologies must be developed and deployed. This is a major task of the Chemistry Analysis Technology Support (CATS) group of the Hanford Corporation. New measurement methods are required for efficient and economical resolution of tank waste safety, waste retrieval, and disposal issues. These development and deployment activities are performed in cooperation with Waste Management Federal Services of Hanford, Inc. This paper provides an overview of current analytical technologies in progress. The high-level waste at the Hanford Site is chemically complex because of the numerous processes used in past nuclear fuel reprocessing there, and a variety of technologies is required for effective characterization. Programmatic and laboratory operational needs drive the selection of new technologies for characterizing Hanford Site high-level waste, and these technologies are developed for deployment in laboratories, hot cells or in the field. New physical methods, such as the propagating reactive systems screening tool (PRSST) to measure the potential for self-propagating reactions in stored wastes, are being implemented. Technology for sampling and measuring gases trapped within the waste matrix is being used to evaluate flammability hazards associated with gas releases from stored wastes. Application of new inductively coupled plasma and laser ablation mass spectrometry systems at the Hanford Site's 222-S Laboratory will be described. A Raman spectroscopy probe mounted in a cone penetrometer to measure oxyanions in wastes or soils will be described. The Hanford Site has used large volumes of organic complexants and acids in processing waste, and capillary zone electrophoresis (CZE) methods have been developed for determining several of the major organic components in complex waste tank matrices. The principles involved, system installation, and results from

Application of SYNROC to high-level defense wastes

International Nuclear Information System (INIS)

Tewhey, J.D.; Hoenig, C.L.; Newkirk, H.W.; Rozsa, R.B.; Coles, D.G.; Ryerson, F.J.

1981-01-01

The SYNROC method for immobilization of high-level nuclear reactor wastes is currently being applied to US defense wastes in tank storage at Savannah River, South Carolina. The minerals zirconolite, perovskite, and hollandite are used in SYNROC D formulations to immobilize fission products and actinides that comprise up to 10% of defense waste sludges and coexisting solutions. Additional phase in SYNROC D are nepheline, the host phase for sodium; and spinel, the host for excess aluminum and iron. Up to 70 wt % of calcined sludge can be incorporated with 30 wt % of SYNROC additives to produce a waste form consisting of 10% nepheline, 30% spinel, and approximately 20% each of the radioactive waste-bearing phases. Urea coprecipitation and spray drying/calcining methods have been used in the laboratory to produce homogeneous, reactive ceramic powders. Hot pressing and sintering at temperatures from 1000 to 1100 0 C result in waste form products with greater than 97% of theoretical density. Hot isostatic pressing has recently been implemented as a processing alternative. Characterization of waste-form mineralogy has been done by means of XRD, SEM, and electron microprobe. Leaching of SYNROC D samples is currently being carried out. Assessment of radiation damage effects and physical properties of SYNROC D will commence in FY81

Waste Treatment of Acidic Solutions from the Dissolution of Irradiated LEU Targets for 99-Mo Production

Energy Technology Data Exchange (ETDEWEB)

Bakel, Allen J. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Conner, Cliff [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Quigley, Kevin [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division; Vandegrift, George F. [Argonne National Lab. (ANL), Argonne, IL (United States). Nuclear Engineering Division

2016-10-01

One of the missions of the Reduced Enrichment for Research and Test Reactors (RERTR) program (and now the National Nuclear Security Administrations Material Management and Minimization program) is to facilitate the use of low enriched uranium (LEU) targets for ⁹⁹Mo production. The conversion from highly enriched uranium (HEU) to LEU targets will require five to six times more uranium to produce an equivalent amount of ⁹⁹Mo. The work discussed here addresses the technical challenges encountered in the treatment of uranyl nitrate hexahydrate (UNH)/nitric acid solutions remaining after the dissolution of LEU targets. Specifically, the focus of this work is the calcination of the uranium waste from ⁹⁹Mo production using LEU foil targets and the Modified Cintichem Process. Work with our calciner system showed that high furnace temperature, a large vent tube, and a mechanical shield are beneficial for calciner operation. One- and two-step direct calcination processes were evaluated. The high-temperature one-step process led to contamination of the calciner system. The two-step direct calcination process operated stably and resulted in a relatively large amount of material in the calciner cup. Chemically assisted calcination using peroxide was rejected for further work due to the difficulty in handling the products. Chemically assisted calcination using formic acid was rejected due to unstable operation. Chemically assisted calcination using oxalic acid was recommended, although a better understanding of its chemistry is needed. Overall, this work showed that the two-step direct calcination and the in-cup oxalic acid processes are the best approaches for the treatment of the UNH/nitric acid waste solutions remaining from dissolution of LEU targets for ⁹⁹Mo production.

Final report, Task 2: alternative waste management options, Nuclear Fuel Services, Inc., high level waste

International Nuclear Information System (INIS)

1978-01-01

Of the alternatives considered for disposal of the high-level waste in tanks 8D2 and 8D4, the following process is recommended: homogenization of the contents of tank 8D2, centrifugation of the sludge and supernate, mixing of the 8D4 acid waste with the centrifuged sludge, and converting the mixture to a borosilicate glass using the Hanford spray calciner/in-can melter

ICPP calcined solids storage facility closure study. Volume III: Engineering design files

International Nuclear Information System (INIS)

1998-02-01

The following information was calculated to support cost estimates and radiation exposure calculations for closure activities at the Calcined Solids Storage Facility (CSSF). Within the estimate, volumes were calculated to determine the required amount of grout to be used during closure activities. The remaining calcine on the bin walls, supports, piping, and floor was also calculated to approximate the remaining residual calcine volumes at different stages of the removal process. The estimates for remaining calcine and vault void volume are higher than what would actually be experienced in the field, but are necessary for bounding purposes. The residual calcine in the bins may be higher than was is experienced in the field as it was assumed that the entire bin volume is full of calcine before removal activities commence. The vault void volumes are higher as the vault roof beam volumes were neglected. The estimations that follow should be considered rough order of magnitude, due to the time constraints as dictated by the project's scope of work. Should more accurate numbers be required, a new analysis would be necessary

ICPP calcined solids storage facility closure study. Volume III: Engineering design files

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-02-01

The following information was calculated to support cost estimates and radiation exposure calculations for closure activities at the Calcined Solids Storage Facility (CSSF). Within the estimate, volumes were calculated to determine the required amount of grout to be used during closure activities. The remaining calcine on the bin walls, supports, piping, and floor was also calculated to approximate the remaining residual calcine volumes at different stages of the removal process. The estimates for remaining calcine and vault void volume are higher than what would actually be experienced in the field, but are necessary for bounding purposes. The residual calcine in the bins may be higher than was is experienced in the field as it was assumed that the entire bin volume is full of calcine before removal activities commence. The vault void volumes are higher as the vault roof beam volumes were neglected. The estimations that follow should be considered rough order of magnitude, due to the time constraints as dictated by the project`s scope of work. Should more accurate numbers be required, a new analysis would be necessary.

Mixed Waste Integrated Program interim evaluation report on thermal treatment technologies

International Nuclear Information System (INIS)

Gillins, R.L.; DeWitt, L.M.; Wollerman, A.L.

1993-02-01

The Mixed Waste Integrated Program (MWIP) is one of several US Department of Energy (DOE) integrated programs established to organize and coordinate throughout the DOE complex the development of technologies for treatment of specific waste categories. The goal of the MWIP is to develop and deploy appropriate technologies for -the treatment of DOE mixed low-level and alpha-contaminated wastes in order to bring all affected DOE installations and projects into compliance with environmental laws. Evaluation of treatment technologies by the MWIP will focus on meeting waste form performance requirements for disposal. Thermal treatment technologies were an early emphasis for the MWIP because thermal treatment is indicated (or mandated) for many of the hazardous constituents in DOE mixed waste and because these technologies have been widely investigated for these applications. An advisory group, the Thermal Treatment Working Group (TTWG), was formed during the program's infancy to assist the MWIP in evaluating and prioritizing thermal treatment technologies suitable for development. The results of the overall evaluation scoring indicate that the four highest-rated technologies were rotary kilns, slagging kilns, electric-arc furnaces, and plasma-arc furnaces. The four highest-rated technologies were all judged to be applicable on five of the six waste streams and are the only technologies in the evaluation with this distinction. Conclusions as to the superiority of one technology over others are not valid based on this preliminary study, although some general conclusions can be drawn

Mercury emissions control technologies for mixed waste thermal treatment

International Nuclear Information System (INIS)

Chambers, A.; Knecht, M.; Soelberg, N.; Eaton, D.

1997-01-01

EPA has identified wet scrubbing at low mercury feedrates, as well as carbon adsorption via carbon injection into the offgas or via flow through fixed carbon beds, as control technologies that can be used to meet the proposed Maximum Achievable Control Technology (MACT) rule limit for mercury emissions from hazardous waste incinerators. DOE is currently funding demonstrations of gold amalgamation that may also control mercury to the desired levels. Performance data from a variety of sources was reviewed to determine ranges of achievable mercury control. Preliminary costs were estimated for using these technologies to control mercury emissions from mixed waste incineration. Mercury emissions control for mixed waste incineration may need to be more efficient than for incineration of other hazardous wastes because of higher mercury concentrations in some mixed waste streams. However, mercury control performance data for wet scrubbing and carbon adsorption is highly variable. More information is needed to demonstrate control efficiencies that are achievable under various design and operating conditions for wet scrubbing, carbon adsorption, and gold amalgamation technologies. Given certain assumptions made in this study, capital costs, operating costs, and lifecycle costs for carbon injection, carbon beds, and gold amalgamation generally vary for different assumed mercury feedrates and for different offgas flowrates. Assuming that these technologies can in fact provide the necessary mercury control performance, each of these technologies may be less costly than the others for certain mercury feedrates and the offgas flowrates

Waste disposal technology transfer matching requirement clusters for waste disposal facilities in China

International Nuclear Information System (INIS)

Dorn, Thomas; Nelles, Michael; Flamme, Sabine; Jinming, Cai

2012-01-01

Highlights: ► We outline the differences of Chinese MSW characteristics from Western MSW. ► We model the requirements of four clusters of plant owner/operators in China. ► We examine the best technology fit for these requirements via a matrix. ► Variance in waste input affects result more than training and costs. ► For China technology adaptation and localisation could become push, not pull factors. - Abstract: Even though technology transfer has been part of development aid programmes for many decades, it has more often than not failed to come to fruition. One reason is the absence of simple guidelines or decision making tools that help operators or plant owners to decide on the most suitable technology to adopt. Practical suggestions for choosing the most suitable technology to combat a specific problem are hard to get and technology drawbacks are not sufficiently highlighted. Western counterparts in technology transfer or development projects often underestimate or don’t sufficiently account for the high investment costs for the imported incineration plant; the differing nature of Chinese MSW; the need for trained manpower; and the need to treat flue gas, bunker leakage water, and ash, all of which contain highly toxic elements. This article sets out requirements for municipal solid waste disposal plant owner/operators in China as well as giving an attribute assessment for the prevalent waste disposal plant types in order to assist individual decision makers in their evaluation process for what plant type might be most suitable in a given situation. There is no ‘best’ plant for all needs and purposes, and requirement constellations rely on generalisations meaning they cannot be blindly applied, but an alignment of a type of plant to a type of owner or operator can realistically be achieved. To this end, a four-step approach is suggested and a technology matrix is set out to ease the choice of technology to transfer and avoid past errors. The four

Environmental, technical and technological aspects of hazardous waste management in Poland

Science.gov (United States)

Pyssa, Justyna

2017-10-01

The issue of recovery and disposal of hazardous waste is not a new concern. The waste comes from various processes and technologies and therefore the bigger emphasis should be placed on reducing quantities of generated hazardous waste (which is often connected with changes in the technology of manufacturing a given product) and limitation of their negative influence on natural environment. Plants specializing in waste processing processes should meet the so-called cardinal triad of conditions deciding on the full success of investment, and namely: economic effectiveness, ecological efficiency and social acceptance. The structure of generation of hazardous waste in EU-28 has been presented in the paper. Methods of hazardous waste disposal in Poland have been discussed. Economic and ecological criteria for the selection of technology of hazardous waste disposal have been analyzed. The influence of the hazardous waste on the environment is also presented. For four groups of waste, which are currently stored, alternative methods of disposal have been proposed.

Systems analysis support to the waste management technology center

International Nuclear Information System (INIS)

Rivera, A.L.; Osborne-Lee, I.W.; DePaoli, S.M.

1988-01-01

This paper describes a systems analysis concept being developed in support of waste management planning and analysis activities for Martin Marietta Energy Systems, Inc. (Energy Systems), sites. This integrated systems model serves as a focus for the accumulation and documentation of technical and economic information from current waste management practices, improved operations projects, remedial actions, and new system development activities. The approach is generic and could be applied to a larger group of sites. This integrated model is a source of technical support to waste management groups in the Energy Systems complex for integrated waste management planning and related technology assessment activities. This problem-solving methodology for low-level waste (LLW) management is being developed through the Waste Management Technology Center (WMTC) for the Low-Level Waste Disposal, Development, and Demonstration (LLWDDD) Program. In support of long-range planning activities, this capability will include the development of management support tools such as specialized systems models, data bases, and information systems. These management support tools will provide continuing support in the identification and definition of technical and economic uncertainties to be addressed by technology demonstration programs. Technical planning activities and current efforts in the development of this system analysis capability for the LLWDDD Program are presented in this paper

UJV line for research into radioactive wastes solidification

International Nuclear Information System (INIS)

Neumann, L.; Feist, I.; Kepak, F.; Nachmilner, L.; Napravnik, J.; Novak, M.; Pecak, V.; Vojtech, O.

1985-01-01

An experimental line with a capacity of 0.01 m 3 /h was developed and built for research of the solidification of liquid radioactive wastes at the Nuclear Research Institute. The line allows the research and pilot plant testing of processes based on vitrification but also on other procedures including calcination. It consists of a horizontal calciner, a resistance melting unit, a homogenization device for research into cementation of the calcinate, and equipment for the disposal of gaseous emissions. The facility is provided with a control console which allows remote control and the control of all basic operating parameters. The design of the line allows its eventual completion with other equipment. (Z.M.)

Mission Need Statement: Calcine Disposition Project Major Systems Acquisition Project

International Nuclear Information System (INIS)

J. T. Beck

2007-01-01

This document identifies the need to establish the Calcine Disposition Project to determine and implement the final disposition of calcine including characterization, retrieval, treatment (if necessary), packaging, loading, onsite interim storage pending shipment to a repository or interim storage facility, and disposition of related facilities

Recovery of gold and uranium from calcines

Energy Technology Data Exchange (ETDEWEB)

Livesey-Goldblatt, E.

1981-10-06

The invention concerns the recovery of non-ferrous metals, such as gold, uranium or the like from iron oxide containing calcines which have the non-ferrous metal present in solid solution and/or encapsulated within the iron oxide. The calcine is reacted, while stirring vigorously, with sulphuric acid or another strong inorganic acid to cause the iron to form the ferric salt. The material obtained is mixed with water and the liquid and solid phases are separated from each other. The non-ferrous metal is then obtained from at least one of these phases by leaching, or the like.

Field test plan: Buried waste technologies, Fiscal Year 1995

International Nuclear Information System (INIS)

Heard, R.E.; Hyde, R.A.; Engleman, V.S.; Evans, J.D.; Jackson, T.W.

1995-06-01

The US Department of Energy, Office of Technology Development, supports the applied research, development, demonstration, testing, and evaluation of a suite of advanced technologies that, when integrated with commercially available baseline technologies, form a comprehensive remediation system for the effective and efficient remediation of buried waste. The Fiscal Year 1995 effort is to deploy and test multiple technologies from four functional areas of buried waste remediation: site characterization, waste characterization, retrieval, and treatment. This document is the basic operational planning document for the deployment and testing of the technologies that support the field testing in Fiscal Year 1995. Discussed in this document are the scope of the tests; purpose and objective of the tests; organization and responsibilities; contingency plans; sequence of activities; sampling and data collection; document control; analytical methods; data reduction, validation, and verification; quality assurance; equipment and instruments; facilities and utilities; health and safety; residuals management; and regulatory management

RECONSTRUCTION OF CALCINED Zn -Al LAYERED DOUBLE HYDROXIDES DURING TETRACYCLINE ADSORPSION

Directory of Open Access Journals (Sweden)

G. M. Starukh

2015-12-01

Full Text Available Zn-Al mixed oxides containing ZnO different degree crystallinity were obtained by calcinations of Zn-Al layered double hydroxides (LDHs. The reconstruction of calcined Zn-Al LDHs has been performed under stirring in aqueous suspensions. The assynthesized LDHs, its decomposition products, as well as the reconstructed solids upon hydration were characterized by XRD, N2adsorption, differential and thermal gravimetric analysis. It was found that the ability of Zn-Al LDHs to recover a layered structure under the hydration of mixed oxides depends on the degree of ZnO crystallinity. The partial reconstruction of Zn-Al layered structure occurs in tetracycline solutions irrespective to the degree of ZnO crystallinity in calcined LDHs. Calcined Zn-Al LDHs demonstrate the higher adsorption capacity to tetracycline in comparison with as-prepared Zn-Al LDHs. The adsorption of TC on calcined and uncalcined ZnAl LDHs occurs on the centers of one particular type. It is suggested that surface complexation of the A-ring ligand of TC with Al-OH centers takes place.

Ceramic process and plant design for high-level nuclear waste immobilization

International Nuclear Information System (INIS)

Grantham, L.F.; McKisson, R.L.; De Wames, R.E.; Guon, J.; Flintoff, J.F.; McKenzie, D.E.

1983-01-01

In the last 3 years, significant advances in ceramic technology for high-level nuclear waste solidification have been made. Product quality in terms of leach-resistance, compositional uniformity, structural integrity, and thermal stability promises to be superior to borosilicate glass. This paper addresses the process effectiveness and preliminary designs for glass and ceramic immobilization plants. The reference two-step ceramic process utilizes fluid-bed calcination (FBC) and hot isostatic press (HIP) consolidation. Full-scale demonstration of these well-developed processing steps has been established at DOE and/or commercial facilities for processing radioactive materials. Based on Savannah River-type waste, our model predicts that the capital and operating cost for the solidification of high-level nuclear waste is about the same for the ceramic and glass options. However, when repository costs are included, the ceramic option potentially offers significantly better economics due to its high waste loading and volume reduction. Volume reduction impacts several figures of merit in addition to cost such as system logistics, storage, transportation, and risk. The study concludes that the ceramic product/process has many potential advantages, and rapid deployment of the technology could be realized due to full-scale demonstrations of FBC and HIP technology in radioactive environments. Based on our finding and those of others, the ceramic innovation not only offers a viable backup to the glass reference process but promises to be a viable future option for new high-level nuclear waste management opportunities

Innovative waste treatment and conditioning technologies at nuclear power plants

International Nuclear Information System (INIS)

2006-05-01

The objective of this publication is to provide Member States with information on the most innovative technologies and strategies used in waste treatment and conditioning. At present, some of those technologies and strategies might not be widely implemented at nuclear power plants (NPP), but they have an important potential for their use as part of the long range NPP, utility, or national strategy. Thus, the target audience is those decision makers at the national and organizational level responsible for selecting waste processing technologies and strategies over a period of three to ten years. Countries and individual nuclear plants have limited financial resources which can be applied toward radioactive waste processing (treatment and conditioning). They are challenged to determine which of the many available technologies and strategies are best suited to meet national or local needs. This publication reduces the selection of processes for wastes generated by nuclear power plants to those technologies and strategies which are considered innovative. The report further identifies the key benefits which may derive from the adoption of those technologies, the different waste streams to which each technology is relevant, and the limitations of the technologies. The technologies and strategies identified have been evaluated to differentiate between (1) predominant technologies (those that are widely practiced in multiple countries or a large number of nuclear plants), and (2) innovative technologies (those which are not so widely used but are considered to offer benefits which make them suitable for broader application across the industry). Those which fall into the second category are the primary focus of this report. Many IAEA publications address the technical aspects of treatment and conditioning for radioactive wastes, covering research, technological advances, and safety issues. These studies and reports primarily target the research and technical staff of a