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Sample records for solidification processes tru

  1. Solidification of TRU wastes in a ceramic matrix

    International Nuclear Information System (INIS)

    Loida, A.; Schubert, G.

    1991-01-01

    Aluminumsilicate based ceramic materials have been evaluated as an alternative waste form for the incorporation of TRU wastes. These waste forms are free of water and - cannot generate hydrogen radiolyticly, - they show good compatibility between the compounds of the waste and the matrix, - they are resistent against aqueous solutions, heat and radiation. R and D-work has been performed to demonstrate the suitability of this waste form for the immobilization of TRU-wastes. Four kinds of original TRU-waste streams and a mixture of all of them have been immobilized by ceramization, using glove box and remote operation technique as well. Clay minerals, (kaolinite, bentonite) and reactive corundum were selected as ceramic raw materials (KAB 78) in an appropriate ratio yielding 78 wt% Al 2 O 3 and 22 wt%SiO 2 . The main process steps are (i) pretreatment of the liquid waste (concentration, denitration, neutralization, solid- liquid separation), (ii) mixing with ceramic raw materials and forming, (iii) heat treatment with T max. of 1300 0 C for 15 minutes. The waste load of the ceramic matrix has been increased gradually from 20 to 50, in some cases to 60 wt.%

  2. Chemical radwaste solidification processes

    International Nuclear Information System (INIS)

    Malloy, C.W.

    1979-01-01

    Some of these processes and their problems are briefly reviewed: early cement systems; urea-formaldehyde; Dow solidification process; low-viscosity chemical agents (POLYPAC); and water-extensible polyester. 9 refs

  3. Process gas solidification system

    International Nuclear Information System (INIS)

    1980-01-01

    A process for withdrawing gaseous UF 6 from a first system and directing same into a second system for converting the gas to liquid UF 6 at an elevated temperature, additionally including the step of withdrawing the resulting liquid UF 6 from the second system, subjecting it to a specified sequence of flash-evaporation, cooling and solidification operations, and storing it as a solid in a plurality of storage vessels. (author)

  4. Solidification process for sludge residue

    International Nuclear Information System (INIS)

    Pearce, K.L.

    1998-01-01

    This report investigates the solidification process used at 100-N Basin to solidify the N Basin sediment and assesses the N Basin process for application to the K Basin sludge residue material. This report also includes a discussion of a solidification process for stabilizing filters. The solidified matrix must be compatible with the Environmental Remediation Disposal Facility acceptance criteria

  5. Thermal processing systems for TRU mixed waste

    International Nuclear Information System (INIS)

    Eddy, T.L.; Raivo, B.D.; Anderson, G.L.

    1992-01-01

    This paper presents preliminary ex situ thermal processing system concepts and related processing considerations for remediation of transuranic (TRU)-contaminated wastes (TRUW) buried at the Radioactive Waste Management Complex (RWMC) of the Idaho National Engineering Laboratory (INEL). Anticipated waste stream components and problems are considered. Thermal processing conditions required to obtain a high-integrity, low-leachability glass/ceramic final waste form are considered. Five practical thermal process system designs are compared. Thermal processing of mixed waste and soils with essentially no presorting and using incineration followed by high temperature melting is recommended. Applied research and development necessary for demonstration is also recommended

  6. Advances in Solidification Processing

    Directory of Open Access Journals (Sweden)

    Hugo F. Lopez

    2015-08-01

    Full Text Available Melt solidification is the shortest and most viable route to obtain components, starting from the design to the finished products. Hence, a sound knowledge of the solidification of metallic materials is essential for the development of advanced structural metallic components that drive modern technological societies. As a result, there have been innumerable efforts and full conferences dedicated to this important subject [1–6]. In addition, there are various scientific journals fully devoted to investigating the various aspects which give rise to various solidification microstructures [7–9]. [...

  7. Centralized processing of contact-handled TRU waste feasibility analysis

    International Nuclear Information System (INIS)

    1986-12-01

    This report presents work for the feasibility study of central processing of contact-handled TRU waste. Discussion of scenarios, transportation options, summary of cost estimates, and institutional issues are a few of the subjects discussed

  8. EXAMPLE OF A RISK-BASED DISPOSAL APPROVAL: SOLIDIFICATION OF HANFORD SITE TRANSURANIC (TRU) WASTE

    International Nuclear Information System (INIS)

    PRIGNANO AL

    2007-01-01

    The Hanford Site requested, and the U.S. Environmental Protection Agency (EPA) Region 10 approved, a Toxic Substances Control Act of 1976 (TSCA) risk-based disposal approval (RBDA) for solidifying approximately four cubic meters of waste from a specific area of one of the K East Basin: the North Loadout Pit (NLOP). The NLOP waste is a highly radioactive sludge that contained polychlorinated biphenyls (PCBs) regulated under TSCA. The prescribed disposal method for liquid PCB waste under TSCA regulations is either thermal treatment or decontamination. Due to the radioactive nature of the waste, however, neither thermal treatment nor decontamination was a viable option. As a result, the proposed treatment consisted of solidifying the material to comply with waste acceptance criteria at the Waste Isolation Pilot Plant (WPP) in Carlsbad, New Mexico, or possibly the Environmental Restoration Disposal Facility at the Hanford Site, depending on the resulting transuranic (TRU) content of the stabilized waste. The RBDA evaluated environmental risks associated with potential airborne PCBs. In addition, the RBDA made use of waste management controls already in place at the treatment unit. The treatment unit, the T Plant Complex, is a Resource Conservation and Recovery Act of 1976 (RCRA)-permitted facility used for storing and treating radioactive waste. The EPA found that the proposed activities did not pose an unreasonable risk to human health or the environment. Treatment took place from October 26,2005 to June 9,2006, and 332 208-liter (55-gallon) containers of solidified waste were produced. All treated drums assayed to date are TRU and will be disposed at WIPP

  9. Radiological Design Summary Report for TRU Vent and Purge Process

    International Nuclear Information System (INIS)

    Taus, L.B.

    2004-01-01

    This report contains top-level requirements for the various areas of radiological protection for workers. Detailed quotations of the requirements for applicable regulatory documents can be found in the accompanying Implementation Guide. For the purposes of demonstrating compliance with these requirements, per Engineering Standard 01064, shall consider / shall evaluate indicates that the designer must examine the requirement for the design and either incorporate or provide a technical justification as to why the requirement is not incorporated. The Transuranic Vent and Purge process is not a project, but is considered a process change. This process has been performed successfully by Solid Waste on lower activity TRU drums. This summary report applies a graded approach and describes how the Transuranic Vent and Purge process meets each of the applicable radiological design criteria and requirements specified in Manual WSRC-TM-95-1, Engineering Standard Number 01064

  10. TRU waste processing comparison: slagging pyrolysis versus modified glassmaker

    International Nuclear Information System (INIS)

    Bonner, W.F.; Cox, N.D.; Hootman, H.E.; Nelson, D.C.; Pye, D.

    1980-03-01

    A task force was assembled to make a technical comparison of the expected performance of two processing systems potentially applicable for treating TRU waste at the Idaho National Engineering Laboratory. One system contained a slagging pyrolysis incinerator; the other a modified Penberthy Electromelt glassmaker. Although the glassmaker technology is essentially undeveloped, it was assumed that the glassmaker could eventually be modified to operate as a combined waste incinerator and melter; that is, to perform the same functions as a slagger. Using a decision analysis methodology to evaluate figures-of-merit, the task force found no significant difference in the performance of the two systems. Some areas for future R and D efforts are recommended for both types of incinerators

  11. Sandia solidification process: a broad range aqueous waste solidification method

    International Nuclear Information System (INIS)

    Lynch, R.W.; Dosch, R.G.; Kenna, B.T.; Johnstone, J.K.; Nowak, E.J.

    1976-01-01

    New ion-exchange materials of the hydrous oxide type were developed for solidifying aqueous radioactive wastes. These materials have the general formula M[M'/sub x/O/sub y/H/sub z/]/sub n/, where M is an exchangeable cation of charge +n and M' may be Ti; Nb; Zr, or Ta. Affinities for polyvalent cations were found to be very high and ion-exchange capacities large (e.g., 4.0--4.5 meq/g for NaTi 2 O 5 H depending on moisture content). The effectiveness of the exchangers for solidifying high-level waste resulting from reprocessing light-water reactor fuel was demonstrated in small-scale tests. Used in conjunction with anion exchange resin, these materials reduced test solution radioactivity from approximately 0.2 Ci/ml to as low as approximately 2 nCi/ml. The residual radioactivity was almost exclusively due to 106 Ru and total α-activity was only a few pCi/ml. Alternative methods of consolidating the solidified waste were evaluated using nonradioactive simulants. Best results were obtained by pressure-sintering which yielded essentially fully dense ceramics, e.g., titanate/titania ceramics with bulk density as high as 4.7 g/cm 3 , waste oxide content as high as 1.2 g/cm 3 , and leach resistance comparable to good borosilicate glass. Based on the above results, a baseline process for solidifying high-level waste was defined and approximate economic analyses indicated costs were not prohibitive. Additional tests have demonstrated that, if desired, operating conditions could be modified to allow recovery of radiocesium (and perhaps other isotopes) during solidification of the remaining constituents of high-level waste. Preliminary tests have also shown that these materials offer promise for treating tank-stored neutralized wastes

  12. Solidification processing of high-Tc superconductors

    CERN Document Server

    Shiohara, Y; Nakamura, Y; Izumi, T

    2001-01-01

    Recent progress in the solidification processing of RE-system (RE:Y, Sm, Nd etc.) oxide superconducting materials is reviewed. The superconducting YBa/sub 2/Cu/sub 3/O/sub y/(Y123) phase is solidified from Y/sub 2/BaCuO/sub 5/(Y211) and liquid phases, by a peritectic reaction. The solidified micro and macro structure can not be explained by the peritectic reaction with diffusion in the solid but rather by diffusion in the liquid. A solidification model for this reaction is developed. It is confirmed that the prediction from the model calculation is in good agreement with the experimental results. Furthermore, the basic idea is expanded to develop a novel single crystal pulling process. Y211 powders were placed at the bottom of the crucible as the solute source for the growth and a BaO-CuO composite (Ba to Cu cation ratio was 3 to 5) was placed on the layer of Y211 powders. Temperature gradient was provided in the melt. Large bulk single crystals were obtained by this technique, and the growth mechanism was al...

  13. An approach for the reasonable TRU waste management in NUCEF

    International Nuclear Information System (INIS)

    Mineo, H.; Dojiri, S.; Takeshita, I.; Tsujino, T.; Matsumura, T.; Nishizawa, I.; Sugikawa, S.

    1995-01-01

    The Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF) has started its hot operation at the beginning of 1995, where TRU (transuranic) elements are used. The management of TRU waste arisen in the facility is very important issue. Liquid and solid wastes containing TRU elements are generated mainly from the Fuel Treatment System for critical experiments and from the researches of reprocessing process and TRU waste management for reprocessing plants using hot cells and glove-boxes. The TRU waste management in NUCEF is based on the classification of waste, and is to maximize the recycle of reagents and the reuse of TRU elements separated from the waste, as well as to reduce the waste volume and to lower the risk of waste by advanced separation and solidification. In the future, the separation and solidification of TRU elements in the tanks of liquid waste, and the classification and discrimination of solid wastes, will be carried out applying the outcomes of the development by the researches in NUCEF. (authors)

  14. The TRUEX [TRansUranium EXtraction] process and the management of liquid TRU [transuranic] waste

    International Nuclear Information System (INIS)

    Schulz, W.W.; Horwitz, E.P.

    1987-01-01

    The TRUEX process is a new generic liquid-liquid extraction process for removal of all actinides from acidic nitrate or chloride nuclear waste solutions. Because of its high efficiency and great flexibility, the TRUEX process appears destined to be widely used in the US and possibly in other countries for cost-effective management and disposal of transuranic (TRU) wastes. In the US, TRU wastes are those that contain ≥3.7 x 10 6 Bq/kg) of TRU elements with half-lives greater than 20 y. This paper gives a brief review of the relevant chemistry and summarizes the current status of development and deployment of the TRUEX (TRansUranium EXtraction) process flowsheets to treat specific acidic waste solutions at several US Department of Energy sites. 19 refs., 4 figs., 4 tabs

  15. Solidification process for toxic and hazardous wastes. Second part: Cement solidification matrices

    International Nuclear Information System (INIS)

    Donato, A.; Arcuri, L.; Dotti, M.; Pace, A.; Pietrelli, L.; Ricci, G.; Basta, M.; Cali, V.; Pagliai, V.

    1989-05-01

    This paper reports the second part of a general study carried out at the Nuclear Fuel Division aiming at verifying the possible application of the radioactive waste solidification processes to industrial hazardous wastes (RTN). The cement solidification of several RTN types has been taken into consideration, both from the technical and from the economic point of view. After a short examination of the Italian juridical and economical situation in the field, which demonstrates the need of the RTN solidification, the origin and characteristics of the RTN considered in the study and directly provided by the producing industries are reviewed. The laboratory experimental results of the cementation of RTN produced by gold manufacturing industries and by galvanic industries are reported. The cementation process can be considered a very effective mean for reducing both the RTN management costs and the environmental impact of RTN disposal. (author)

  16. Comparative assessment of TRU waste forms and processes. Volume I. Waste form and process evaluations

    International Nuclear Information System (INIS)

    Ross, W.A.; Lokken, R.O.; May, R.P.; Roberts, F.P.; Timmerman, C.L.; Treat, R.L.; Westsik, J.H. Jr.

    1982-09-01

    This study provides an assesses seven waste forms and eight processes for immobilizing transuranic (TRU) wastes. The waste forms considered are cast cement, cold-pressed cement, FUETAP (formed under elevated temperature and pressure) cement, borosilicate glass, aluminosilicate glass, basalt glass-ceramic, and cold-pressed and sintered silicate ceramic. The waste-immobilization processes considered are in-can glass melting, joule-heated glass melting, glass marble forming, cement casting, cement cold-pressing, FUETAP cement processing, ceramic cold-pressing and sintering, basalt glass-ceramic processing. Properties considered included gas generation, chemical durability, mechanical strength, thermal stability, and radiation stability. The ceramic products demonstrated the best properties, except for plutonium release during leaching. The glass and ceramic products had similar properties. The cement products generally had poorer properties than the other forms, except for plutonium release during leaching. Calculations of the Pu release indicated that the waste forms met the proposed NRC release rate limit of 1 part in 10 5 per year in most test conditions. The cast-cement process had the lowest processing cost, followed closely by the cold-pressed and FUETAP cement processes. Joule-heated glass melting had the lower cost of the glass processes. In-can melting in a high-quality canister had the highest cost, and cold-pressed and sintered ceramic the second highest. Labor and canister costs for in-can melting were identified. The major contributor to costs of disposing of TRU wastes in a defense waste repository is waste processing costs. Repository costs could become the dominant cost for disposing of TRU wastes in a commercial repository. It is recommended that cast and FUETAP cement and borosilicate glass waste-form systems be considered. 13 figures, 16 tables

  17. NPP radioactive waste processing and solidification

    International Nuclear Information System (INIS)

    Nikiforov, A.S.; Polyakov, A.S.; Zakharova, K.P.

    1983-01-01

    The problems of proce-sing NPP intermediate level- and low-level liquid radioactive wastes (LRW) are considered. Various methods are compared of LWR solidification on the base of bituminization, cement grouting and inclusion into synthetic resins. It is concluded that the considered methods ensure radioactive radionuclides effluents into open hydronetwork at the level below the sanitary, standards

  18. STRONTIUM & TRANSURANIC (TRU) SEPARATION PROCESS IN THE DOUBLE SHELL TANK (DST) SYSTEM

    Energy Technology Data Exchange (ETDEWEB)

    JOHNSON; SWANSON; BOECHLER

    2005-06-10

    The supernatants stored in tanks 241-AN-102 (AN-102) and 241-AN-107 (AN-107) contain soluble strontium-90 ({sup 90}Sr) and transuranic (TRU) elements that require removal prior to vitrification to comply with the Waste Treatment and Immobilization Plant (WTP) immobilized low-activity waste (ILAW) specification and with the 1997 agreement with the Nuclear Regulatory Commission on incidental waste. A precipitation process has been developed and tested with tank waste samples and simulants using strontium nitrate (Sr(NO{sub 3}){sub 2}) and sodium permanganate (NaMnO{sub 4}) to separate {sup 90}Sr and TRU from these wastes. This report evaluates removing Sr/TRU from AN-102 and AN-107 supernates in the DST system before delivery to the WTP. The in-tank precipitation is a direct alternative to the baseline WTP process, using the same chemical separations. Implementing the Sr/TRU separation in the DST system beginning in 2012 provides {approx}6 month schedule advantage to the overall mission, without impacting the mission end date or planned SST retrievals.

  19. Solidification processing of monotectic alloy matrix composites

    Science.gov (United States)

    Frier, Nancy L.; Shiohara, Yuh; Russell, Kenneth C.

    1989-01-01

    Directionally solidified aluminum-indium alloys of the monotectic composition were found to form an in situ rod composite which obeys a lambda exp 2 R = constant relation. The experimental data shows good agreement with previously reported results. A theoretical boundary between cellular and dendritic growth conditions was derived and compared with experiments. The unique wetting characteristics of the monotectic alloys can be utilized to tailor the interface structure in metal matrix composites. Metal matrix composites with monotectic and hypermonotectic Al-In matrices were made by pressure infiltration, remelted and directionally solidified to observe the wetting characteristics of the alloys as well as the effect on structure of solidification in the constrained field of the fiber interstices. Models for monotectic growth are modified to take into account solidification in these constrained fields.

  20. Waste Disposition Issues and Resolutions at the TRU Waste Processing Center at Oak Ridge TN

    International Nuclear Information System (INIS)

    Gentry, R.

    2009-01-01

    This paper prepared for the Waste Management Conference 2009 provides lessons learned from the Transuranic (TRU) Waste Processing Center (TWPC) associated with development of approaches used to certify and ensure disposition of problematic TRU wastes at the Waste Isolation Pilot Plant (WIPP) site. The TWPC is currently processing the inventory of available waste TRU waste at the Oak Ridge National Lab (ORNL). During the processing effort several waste characteristics were identified/discovered that did not conform to the normal standards and processes for disposal at WIPP. Therefore, the TWPC and ORNL were challenged with determining a path forward for this problematic, special case TRU wastes to ensure that they can be processed, packaged, and shipped to WIPP. Additionally, unexpected specific waste characteristics have challenged the project to identify and develop processing methods to handle problematic waste. The TWPC has several issues that have challenged the projects ability to process RH Waste. High Neutron Dose Rate resulting from both Californium and Curium in the waste stream challenge the RH-TRU 72-B limit for dose rate measured from the side of the package under normal conditions of transport, as specified in Chapter 5.0 of the RH-TRU 72-B SAR (i.e., ≤10 mrem/hour at 2 meters). Difficult to process waste in the hot cell has introduced processing and handling difficulties included problems associated with the disposition of prohibited items that fall out of the waste stream such as liquids, aerosol cans, etc. Lastly, multiple waste streams require characterization and AK challenge the ability to generate dose-to curie models for the waste. Repackaging is one solution to the high neutron dose rate issue. In parallel, an effort is underway to request a change to the TRAMPAC requirements to allow shielding in the drum or canister to reduce the impact of the high neutron dose rates. Due diligence on supporting AK efforts is important in ensuring adequate

  1. Demonstration of Entrained Solids and Sr/TRU Removal Processes with Archived AN-107 Waste

    International Nuclear Information System (INIS)

    Hallen, R.T.; Brooks, K.P.; Jagoda, L.K.

    2000-01-01

    Archived AN-107 waste was used to evaluate entrained solids removal, Sr/TRU decontamination of supernatant, and Sr/TRU solids removal. Even though most of the entrained solids had been previously removed from the archived sample, the residual entrained solids rapidly fouled the filter element resulting in very poor filter performance. An attempt to run at higher pressure resulted in more fouling, and reduced filter performance. Filtration efforts to remove entrained solids were abandoned and the waste was treated for Sr/TRU removal with the entrained solids present. The new processing scheme for Sr/TRU removal involving precipitation by added strontium and permanganate worked well. The decontamination factors for Sr and TRU components were significantly greater than the ILAW DF requirements for higher reagent concentrations of 1M hydroxide, 0.075M Sr, and 0.05M permanganate and lower reagent concentrations of 0.8M hydroxide, 0.05M Sr, and 0.03M permanganate. These results support the use of lower concentration of reagent additions in future tests. Optimization studies should be conducted to examine the reduction in added hydroxide from 1M to 0.5 M, reduction of Sr from 0.075M to 0.05M, and reduction in permanganate from 0.05M to 0.03M and the impact this reduction has on filtration performance with new samples from Tank AN-107. The combined entrained solids and Sr/TRU precipitate were successfully filtered in the single element, crossflow filtration unit. The filtrate flux was high, >0.1 gpm/ft 2 , at the initial test conditions of 53 psi and 11.2ft/s for the treated archived AN-107 sample. The filter flux rate dropped significantly with time as testing progressed and appears to be a result of shearing the agglomerated solids and fouling of the filter element by the resulting fine particles. The relatively low clean water flux rates obtained at the end of the test also indicate filter fouling. Chemical cleaning was required to restore clean water flux rates to pre

  2. Solidification processing of intermetallic Nb-Al alloys

    Science.gov (United States)

    Smith, Preston P.; Oliver, Ben F.; Noebe, Ronald D.

    1992-01-01

    Several Nb-Al alloys, including single-phase NbAl3 and the eutectic of Nb2Al and NbAl3, were prepared either by nonconsumable arc melting in Ar or by zone processing in He following initial induction melting and rod casting, and the effect of the solidification route on the microstructure and room-temperature mechanical properties of these alloys was investigated. Automated control procedures and melt conditions for directional solidification of NbAl3 and the Nb2Al/Nb3Al eutectic were developed; high purity and stoichiometry were obtained. The effects of ternary additions of Ti and Ni are described.

  3. Transuranic Waste Processing Center (TWPC) Legacy Tank RH-TRU Sludge Processing and Compliance Strategy - 13255

    Energy Technology Data Exchange (ETDEWEB)

    Rogers, Ben C.; Heacker, Fred K.; Shannon, Christopher [Wastren Advantage, Inc., Transuranic Waste Processing Center, 100 WIPP Road, Lenoir City, Tennessee 37771 (United States); and others

    2013-07-01

    The U.S. Department of Energy (DOE) needs to safely and efficiently treat its 'legacy' transuranic (TRU) waste and mixed low-level waste (LLW) from past research and defense activities at the Oak Ridge National Laboratory (ORNL) so that the waste is prepared for safe and secure disposal. The TWPC operates an Environmental Management (EM) waste processing facility on the Oak Ridge Reservation (ORR). The TWPC is classified as a Hazard Category 2, non-reactor nuclear facility. This facility receives, treats, and packages low-level waste and TRU waste stored at various facilities on the ORR for eventual off-site disposal at various DOE sites and commercial facilities. The Remote Handled TRU Waste Sludge held in the Melton Valley Storage Tanks (MVSTs) was produced as a result of the collection, treatment, and storage of liquid radioactive waste originating from the ORNL radiochemical processing and radioisotope production programs. The MVSTs contain most of the associated waste from the Gunite and Associated Tanks (GAAT) in the ORNL's Tank Farms in Bethel Valley and the sludge (SL) and associated waste from the Old Hydro-fracture Facility tanks and other Federal Facility Agreement (FFA) tanks. The SL Processing Facility Build-outs (SL-PFB) Project is integral to the EM cleanup mission at ORNL and is being accelerated by DOE to meet updated regulatory commitments in the Site Treatment Plan. To meet these commitments a Baseline (BL) Change Proposal (BCP) is being submitted to provide continued spending authority as the project re-initiation extends across fiscal year 2012 (FY2012) into fiscal year 2013. Future waste from the ORNL Building 3019 U-233 Disposition project, in the form of U-233 dissolved in nitric acid and water, down-blended with depleted uranyl nitrate solution is also expected to be transferred to the 7856 MVST Annex Facility (formally the Capacity Increase Project (CIP) Tanks) for co-processing with the SL. The SL-PFB project will construct

  4. Transuranic Waste Processing Center (TWPC) Legacy Tank RH-TRU Sludge Processing and Compliance Strategy - 13255

    International Nuclear Information System (INIS)

    Rogers, Ben C.; Heacker, Fred K.; Shannon, Christopher

    2013-01-01

    The U.S. Department of Energy (DOE) needs to safely and efficiently treat its 'legacy' transuranic (TRU) waste and mixed low-level waste (LLW) from past research and defense activities at the Oak Ridge National Laboratory (ORNL) so that the waste is prepared for safe and secure disposal. The TWPC operates an Environmental Management (EM) waste processing facility on the Oak Ridge Reservation (ORR). The TWPC is classified as a Hazard Category 2, non-reactor nuclear facility. This facility receives, treats, and packages low-level waste and TRU waste stored at various facilities on the ORR for eventual off-site disposal at various DOE sites and commercial facilities. The Remote Handled TRU Waste Sludge held in the Melton Valley Storage Tanks (MVSTs) was produced as a result of the collection, treatment, and storage of liquid radioactive waste originating from the ORNL radiochemical processing and radioisotope production programs. The MVSTs contain most of the associated waste from the Gunite and Associated Tanks (GAAT) in the ORNL's Tank Farms in Bethel Valley and the sludge (SL) and associated waste from the Old Hydro-fracture Facility tanks and other Federal Facility Agreement (FFA) tanks. The SL Processing Facility Build-outs (SL-PFB) Project is integral to the EM cleanup mission at ORNL and is being accelerated by DOE to meet updated regulatory commitments in the Site Treatment Plan. To meet these commitments a Baseline (BL) Change Proposal (BCP) is being submitted to provide continued spending authority as the project re-initiation extends across fiscal year 2012 (FY2012) into fiscal year 2013. Future waste from the ORNL Building 3019 U-233 Disposition project, in the form of U-233 dissolved in nitric acid and water, down-blended with depleted uranyl nitrate solution is also expected to be transferred to the 7856 MVST Annex Facility (formally the Capacity Increase Project (CIP) Tanks) for co-processing with the SL. The SL-PFB project will construct and install

  5. TRUEX process: a new dimension in management of liquid TRU wastes

    International Nuclear Information System (INIS)

    Schulz, W.W.; Horwitz, E.P.

    1986-01-01

    The TRUEX process is one of the, if not the, most exciting and potentially useful nuclear separations processes to be developed since the PUREX process was developed and applied in the 1950s. Attesting to its potential widespread use, Rockwell Hanford and ANL investigators, in a joint effort, are developing and testing TRUEX process flow sheets for removal of TRU elements from several Hanford Site wastes including the Plutonium Finishing Plant and complexed concentrate wastes. The TRUEX process also appears to be well suited to removal of plutonium and Am from aqueous chloride wastes generated during plutonium processing operations at the Los Alamos National Lab. (LANL); collaborative efforts between LANL and ANL scientists to develop and demonstrate TRUEX process flow sheets for treatment of LANL site chloride wastes are currently under way

  6. A Cask Processing Enclosure for the TRU Waste Processing Center - 13408

    Energy Technology Data Exchange (ETDEWEB)

    Newman, John T.; Mendez, Nicholas [IP Systems, Inc., 2685 Industrial Lane, Broomfield, Colorado 80020 (United States)

    2013-07-01

    This paper will discuss the key elements considered in the design, construction, and use of an enclosure system built for the TRU Waste Processing Center (TWPC). The TWPC system is used for the repackaging and volume reduction of items contaminated with radioactive material, hazardous waste and mixed waste. The modular structural steel frame and stainless steel skin was designed for rapid field erection by the use of interchangeable self-framing panel sections to allow assembly of a sectioned containment building and for ease of field mobility. The structure was installed on a concrete floor inside of an outer containment building. The major sections included an Outer Cask Airlock, Inner Cask Airlock, Cask Process Area, and Personnel Airlocks. Casks in overpacks containing transuranic waste are brought in via an inter-site transporter. The overpack lid is removed and the cask/overpack is transferred into the Outer Cask Airlock. A contamination cover is installed on the overpack body and the Outer Cask Airlock is closed. The cask/overpack is transferred into the Inner Cask Airlock on a cask bogie and the Inner Cask Airlock is closed. The cask lid is removed and the cask is transferred into the Cask Process Area where it is placed on a cask tilting station. Once the Cask Processing Area is closed, the cask tilt station is activated and wastes are removed, size reduced, then sorted and re-packaged into drums and standard waste boxes through bag ports. The modular system was designed and built as a 'Fast Track' project at IP Systems in Broomfield Colorado and then installed and is currently in use at the DOE TWPC located near Oak Ridge, Tennessee. (authors)

  7. Study of plastic solidification process on solid radioactive waste treatment

    International Nuclear Information System (INIS)

    Jing Weiguan; Zhang Yinsheng; Qian Wenju

    1994-01-01

    Comparisons between the plastic solidification conditions of incinerated ash and waste cation resin by using thermosetting plastic polyvinyl chloride (PVC), polystyrene (PS) and polyethylene (PE), and identified physico-chemical properties and irradiation resistance of solidified products were presented. These solidified products have passed through different tests as compression strength, leachability, durability, stability, permeability and irradiation resistance (10 6 Gy) etc. The result showed that the solidified products possessed stable properties and met the storage requirement. The waste tube of radioimmunoassay, being used as solidification medium to contain incinerated ash, had good mechanical properties and satisfactory volume reduction. This process may develop a new way for disposal solid radioactive waste by means of re-using waste

  8. Modeling of solidification of MMC composites during gravity casting process

    Directory of Open Access Journals (Sweden)

    R. Zagórski

    2013-04-01

    Full Text Available The paper deals with computer simulation of gravity casting of the metal matrix composites reinforced with ceramics (MMC into sand mold. The subject of our interest is aluminum matrix composite (AlMMC reinforced with ceramic particles i.e. silicon carbide SiC and glass carbon Cg. The created model describes the process taking into account solidification and its influence on the distribution of reinforcement particles. The computer calculation has been carried out in 2D system with the use of Navier-Stokes equations using ANSYS FLUENT 13. The Volume of Fluid approach (VOF and enthalpy method have been used to model the air-fluid free surface (and also volume fraction of particular continuous phases and the solidification of the cast, respectively.

  9. Solidification in direct metal deposition by LENS processing

    Science.gov (United States)

    Hofmeister, William; Griffith, Michelle

    2001-09-01

    Thermal imaging and metallographic analysis were used to study Laser Engineered Net Shaping (LENS™) processing of 316 stainless steel and H13 tool steel. The cooling rates at the solid-liquid interface were measured over a range of conduction conditions. The length scale of the molten zone controls cooling rates during solidification in direct metal deposition. In LENS processing, the molten zone ranges from 0.5 mm in length to 1.5 mm, resulting in cooling rates at the solid-liquid interface ranging from 200 6,000 Ks-1.

  10. A process of spent nuclear fuel treatment with the interim storage of TRU by use amidic extractants

    International Nuclear Information System (INIS)

    Tachimori, Shoichi; Suzuki, Shinichi; Sasaki, Yuji

    2001-01-01

    A new chemical process, ARTIST process, is proposed for the treatment of spent nuclear fuel. The main concept of the ARTIST process is to recover and stock separately all actinides, uranium and a mixture of transuranics, and to dispose fission products. The process composed of two main steps, a uranium exclusive isolation and a total recovery of transuranium elements (TRU); which copes with the nuclear non-proliferation measures, and additional processes. Both actinide products are solidified by calcination and allowed to the interim storage for future utilization. These separations are achieved by use of amidic extractants in accord with the CHON principle. The technical feasibility of the ARTIST process was explained by the experimental results of both the branched-alkyl monoamides in extracting uranium and suppressing the extraction of tetravalent actinides due to the steric effect and the diglycolic amide in thorough extraction of all TRU by tridentate coordination. When these TRU are requested to put into reactors, LWR or FBR, for power generation or the Accelerator-Driven System (ADS) for transmutation, lanthanides are to be removed from TRU by utilizing a soft nitrogen donor ligand. (author)

  11. Solidification process of a tool steel with niobium

    International Nuclear Information System (INIS)

    Makray, E.T.; Bresciani Filho, E.; Martinez Nazar, A.M.

    1984-01-01

    The solidification process of M2 high speed steel where tungsten was totally substituted by niobium was analysed. It occurs through a eutectic type reaction, in four steps. It was verified that one can apply the Coupled Zone Concept to explain the solification mechanism of this alloy: there is a primary phase (NbC), which is envolved by the other phase (ferrite) as a halo in order to send the composition back to the coupled growth region, where the binary eutectic forms. The last step is the formation of other compounds at the grain boundary. (Author) [pt

  12. Method of processing solidification product of radioactive waste

    International Nuclear Information System (INIS)

    Daime, Fumiyoshi.

    1988-01-01

    Purpose: To improve the long-time stability of solidification products by providing solidification products with liquid tightness, gas tightness, abrasion resistance, etc., of the products in the course of the solidification for the treatment of radioactive wastes. Method: The surface of solidification products prepared by mixing solidifying agents with powder or pellets is entirely covered with high molecular polymer such as epoxy resin. The epoxy resin has excellent properties such as radiation-resistance, heat resistance, water proofness and chemical resistance, as well as have satisfactory mechanical properties. This can completely isolate the solidification products of radioactive wastes from the surrounding atmosphere. (Yoshino, Y.)

  13. Microstructure and Corrosion Resistance Property of a Zn-AI-Mg Alloy with Different Solidification Processes

    Directory of Open Access Journals (Sweden)

    Jiang Guang-rui

    2017-01-01

    Full Text Available Zn-Al-Mg alloy coating attracted much attention due to its high corrosion resistance properties, especially high anti-corrosion performance at the cut edge. As the Zn-Al-Mg alloy coating was usually produced by hot-dip galvanizing method, solidification process was considered to influence its microstructure and corrosion properties. In this work, a Zn-Al-Mg cast alloy was melted and cooled to room temperature with different solidification processes, including water quench, air cooling and furnace cooling. Microstructure of the alloy with different solidification processes was characterized by scanning electron microscopy (SEM. Result shows that the microstructure of the Zn-Al-Mg alloy are strongly influenced by solidification process. With increasing solidification rate, more Al is remained in the primary crystal. Electrochemical analysis indicates that with lowering solidification rate, the corrosion current density of the Zn-Al-Mg alloy decreases, which means higher corrosion resistance.

  14. Real-Time Detection Methods to Monitor TRU Compositions in UREX+Process Streams

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean; Charlton, William; Indacochea, J Ernesto; taleyarkhan, Rusi; Pereira, Candido

    2013-03-01

    The U.S. Department of Energy has developed advanced methods for reprocessing spent nuclear fuel. The majority of this development was accomplished under the Advanced Fuel Cycle Initiative (AFCI), building on the strong legacy of process development R&D over the past 50 years. The most prominent processing method under development is named UREX+. The name refers to a family of processing methods that begin with the Uranium Extraction (UREX) process and incorporate a variety of other methods to separate uranium, selected fission products, and the transuranic (TRU) isotopes from dissolved spent nuclear fuel. It is important to consider issues such as safeguards strategies and materials control and accountability methods. Monitoring of higher actinides during aqueous separations is a critical research area. By providing on-line materials accountability for the processes, covert diversion of the materials streams becomes much more difficult. The importance of the nuclear fuel cycle continues to rise on national and international agendas. The U.S. Department of Energy is evaluating and developing advanced methods for safeguarding nuclear materials along with instrumentation in various stages of the fuel cycle, especially in material balance areas (MBAs) and during reprocessing of used nuclear fuel. One of the challenges related to the implementation of any type of MBA and/or reprocessing technology (e.g., PUREX or UREX) is the real-time quantification and control of the transuranic (TRU) isotopes as they move through the process. Monitoring of higher actinides from their neutron emission (including multiplicity) and alpha signatures during transit in MBAs and in aqueous separations is a critical research area. By providing on-line real-time materials accountability, diversion of the materials becomes much more difficult. The objective of this consortium was to develop real time detection methods to monitor the efficacy of the UREX+ process and to safeguard the separated

  15. Development of an integrated facility for processing TRU solid wastes at the Savannah River Plant

    International Nuclear Information System (INIS)

    Boersma, M.D.; Hootman, H.E.; Permar, P.H.

    1977-01-01

    An integrated facility is being designed for processing solid wastes contaminated with long-lived alpha emitting (TRU) nuclides; this waste has been stored retrievably at the Savannah River Plant since 1965. The stored waste, having a volume of 10 4 m 3 and containing 3 x 10 5 Ci of transuranics, consists of both mixed combustible trash and failed and obsolete equipment primarily from transuranic production and associated laboratory operations. The facility for processing solid transuranic waste will consist of five processing modules: (1) unpackaging, sorting, and assaying; (2) treatment of combustibles by controlled air incineration; (3) size reduction of noncombustibles by plasma-arc cutting followed by decontamination by electropolishing; (4) fixation of the processed waste in cement; and (5) packaging for shipment to a federal repository. The facility is projected for construction in the mid-1980's. Pilot facilities, sized to manage currently generated wastes, will also demonstrate the key process steps of incineration of combustibles and size reduction/decontamination of noncombustibles; these facilities are projected for 1980-81. Development programs leading to these extensive new facilities are described

  16. Solidification, processing and properties of ductile cast iron

    DEFF Research Database (Denmark)

    Tiedje, Niels Skat

    2010-01-01

    Ductile cast iron has been an important engineering material in the past 50 years. In that time, it has evolved from a complicated material that required the foundry metallurgist's highest skill and strict process control to being a commonly used material that can easily be produced with modern...... of the latest years of research indicate that ductile cast iron in the future will become a highly engineered material in which strict control of a range of alloy elements combined with intelligent design and highly advanced processing allows us to target properties to specific applications to a much higher...... degree than we have seen previously. It is the aim of the present paper to present ductile iron as a modern engineering material and present the many different possibilities that the material hides. Focus will be on the latest research in solidification and melt treatment. But for completeness...

  17. Repackaging of High Fissile TRU Waste at the Transuranic Waste Processing Center - 13240

    Energy Technology Data Exchange (ETDEWEB)

    Oakley, Brian; Heacker, Fred [WAI, TRU Waste Processing Center, 100 WIPP Road Lenoir City, TN 37771 (United States); McMillan, Bill [DOE, Oak Ridge Operations, Bldg. 2714, Oak Ridge, TN 37830 (United States)

    2013-07-01

    Twenty-six drums of high fissile transuranic (TRU) waste from Oak Ridge National Laboratory (ORNL) operations were declared waste in the mid-1980's and placed in storage with the legacy TRU waste inventory for future treatment and disposal at the Waste Isolation Pilot Plant (WIPP). Repackaging and treatment of the waste at the TRU Waste Packaging Center (TWPC) will require the installation of additional equipment and capabilities to address the hazards for handling and repackaging the waste compared to typical Contact Handled (CH) TRU waste that is processed at the TWPC, including potential hydrogen accumulation in legacy 6M/2R packaging configurations, potential presence of reactive plutonium hydrides, and significant low energy gamma radiation dose rates. All of the waste is anticipated to be repackaged at the TWPC and certified for disposal at WIPP. The waste is currently packaged in multiple layers of containers which presents additional challenges for repackaging activities due to the potential for the accumulation of hydrogen gas in the container headspace in quantities than could exceed the Lower Flammability Limit (LFL). The outer container for each waste package is a stainless steel 0.21 m{sup 3} (55-gal) drum which contains either a 0.04 m{sup 3} or 0.06 m{sup 3} (10-gal or 15-gal) 6M drum. The inner 2R container in each 6M drum is ∼12 cm (5 in) outside diameter x 30-36 cm (12-14 in) long and is considered to be a > 4 liter sealed container relative to TRU waste packaging criteria. Inside the 2R containers are multiple configurations of food pack cans, pipe nipples, and welded capsules. The waste contains significant quantities of high burn-up plutonium oxides and metals with a heavy weight percentage of higher atomic mass isotopes and the subsequent in-growth of significant quantities of americium. Significant low energy gamma radiation is expected to be present due to the americium in-growth. Radiation dose rates on inner containers are estimated

  18. Demonstration of a remotely operated TRU waste size-reduction and material handling process

    International Nuclear Information System (INIS)

    Stewart, J.A. III; Schuler, T.F.; Ward, C.R.

    1986-01-01

    Noncombustible Pu-238 and Pu-239 waste is generated as a result of normal operation and decommissioning activity at the Savannah River Plant and is being retrievably stored at the site. As part of the long-term plan to process the stored waste and current waste for permanent disposal, a remote size-reduction and material handling process is being tested at Savannah River Laboratory to provide design support for the plant TRU Waste Facility scheduled to be completed in 1993. The process consists of a large, low-speed shredder and material handling system, a remote worktable, a bagless transfer system, and a robotically controlled manipulator, or Telerobot. Initial testing of the shredder and material handling system and a cycle test of the bagless transfer system were completed. Initial Telerobot run-in and system evaluation was completed. User software was evaluated and modified to support complete menu-driven operation. Telerobot prototype size-reduction tooling was designed and successfully tested. Complete nonradioactive testing of the equipment is scheduled to be completed in 1987

  19. Heat transfer and solidification processes of alloy melt with undercooling: I. Experimental results

    International Nuclear Information System (INIS)

    Yoshioka, Hideaki; Tada, Yukio; Kunimine, Kanji; Furuichi, Taira; Hayashi, Yujiro

    2006-01-01

    The solidification process of Pb-Sn and Bi-Sn alloy melts is discussed to obtain a basic understanding of the essential phenomena of solidification with undercooling. First, from macroscopic observations, it is shown that the solidification process consists of the following three stages: (1) free growth with recalescence dissipation of thermal undercooling (2) expansion of crystals with the relaxation of constitutional undercooling or with the recovering process of interrupted quasi-steady heat conduction, and (3) equilibrium solidification. The specific features of free growth under non-uniform undercooling are also shown by comparison with the Lipton, Glicksman, and Kurz model. Next, from microscopic observations, the distribution of the solute concentration and the change of crystal morphology in the solidified materials were investigated quantitatively using scanning electron microscopy and energy-dispersive spectroscopy. Finally, the solidification path during the above three fundamental processes is dynamically represented on phase diagrams

  20. Status of microwave process development for RH-TRU [remote-handled transuranic] wastes at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    White, T.L.; Youngblood, E.L.; Berry, J.B.; Mattus, A.J.

    1990-01-01

    The Oak Ridge National Laboratory (ORNL) Waste Handling and Packaging Plant is developing a microwave process to reduce and solidify remote-handled transuranic (RH-TRU) liquids and sludges presently stored in large tanks at ORNL. Testing has recently begun on an in-drum microwave process using nonradioactive RH-TRU surrogates. The microwave process development effort has focused on an in-drum process to dry the RH-TRU liquids and sludges in the final storage container and then melt the salt residues to form a solid monolith. A 1/3-scale proprietary microwave applicator was designed, fabricated, and tested to demonstrate the essential features of the microwave design and to provide input into the design of the full-scale applicator. The microwave fields are uniform in one dimension to reduce the formation of hot spots on the microwaved wasteform. The final wasteform meets the waste acceptance criteria for the Waste Isolation Pilot Plant, a federal repository for defense transuranic wastes near Carlsbad, New Mexico. 7 refs., 1 fig., 1 tab

  1. Sandia solidification process: consolidation and characterization. Part I. Consolidation studies

    International Nuclear Information System (INIS)

    Johnstone, J.K.

    1978-05-01

    The consolidation behavior of a complex polycrystalline ceramic nuclear waste form composed of titanates, zeolite, and metallic silicon was studied. Initial solidification takes place by an ion exchange process. The resulting powder exhibits a large surface area, approximately 350 m 2 /g, and several decomposition, crystallization and phase change reactions from room temperature to 1100 0 C. In spite of the large surface area, consolidation by cold pressing and atmospheric sintering to 1100 0 C was not satisfactory. Vacuum hot pressing was found to produce fully dense pellets (less than 1% residual porosity) under very mild conditions, 6.9 MPa (1100 psi) and 1100 0 C. The dominant densification mechanism was viscous flow. Under less than optimum hot pressing conditions, three stages of densification were observed. Initial densification took place by particle rearrangement which was described with a viscous flow model. Second stage densification occurred by a solution-precipitation process controlled by a phase boundary dissolution reaction. In several cases, a third, final densification stage was observed. Detailed studies describe the effects of heating rate, processing temperature, pressure, residence time, atmosphere, composition, heat treatment, and the addition of consolidation aids on the densification behavior. In addition, fully radioactive high level mixed fission product titanate/waste pellets (1.27 cm diameter) were hot pressed at Oak Ridge National Laboratory to demonstrate the feasibility of such a process in a remotely operated hot cell. High density uniform pellets were obtained

  2. Modelling of solidification processing and continuous strip casting for copper-base alloys

    Energy Technology Data Exchange (ETDEWEB)

    Mahmoudi, Jafar [Royal Inst. of Tech., Stockholm (Sweden). Dept. of Materials Processing

    2000-04-01

    An experimental and numerical study was carried out to investigate the solidification process in a copper continuous strip casting process. Heat flow and solidification process has been experimentally studied. Cooling curves during solidification were registered using a thermocouple of type K connected to a data acquisition system. Temperature measurements in the mould and cooling water were also performed. The numerical model considers a generalized set of mass, momentum and heat equations that is valid for the solid, liquid and solidification interval in the cast. A k-{epsilon} turbulence model, produced with the commercial program CFX, is used to analyse the solidification process of pure copper in the mould region of the caster. The fluid flow, temperature and heat flux distributions in the mould region of the caster were computed. The shape and location of the solidification front were also determined. The effects of the parameters such as heat transfer coefficient, casting speed, casting temperature, heat of fusion and specific heat on the shape and location of the solidification front and the heat transport at the mould-cast interface were investigated. The predicted temperature and heat flux distributions were compared with experimental measurements, and reasonable agreement was obtained. The solidification behaviour of pure copper and different copper base alloys has been studied. A series of solidification experiments using DTA furnace, mirror furnace and levitation technique were performed on different copper-base alloys. The undercooling, cooling rates of the liquid and the solid states, solidification times and temperatures were evaluated from the curves. The cooling curves for different samples were simulated using a FEM solidification program. It was found that the calculated values of the heat of fusion were much lower than the tabulated ones. The fraction of solid formed before quenching, in the DTA experiments, has been observed to be much higher

  3. Superior metallic alloys through rapid solidification processing (RSP) by design

    Energy Technology Data Exchange (ETDEWEB)

    Flinn, J.E. [Idaho National Engineering Laboratory, Idaho Falls, ID (United States)

    1995-05-01

    Rapid solidification processing using powder atomization methods and the control of minor elements such as oxygen, nitrogen, and carbon can provide metallic alloys with superior properties and performance compared to conventionally processing alloys. Previous studies on nickel- and iron-base superalloys have provided the baseline information to properly couple RSP with alloy composition, and, therefore, enable alloys to be designed for performance improvements. The RSP approach produces powders, which need to be consolidated into suitable monolithic forms. This normally involves canning, consolidation, and decanning of the powders. Canning/decanning is expensive and raises the fabrication cost significantly above that of conventional, ingot metallurgy production methods. The cost differential can be offset by the superior performance of the RSP metallic alloys. However, without the performance database, it is difficult to convince potential users to adopt the RSP approach. Spray casting of the atomized molten droplets into suitable preforms for subsequent fabrication can be cost competitive with conventional processing. If the fine and stable microstructural features observed for the RSP approach are preserved during spray casing, a cost competitive product can be obtained that has superior properties and performance that cannot be obtained by conventional methods.

  4. Toxic and hazardous waste disposal. Volume 1. Processes for stabilization/solidification

    International Nuclear Information System (INIS)

    Pojasek, R.B.

    1979-01-01

    Processes for the stabilization and/or solidification of toxic, hazardous, and radioactive wastes are reviewed. The types of wastes classified as hazardous are defined. The following processes for the solidification of hazardous wastes are described: lime-based techniques; thermoplastic techniques; organic polymer techniques; and encapsulation. The following processes for the solidification of high-level radioactive wastes are described: calcination; glassification; and ceramics. The solidification of low-level radioactive wastes with asphalt, cement, and polymeric materials is also discussed. Other topics covered include: the use of an extruder/evaporator to stabilize and solidify hazardous wastes; effect disposal of fine coal refuse and flue gas desulfurization slurries using Calcilox additive stabilization; the Terra-Tite Process; the Petrifix Process; the SFT Terra-Crete Process; Sealosafe Process; Chemfix Process; and options for disposal of sulfur oxide wastes

  5. High linear energy transfer degradation studies simulating alpha radiolysis of TRU solvent extraction processes

    Energy Technology Data Exchange (ETDEWEB)

    Pearson, Jeremy [Department of Chemical Engineering and Materials Science - University of California Irvine, 916 Engineering Tower, Irvine, CA, 92697 (United States); Miller, George [Department of Chemistry- University of California Irvine, 2046D PS II, Irvine, CA, 92697 (United States); Nilsson, Mikael [Department of Chemical Engineering and Materials Science - University of California Irvine, 916 Engineering Tower, Irvine, CA, 92697 (United States)

    2013-07-01

    Treatment of used nuclear fuel through solvent extraction separation processes is hindered by radiolytic damage from radioactive isotopes present in used fuel. The nature of the damage caused by the radiation may depend on the radiation type, whether it be low linear energy transfer (LET) such as gamma radiation or high LET such as alpha radiation. Used nuclear fuel contains beta/gamma emitting isotopes but also a significant amount of transuranics which are generally alpha emitters. Studying the respective effects on matter of both of these types of radiation will allow for accurate prediction and modeling of process performance losses with respect to dose. Current studies show that alpha radiation has milder effects than that of gamma. This is important to know because it will mean that solvent extraction solutions exposed to alpha radiation may last longer than expected and need less repair and replacement. These models are important for creating robust, predictable, and economical processes that have strong potential for mainstream adoption on the commercial level. The effects of gamma radiation on solvent extraction ligands have been more extensively studied than the effects of alpha radiation. This is due to the inherent difficulty in producing a sufficient and confluent dose of alpha particles within a sample without leaving the sample contaminated with long lived radioactive isotopes. Helium ion beam and radioactive isotope sources have been studied in the literature. We have developed a method for studying the effects of high LET radiation in situ via {sup 10}B activation and the high LET particles that result from the {sup 10}B(n,a){sup 7}Li reaction which follows. Our model for dose involves solving a partial differential equation representing absorption by 10B of an isentropic field of neutrons penetrating a sample. This method has been applied to organic solutions of TBP and CMPO, two ligands common in TRU solvent extraction treatment processes. Rates

  6. THE SITE DEMONSTRATION OF CHEMFIX SOLIDIFICATION/ STABILIZATION PROCESS AT THE PORTABLE EQUIPMENT SALVAGE COMPANY SITE

    Science.gov (United States)

    A demonstration of the GHEMFIX solidification/stabilization process was conducted under the United States Environmental Protection Agency`s (EPA) Superfund Innovative Technology Evaluation (SITE) program. The demonstration was conducted in March 1989, at the Portable Equipment Sa...

  7. Application of the dual reciprocity boundary element method for numerical modelling of solidification process

    Directory of Open Access Journals (Sweden)

    E. Majchrzak

    2008-12-01

    Full Text Available The dual reciprocity boundary element method is applied for numerical modelling of solidification process. This variant of the BEM is connected with the transformation of the domain integral to the boundary integrals. In the paper the details of the dual reciprocity boundary element method are presented and the usefulness of this approach to solidification process modelling is demonstrated. In the final part of the paper the examples of computations are shown.

  8. 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

  9. Preliminary evaluation of alternative waste form solidification processes. Volume I. Identification of the processes

    International Nuclear Information System (INIS)

    Treat, R.L.; Nesbitt, J.F.; Blair, H.T.; Carter, J.G.; Gorton, P.S.; Partain, W.L.; Timmerman, C.L.

    1980-04-01

    This document contains preconceptual design data on 11 processes for the solidification and isolation of nuclear high-level liquid wastes (HLLW). The processes are: in-can glass melting (ICGM) process, joule-heated glass melting (JHGM) process, glass-ceramic (GC) process, marbles-in-lead (MIL) matrix process, supercalcine pellets-in-metal (SCPIM) matrix process, pyrolytic-carbon coated pellets-in-metal (PCCPIM) matrix process, supercalcine hot-isostatic-pressing (SCHIP) process, SYNROC hot-isostatic-pressing (SYNROC HIP) process, titanate process, concrete process, and cermet process. For the purposes of this study, it was assumed that each of the solidification processes is capable of handling similar amounts of HLLW generated in a production-sized fuel reprocessing plant. It was also assumed that each of the processes would be enclosed in a shielded canyon or cells within a waste facility located at the fuel reprocessing plant. Finally, it was assumed that all of the processes would be subject to the same set of regulations, codes and standards. Each of the solidification processes converts waste into forms that may be acceptable for geological disposal. Each process begins with the receipt of HLLW from the fuel reprocessing plant. In this study, it was assumed that the original composition of the HLLW would be the same for each process. The process ends when the different waste forms are enclosed in canisters or containers that are acceptable for interim storage. Overviews of each of the 11 processes and the bases used for their identification are presented in the first part of this report. Each process, including its equipment and its requirements, is covered in more detail in Appendices A through K. Pertinent information on the current state of the art and the research and development required for the implementation of each process are also noted in the appendices

  10. Geometrical modulus of a casting and its influence on solidification process

    Directory of Open Access Journals (Sweden)

    F. Havlicek

    2011-10-01

    Full Text Available Object: The work analyses the importance of the known criterion for evaluating the controlled solidification of castings, so called geometrical modulus defined by N. Chvorinov as the first one. Geometrical modulus influences the solidification process. The modulus has such specificity that during the process of casting formation it is not a constant but its initial value decreases with the solidification progress because the remaining melt volume can decrease faster than its cooling surface.Methodology: The modulus is determined by a simple calculation from the ratio of the casting volume after pouring the metal in the mould to the cooled mould surface. The solidified metal volume and the cooled surface too are changed during solidification. That calculation is much more complicated. Results were checked up experimentally by measuring the temperatures in the cross-section of heavy steel castings during cooling them.Results: The given experimental results have completed the original theoretical calculations by Chvorinov and recent researches done with use of numerical calculations. The contribution explains how the geometrical modulus together with the thermal process in the casting causes the higher solidification rate in the axial part of the casting cross-section and shortening of solidification time. Practical implications: Change of the geometrical modulus negatively affects the casting internal quality. Melt feeding by capillary filtration in the dendritic network in the casting central part decreases and in such a way the shrinkage porosity volume increases. State of stress character in the casting is changed too and it increases.

  11. Evaluation of process alternatives for solidification of the West Valley high-level liquid wastes

    International Nuclear Information System (INIS)

    Holton, L.K.; Larson, D.E.

    1982-01-01

    The Department of Energy (DOE) established the West Valley Solidification Project (WVSP) in 1980. The project purpose is to demonstrate removal and solidification of the high-level liquid wastes (HLLW) presently stored in tanks at the Western New York Nuclear Service Center (WNYNSC), West Valley, New York. As part of this effort, the Pacific Northwest Laboratory (PNL) conducted a study to evaluate process alternatives for solidifcation of the WNYNSC wastes. Two process approaches for waste handling before solidification, together with solidification processes for four terminal and four interim waste forms, were considered. The first waste-handling approach, designated the salt/sludge separation process, involves separating the bulk of the nonradioactive nuclear waste constituents from the radioactive waste constituents, and the second waste-handling approach, designated the combined-waste process, involves no waste segregation prior to solidification. The processes were evaluated on the bases of their (1) readiness for plant startup by 1987, (2) relative technical merits, and (3) process cost. The study has shown that, based on these criteria, the salt/sludge separation process with a borosilicate glass waste form is preferred when producing a terminal waste form. It was also concluded that if an interim waste form is to be used, the preferred approach would be the combined waste process with a fused-salt waste form

  12. Method of processing radioactive liquid wastes by solidification with cement

    International Nuclear Information System (INIS)

    Yasumura, Keijiro; Matsuura, Hiroyuki.

    1975-01-01

    Object: To subject radioactive liquid wastes to a cement solidification treatment after heating and drying it by a thin film scrape-off drier to render it into the form of power, and then molding it into pellets for the treatment. Structure: Radioactive liquid wastes discharged from a nuclear power plant or nuclear reactor are supplied through a storage tank into a thin film scrape-off drier. In the drier, the radioactive liquid wastes are heated to separate the liquid, and the residue is taken out as dry powder from the scrape-off apparatus. The powder obtained in this way is molded into pellets of a desired form. These pellets are then packed in a drum can or similar container, into which cement paste is then poured for solidification. (Moriyama, K.)

  13. Development of sodium disposal technology. Experiment of sodium compound solidification process

    International Nuclear Information System (INIS)

    Matsumoto, Toshiyuki; Ohura, Masato; Yatoh, Yasuo

    2007-07-01

    A large amount of sodium containing radioactive waste will come up at the time of final shutdown/decommission of FBR plant. The radioactive waste is managed as solid state material in a closed can in Japan. As for the sodium, there is no established method to convert the radioactive sodium to solid waste. Further, the sodium is highly reactive. Thus, it is recommended to convert the sodium to a stable substance before the solidification process. One of the stabilizing methods is conversion of sodium into sodium hydroxide solution. These stabilization and solidification processes should be safe, economical, and efficient. In order to develop such sodium disposal technology, nonradioactive sodium was used and a basic experiment was performed. Waste-fluid Slag Solidification method was employed as the solidification process of sodium hydroxide solution. Experimental parameters were mixing ratio of the sodium hydroxide and the slag solidification material, temperature and concentration of the sodium hydroxide. The best parameters were obtained to achieve the maximum filling ratio of the sodium hydroxide under a condition of enough high compressive strength of the solidified waste. In a beaker level test, the solidified waste was kept in a long term and it was shown that there was no change of appearance, density, and also the compressive strength was kept at a target value. In a real scale test, homogeneous profiles of the density and the compressive strength were obtained. The compressive strength was higher than the target value. It was shown that the Waste-fluid Slag Solidification method can be applied to the solidification process of the sodium hydroxide solution, which was produced by the stabilization process. (author)

  14. Effect of process parameters on hardness, temperature profile and solidification of different layers processed by direct metal laser sintering (DMLS)

    International Nuclear Information System (INIS)

    Ahmed, Sazzad Hossain; Mian, Ahsan; Srinivasan, Raghavan

    2016-01-01

    In DMLS process objects are fabricated layer by layer from powdered material by melting induced by a controlled laser beam. Metallic powder melts and solidifies to form a single layer. Solidification map during layer formation is an important route to characterize micro-structure and grain morphology of sintered layer. Generally, solidification leads to columnar, equiaxed or mixture of these two types grain morphology depending on solidification rate and thermal gradient. Eutectic or dendritic structure can be formed in fully equiaxed zone. This dendritic growth has a large effect on material properties. Smaller dendrites generally increase ductility of the layer. Thus, materials can be designed by creating desired grain morphology in certain regions using DMLS process. To accomplish this, hardness, temperature distribution, thermal gradient and solidification cooling rate in processed layers will be studied under change of process variables by using finite element analysis, with specific application to Ti-6Al-4V.

  15. Effect of process parameters on hardness, temperature profile and solidification of different layers processed by direct metal laser sintering (DMLS)

    Science.gov (United States)

    Ahmed, Sazzad Hossain; Mian, Ahsan; Srinivasan, Raghavan

    2016-07-01

    In DMLS process objects are fabricated layer by layer from powdered material by melting induced by a controlled laser beam. Metallic powder melts and solidifies to form a single layer. Solidification map during layer formation is an important route to characterize micro-structure and grain morphology of sintered layer. Generally, solidification leads to columnar, equiaxed or mixture of these two types grain morphology depending on solidification rate and thermal gradient. Eutectic or dendritic structure can be formed in fully equiaxed zone. This dendritic growth has a large effect on material properties. Smaller dendrites generally increase ductility of the layer. Thus, materials can be designed by creating desired grain morphology in certain regions using DMLS process. To accomplish this, hardness, temperature distribution, thermal gradient and solidification cooling rate in processed layers will be studied under change of process variables by using finite element analysis, with specific application to Ti-6Al-4V.

  16. Effect of process parameters on hardness, temperature profile and solidification of different layers processed by direct metal laser sintering (DMLS)

    Energy Technology Data Exchange (ETDEWEB)

    Ahmed, Sazzad Hossain; Mian, Ahsan, E-mail: ahsan.mian@wright.edu; Srinivasan, Raghavan [Department of Mechanical and Materials Engineering, Wright State University, Dayton, Ohio 45435 (United States)

    2016-07-12

    In DMLS process objects are fabricated layer by layer from powdered material by melting induced by a controlled laser beam. Metallic powder melts and solidifies to form a single layer. Solidification map during layer formation is an important route to characterize micro-structure and grain morphology of sintered layer. Generally, solidification leads to columnar, equiaxed or mixture of these two types grain morphology depending on solidification rate and thermal gradient. Eutectic or dendritic structure can be formed in fully equiaxed zone. This dendritic growth has a large effect on material properties. Smaller dendrites generally increase ductility of the layer. Thus, materials can be designed by creating desired grain morphology in certain regions using DMLS process. To accomplish this, hardness, temperature distribution, thermal gradient and solidification cooling rate in processed layers will be studied under change of process variables by using finite element analysis, with specific application to Ti-6Al-4V.

  17. Solidification of HLLW by glass-ceramic process

    International Nuclear Information System (INIS)

    Oguino, N.; Masuda, S.; Tsunoda, N.; Yamanaka, T.; Ninomiya, M.; Sakane, T.; Nakamura, S.; Kawamura, S.

    1979-01-01

    The compositions of glass-ceramics for the solidification of HLLW were studied, and the glass-ceramics in the diopside system was concluded to be the most suitable. Compared with the properties of HLW borosilicate glasses, those of diopside glass-ceramic were thought to be almost equal in leach rate and superior in thermal stability and mechanical strength. It was also found that the glass in this system can be crystallized simply by pouring it into a thermally insulated canister and then allowing it to cool to room temperature. 2 figures, 5 tables

  18. PREFACE: Third International Conference on Advances in Solidification Processes (ICASP - 3)

    Science.gov (United States)

    Zimmermann, Gerhard; Ratke, Lorenz

    2012-01-01

    The 3rd International Conference on Advances in Solidification Processes was held in the Rolduc Abbey in the Netherlands a few kilometres away from Aachen. Around 200 scientists from 24 countries come in for the four day meeting. They found a stimulating but also relaxing environment and atmosphere, with beautiful weather and the medieval abbey inviting for walks, discussions, sitting outside and drinking a beer or wine. The contributions given at the conference reflected recent advances in various topics of solidification processes, ranging from fundamental aspects to applied casting technologies. In 20 oral sessions and a large poster session innovative results of segregation phenomena, microstructure evolution, nucleation and growth, phase formation, polyphase solidification, rapid solidification and welding, casting technology, thermophysics of molten alloys, solidification with forced melt flow and growth of single crystals and superalloys together with innovative diagnostic techniques were presented. Thereby, findings from experiments as well as from numerical modeling on different lengths scales were jointly discussed and contribute to new insight in solidification behaviour. The papers presented in this open access proceedings cover about half the oral and poster presentations given. They were carefully reviewed as in classical peer reviewed journals by two independent referees and most of them were revised and thus improved according to the reviewers comments. We think that this collection of papers presented at ICASP-3 gives an impression of the excellent contributions made. The papers embrace both the basic and applied aspects of solidification. We especially wish to express our appreciation for the team around Georg Schmitz and Margret Nienhaus organising this event and giving us their valued advice and support at every stage in preparing the conference. We also thank Lokasenna Lektorat for taking the task of checking all language-associated issues and

  19. Ductile failure in upsetting of a rapid-solidification-processed aluminium alloy

    NARCIS (Netherlands)

    Habraken, F.A.C.M.; Dautzenberg, J.H.

    1993-01-01

    Cold upset-tests have been performed on a Rapid Solidification Processed (RSP) aluminium-alloy, produced by the ‘melt-spun ribbons’-process out of 70% car-scrap and 30% primary scrap. The ribbons are hot extruded, resulting in 29 mm diameter bar. Its properties regarding plastic flow and fracture

  20. Process control of Low and Intermediate-level radioactive wastes solidification

    International Nuclear Information System (INIS)

    1993-01-01

    Safety guidelines issued by the Spanish Council of Nuclear Safety (CSN) with basic criteria which must be adopted for the control of the Process for wastes solidification, establishing, in addition, a series of protocols and basic contents to assist the elaboration of Process Control Programs

  1. Solidification process for toxic and hazardous wastes. Second part: Cement solidification matrices; Inertizzazione di rifiuti tossici e nocivi (RTN). Parte seconda: Inertizzazione in matrici cementizie

    Energy Technology Data Exchange (ETDEWEB)

    Donato, A; Arcuri, L; Dotti, M; Pace, A; Pietrelli, L; Ricci, G [ENEA - Dipartimento Ciclo del Combustibile, Centro Ricerche Energia, Casaccia (Italy); Basta, M; Cali, V; Pagliai, V [ENEA - Dipartimento Ciclo del Combustibile, Centro Ricerche Energia, Saluggia (Italy)

    1989-05-15

    This paper reports the second part of a general study carried out at the Nuclear Fuel Division aiming at verifying the possible application of the radioactive waste solidification processes to industrial hazardous wastes (RTN). The cement solidification of several RTN types has been taken into consideration, both from the technical and from the economic point of view. After a short examination of the Italian juridical and economical situation in the field, which demonstrates the need of the RTN solidification, the origin and characteristics of the RTN considered in the study and directly provided by the producing industries are reviewed. The laboratory experimental results of the cementation of RTN produced by gold manufacturing industries and by galvanic industries are reported. The cementation process can be considered a very effective mean for reducing both the RTN management costs and the environmental impact of RTN disposal. (author)

  2. Processing and solidification of Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Kelley, J.A.

    1981-01-01

    The entire flowsheet for processing and solidification of Savannah River Plant (SRP) high-level wastes has been demonstrated. A new small-scale integrated pilot plant is operating with actual radioactive wastes, and large-scale equipment is being demonstrated with nonradioactive simulated wastes. Design of a full-scale waste solidification plant is in progress. Plant construction is expected to begin in 1983, and startup is anticipated in 1988. The plant will poduce about 500 cans of glass per year with each can containing about 1.5 tons of glass

  3. Solidification method for organic solution and processing method of aqueous solution

    International Nuclear Information System (INIS)

    Kamoshida, Mamoru; Fukazawa, Tetsuo; Yazawa, Noriko; Hasegawa, Toshihiko

    1998-01-01

    The relative dielectric constant of an organic solution containing polar ingredients is controlled to 13 or less to enable its solidification. The polarity of the organic solution can be evaluated quantitatively by using the relative dielectric constant. If the relative dielectric constant is high, it can be controlled by dilution using a non-polar organic solvent of low relative dielectric constant. With such procedures, solidification can be conducted by using an economical 12-hydroxy stearic acid, process of liquid wastes can be facilitated and the safety can be ensured. (T.M.)

  4. Oxygen and carbon transfer during solidification of semiconductor grade silicon in different processes

    Science.gov (United States)

    Ribeyron, P. J.; Durand, F.

    2000-03-01

    A model is established for comparing the solute distribution resulting from four solidification processes currently applied to semiconductor grade silicon: Czochralski pulling (CZ), floating zone (FZ), 1D solidification and electromagnetic continuous pulling (EMCP). This model takes into account solid-liquid interface exchange, evaporation to or contamination by the gas phase, container dissolution, during steady-state solidification, and in the preliminary preparation of the melt. For simplicity, the transfers are treated in the crude approximation of perfectly mixed liquid and boundary layers. As a consequence, only the axial ( z) distribution can be represented. Published data on oxygen and carbon transfer give a set of acceptable values for the thickness of the boundary layers. In the FZ and EMCP processes, oxygen evaporation can change the asymptotic behaviour of the reference Pfann law. In CZ and in 1D-solidification, a large variety of solute profile curves can be obtained, because they are very sensitive to the balance between crucible dissolution and evaporation. The CZ process clearly brings supplementary degrees of freedom via the geometry of the crucible, important for the dissolution phenomena, and via the rotation rate of the crystal and of the crucible, important for acting on transfer kinetics.

  5. Preliminary evaluation of alternative waste form solidification processes. Volume II. Evaluation of the processes

    International Nuclear Information System (INIS)

    1980-08-01

    This Volume II presents engineering feasibility evaluations of the eleven processes for solidification of nuclear high-level liquid wastes (HHLW) described in Volume I of this report. Each evaluation was based in a systematic assessment of the process in respect to six principal evaluation criteria: complexity of process; state of development; safety; process requirements; development work required; and facility requirements. The principal criteria were further subdivided into a total of 22 subcriteria, each of which was assigned a weight. Each process was then assigned a figure of merit, on a scale of 1 to 10, for each of the subcriteria. A total rating was obtained for each process by summing the products of the subcriteria ratings and the subcriteria weights. The evaluations were based on the process descriptions presented in Volume I of this report, supplemented by information obtained from the literature, including publications by the originators of the various processes. Waste form properties were, in general, not evaluated. This document describes the approach which was taken, the developent and application of the rating criteria and subcriteria, and the evaluation results. A series of appendices set forth summary descriptions of the processes and the ratings, together with the complete numerical ratings assigned; two appendices present further technical details on the rating process

  6. Cadarache LOR (liquides organiques radioactifs) treatment by a solidification process using NOCHAR polymers

    International Nuclear Information System (INIS)

    Vaudey, Claire-Emilie; Renou, Sebastien; Kelley, Dennis; Cochaud, Chantal; Serrano, Roger

    2013-01-01

    In France, two options can be considered to handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW). The first one is the incineration at CENTRACO facility and the second one is the disposal at ANDRA sites. The waste acceptance in these two channels is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the channel specifications. If the waste characteristics and the channel specifications (presence of significant quantities of halogens, complexing agents, organic components... or/and high activity limits) are incompatible, an alternative solution have to be identify. It consists of a waste pre-treatment process. For Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. They are composed of a mix of organic liquids and water: for the first one, 19 % of organic compounds (xylene, mesitylene, diphenyloxazole, TBP...) and 86.9 % of water, and for the second one, 23 % of organic compounds (TBP...) and 77 % of water. They contain halogens (chlorine and fluorine), complexants agents (nitrate, sulphate, oxalate and formate) and have got αβγ spectra with mass activities equal to some 100 Bq/g. Therefore, tritium is also present. As a consequence, in order for storage acceptance at the ANDRA site, it is necessary to pre-treat the waste. An adequate solution seems to be a solidification process using NOCHAR polymers. Indeed, NOCHAR polymers correspond to an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing ...) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and

  7. NOCHAR Polymers: An Aqueous and Organic Liquid Solidification Process for Cadarache LOR (Liquides Organiques Radioactifs) - 13195

    International Nuclear Information System (INIS)

    Vaudey, Claire-Emilie; Renou, Sebastien; Porco, Julien; Kelley, Dennis; Cochaud, Chantal; Serrano, Roger

    2013-01-01

    To handle the Very Low Level Waste (VLLW) and the Low Level Waste (LLW) in France, two options can be considered: the incineration at CENTRACO facility and the disposal facility on ANDRA sites. The waste acceptance in these radwaste routes is dependent upon the adequacy between the waste characteristics (physical chemistry and radiological) and the radwaste route specifications. If the waste characteristics are incompatible with the radwaste route specifications (presence of significant quantities of chlorine, fluorine, organic component etc or/and high activity limits), it is necessary to find an alternative solution that consists of a waste pre-treatment process. In the context of the problematic Cadarache LOR (Liquides Organiques Radioactifs) waste streams, two radioactive scintillation cocktails have to be treated. The first one is composed of organic liquids at 13.1 % (diphenyloxazol, mesitylene, TBP, xylene) and water at 86.9 %. The second one is composed of TBP at 8.6 % and water at 91.4 %. They contain chlorine, fluorine and sulphate and have got alpha/beta/gamma spectra with mass activities equal to some kBq.g -1 . Therefore, tritium is present and creates the second problematic waste stream. As a consequence, in order for disposal acceptance at the ANDRA site, it is necessary to pre-treat the waste. The NOCHAR polymers as an aqueous and organic liquid solidification process seem to be an adequate solution. Indeed, these polymers constitute an important variety of products applied to the treatment of radioactive aqueous and organic liquids (solvent, oil, solvent/oil mixing etc) and sludge through a mechanical and chemical solidification process. For Cadarache LOR, N910 and N960 respectively dedicated to the organic and aqueous liquids solidification are considered. With the N910, the organic waste solidification occurs in two steps. As the organic liquid travels moves through the polymer strands, the strands swell and immobilise the liquid. Then as the

  8. Optimization of heat transfer during the directional solidification process of 1600 kg silicon feedstock

    Science.gov (United States)

    Hu, Chieh; Chen, Jyh Chen; Nguyen, Thi Hoai Thu; Hou, Zhi Zhong; Chen, Chun Hung; Huang, Yen Hao; Yang, Michael

    2018-02-01

    In this study, the power ratio between the top and side heaters and the moving velocity of the side insulation are designed to control the shape of the crystal-melt interface during the growth process of a 1600 kg multi-crystalline silicon ingot. The power ratio and insulation gap are adjusted to ensure solidification of the melt. To ensure that the crystal-melt interface is slightly convex in relation to the melt during the entire solidification process, the power ratio should be augmented gradually in the initial stages while being held to a constant value in the middle stages. Initially the gap between the side and the bottom insulation is kept small to reduce thermal stress inside the seed crystals. However, the growth rate will be slow in the early stages of the solidification process. Therefore, the movement of the side insulation is fast in the initial stages but slower in the middle stages. In the later stages, the side insulation gap is fixed. With these modifications, the convexity of the crystal-melt interface in relation to the melt can be maintained during the growth process with an approximately 41% reduction in the thermal stress inside the growing ingot and an 80% reduction in dislocation density along the center line of the ingot compared with the original case.

  9. MCWASP XIV: International Conference on Modelling of Casting, Welding and Advanced Solidification Processes

    International Nuclear Information System (INIS)

    Yasuda, H

    2015-01-01

    The current volume represents contributed papers of the proceedings of the 14th international conference on ''Modeling of Casting, Welding and Advanced Solidification Processes (MCWASP XIV)'', Yumebutai International Conference Center, Awaji island, Hyogo, Japan on 21 – 26 June, 2016. The first conference of the series 'Modeling of Casting, Welding and Advanced Solidification Processes (MCWASP)' was started up in 1980, and this is the 14th conference. The participants are more than 100 scientists from industry and academia, coming from 19 countries. In the conference, we have 5 invited, 70 oral and 31 poster presentations on different aspects of the modeling. The conference deals with various casting processes (Ingot / shape casting, continuous casting, direct chill casting and welding), fundamental phenomena (nucleation and growth, dendritic growth, eutectic growth, micro-, meso- and macrostructure formation and defect formation), coupling problems (electromagnetic interactions, application of ultrasonic wave), development of experimental / computational methods and so on. This volume presents the cutting-edge research in the modeling of casting, welding and solidification processes. I would like to thank MAGMA Giessereitechnologie GmbH, Germany and SCSK Corporation, Japan for supporting the publication of contributed papers. Hideyuki Yasuda Conference Chairman Department of Materials Science and Engineering, Kyoto University Japan (preface)

  10. Process Description for the Retrieval of Earth Covered Transuranic (TRU) Waste Containers at the Hanford Site

    International Nuclear Information System (INIS)

    DEROSA, D.C.

    2000-01-01

    This document describes process and operational options for retrieval of the contact-handled suspect transuranic waste drums currently stored below grade in earth-covered trenches at the Hanford Site. Retrieval processes and options discussed include excavation, container retrieval, venting, non-destructive assay, criticality avoidance, incidental waste handling, site preparation, equipment, and shipping

  11. Process Description for the Retrieval of Earth Covered Transuranic (TRU) Waste Containers at the Hanford Site

    Energy Technology Data Exchange (ETDEWEB)

    DEROSA, D.C.

    2000-01-13

    This document describes process and operational options for retrieval of the contact-handled suspect transuranic waste drums currently stored below grade in earth-covered trenches at the Hanford Site. Retrieval processes and options discussed include excavation, container retrieval, venting, non-destructive assay, criticality avoidance, incidental waste handling, site preparation, equipment, and shipping.

  12. TruMicro Series 2000 sub-400 fs class industrial fiber lasers: adjustment of laser parameters to process requirements

    Science.gov (United States)

    Kanal, Florian; Kahmann, Max; Tan, Chuong; Diekamp, Holger; Jansen, Florian; Scelle, Raphael; Budnicki, Aleksander; Sutter, Dirk

    2017-02-01

    and multi-level quad-loop stabilization of the output power of the laser.2 In addition to the well-established platform latest developments addressed single-pulse energies up to 50 μJ and made femtosecond pulse durations available for the TruMicro Series 2000. Beyond these stabilization aspects this laser architecture together with other optical modules and combined with smart laser control software enables process-driven adjustments of the parameters (e. g. repetition rate, multi-pulse functionalities, pulse energy, pulse duration) by external signals, which will be presented in this work.

  13. Improvements to a uranium solidification process by in-plant testing

    International Nuclear Information System (INIS)

    Rindfleisch, J.A.

    1984-01-01

    When a process is having operational or equipment problems, often there is not enough time or money available for an extensive pilot plant program. This is when in-plant testing becomes imperative. One such process at the Idaho Chemical Processing Plant (ICPP) to undergo such an in-plant testing program was the uranium product solidification (denitrator) system. The testing program took approximately six months of in-plant testing that would have required at least two years of pilot plant preparation and operation to obtain the same information. This paper describes the results of the testing program, and the equipment and procedural changes

  14. Influences on Distribution of Solute Atoms in Cu-8Fe Alloy Solidification Process Under Rotating Magnetic Field

    Science.gov (United States)

    Zou, Jin; Zhai, Qi-Jie; Liu, Fang-Yu; Liu, Ke-Ming; Lu, De-Ping

    2018-05-01

    A rotating magnetic field (RMF) was applied in the solidification process of Cu-8Fe alloy. Focus on the mechanism of RMF on the solid solution Fe(Cu) atoms in Cu-8Fe alloy, the influences of RMF on solidification structure, solute distribution, and material properties were discussed. Results show that the solidification behavior of Cu-Fe alloy have influenced through the change of temperature and solute fields in the presence of an applied RMF. The Fe dendrites were refined and transformed to rosettes or spherical grains under forced convection. The solute distribution in Cu-rich phase and Fe-rich phase were changed because of the variation of the supercooling degree and the solidification rate. Further, the variation in solute distribution was impacted the strengthening mechanism and conductive mechanism of the material.

  15. Modeling of multiphase flow with solidification and chemical reaction in materials processing

    Science.gov (United States)

    Wei, Jiuan

    Understanding of multiphase flow and related heat transfer and chemical reactions are the keys to increase the productivity and efficiency in industrial processes. The objective of this thesis is to utilize the computational approaches to investigate the multiphase flow and its application in the materials processes, especially in the following two areas: directional solidification, and pyrolysis and synthesis. In this thesis, numerical simulations will be performed for crystal growth of several III-V and II-VI compounds. The effects of Prandtl and Grashof numbers on the axial temperature profile, the solidification interface shape, and melt flow are investigated. For the material with high Prandtl and Grashof numbers, temperature field and growth interface will be significantly influenced by melt flow, resulting in the complicated temperature distribution and curved interface shape, so it will encounter tremendous difficulty using a traditional Bridgman growth system. A new design is proposed to reduce the melt convection. The geometric configuration of top cold and bottom hot in the melt will dramatically reduce the melt convection. The new design has been employed to simulate the melt flow and heat transfer in crystal growth with large Prandtl and Grashof numbers and the design parameters have been adjusted. Over 90% of commercial solar cells are made from silicon and directional solidification system is the one of the most important method to produce multi-crystalline silicon ingots due to its tolerance to feedstock impurities and lower manufacturing cost. A numerical model is developed to simulate the silicon ingot directional solidification process. Temperature distribution and solidification interface location are presented. Heat transfer and solidification analysis are performed to determine the energy efficiency of the silicon production furnace. Possible improvements are identified. The silicon growth process is controlled by adjusting heating power and

  16. Continuum simulation of heat transfer and solidification behavior of AlSi10Mg in Direct Metal Laser Sintering Process

    Science.gov (United States)

    Ojha, Akash; Samantaray, Mihir; Nath Thatoi, Dhirendra; Sahoo, Seshadev

    2018-03-01

    Direct Metal Laser Sintering (DMLS) process is a laser based additive manufacturing process, which built complex structures from powder materials. Using high intensity laser beam, the process melts and fuse the powder particles makes dense structures. In this process, the laser beam in terms of heat flux strikes the powder bed and instantaneously melts and joins the powder particles. The partial solidification and temperature distribution on the powder bed endows a high cooling rate and rapid solidification which affects the microstructure of the build part. During the interaction of the laser beam with the powder bed, multiple modes of heat transfer takes place in this process, that make the process very complex. In the present research, a comprehensive heat transfer and solidification model of AlSi10Mg in direct metal laser sintering process has been developed on ANSYS 17.1.0 platform. The model helps to understand the flow phenomena, temperature distribution and densification mechanism on the powder bed. The numerical model takes into account the flow, heat transfer and solidification phenomena. Simulations were carried out for sintering of AlSi10Mg powders in the powder bed having dimension 3 mm × 1 mm × 0.08 mm. The solidification phenomena are incorporated by using enthalpy-porosity approach. The simulation results give the fundamental understanding of the densification of powder particles in DMLS process.

  17. Leaching properties of solidified TRU waste forms

    International Nuclear Information System (INIS)

    Colombo, P.; Neilson, R.M. Jr.

    1979-01-01

    Safety analysis of waste forms requires an estimate of the ability of these forms to retain activity in the disposal environment. This program of leaching tests will determine the leaching properties of TRU contaminated incinerator ash waste forms using hydraulic cement, urea--formaldehyde, bitumen, and vinyl ester--styrene as solidification agents. Three types of leaching tests will be conducted, including both static and flow rate. Five generic groundwaters will be used. Equipment and procedures are described. Experiments have been conducted to determine plate out of 239 Pu, counter efficiency, and stability of counting samples

  18. Radioactive gas solidification apparatus

    International Nuclear Information System (INIS)

    Kobayashi, Yoshihiro; Seki, Eiji; Yabu, Tomohiko; Matsunaga, Hiroyuki.

    1990-01-01

    Handling of a solidification container from the completion for the solidifying processing to the storage of radioactive gases by a remote control equipment such as a manipulator requires a great cost and is difficult to realize. In a radioactive gas solidification device for injection and solidification in accumulated layers of sputtered metals by glow discharge, radiation shieldings are disposed surrounding the entire container, and cooling water is supplied to a cooling vessel formed between the container and the shielding materials. The shielding materials are divided into upper and lower shielding materials, so that solidification container can be taken out from the shielding materials. As a result, the solidification container after the solidification of radioactive gases can be handled with ease. Further, after-heat can be removed effectively from the ion injection electrode upon solidifying treatment upon storage, to attain a radioactive gas solidifying processing apparatus which is safe, economical and highly reliable. (N.H.)

  19. Solidification of high level liquid waste (HLLW) into ceramics by sintering process

    International Nuclear Information System (INIS)

    Masuda, Sumio; Oguino, Naohiko; Tsunoda, Naomi; O-oka, Kazuo; Ohta, Takao.

    1979-01-01

    One of the alternatives to vitrified solid which is acceptable and well characterized for storing radioactive HLLW with desirable long-term stability is ceramics. On the other hand, the solidification process of highly radioactive wastes should be simple and suitable for continuous production. On the above described basis, the authors have made preliminary study on the production of sintered ceramics by the addition of several oxides to HLLW. The simulated waste and additive oxides were pressed in a mold to make the preforms of 50 mm diameter and 10 to 15 mm thick. The preforms were then normally sintered at temperature from 1000 to 1400 deg C for 2 to 4 hours. The characterization of the sintered solids revealed the following facts. (1) X-ray diffraction analysis showed that the expected crystals were formed by normal-sintering as well as by hot-pressing. (2) The bulk density of the ceramics by normal-sintering was around 90 to 95% of the assumed theoretical values. (3) The leach-rate of the solids was affected by the bulk density. (4) Other properties of the solids, such as thermal expansion or thermal conductivity, are dominantly determined by those of main crystals in the solids. Sintering process is generally simple and productive as far as normal sintering is concerned. However, hot-pressing is an intermittent and time consuming process. From this fact, the authors intended to adopt the normal sintering process for the ceramic solidification of high level liquid wastes. (Wakatsuki, Y.)

  20. Strain Measurement in Aluminium Alloy during the Solidification Process Using Embedded Fibre Bragg Gratings.

    Science.gov (United States)

    Weraneck, Klaus; Heilmeier, Florian; Lindner, Markus; Graf, Moritz; Jakobi, Martin; Volk, Wolfram; Roths, Johannes; Koch, Alexander W

    2016-11-04

    In recent years, the observation of the behaviour of components during the production process and over their life cycle is of increasing importance. Structural health monitoring, for example of carbon composites, is state-of-the-art research. The usage of Fibre Bragg Gratings (FBGs) in this field is of major advantage. Another possible area of application is in foundries. The internal state of melts during the solidification process is of particular interest. By using embedded FBGs, temperature and stress can be monitored during the process. In this work, FBGs were embedded in aluminium alloys in order to observe the occurring strain. Two different FBG positions were chosen in the mould in order to compare its dependence. It was shown that FBGs can withstand the solidification process, although a compression in the range of one percent was measured, which is in agreement with the literature value. Furthermore, different lengths of the gratings were applied, and it was shown that shorter gratings result in more accurate measurements. The obtained results prove that FBGs are applicable as sensors for temperatures up to 740 °C.

  1. The cement solidification systems at LANL

    International Nuclear Information System (INIS)

    Veazey, G.W.

    1990-01-01

    There are two major cement solidification systems at Los Alamos National Laboratory. Both are focused primarily around treating waste from the evaporator at TA-55, the Plutonium Processing Facility. The evaporator receives the liquid waste stream from TA-55's nitric acid-based, aqueous-processing operations and concentrates the majority of the radionuclides in the evaporator bottoms solution. This is sent to the TA-55 cementation system. The evaporator distillate is sent to the TA-50 facility, where the radionuclides are precipitated and then cemented. Both systems treat TRU-level waste, and so are operated according to the criteria for WIPP-destined waste, but they differ in both cement type and mixing method. The TA-55 systems uses Envirostone, a gypsum-based cement and in-drum prop mixing; the TA-50 systems uses Portland cement and drum tumbling for mixing

  2. Microstructure engineering of TiAl-based refractory intermetallics within power-down directional solidification process

    International Nuclear Information System (INIS)

    Kartavykh, A.V.; Tcherdyntsev, V.V.; Gorshenkov, M.V.; Kaloshkin, S.D.

    2014-01-01

    Highlights: ► VGF power-down technique is suitable for TiAl-based alloys solidification with tailored microstructure. ► Both columnar-dendrite and granular structures are created in Ti–46Al–8Nb ingots. ► Granular microstructure has been refined with TiB 2 addition to the melt. ► TiB 2 re-precipitate into (Ti,Nb)B particles, those acting as point seeds for fine equiaxed grains nucleation. -- Abstract: The work is aimed at the study of the formation and refinement of primary microstructure appearing in the refractory lightweight structural TiAl-based alloy of Ti–46Al–8Nb (at.%) nominal composition. For tailored microstructure development, the Directional Solidification (DS) of pre-synthesized alloy was performed in the vertical multizone resistive electro-furnace by power-down technique in pure argon environment. Both columnar-dendrite, and equiaxed-granular reproducible as-cast microstructures have been produced in DS ingots, basing on Columnar-to-Equiaxed Transition (CET) diagram and experimental exploration. Particular attention was paid further to equiaxed microstructure improvement by combination of modifying doping of alloy with boron grain refiner and DS processing. As a result the perfect inoculated microstructure of Ti–44Al–7Nb–2B (at.%) ingots was produced with 100 μm mean grain diameter, low scattering of dimensional grain characteristics and high tolerance to DS process parameters variation

  3. Lamellar boundary alignment of DS-processed TiAl-W alloys by a solidification procedure

    Science.gov (United States)

    Jung, In-Soo; Oh, Myung-Hoon; Park, No-Jin; Kumar, K. Sharvan; Wee, Dang-Moon

    2007-12-01

    In this study, a β solidification procedure was used to align the lamellae in a Ti-47Al-2W (at.%) alloy parallel to the growth direction. The Bridgman technique and the floating zone process were used for directional solidification. The mechanical properties of the directionally solidified alloy were evaluated in tension at room temperature and at 800°C. At a growth rate of 30 mm/h (with the floating zone approach), the lamellae were well aligned parallel to the growth direction. The aligned lamellae yielded excellent room temperature tensile ductility. The tensile yield strength at 800°C was similar to that at room temperature. The orientation of the γ lamellar laths in the directionally solidified ingots, which were manufactured by means of a floating zone process, was identified with the aid of electron backscattered diffraction analysis. On the basis of this analysis, the preferred growth direction of the bcc-β dendrites that formed at high temperatures close to the melting point was inferred to be [001]β at a growth rate of 30 mm/h and [111]β at a growth rate of 90 mm/h.

  4. Large Eddy Simulation of Transient Flow, Solidification, and Particle Transport Processes in Continuous-Casting Mold

    Science.gov (United States)

    Liu, Zhongqiu; Li, Linmin; Li, Baokuan; Jiang, Maofa

    2014-07-01

    The current study developed a coupled computational model to simulate the transient fluid flow, solidification, and particle transport processes in a slab continuous-casting mold. Transient flow of molten steel in the mold is calculated using the large eddy simulation. An enthalpy-porosity approach is used for the analysis of solidification processes. The transport of bubble and non-metallic inclusion inside the liquid pool is calculated using the Lagrangian approach based on the transient flow field. A criterion of particle entrapment in the solidified shell is developed using the user-defined functions of FLUENT software (ANSYS, Inc., Canonsburg, PA). The predicted results of this model are compared with the measurements of the ultrasonic testing of the rolled steel plates and the water model experiments. The transient asymmetrical flow pattern inside the liquid pool exhibits quite satisfactory agreement with the corresponding measurements. The predicted complex instantaneous velocity field is composed of various small recirculation zones and multiple vortices. The transport of particles inside the liquid pool and the entrapment of particles in the solidified shell are not symmetric. The Magnus force can reduce the entrapment ratio of particles in the solidified shell, especially for smaller particles, but the effect is not obvious. The Marangoni force can play an important role in controlling the motion of particles, which increases the entrapment ratio of particles in the solidified shell obviously.

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  6. Solidification of low-level radioactive liquid waste using a cement-silicate process

    International Nuclear Information System (INIS)

    Grandlund, R.W.; Hayes, J.F.

    1979-01-01

    Extensive use has been made of silicate and Portland cement for the solidification of industrial waste and recently this method has been successfully used to solidify a variety of low level radioactive wastes. The types of wastes processed to date include fuel fabrication sludges, power reactor waste, decontamination solution, and university laboratory waste. The cement-silicate process produces a stable solid with a minimal increase in volume and the chemicals are relatively inexpensive and readily available. The method is adaptable to either batch or continuous processing and the equipment is simple. The solid has leaching characteristics similar to or better than plain Portland cement mixtures and the leaching can be further reduced by the use of ion-exchange additives. The cement-silicate process has been used to solidify waste containing high levels of boric acid, oils, and organic solvents. The experience of handling the various types of liquid waste with a cement-silicate system is described

  7. INVESTIGATION OF CEMENT CONCRETE CONGLOMERATE SOLIDIFICATION PROCESS BY IMPEDANCE SPECTROSCOPY METHOD

    Directory of Open Access Journals (Sweden)

    S. N. Bandarenka

    2015-01-01

    Full Text Available One of the most prospective directions in preservation  and increase of service live of  road pavements is a construction of  automobile roads with cement concrete surface. Modern tendencies for provision of road construction quality presuppose a necessity to control processes of solidification and subsequent destruction of the material while forming and using cement concrete conglomerate being considered as a basic element of the road surface.  Multiyear practical experience of  automobile road operation using cement concrete pavements reveals an importance for monitoring  such processes as formation and destruction of cement concrete materials. An impedance spectroscopy method has been tried out and proposed as a tool for solution of the given problem.Experimental samples of cement concrete have been prepared for execution of tests, graded silica sand and granite chippings with particle size from 0.63 to 2.5 mm have been used as a fine aggregate in the samples. Dependencies of resistance (impedance on AC-current frequency  have been studied for samples of various nature and granulometric composition. The Gamry  G300 potentiostat has been used for measurement of complex impedance value. A spectrum analysis and calculation of equivalent circuit parameters calculation have been carried out while using EIS Spectrum Analyzer program.Comparison of impedance spectra for the prepared cement concrete samples have made it possible to reveal tendencies in changing spectrum parameters during solidification and subsequent contact with moisture in respect of every type of the sample. An equivalent electrical circuit has been developed that  characterizes physical and chemical processes which are accompanied by charge transfer in cement concrete conglomerate. The paper demonstrates a possibility to use an impedance spectroscopy for solution of a number of actual problems in the field of cement concrete technology problems. Particularly, the problems

  8. Solidification microstructure development

    Indian Academy of Sciences (India)

    Unknown

    A majority of manufacturing processes involve melting and solidification of metals and ... In such a case (for example, chill casting), the solidification thickness (S) is ... (5). Here, LX is the system length scale in one dimension and DS is the solute diffusivity in solid. Thermal and solutal diffusivities are finite and usually very ...

  9. Superplastic forming of rapid solidification processed Al-4Li-0.2Zr

    International Nuclear Information System (INIS)

    Meschter, P.J.; Lederich, R.J.; Sastry, S.M.L.

    1987-01-01

    Aluminum-4 wt pct lithium alloys are attractive as structural materials because they are 13 to 14 pct less dense and have 25 pct larger elastic moduli than high-strength 2XXX-and 7XXX-series aluminum alloys. These low-density alloys can be produced only by rapid solidification processing (RSP). Successful RSP of Al-4Li-0.2Zr, Al-4Li-1Mg-0.2Zr, and Al-4Li-1Cu-0.2Zr alloys with strengths similar to that of 7075-T76 has recently been demonstrated. Net-shaped processing techniques such as superplastic forming are capable of producing complex structural elements while minimizing usage of expensive material; thus, these techniques are particularly applicable to Al-Li alloys. The purpose of this study was to determine the conditions of strain rate and temperature under which RSP Al-4Li alloys could be superplastically formed

  10. A comparison of acoustic levitation with microgravity processing for containerless solidification of ternary Al-Cu-Sn alloy

    Science.gov (United States)

    Yan, N.; Hong, Z. Y.; Geng, D. L.; Wei, B.

    2015-07-01

    The containerless rapid solidification of liquid ternary Al-5 %Cu-65 %Sn immiscible alloy was accomplished at both ultrasonic levitation and free fall conditions. A maximum undercooling of 185 K (0.22 T L) was obtained for the ultrasonically levitated alloy melt at a cooling rate of about 122 K s-1. Meanwhile, the cooling rate of alloy droplets in drop tube varied from 102 to 104 K s-1. The macrosegregation was effectively suppressed through the complex melt flow under ultrasonic levitation condition. In contrast, macrosegregation became conspicuous and core-shell structures with different layers were formed during free fall. The microstructure formation mechanisms during rapid solidification at containerless states were investigated in comparison with the conventional static solidification process. It was found that the liquid phase separation and structural growth kinetics may be modulated by controlling both alloy undercooling and cooling rate.

  11. Superconducting properties of single-crystal Nb sphere formed by large-undercooling solidification process

    Energy Technology Data Exchange (ETDEWEB)

    Takeya, H.; Sung, Y.S.; Hirata, K.; Togano, K

    2003-10-15

    An electrostatic levitation (ESL) system has been used for investigating undercooling effects on superconducting materials. In this report, preliminary experiments on Nb (melting temperature: T{sub m}=2477 deg. C) have been performed by melting Nb in levitation using 150 and 250 W Nd-YAG lasers. Since molten Nb is solidified without any contact in a high vacuum condition, a significantly undercooled state up to 400 deg. C is maintained before recalescence followed by solidification. Spherical single crystals of Nb are formed by the ESL process due to the suppression of heterogeneous nucleation. The field dependence of magnetization of Nb shows a reversible behavior as an ideal type II superconductor, implying that it contains almost no flux-pinning centers.

  12. Solidified structure and leaching properties of metallurgical wastewater treatment sludge after solidification/stabilization process.

    Science.gov (United States)

    Radovanović, Dragana Đ; Kamberović, Željko J; Korać, Marija S; Rogan, Jelena R

    2016-01-01

    The presented study investigates solidification/stabilization process of hazardous heavy metals/arsenic sludge, generated after the treatment of the wastewater from a primary copper smelter. Fly ash and fly ash with addition of hydrated lime and Portland composite cement were studied as potential binders. The effectiveness of the process was evaluated by unconfined compressive strength (UCS) testing, leaching tests (EN 12457-4 and TCLP) and acid neutralization capacity (ANC) test. It was found that introduction of cement into the systems increased the UCS, led to reduced leaching of Cu, Ni and Zn, but had a negative effect on the ANC. Gradual addition of lime resulted in decreased UCS, significant reduction of metals leaching and high ANC, due to the excess of lime that remained unreacted in pozzolanic reaction. Stabilization of more than 99% of heavy metals and 90% of arsenic has been achieved. All the samples had UCS above required value for safe disposal. In addition to standard leaching tests, solidificates were exposed to atmospheric conditions during one year in order to determine the actual leaching level of metals in real environment. It can be concluded that the EN 12457-4 test is more similar to the real environmental conditions, while the TCLP test highly exaggerates the leaching of metals. The paper also presents results of differential acid neutralization (d-AN) analysis compared with mineralogical study done by scanning electron microscopy and X-ray diffraction analysis. The d-AN coupled with Eh-pH (Pourbaix) diagrams were proven to be a new effective method for analysis of amorphous solidified structure.

  13. Proceedings of the international seminar on chemistry and process engineering for high-level liquid waste solidification

    International Nuclear Information System (INIS)

    Odoj, R.; Merz, E.

    1981-06-01

    The proceedings record a very distinct phase of the chemistry and process engineering for high-level liquid waste solidification in the past years. The main purpose is to provide solutions which guarantee sufficient safe and economically acceptable measure causing no adverse consequence to man and his environment. (DG)

  14. Leaching of solidified TRU-contaminated incinerator ash

    International Nuclear Information System (INIS)

    Fuhrmann, M.; Colombo, P.

    1984-01-01

    Leach rate and cumulative fractional releases of plutonium were determined for a series of laboratory-scale waste forms containing transuranic (TRU) contaminated incinerator ash. The solidification agents from which these waste forms were produced are commercially available materials for radioactive waste disposal. The leachants simulate groundwaters with chemical compositions that are indiginous to different geological media proposed for repositories. In this study TRU-contaminated ash was incorporated into waste forms fabricated with portland type I cement, urea-formaldehyde, polyester-styrene or Pioneer 221 bitumen. The ash was generated at the dual-chamber incinerator at the Rocky Flats Plant. These waste forms contained between 1.25 x 10 -2 and 4.4 x 10 -2 Ci (depending on the solidification agent) of mixed TRU isotopes comprised primarily of 239 Pu and 240 Pu. Five leachant solutions were prepared consisting of: (1) demineralized water, (2) simulated brine, (3) simplified sodium-dominated groundwater (30 meq NaCl/liter), (4) simplified calcium-dominated groundwater (30 meq CaCl 2 /liter), and (5) simplified bicarbonate-dominated groundwater (30 meq NaHCO 3 /liter). Cumulative fractional releases were found to vary significantly with different leachants and solidification agents. In all cases waste forms leached in brine gave the lowest leach rates. Urea-formaldehyde had the greatest release of radionuclides while polyester-styrene and portland cement had approximately equivalent fractional releases. Cement cured for 210 days retained radionuclides three times more effectively than cement cured only 30 days

  15. Comparative assessment of TRU waste forms and processes. Volume II. Waste form data, process descriptions, and costs

    International Nuclear Information System (INIS)

    Ross, W.A.; Lokken, R.O.; May, R.P.; Roberts, F.P.; Thornhill, R.E.; Timmerman, C.L.; Treat, R.L.; Westsik, J.H. Jr.

    1982-09-01

    This volume contains supporting information for the comparative assessment of the transuranic waste forms and processes summarized in Volume I. Detailed data on the characterization of the waste forms selected for the assessment, process descriptions, and cost information are provided. The purpose of this volume is to provide additional information that may be useful when using the data in Volume I and to provide greater detail on particular waste forms and processes. Volume II is divided into two sections and two appendixes. The first section provides information on the preparation of the waste form specimens used in this study and additional characterization data in support of that in Volume I. The second section includes detailed process descriptions for the eight processes evaluated. Appendix A lists the results of MCC-1 leach test and Appendix B lists additional cost data. 56 figures, 12 tables

  16. Melting, solidification, remelting, and separation of glass and metal

    International Nuclear Information System (INIS)

    Ebadian, M.A.; Xin, R.C.; Liu, Y.Z.

    1998-01-01

    Several high-temperature vitrification technologies have been developed for the treatment of a wide range of mixed waste types in both the low-level waste and transuranic (TRU) mixed waste categories currently in storage at DOE sites throughout the nation. The products of these processes are an oxide slag phase and a reduced metal phase. The metal phase has the potential to be recycled within the DOE Complex. Enhanced slag/metal separation methods are needed to support these processes. This research project involves an experimental investigation of the melting, solidification, remelting, and separation of glass and metal and the development of an efficient separation technology. The ultimate goal of this project is to find an efficient way to separate the slag phase from the metal phase in the molten state. This two-year project commenced in October 1995 (FY96). In the first fiscal year, the following tasks were accomplished: (1) A literature review and an assessment of the baseline glass and metal separation technologies were performed. The results indicated that the baseline technology yields a high percentage of glass in the metal phase, requiring further separation. (2) The main melting and solidification system setup was established. A number of melting and solidification tests were conducted. (3) Temperature distribution, solidification patterns, and flow field in the molten metal pool were simulated numerically for the solidification processes of molten aluminum and iron steel. (4) Initial designs of the laboratory-scale DCS and CS technologies were also completed. The principal demonstration separation units were constructed. (5) An application for a patent for an innovative liquid-liquid separation technology was submitted and is pending

  17. In situ synchrotron x-ray characterization of microstructure formation in solidification processing of Al-based metallic alloys

    International Nuclear Information System (INIS)

    Billia, Bernard; Nguyen-Thi, Henri; Mangelinck-Noel, Nathalie

    2010-01-01

    The microstructure formed during the solidification step has a major influence on the properties of materials processed by major techniques (casting, welding ...). In situ and real-time characterization by synchrotron X-ray imaging is the method of choice to unveil the dynamical formation of the solidification microstructure in metallic alloys, and thus provide precise data for the critical validation of the theoretical predictions that is needed for sound advancement of modeling and numerical simulation. After a description of the experimental procedure used at the European Synchrotron Radiation Facility (ESRF), dynamical phenomena in the formation of the grain structure and dendritic or equiaxed solidification microstructure in Al-based alloys are presented. Beyond fluid flow interaction, earth gravity induces stresses, deformation and fragmentation in the dendritic mush. Settling of dendrite arms and equiaxed grains thus occurs, in particular in the columnar to equiaxed transition. Other types of stresses and strains are caused by the mere formation of the solidification microstructure itself. In white-beam X-ray topography, stresses and strains are manifested by specific contrasts and breaking of the Laue images into several pieces. Finally, quantitative analysis of the grey level in radiographs enables the analysis of solute segregation, which noticeably results in solutal poisoning of growth when equiaxed grains are interacting. (author)

  18. Transport processes in directional solidification and their effects on microstructure development

    Science.gov (United States)

    Mazumder, Prantik

    The processing of materials with unique electronic, mechanical, optical and thermal properties plays a crucial role in modern technology. The quality of these materials depend strongly on the microstructures and the solute/dopant fields in the solid product, that are strongly influenced by the intricate coupling of heat and mass transfer and melt flow in the growth systems. An integrated research program is developed that include precisely characterized experiments and detailed physical and numerical modeling of the complex transport and dynamical processes. Direct numerical simulation of the solidification process is carried out that takes into account the unsteady thermo-solutal convection in the vertical Bridgman crystal growth system, and accurately models the thermal interaction between the furnace and the ampoule by appropriately using experimentally measured thermal profiles. The flow instabilities and transitions and the nonlinear evolution following the transitions are investigated by time series and flow pattern analysis. A range of complex dynamical behavior is predicted with increasing thermal Rayleigh number. The route to chaos appears as: steady convection --> transient mono-periodic --> transient bi-periodic --> transient quasiperiodic --> transient intermittent oscillation- relaxation --> stable intermittent oscillation-relaxation attractor. The spatio-temporal dynamics of the melt flow is found to be directly related to the spatial patterns observed experimentally in the solidified crystals. The application of the model to two phase Sn-Cd peritectic alloys showed that a new class of tree-like oscillating microstructure develops in the solid phase due to unsteady thermo-solutal convection in the liquid melt. These oscillating layered structures can give the illusion of band structures on a plane of polish. The model is applied to single phase solidification in the Al-Cu and Pb-Sn systems to characterize the effect of convection on the macroscopic

  19. Solidification of metal chloride waste from pyrochemical process via dechlorination-chlorination reaction system

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.S.; Cho, I.H.; Lee, K.R.; Choi, J.H.; Eun, H.C.; Kim, I.T.; Park, G.I. [Korea Atomic Energy Research Inst., Deajeon (Korea, Republic of)

    2014-07-01

    The metal chloride wastes generated from the pyro-chemical process to recover uranium and TRUs has been considered as a problematic waste due to the high volatility and low compatibility with conventional silicate glass. Our research group has suggested the dechlorination approach for the solidification of this kind of waste by using a synthetic composite, SAP (SiO{sub 2}-Al{sub 2}O{sub 3}-P{sub 2}O{sub 5}). During the dechlorination, metal elements are chemically interacted with the inorganic composite, SAP, while chlorine is vaporized as gaseous chlorine. Metal elements in the salt were immobilized into phosphate and silicate glass which are uniformly distributed in tens of nm scale. During the dechlorination, gaseous chlorine is captured by Li{sub 2}O-Li{sub 2}O{sub 2} composite that can be converted into metal chloride (LiCl). About 98wt% of oxide composite was converted into LiCl that can be used as an electrolyte in the electrochemical process. The method suggested in this study can provide a chance to minimize the waste volume for the final disposal of salt wastes from a pyro-chemical process. (author)

  20. A solidification/stabilization process for wastewater treatment sludge from a primary copper smelter

    Directory of Open Access Journals (Sweden)

    Ivšić-Bajčeta Dragana

    2013-01-01

    Full Text Available Wastewater treatment sludge from primary copper smelter is characterized as hazardous waste that requires treatment prior disposal due to significant amount of heavy metals and arsenic. The aim of the presented study was to investigate the feasibility and the effectiveness of solidification/stabilization process of the sludge using fly ash and lime as binders. The effectiveness of the process was evaluated by Unconfined Compressive Strength (UCS testing, leaching tests (EN 12457-4 and Toxicity Characteristic Leaching Procedure (TCLP and Acid Neutralization Capacity (ANC test. All samples reached target UCS of 0.35 MPa. Calcium to silicon concentration ratio (cCa/cSi, determined by X-Ray Fluorescence (XRF analysis, was identified as main factor governing strength development. Inductively coupled plasma-optical emission spectrometry (ICP-OES analyses of solutions after leaching tests showed excellent stabilization of Cu, Ni, Pb and Zn (above 99 % and arsenic (above 90 % in samples with high Ca(OH2 content. Results of ANC test indicated that buffering capacity of solidified material linearly depended on Ca concentration in FA and lime. Sample with 20 % of binder heaving 50 % of FA and 50 % of lime met all requirements to be safely disposed. [Projekat Ministarstva nauke Republike Srbije, br. 34033

  1. TRU assay system and measurements

    International Nuclear Information System (INIS)

    Brodzinski, R.L.

    1984-02-01

    The measurement of the transuranic content of nuclear products or process residues has become increasingly important for the recovery of fissionable material from spent fuel elements, the identification of commercial fuel elements which have not yet reached full burnup, the measurement and recovery of transuranics from discarded or stored waste materials, the determination of the transuranic content in high gamma activity waste material scheduled for disposal, compliance with 10CFR61 by land burial operators/shippers, and the satisfaction of accountability requirements. Active neutron interrogation techniques measure either the prompt neutrons or the beta delayed neutrons from fission products following induced fission. These techniques normally only measure fissile transuranics ( 235 U, 239 Pu, and 241 Pu) and are commonly applied only to contact handleable waste. Passive neutron interrogation techniques, on the other hand, are capable of measuring all transuranics except 235 U with adequate sensitivity and will work on both contact handleable and high gamma activity wastes. Since the passive techniques are senstitive to a wider spectrum of transuranic isotopes than the active techniques, substantially less complex and less expensive than the active systems, and they have proven techniques for measuring small quantities of TRU in high gamma activity packages, the passive neutron TRU assay technology was chosen for development into the instruments discussed in this paper

  2. Formation and magic number characteristics of clusters formed during solidification processes

    International Nuclear Information System (INIS)

    Liu Rangsu; Dong Kejun; Tian Zean; Liu Hairong; Peng Ping; Yu Aibing

    2007-01-01

    A molecular dynamics simulation study has been performed for a large-sized system consisting of 10 6 liquid metal Al atoms to investigate the formation and magic number characteristics of various clusters formed during solidification processes. The cluster-type index method (CTIM) is adopted to describe various types of cluster by basic clusters. It is demonstrated that the icosahedral cluster (12 0 12 0) is the most important basic cluster, and that it plays a critical role in the microstructure transition. A new statistical method has been proposed to classify the clusters as some group levels according to the numbers of basic clusters contained in each cluster. The magic numbers can be determined by the respective peak value positions of different group levels of clusters, and the magic number sequence in the system is 13, 19, 25(27), 31(33), 38(40), 42(45), 48(51), 55(59), 61(65), 67,... the numbers in the brackets are the second magic number of the corresponding group levels of clusters. This magic number sequence is in good agreement with the experimental results obtained by Schriver and Harris et al, and the experimental results can be reasonably well explained

  3. Disaster forming reasons on fire explosion at an asphalt solidification processing facility

    International Nuclear Information System (INIS)

    Hasegawa, Kazutoshi; Li Yongfu; Sun Jinhua

    2002-01-01

    Disaster forming reasons on fire explosion accident at an asphalt solidification processing facility of the Power Reactor and Nuclear Fuel Development Corporation formed on 1997 was elucidated. Mixture of salts composing of nitrates, nitrites, and so on with asphalt was filled into a drum at about 180 centigrade, and generated disaster during its natural cooling after about 20 hours. Its reason consisted in change of production condition to make liquid wastes of batches 29 and 30 producing the mixture to contain about 7.7 g/L of salts and liquid wastes supplying rate to reduce to about 160 mL/h. The liquid wastes were mixed with asphalt heated to temperature of about 250 centigrade, when it contained a lot of NaHCO 3 into the salts particles on filling the mixture because moisture was evaporated more rapidly under pressure of phosphates based on the change of production condition. NaHCO 3 directly decomposed to make the salts particles porous and to form a weak redox reaction based on boundary reaction appearing at temperature range from 160 to 200 centigrades. By this reaction, the mixture filled into drum generated thermal accumulation to fire the mixture. (G.K.)

  4. Evaluation of mechanical properties for spherical magnetic regenerator materials fabricated by rapid solidification process

    International Nuclear Information System (INIS)

    Okamura, M.; Sori, N.; Saito, A.

    1997-01-01

    Various magnetic regenerator materials, such as Er 3 Ni, Er 3 Co and ErNi, are fabricated in the form of a spherical particle by a rapid solidification process. 4 K level refrigeration has been obtained by a GM refrigerator using these materials. However, the magnetic regenerator materials are considered brittle, as they are intermetallic compounds. It is important to evaluate the mechanical properties of these materials to confirm reliability as a regenerator material. In this paper, experimental results of compression and vibration tests for magnetic regenerator materials are described. The technical point of this study is to use spherical particles as test samples. The compressive stress of 20 MPa was applied to these spherical particles and no fractured spheres were observed. Similarly, no fractured spheres were found after the vibration test, in which the maximum acceleration was 30 X 9.8 m/s 2 and the number of vibration times was 1 X 10 6 , insofar as there was no room to stir spherical particles in a regenerator. In practice, the reliability of magnetic regenerator materials has been confirmed by a long-run test of 7,000 h in a usual GM refrigerator

  5. Process envelopes for stabilisation/solidification of contaminated soil using lime-slag blend.

    Science.gov (United States)

    Kogbara, Reginald B; Yi, Yaolin; Al-Tabbaa, Abir

    2011-09-01

    Stabilisation/solidification (S/S) has emerged as an efficient and cost-effective technology for the treatment of contaminated soils. However, the performance of S/S-treated soils is governed by several intercorrelated variables, which complicates the optimisation of the treatment process design. Therefore, it is desirable to develop process envelopes, which define the range of operating variables that result in acceptable performance. In this work, process envelopes were developed for S/S treatment of contaminated soil with a blend of hydrated lime (hlime) and ground granulated blast furnace slag (GGBS) as the binder (hlime/GGBS = 1:4). A sand contaminated with a mixture of heavy metals and petroleum hydrocarbons was treated with 5%, 10% and 20% binder dosages, at different water contents. The effectiveness of the treatment was assessed using unconfined compressive strength (UCS), permeability, acid neutralisation capacity and contaminant leachability with pH, at set periods. The UCS values obtained after 28 days of treatment were up to ∼800 kPa, which is quite low, and permeability was ∼10(-8) m/s, which is higher than might be required. However, these values might be acceptable in some scenarios. The binder significantly reduced the leachability of cadmium and nickel. With the 20% dosage, both metals met the waste acceptance criteria for inert waste landfill and relevant environmental quality standards. The results show that greater than 20% dosage would be required to achieve a balance of acceptable mechanical and leaching properties. Overall, the process envelopes for different performance criteria depend on the end-use of the treated material.

  6. TRU partnership-benefits to the national TRU program

    International Nuclear Information System (INIS)

    Lippis, J.; Lott, S.A.

    1995-01-01

    Because increased regulatory authority has been given to the states, the management of transuranic (TRU) wastes varies considerably. One effective tool for facilitating better communications, coordination, and cooperation among the generator/storage sites is the formation of topic specific interface working groups. The National TRU Program supports these groups, and in 1994, a policy was adopted to manage these interface working groups

  7. Progress on Numerical Modeling of the Dispersion of Ceramic Nanoparticles During Ultrasonic Processing and Solidification of Al-Based Nanocomposites

    Science.gov (United States)

    Zhang, Daojie; Nastac, Laurentiu

    2016-12-01

    In present study, 6061- and A356-based nano-composites are fabricated by using the ultrasonic stirring technology (UST) in a coreless induction furnace. SiC nanoparticles are used as the reinforcement. Nanoparticles are added into the molten metal and then dispersed by ultrasonic cavitation and acoustic streaming assisted by electromagnetic stirring. The applied UST parameters in the current experiments are used to validate a recently developed magneto-hydro-dynamics (MHD) model, which is capable of modeling the cavitation and nanoparticle dispersion during UST processing. The MHD model accounts for turbulent fluid flow, heat transfer and solidification, and electromagnetic field, as well as the complex interaction between the nanoparticles and both the molten and solidified alloys by using ANSYS Maxwell and ANSYS Fluent. Molecular dynamics (MD) simulations are conducted to analyze the complex interactions between the nanoparticle and the liquid/solid interface. The current modeling results demonstrate that a strong flow can disperse the nanoparticles relatively well during molten metal and solidification processes. MD simulation results prove that ultrafine particles (10 nm) will be engulfed by the solidification front instead of being pushed, which is beneficial for nano-dispersion.

  8. Crystallization characteristics of cast aluminum alloys during a unidirectional solidification process

    Energy Technology Data Exchange (ETDEWEB)

    Okayasu, Mitsuhiro, E-mail: mitsuhiro.okayasu@utoronto.ca; Takeuchi, Shuhei

    2015-05-01

    The crystal orientation characteristics of cast Al–Si, Al–Cu and Al–Mg alloys produced by a unidirectional solidification process are examined. Two distinct crystal orientation patterns are observed: uniform and random formation. A uniform crystal orientation is created by columnar growth of α-Al dendrites in the alloys with low proportions of alloying element, e.g., the Al–Si alloy (with Si <12.6%) and the Al–Cu and Al–Mg alloys (with Cu and Mg <2%). A uniformly organized crystal orientation with [100] direction is created by columnar growth of α-Al dendrites. With increasing proportion of alloying element (>2% Cu or Mg), the uniform crystal orientations collapse in the Al–Cu and Al–Mg alloys, owing to interruption of the columnar α-Al dendrite growth as a result of different dynamics of the alloying atoms and the creation of a core for the eutectic phases. For the hypo-eutectic Al–Si alloys, a uniform crystal orientation is obtained. In contrast, a random orientation can be detected in the hyper-eutectic Al–Si alloy (15% Si), which results from interruption of the growth of the α-Al dendrites due to precipitation of primary Si particles. There is no clear effect of crystal formation on ultimate tensile strength (UTS), whereas crystal orientation does influence the material ductility, with the alloys with a uniform crystal orientation being elongated beyond their UTS points and with necking occurring in the test specimens. In contrast, the alloys with a nonuniform crystal orientation are not elongated beyond their UTS points.

  9. Crystallization characteristics of cast aluminum alloys during a unidirectional solidification process

    International Nuclear Information System (INIS)

    Okayasu, Mitsuhiro; Takeuchi, Shuhei

    2015-01-01

    The crystal orientation characteristics of cast Al–Si, Al–Cu and Al–Mg alloys produced by a unidirectional solidification process are examined. Two distinct crystal orientation patterns are observed: uniform and random formation. A uniform crystal orientation is created by columnar growth of α-Al dendrites in the alloys with low proportions of alloying element, e.g., the Al–Si alloy (with Si <12.6%) and the Al–Cu and Al–Mg alloys (with Cu and Mg <2%). A uniformly organized crystal orientation with [100] direction is created by columnar growth of α-Al dendrites. With increasing proportion of alloying element (>2% Cu or Mg), the uniform crystal orientations collapse in the Al–Cu and Al–Mg alloys, owing to interruption of the columnar α-Al dendrite growth as a result of different dynamics of the alloying atoms and the creation of a core for the eutectic phases. For the hypo-eutectic Al–Si alloys, a uniform crystal orientation is obtained. In contrast, a random orientation can be detected in the hyper-eutectic Al–Si alloy (15% Si), which results from interruption of the growth of the α-Al dendrites due to precipitation of primary Si particles. There is no clear effect of crystal formation on ultimate tensile strength (UTS), whereas crystal orientation does influence the material ductility, with the alloys with a uniform crystal orientation being elongated beyond their UTS points and with necking occurring in the test specimens. In contrast, the alloys with a nonuniform crystal orientation are not elongated beyond their UTS points

  10. The fundamental study for Y2O3 stabilized ZrO2 containing simulated TRU

    International Nuclear Information System (INIS)

    Kuramoto, Ken-ichi; Kamizono, Hiroshi; Hayakawa, Issei; Muraoka, Susumu; Yanagi, Tadashi.

    1991-06-01

    Borosilicate glass waste form is considered to be the most suitable material for the immobilization of high-level nuclear waste (HLW). However when the salt-free process on Purex method is adopted and the group partitioning technique of HLW is completely developed, ceramic waste forms which are excellent in thermal stability seem better for the immobilization of hazardous TRU elements. This work is a fundamental study on the solidification of TRU with Y 2 O 3 -stabilized ZrO 2 . In this work, Ce and Nd were used as substitutes for Pu and Am or Cm. TZ-8Y (submicron powder) and designed Ce(NO 3 ) 3 or Nd(NO 3 ) 3 solution were mixed into paste. After dried, the paste was pelletized by the rubber press, and then sintered at 1400degC for 16h. Densities of the sintered pellets were measured and their microstructure was observed by X-ray diffraction and scanning electron microscopy. The results showed that (1) the relative density of ceramic pellet sample was as high as 96.4%, (2) each element was distributed homogeneously and only cubic phase existed. From leach tests in nitric acid and distilled water at 150degC, those ceramic pellet samples showed aqueous corrosion rates which were about 10 2 to 10 3 times lower than that of a glass waste form(PO500). (author)

  11. Hybrid Microwave Treatment of SRS TRU and Mixed Wastes

    International Nuclear Information System (INIS)

    Wicks, G.G.

    1999-01-01

    A new process, using hybrid microwave energy, has been developed as part of the Strategic Research and Development program and successfully applied to treatment of a wide variety of non-radioactive materials, representative of SRS transuranic (TRU) and mixed wastes. Over 35 simulated (non-radioactive) TRU and mixed waste materials were processed individually, as well as in mixed batches, using hybrid microwave energy, a new technology now being patented by Westinghouse Savannah River Company (WSRC)

  12. In situ observations of solidification processes in γ-TiAl alloys by synchrotron radiation

    International Nuclear Information System (INIS)

    Shuleshova, Olga; Holland-Moritz, Dirk; Loeser, Wolfgang; Voss, Andrea; Hartmann, Helena; Hecht, Ulrike; Witusiewicz, Victor T.; Herlach, Dieter M.; Buechner, Bernd

    2010-01-01

    In situ observations of phase transformations involving melts are performed using energy-dispersive diffraction of synchrotron X-rays on electromagnetically levitated γ-TiAl alloys containing Nb. The determined primary solidification modes, confirmed by microstructure analysis, delivered new reliable data about the boundary of the α(Ti) solidification domain, which differs in the various Ti-Al-Nb phase diagram descriptions. These data have been used for a reassessment of the thermodynamic database of the ternary Ti-Al-Nb system. The new description realistically reflects the experimental findings. Liquidus and solidus temperatures determined by the pyrometric method agree fairly well with the calculated values. Direct experimental information on the nature of the reactions along the univariant lines is provided.

  13. Optimization of Magnetically Driven Directional Solidification of Silicon Using Artificial Neural Networks and Gaussian Process Models

    Czech Academy of Sciences Publication Activity Database

    Dropka, N.; Holeňa, Martin

    2017-01-01

    Roč. 471, 1 August (2017), s. 53-61 ISSN 0022-0248 R&D Projects: GA ČR GA17-01251S Institutional support: RVO:67985807 Keywords : computer simulation * fluid flows * magnetic fields * directional solidification * semiconducting silicon Subject RIV: IN - Informatics, Computer Science OBOR OECD: Condensed matter physics (including formerly solid state physics, supercond.) Impact factor: 1.751, year: 2016

  14. Solidification of refractory materials processed in the ultra high vacuum drop tube at the CEREM-Grenoble

    International Nuclear Information System (INIS)

    Vinet, B.; Cini, E.; Tournier, S.; Cortella, L.

    1994-01-01

    Undercooling experiments have been performed on refractory materials by processing in a high drop-tube. Emphasis is put on general aspects of surface and bulk microstructures of solidified droplets, including pure metals (W, Re, Ta, Mo, Nb, Ir, Zr) and alloys (W-Re, Mo-Re, Ta-Zr). It is shown that recrystallization often causes polycrystallinity. Moreover, the microstructure is closely related to undercooling amount prior solidification. The effect of secondary cooling on microstructure can also be studied by quenching the samples in solid tin at the end of free-fall. (authors). 20 refs., 11 figs

  15. Polymer solidification national program

    International Nuclear Information System (INIS)

    Kalb, P.D.; Colombo, P.

    1993-04-01

    Brookhaven National Laboratory (BNL) has developed several new and innovative polymer processes for the solidification of low-level radioactive, hazardous and mixed wastes streams. Polyethylene and modified sulfur cement solidification technologies have undergone steady, gradual development at BNL over the past nine years. During this time they have progressed through each of the stages necessary for logical technology maturation: from process conception, parameter optimization, waste form testing, evaluation of long-term durability, economic analysis, and scale-up feasibility. This technology development represents a significant investment which can potentially provide DOE with both short- and long-term savings

  16. Radioactive waste solidification material

    International Nuclear Information System (INIS)

    Nishihara, Yukio; Wakuta, Kuniharu; Ishizaki, Kanjiro; Koyanagi, Naoaki; Sakamoto, Hiroyuki; Uchida, Ikuo.

    1992-01-01

    The present invention concerns a radioactive waste solidification material containing vermiculite cement used for a vacuum packing type waste processing device, which contains no residue of calcium hydroxide in cement solidification products. No residue of calcium hydroxide means, for example, that peak of Ca(OH) 2 is not recognized in an X ray diffraction device. With such procedures, since calcium sulfoaluminate clinker and Portland cement themselves exhibit water hardening property, and slugs exhibit hydration activity from the early stage, the cement exhibits quick-hardening property, has great extension of long term strength, further, has no shrinking property, less dry- shrinkage, excellent durability, less causing damages such as cracks and peeling as processing products of radioactive wastes, enabling to attain highly safe solidification product. (T.M.)

  17. Al-Si-Re Alloys Cast by the Rapid Solidification Process / Stopy Al-Si-Re Odlewane Metodą Rapid Solidification

    Directory of Open Access Journals (Sweden)

    Szymanek M.

    2015-12-01

    Full Text Available The aim of the studies described in this article was to present the effect of rare earth elements on aluminium alloys produced by an unconventional casting technique. The article gives characteristics of the thin strip of Al-Si-RE alloy produced by Rapid Solidification (RS. The effect of rare earth elements on structure refinement, i.e. on the size of near-eutectic crystallites in an aluminium-silicon alloy, was discussed. To determine the size of crystallites, the Scherrer X-ray diffraction method was used. The results presented capture relationships showing the effect of variable casting parameters and chemical composition on microstructure of the examined alloys. Rapid Solidification applied to Al-Si alloys with the addition of mischmetal (Ce, La, Ne, Pr refines their structure.

  18. Microstructures and mechanical responses of powder metallurgy non-combustive magnesium extruded alloy by rapid solidification process in mass production

    International Nuclear Information System (INIS)

    Kondoh, Katsuyoshi; Hamada, EL-Sayed Ayman; Imai, Hisashi; Umeda, Junko; Jones, Tyrone

    2010-01-01

    Spinning Water Atomization Process (SWAP), which was one of the rapid solidification processes, promised to produce coarse non-combustible magnesium alloy powder with 1-4 mm length, having fine α-Mg grains and Al 2 Ca intermetallic compounds. It had economical and safe benefits in producing coarse Mg alloy powders with very fine microstructures in the mass production process due to its extreme high solidification rate compared to the conventional atomization process. AMX602 (Mg-6%Al-0.5%Mn-2%Ca) powders were compacted at room temperature. Their green compacts with a relative density of about 85% were heated at 573-673 K for 300 s in Ar gas atmosphere, and immediately consolidated by hot extrusion. Microstructure observation and evaluation of mechanical properties of the extruded AMX602 alloys were carried out. The uniform and fine microstructures with grains less than 0.45-0.8 μm via dynamic recrystallization during hot extrusion were observed, and were much small compared to the extruded AMX602 alloy fabricated by using cast ingot. The extremely fine intermetallic compounds 200-500 nm diameter were uniformly distributed in the matrix of powder metallurgy (P/M) extruded alloys. These microstructures caused excellent mechanical properties of the wrought alloys. For example, in the case of AMX602 alloys extruded at 573 K, the tensile strength (TS) of 447 MPa, yield stress (YS) of 425 MPa and 9.6% elongation were obtained.

  19. The role of ultrasonic cavitation in refining the microstructure of aluminum based nanocomposites during the solidification process.

    Science.gov (United States)

    Xuan, Yang; Nastac, Laurentiu

    2018-02-01

    Recent studies showed that the microstructure and mechanical properties of aluminum based nanocomposites can be significantly improved when ultrasonic cavitation and solidification processing is used. This is because ultrasonic cavitation processing plays an important role not only in degassing and dispersion of the nanoparticles, but also in breaking up the dendritic grains and refining the as-cast microstructure. In the present study, A356 alloy and Al 2 O 3 nanoparticles are used as the matrix alloy and the reinforcement, respectively. Nanoparticles were added into the molten A356 alloy and dispersed via ultrasonic cavitation processing. Ultrasonic cavitation was applied over various temperature ranges during molten alloy cooling and solidification to investigate the grain structure formation and the nanoparticle dispersion behavior. Optical Microscopy and Scanning Electron Microscopy were used to investigate in detail the differences in the microstructure characteristics and the nanoparticle distribution. Experimental results indicated that the ultrasonic cavitation processing and Al 2 O 3 nanoparticles play an important role for microstructure refinement. In addition, it was shown in this study that the Al 2 O 3 nanoparticles modified the eutectic phase. Copyright © 2017 Elsevier B.V. All rights reserved.

  20. Behavior of nuclides at plasma melting of TRU wastes

    International Nuclear Information System (INIS)

    Amakawa, Tadashi; Adachi, Kazuo

    2001-01-01

    Arc plasma heating technique can easily be formed at super high temperature, and can carry out stable heating without any effect of physical and chemical properties of the wastes. By focussing to these characteristics, this technique was experimentally investigated on behavior of TRU nuclides when applying TRU wastes forming from reprocessing process of used fuels to melting treatment by using a mimic non-radioactive nuclide. At first, according to mechanism determining the behavior of TRU nuclides, an element (mimic nuclide) to estimate the behavior was selected. And then, to zircaloy with high melting point or steel can simulated to metal and noncombustible wastes and fly ash, the mimic nuclide was added, prior to melting by using the arc plasma heating technique. As a result, on a case of either melting sample, it was elucidated that the nuclides hardly moved into their dusts. Then, the technique seems to be applicable for melting treatment of the TRU wastes. (G.K.)

  1. Low-level radwaste solidification

    International Nuclear Information System (INIS)

    Naughton, M.D.; Miller, C.C.; Nelson, R.A.; Tucker, R.F.

    1983-01-01

    This paper reports on a study of ''Advanced Low-Level Radioactive Waste Treatment Systems'' conducted under an EPRI contract. The object of the study is to identify advanced lowlevel radwaste treatment systems that are commercially available or are expected to be in the near future. The current state-ofthe-art in radwaste solidification technology is presented. Related processing technologies, such as the compaction of dry active waste (DAW), containers available for radwaste disposal, and the regulatory aspects of radwaste transportation and solidification, are described. The chemical and physical properties of the currently acceptable solidification agents, as identified in the Barnwell radwaste burial site license, are examined. The solidification agents investigated are hydraulic cements, thermoplastic polymers, and thermosetting polymers. It is concluded that solidification processes are complex and depend not only on the chemical and physical properties of the binder material and the waste, but also on how these materials are mixed

  2. Cement solidification method for miscellaneous radioactive solid, processing device and processing tool therefor

    International Nuclear Information System (INIS)

    Mihara, Shigeru; Suzuki, Kazunori; Hasegawa, Akira.

    1994-01-01

    A basket made of a metal net and a lid with a spacer constituting a processing tool for processing miscellaneous radioactive solid wastes is formed as a mesh which scarcely passes the miscellaneous solids but pass mortars. The size of the mesh is usually from about 10 to 30mm. Since this mesh allows fine solids approximate to powders such as burning ashes and heat insulation materials, they fall to the bottom of a dram can, to cause corrosion. Then, the corners of the bottom and the bottom of the dram can are coated with cement. The miscellaneous solid wastes are contained, and the lid of a metal net having a spacer at the upper portion thereof is set, a provisional lid is put on, and it is evacuated, and mortars are injected. Since there is a possibility that light and fine radioactive powders are exposed on the surface of the mortars coagulated and hardened by curing, conditioning for further adding mortars is applied for securing the mortars in order to prevent scattering of the radioactive powders. With such procedures, a satisfactory safe solidified products can be formed. (T.M.)

  3. Advancement of Solidification Processing Technology Through Real Time X-Ray Transmission Microscopy: Sample Preparation

    Science.gov (United States)

    Stefanescu, D. M.; Curreri, P. A.

    1996-01-01

    Two types of samples were prepared for the real time X-ray transmission microscopy (XTM) characterization. In the first series directional solidification experiments were carried out to evaluate the critical velocity of engulfment of zirconia particles in the Al and Al-Ni eutectic matrix under ground (l-g) conditions. The particle distribution in the samples was recorded on video before and after the samples were directionally solidified. In the second series samples of the above two type of composites were prepared for directional solidification runs to be carried out on the Advanced Gradient Heating Facility (AGHF) aboard the space shuttle during the LMS mission in June 1996. X-ray microscopy proved to be an invaluable tool for characterizing the particle distribution in the metal matrix samples. This kind of analysis helped in determining accurately the critical velocity of engulfment of ceramic particles by the melt interface in the opaque metal matrix composites. The quality of the cast samples with respect to porosity and instrumented thermocouple sheath breakage or shift could be easily viewed and thus helped in selecting samples for the space shuttle experiments. Summarizing the merits of this technique it can be stated that this technique enabled the use of cast metal matrix composite samples since the particle location was known prior to the experiment.

  4. Plastic solidification of radioactive wastes

    International Nuclear Information System (INIS)

    Moriyama, Noboru

    1981-01-01

    Over 20 years have elapsed after the start of nuclear power development, and the nuclear power generation in Japan now exceeds the level of 10,000 MW. In order to meet the energy demands, the problem of the treatment and disposal of radioactive wastes produced in nuclear power stations must be solved. The purpose of the plastic solidification of such wastes is to immobilize the contained radionuclides, same as other solidification methods, to provide the first barrier against their move into the environment. The following matters are described: the nuclear power generation in Japan, the radioactive wastes from LWR plants, the position of plastic solidification, the status of plastic solidification in overseas countries and in Japan, the solidification process for radioactive wastes with polyethylene, and the properties of solidified products, and the leachability of radionuclides in asphalt solids. (J.P.N.)

  5. Final Hanford Site Transuranic (TRU) Waste Characterization QA Project Plan

    International Nuclear Information System (INIS)

    GREAGER, T.M.

    2000-01-01

    The Quality Assurance Project Plan (QAPjP) has been prepared for waste characterization activities to be conducted by the Transuranic (TRU) Project at the Hanford Site to meet requirements set forth in the Waste Isolation Pilot Plan (WIPP) Hazardous Waste Facility Permit, 4890139088-TSDF, Attachment B, including Attachments B1 through B6 (WAP) (DOE, 1999a). The QAPjP describes the waste characterization requirements and includes test methods, details of planned waste sampling and analysis, and a description of the waste characterization and verification process. In addition, the QAPjP includes a description of the quality assurance/quality control (QA/QC) requirements for the waste characterization program. Before TRU waste is shipped to the WIPP site by the TRU Project, all applicable requirements of the QAPjP shall be implemented. Additional requirements necessary for transportation to waste disposal at WIPP can be found in the ''Quality Assurance Program Document'' (DOE 1999b) and HNF-2600, ''Hanford Site Transuranic Waste Certification Plan.'' TRU mixed waste contains both TRU radioactive and hazardous components, as defined in the WLPP-WAP. The waste is designated and separately packaged as either contact-handled (CH) or remote-handled (RH), based on the radiological dose rate at the surface of the waste container. RH TRU wastes are not currently shipped to the WIPP facility

  6. Thermal treatment for TRU waste sorting

    International Nuclear Information System (INIS)

    Sasaki, Toshiki; Aoyama, Yoshio; Yamashita, Toshiyuki

    2009-03-01

    A thermal treatment that can automatically unpack TRU waste and remove hazardous materials has been developed to reduce the risk of radiation exposure and save operation cost. The thermal treatment is a process of removing plastic wrapping and hazardous material from TRU waste by heating waste at 500 to 700degC. Plastic wrappings of simulated wastes were removed using a laboratory scale thermal treatment system. Celluloses and isoprene rubbers that must be removed from waste for disposal were pyrolyzed by the treatment. Although the thermal treatment can separate lead and aluminum from the waste, a further technical development is needed to separate lead and aluminum. A demonstration scale thermal treatment system that comprises a rotary kiln with a jacket water cooler and a rotating inner cage for lead and aluminum separation is discussed. A clogging prevention system against zinc chloride, a lead and aluminum accumulation system, and a detection system for spray cans that possibly cause explosion and fire are also discussed. Future technology development subjects for the TRU waste thermal treatment system are summarized. (author)

  7. Management of metal-bearing industrial solid waste by stabilization/solidification process

    Energy Technology Data Exchange (ETDEWEB)

    Sunitha, C.; Palanivelu, K. [Anna University, Chennai (India). Centre for Environmental Studies

    2005-07-01

    Metal-bearing sludge from an electroplating industry was immobilised by the solidification stabilisation treatment method. Reduction of the leachability of metals from the waste was studied in different combinations of waste and additives - cement, lime and fly ash. The study revealed that the optimum proportion for cement: metal hydroxide sludge: fly ash as 1:2:2 is the best. The encapsulation efficiency calculated for the metals such as Cu, Cr, Ni, Pb, and Zn was above 92%. The unconfined compressive strength (UCS) for the developed block was found to be 11.5 kg/cm{sup 2} after curing. The toxicity characteristic leach test (TCLP) test reveals that the heavy metal content in the leachate was well below the maximum permissible limit of WHO drinking water standard. 10 refs., 6 tabs.

  8. Accelerating solidification process simulation for large-sized system of liquid metal atoms using GPU with CUDA

    Energy Technology Data Exchange (ETDEWEB)

    Jie, Liang [School of Information Science and Engineering, Hunan University, Changshang, 410082 (China); Li, KenLi, E-mail: lkl@hnu.edu.cn [School of Information Science and Engineering, Hunan University, Changshang, 410082 (China); National Supercomputing Center in Changsha, 410082 (China); Shi, Lin [School of Information Science and Engineering, Hunan University, Changshang, 410082 (China); Liu, RangSu [School of Physics and Micro Electronic, Hunan University, Changshang, 410082 (China); Mei, Jing [School of Information Science and Engineering, Hunan University, Changshang, 410082 (China)

    2014-01-15

    Molecular dynamics simulation is a powerful tool to simulate and analyze complex physical processes and phenomena at atomic characteristic for predicting the natural time-evolution of a system of atoms. Precise simulation of physical processes has strong requirements both in the simulation size and computing timescale. Therefore, finding available computing resources is crucial to accelerate computation. However, a tremendous computational resource (GPGPU) are recently being utilized for general purpose computing due to its high performance of floating-point arithmetic operation, wide memory bandwidth and enhanced programmability. As for the most time-consuming component in MD simulation calculation during the case of studying liquid metal solidification processes, this paper presents a fine-grained spatial decomposition method to accelerate the computation of update of neighbor lists and interaction force calculation by take advantage of modern graphics processors units (GPU), enlarging the scale of the simulation system to a simulation system involving 10 000 000 atoms. In addition, a number of evaluations and tests, ranging from executions on different precision enabled-CUDA versions, over various types of GPU (NVIDIA 480GTX, 580GTX and M2050) to CPU clusters with different number of CPU cores are discussed. The experimental results demonstrate that GPU-based calculations are typically 9∼11 times faster than the corresponding sequential execution and approximately 1.5∼2 times faster than 16 CPU cores clusters implementations. On the basis of the simulated results, the comparisons between the theoretical results and the experimental ones are executed, and the good agreement between the two and more complete and larger cluster structures in the actual macroscopic materials are observed. Moreover, different nucleation and evolution mechanism of nano-clusters and nano-crystals formed in the processes of metal solidification is observed with large

  9. Study of stabilization/solidification processes (of solid porous wastes) based on hydraulic or bituminous binders; Etude des procedes de stabilisation/solidification (des dechets solides poreux) a base de liants hydrauliques ou de liants bitumineux

    Energy Technology Data Exchange (ETDEWEB)

    Sing-Teniere, Ch.

    1998-02-01

    The first part of this thesis presents the regulatory framework and the technical context linked with the study of stabilized/solidified wastes and with the evaluation of stabilization/solidification processes. A presentation of the two type of ultimate wastes under study (a used catalyst and an activated charcoal) and an analysis of the processes is given. The second part is devoted to the experimental characterization of both types of porous wastes. The third part deals with the processing of such wastes using an hydraulic binder. The study stresses on both on the stabilization/solidification efficiency of the process and on the conditions of its implementation. The same work is made for a process that uses a bituminous binder. Some choice criteria for the selection of the better process are deduced from the examination of the overall data collected. The waste characterization methodology is applied six times: two times for the raw wastes, two times for the same wastes processed with an hydraulic binder, and two times for the same wastes processed with a bituminous binder. (J.S.)

  10. IMPROVEMENTS IN HANFORD TRANSURANIC (TRU) PROGRAM UTILIZING SYSTEMS MODELING AND ANALYSES

    International Nuclear Information System (INIS)

    UYTIOCO EM

    2007-01-01

    Hanford's Transuranic (TRU) Program is responsible for certifying contact-handled (CH) TRU waste and shipping the certified waste to the Waste Isolation Pilot Plant (WIPP). Hanford's CH TRU waste includes material that is in retrievable storage as well as above ground storage, and newly generated waste. Certifying a typical container entails retrieving and then characterizing it (Real-Time Radiography, Non-Destructive Assay, and Head Space Gas Sampling), validating records (data review and reconciliation), and designating the container for a payload. The certified payload is then shipped to WIPP. Systems modeling and analysis techniques were applied to Hanford's TRU Program to help streamline the certification process and increase shipping rates

  11. TRU-waste decontamination and size reduction review, June 1983, US DOE/PNC technology exchange

    International Nuclear Information System (INIS)

    Becker, G.W. Jr.

    1983-01-01

    A review of transuranic (TRU) noncombustible waste decontamination and size reduction technology is presented. Electropolishing, vibratory cleaning, and spray decontamination processes developed at Battelle Pacific Northwest Laboratory (PNL) and Savannah River Laboratory (SRL) are highlighted. TRU waste size reduction processes at (PNL), Los Alamos National Laboratory (LANL), the Rocky Flats Plant (RFP), and SRL are also highlighted

  12. Status of ERDA TRU waste packaging study

    International Nuclear Information System (INIS)

    Doty, J.W. Jr.

    1977-01-01

    This paper discusses the status of Task 3 of the TRU Waste Cyclone Drum Incinerator and Treatment System program. This task covers acceptable TRU packaging for interim storage and terminal isolation. The kind of TRU wastes generated by contractors and its transport are discussed. Both drum and box systems are desirable

  13. Solidification method of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Tsutomu; Chino, Koichi; Sasahira, Akira; Ikeda, Takashi

    1992-07-24

    Metal solidification material can completely seal radioactive wastes and it has high sealing effect even if a trace amount of evaporation should be caused. In addition, the solidification operation can be conducted safely by using a metal having a melting point of lower than that of the decomposition temperature of the radioactive wastes. Further, the radioactive wastes having a possibility of evaporation and scattering along with oxidation can be solidified in a stable form by putting the solidification system under an inert gas atmosphere. Then in the present invention, a metal is selected as a solidification material for radioactive wastes, and a metal, for example, lead or tin having a melting point of lower than that of the decomposition temperature of the wastes is used in order to prevent the release of the wastes during the solidification operation. Radioactive wastes which are unstable in air and scatter easily, for example, Ru or the like can be converted into a stable solidification product by conducting the solidification processing under an inert gas atmosphere. (T.M.).

  14. Final environmental assessment: TRU waste drum staging building, Technical Area 55, Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    1996-01-01

    Much of the US Department of Energy's (DOE's) research on plutonium metallurgy and plutonium processing is performed at Los Alamos National Laboratory (LANL), in Los Alamos, New Mexico. LANL's main facility for plutonium research is the Plutonium Facility, also referred to as Technical Area 55 (TA-55). The main laboratory building for plutonium work within the Plutonium Facility (TA-55) is the Plutonium Facility Building 4, or PF-4. This Environmental Assessment (EA) analyzes the potential environmental effects that would be expected to occur if DOE were to stage sealed containers of transuranic (TRU) and TRU mixed waste in a support building at the Plutonium Facility (TA-55) that is adjacent to PF-4. At present, the waste containers are staged in the basement of PF-4. The proposed project is to convert an existing support structure (Building 185), a prefabricated metal building on a concrete foundation, and operate it as a temporary staging facility for sealed containers of solid TRU and TRU mixed waste. The TRU and TRU mixed wastes would be contained in sealed 55-gallon drums and standard waste boxes as they await approval to be transported to TA-54. The containers would then be transported to a longer term TRU waste storage area at TA-54. The TRU wastes are generated from plutonium operations carried out in PF-4. The drum staging building would also be used to store and prepare for use new, empty TRU waste containers

  15. Production of solidified high level wastes: a cost comparison of solidification processes

    International Nuclear Information System (INIS)

    1977-06-01

    Differential cost estimates of the annual operating and maintenance costs and the capital costs for five HLW Waste Solidification Alternates were developed. The annual operating and maintenance cost estimates included the cost of labor, consumables, utilities, shipping casks, shipping and disposal at a federal repository. The capital cost included the cost of the component, installation and building. The differential cost estimates do not include equipment and facilities which are either shared with the reprocessing facility or are common between all of the alternates. Total annual cost differential between the five waste form alternates is summarized in tabular form. The Borosilicate Glass Alternate has the lowest total annual cost. The other alternates have higher costs which range from $6.6 M to $7.4 M per year higher than the Glass alternate with the Supercalcine being the highest cost at $7.4 M per year differential. The major items in the cost estimates are then disposal costs in the operating cost estimates and the HLW Storage Tanks in the capital cost estimates. The Supercalcine Multibarrier Alternate ships 180 canisters per year more than the other alternates and consequently has a significantly higher operating cost. However, off-setting this the Supercalcine Multibarrier Alternate does not require HLW Storage Tanks for decay because of the high heat conductivity of this product and correspondingly the capital cost for this alternate is significantly lower than the other alternates. The radiological risk values are correlated with the cost evaluation normalized to cost ($)/MWe-yr

  16. Finite element modelling of solidification phenomena

    Indian Academy of Sciences (India)

    Unknown

    Abstract. The process of solidification process is complex in nature and the simulation of such process is required in industry before it is actually undertaken. Finite element method is used to simulate the heat transfer process accompanying the solidification process. The metal and the mould along with the air gap formation ...

  17. Vitrification of TRU wastes at Rocky Flats Plant

    International Nuclear Information System (INIS)

    Williams, P.M.; Johnson, A.J.; Ledford, J.A.

    1979-01-01

    Immobilization of incinerator ash and various noncombustible TRU wastes was investigated. In three different research projects borosilicate glass proved to be the best candidate for TRU waste fixation. This glass has excellent chemical durability, long-term stability in the presence of radiation, and will withstand continuous temperatures up to 400 0 C without devitrification. In addition, wastes prepared in the form of glass will attain densities of approximately 2500 kg/m 3 (2.5 g/cc). The free forming method of producing glass buttons provides a very simple, consistent, low maintenance way of producing a final waste form for transporting and either retrievable or permanent storage for TRU waste. The vitrification process produces a durable glass from the low density ash generated by the fluidized bed incinerator process and provides volume and weight reductions that are superior to other fixation processes. This results in decreased transportation and storage costs

  18. Method of storing solidification products

    International Nuclear Information System (INIS)

    Tani, Yutaro.

    1985-01-01

    Purpose: To enable to efficiently and satisfactorily cool and store solidification products of liquid wastes generated from the reactor spent fuel reprocessing process by a simple facility. Method: Liquid wastes generated from the reactor spent fuel reprocessing process are caused to flow from the upper opening to the inside of a spherical canistor. The opening of the spherical canistor is welded with a lid by a remote control and the liquid wastes are tightly sealed within the spherical canistor as glass solidification products. Spherical canistors having the solidification products tightly sealed therein are sent into and stored in a hopper by the remote control. Further, a blower is driven upon storing to suck cooling air from the cooling air intake port to the inside of the hopper to absorb the decay heat of radioactive materials in the solidification products and the air is discharged from the duct and through the stack to the atmosphere. (Kawakami, Y.)

  19. Thermodynamic Modeling of Sr/TRU Removal

    International Nuclear Information System (INIS)

    Felmy, A.R.

    2000-01-01

    This report summarizes the development and application of a thermodynamic modeling capability designed to treat the Envelope C wastes containing organic complexants. A complete description of the model development is presented. In addition, the model was utilized to help gain insight into the chemical processes responsible for the observed levels of Sr, TRU, Fe, and Cr removal from the diluted feed from tank 241-AN-107 which had been treated with Sr and permanganate. Modeling results are presented for Sr, Nd(III)/Eu(III), Fe, Cr, Mn, and the major electrolyte components of the waste (i.e. NO 3 , NO 2 , F,...). On an overall basis the added Sr is predicted to precipitate as SrCO 3 (c) and the MnO 4 - reduced by the NO 2 - and precipitated as a Mn oxide. These effects result in only minor changes to the bulk electrolyte chemistry, specifically, decreases in NO 2 - and CO 3 2- , and increases in NO 3 - and OH - . All of these predictions are in agreement with the experimental observations. The modeling also indicates that the majority of the Sr, TRU's (or Nd(III)/Eu(III)) analogs, and Fe are tied up with the organic complexants. The Sr and permanganate additions are not predicted to effect these chelate complexes significantly owing to the precipitation of insoluble Mn oxides or SrCO 3 . These insoluble phases maintain low dissolved concentrations of Mn and Sr which do not affect any of the other components tied up with the complexants. It appears that the removal of the Fe and TRU'S during the treatment process is most likely as a result of adsorption or occlusion on/into the Mn oxides or SrCO 3 , not as direct displacement from the complexants into precipitates. Recommendations are made for further studies that are needed to help resolve these issues

  20. TRU drum corrosion task team report

    Energy Technology Data Exchange (ETDEWEB)

    Kooda, K.E.; Lavery, C.A.; Zeek, D.P.

    1996-05-01

    During routine inspections in March 1996, transuranic (TRU) waste drums stored at the Radioactive Waste Management Complex (RWMC) were found with pinholes and leaking fluid. These drums were overpacked, and further inspection discovered over 200 drums with similar corrosion. A task team was assigned to investigate the problem with four specific objectives: to identify any other drums in RWMC TRU storage with pinhole corrosion; to evaluate the adequacy of the RWMC inspection process; to determine the precise mechanism(s) generating the pinhole drum corrosion; and to assess the implications of this event for WIPP certifiability of waste drums. The task team investigations analyzed the source of the pinholes to be Hcl-induced localized pitting corrosion. Hcl formation is directly related to the polychlorinated hydrocarbon volatile organic compounds (VOCs) in the waste. Most of the drums showing pinhole corrosion are from Content Code-003 (CC-003) because they contain the highest amounts of polychlorinated VOCs as determined by headspace gas analysis. CC-001 drums represent the only other content code with a significant number of pinhole corrosion drums because their headspace gas VOC content, although significantly less than CC-003, is far greater than that of the other content codes. The exact mechanisms of Hcl formation could not be determined, but radiolytic and reductive dechlorination and direct reduction of halocarbons were analyzed as the likely operable reactions. The team considered the entire range of feasible options, ranked and prioritized the alternatives, and recommended the optimal solution that maximizes protection of worker and public safety while minimizing impacts on RWMC and TRU program operations.

  1. TRU drum corrosion task team report

    International Nuclear Information System (INIS)

    Kooda, K.E.; Lavery, C.A.; Zeek, D.P.

    1996-05-01

    During routine inspections in March 1996, transuranic (TRU) waste drums stored at the Radioactive Waste Management Complex (RWMC) were found with pinholes and leaking fluid. These drums were overpacked, and further inspection discovered over 200 drums with similar corrosion. A task team was assigned to investigate the problem with four specific objectives: to identify any other drums in RWMC TRU storage with pinhole corrosion; to evaluate the adequacy of the RWMC inspection process; to determine the precise mechanism(s) generating the pinhole drum corrosion; and to assess the implications of this event for WIPP certifiability of waste drums. The task team investigations analyzed the source of the pinholes to be Hcl-induced localized pitting corrosion. Hcl formation is directly related to the polychlorinated hydrocarbon volatile organic compounds (VOCs) in the waste. Most of the drums showing pinhole corrosion are from Content Code-003 (CC-003) because they contain the highest amounts of polychlorinated VOCs as determined by headspace gas analysis. CC-001 drums represent the only other content code with a significant number of pinhole corrosion drums because their headspace gas VOC content, although significantly less than CC-003, is far greater than that of the other content codes. The exact mechanisms of Hcl formation could not be determined, but radiolytic and reductive dechlorination and direct reduction of halocarbons were analyzed as the likely operable reactions. The team considered the entire range of feasible options, ranked and prioritized the alternatives, and recommended the optimal solution that maximizes protection of worker and public safety while minimizing impacts on RWMC and TRU program operations

  2. Partitioning of TRU elements from Chinese HLLW

    International Nuclear Information System (INIS)

    Song Chongli; Zhu Yongjun

    1994-04-01

    The partitioning of TRU elements from the Chinese HLLW is feasible. The required D.F. values for producing a waste suitable for land disposal are given. The TRPO process developed in China could be used for this purpose. The research and development of the TRPO process is summarized and the general flowsheet is given. The Chinese HLLW has very high salt concentration. It causes the formation of third phase when contacted with TRPO extractant. The third phase would disappear by diluting the Chinese HLLW to 2∼3 times before extraction. The preliminary experiment shows very attractive results. The separation of Sr and Cs from the Chinese HLLW is also possible. The process is being studied. The partitioning of TRU elements and long lived ratio-nuclides from the Chinese HLLW provides an alternative method for its disposal. The partitioning of the Chinese HLLW could greatly reduce the waste volume, that is needed to be vitrified and to be disposed in to the deep repository, and then would drastically save the overall waste disposal cost

  3. Guidelines for developing certification programs for newly generated TRU waste

    International Nuclear Information System (INIS)

    Whitty, W.J.; Ostenak, C.A.; Pillay, K.K.S.; Geoffrion, R.R.

    1983-05-01

    These guidelines were prepared with direction from the US Department of Energy (DOE) Transuranic (TRU) Waste Management Program in support of the DOE effort to certify that newly generated TRU wastes meet the Waste Isolation Pilot Plant (WIPP) Waste Acceptance Criteria. The guidelines provide instructions for generic Certification Program preparation for TRU-waste generators preparing site-specific Certification Programs in response to WIPP requirements. The guidelines address all major aspects of a Certification Program that are necessary to satisfy the WIPP Waste Acceptance Criteria and their associated Compliance Requirements and Certification Quality Assurance Requirements. The details of the major element of a Certification Program, namely, the Certification Plan, are described. The Certification Plan relies on supporting data and control documentation to provide a traceable, auditable account of certification activities. Examples of specific parts of the Certification Plan illustrate the recommended degree of detail. Also, a brief description of generic waste processes related to certification activities is included

  4. Influence of Crucible Thermal Conductivity on Crystal Growth in an Industrial Directional Solidification Process for Silicon Ingots

    Directory of Open Access Journals (Sweden)

    Zaoyang Li

    2016-01-01

    Full Text Available We carried out transient global simulations of heating, melting, growing, annealing, and cooling stages for an industrial directional solidification (DS process for silicon ingots. The crucible thermal conductivity is varied in a reasonable range to investigate its influence on the global heat transfer and silicon crystal growth. It is found that the crucible plays an important role in heat transfer, and therefore its thermal conductivity can influence the crystal growth significantly in the entire DS process. Increasing the crucible thermal conductivity can shorten the time for melting of silicon feedstock and growing of silicon crystal significantly, and therefore large thermal conductivity is helpful in saving both production time and power energy. However, the high temperature gradient in the silicon ingots and the locally concave melt-crystal interface shape for large crucible thermal conductivity indicate that high thermal stress and dislocation propagation are likely to occur during both growing and annealing stages. Based on the numerical simulations, some discussions on designing and choosing the crucible thermal conductivity are presented.

  5. A microencapsulation process of liquid mercury by sulfur polymer stabilization/solidification technology. Part II: Durability of materials

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Delgado, A.; Guerrero, A.; Lopez, F. A.; Perez, C.; Alguacil, F. J.

    2012-11-01

    Under the European LIFE Program a microencapsulation process was developed for liquid mercury using Sulfur Polymer Stabilization/Solidification (SPSS) technology, obtaining a stable concrete-like sulfur matrix that allows the immobilization of mercury for long-term storage. The process description and characterization of the materials obtained were detailed in Part I. The present document, Part II, reports the results of different tests carried out to determine the durability of Hg-S concrete samples with very high mercury content (up to 30 % w/w). Different UNE and RILEM standard test methods were applied, such as capillary water absorption, low pressure water permeability, alkali/acid resistance, salt mist aging, freeze-thaw resistance and fire performance. The samples exhibited no capillarity and their resistance in both alkaline and acid media was very high. They also showed good resistance to very aggressive environments such as spray salt mist, freeze-thaw and dry-wet. The fire hazard of samples at low heat output was negligible. (Author)

  6. The Advantages of Fixed Facilities in Characterizing TRU Wastes

    International Nuclear Information System (INIS)

    FRENCH, M.S.

    2000-01-01

    In May 1998 the Hanford Site started developing a program for characterization of transuranic (TRU) waste for shipment to the Waste Isolation Pilot Plant (WIPP) in New Mexico. After less than two years, Hanford will have a program certified by the Carlsbad Area Office (CAO). By picking a simple waste stream, taking advantage of lessons learned at the other sites, as well as communicating effectively with the CAO, Hanford was able to achieve certification in record time. This effort was further simplified by having a centralized program centered on the Waste Receiving and Processing (WRAP) Facility that contains most of the equipment required to characterize TRU waste. The use of fixed facilities for the characterization of TRU waste at sites with a long-term clean-up mission can be cost effective for several reasons. These include the ability to control the environment in which sensitive instrumentation is required to operate and ensuring that calibrations and maintenance activities are scheduled and performed as an operating routine. Other factors contributing to cost effectiveness include providing approved procedures and facilities for handling hazardous materials and anticipated contingencies and performing essential evolutions, and regulating and smoothing the work load and environmental conditions to provide maximal efficiency and productivity. Another advantage is the ability to efficiently provide characterization services to other sites in the Department of Energy (DOE) Complex that do not have the same capabilities. The Waste Receiving and Processing (WRAP) Facility is a state-of-the-art facility designed to consolidate the operations necessary to inspect, process and ship waste to facilitate verification of contents for certification to established waste acceptance criteria. The WRAP facility inspects, characterizes, treats, and certifies transuranic (TRU), low-level and mixed waste at the Hanford Site in Washington state. Fluor Hanford operates the $89

  7. Remote Handled TRU Waste Status and Activities and Challenges at the Hanford Site

    International Nuclear Information System (INIS)

    MCKENNEY, D.E.

    2000-01-01

    A significant portion of the Department of Energy's forecast volume of remote-handled (RH) transuranic (TRU) waste will originate from the Hanford Site. The forecasted Hanford RH-TRU waste volume of over 2000 cubic meters may constitute over one-third of the forecast inventory of RH-TRU destined for disposal at the Waste Isolation Pilot Plant (WIPP). To date, the Hanford TRU waste program has focused on the retrieval, treatment and certification of the contact-handled transuranic (CH-TRU) wastes. This near-term focus on CH-TRU is consistent with the National TRU Program plans and capabilities. The first shipment of CH-TRU waste from Hanford to the WIPP is scheduled early in Calendar Year 2000. Shipments of RH-TRU from Hanford to the WIPP are scheduled to begin in Fiscal Year 2006 per the National TRU Waste Management Plan. This schedule has been incorporated into milestones within the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement). These Tri-Party milestones (designated the ''M-91'' series of milestones) relate to development of project management plans, completion of design efforts, construction and contracting schedules, and initiation of process operations. The milestone allows for modification of an existing facility, construction of a new facility, and/or commercial contracting to provide the capabilities for processing and certification of RH-TRU wastes for disposal at the WIPP. The development of a Project Management Plan (PMP) for TRU waste is the first significant step in the development of a program for disposal of Hanford's RH-TRU waste. This PMP will address the path forward for disposition of waste streams that cannot be prepared for disposal in the Hanford Waste Receiving and Processing facility (a contact-handled, small container facility) or other Site facilities. The PMP development effort has been initiated, and the PMP will be provided to the regulators for their approval by June 30, 2000. This plan will detail the

  8. Solidification paths of multicomponent monotectic aluminum alloys

    Energy Technology Data Exchange (ETDEWEB)

    Mirkovic, Djordje; Groebner, Joachim [Clausthal University of Technology, Institute of Metallurgy, Robert-Koch-Street 42, D-38678 Clausthal-Zellerfeld (Germany); Schmid-Fetzer, Rainer [Clausthal University of Technology, Institute of Metallurgy, Robert-Koch-Street 42, D-38678 Clausthal-Zellerfeld (Germany)], E-mail: schmid-fetzer@tu-clausthal.de

    2008-10-15

    Solidification paths of three ternary monotectic alloy systems, Al-Bi-Zn, Al-Sn-Cu and Al-Bi-Cu, are studied using thermodynamic calculations, both for the pertinent phase diagrams and also for specific details concerning the solidification of selected alloy compositions. The coupled composition variation in two different liquids is quantitatively given. Various ternary monotectic four-phase reactions are encountered during solidification, as opposed to the simple binary monotectic, L' {yields} L'' + solid. These intricacies are reflected in the solidification microstructures, as demonstrated for these three aluminum alloy systems, selected in view of their distinctive features. This examination of solidification paths and microstructure formation may be relevant for advanced solidification processing of multicomponent monotectic alloys.

  9. Neutron radiography for visualization of liquid metal processes: bubbly flow for CO2 free production of Hydrogen and solidification processes in EM field

    Science.gov (United States)

    Baake, E.; Fehling, T.; Musaeva, D.; Steinberg, T.

    2017-07-01

    The paper describes the results of two experimental investigations aimed to extend the abilities of a neutron radiography to visualize two-phase processes in the electromagnetically (EM) driven melt flow. In the first experiment the Argon bubbly flow in the molten Gallium - a simulation of the CO2 free production of Hydrogen process - was investigated and visualized. Abilities of EM stirring for control on the bubbles residence time in the melt were tested. The second experiment was directed to visualization of a solidification front formation under the influence of EM field. On the basis of the neutron shadow pictures the form of growing ingot, influenced by turbulent flows, was considered. In the both cases rotating permanent magnets were agitating the melt flow. The experimental results have shown that the neutron radiography can be successfully employed for obtaining the visual information about the described processes.

  10. Radiation management at the occurrence of accident and restoration works. Fire and explosion of asphalt solidification processing facility

    Energy Technology Data Exchange (ETDEWEB)

    Miyabe, Kenjiro; Jin, K; Namiki, A; Mizutani, K; Horiuchi, N; Saruta, J [Power Reactor and Nuclear Fuel Development Corp., Health and Safety Division, Tokai, Ibaraki (Japan); Ninomiya, Kazushige [Power Reactor and Nuclear Fuel Development Corp., Tsuruga, Fukui (Japan). Monju Construction Office

    1998-06-01

    Fire and explosion accident in the cell of Asphalt Solidification Processing Facility(ASP) in PNC took placed at March 11 in 1997. Following to the alarm of many radiation monitoring system in the facility, some of workers inhale radioactive materials in their bodies. Indication values of an exhaust monitor installed in the first auxiliary exhaust stack increased suddenly. A large number of windows, doors, and shutters in the facility were raptured by the explosion. A lot of radioactive materials blew up and were released to the outside of the facility. Reinforcement of radiation surveillance function, nose smearing test for the workers and confirmation of contamination situation were implemented on the fire. Investigation of radiation situation, radiation management on the site, exposure management for the workers, surveillance of exhaustion, and restoration works of the damaged radiation management monitoring system were carried out after the explosion. The detailed data of radiation management measures taken during three months after the accident are described in the paper. (M. Suetake)

  11. Compact x-ray microradiograph for in situ imaging of solidification processes: Bringing in situ x-ray micro-imaging from the synchrotron to the laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Rakete, C.; Baumbach, C.; Goldschmidt, A.; Samberg, D.; Schroer, C. G. [Institut fuer Strukturphysik, Technische Universitaet Dresden, D-01062 Dresden (Germany); Breede, F.; Stenzel, C. [Astrium-Space Transportation, Department: TO 611, Claude-Dornier-Strasse, D-88039 Friedrichshafen (Germany); Zimmermann, G.; Pickmann, C. [ACCESS e.V., Intzestrasse 5, D-52072 Aachen (Germany); Houltz, Y.; Lockowandt, C. [Science Services Division, SSC, Box 4207, SE-17104 Solna (Sweden); Svenonius, O.; Wiklund, P. [Scint-X AB, Torshamnsgatan 35, SE-164 40 Kista (Sweden); Mathiesen, R. H. [Inst. for Fysikk, NTNU, N-7491 Trondheim (Norway)

    2011-10-15

    A laboratory based high resolution x-ray radiograph was developed for the investigation of solidification dynamics in alloys. It is based on a low-power microfocus x-ray tube and is potentially appropriate for x-ray diagnostics in space. The x-ray microscope offers a high spatial resolution down to approximately 5 {mu}m. Dynamic processes can be resolved with a frequency of up to 6 Hz. In reference experiments, the setup was optimized to yield a high contrast for AlCu-alloys. With samples of about 150 {mu}m thickness, high quality image sequences of the solidification process were obtained with high resolution in time and space.

  12. From Solidification Processing to Microstructure to Mechanical Properties: A Multi-scale X-ray Study of an Al-Cu Alloy Sample

    Science.gov (United States)

    Tourret, D.; Mertens, J. C. E.; Lieberman, E.; Imhoff, S. D.; Gibbs, J. W.; Henderson, K.; Fezzaa, K.; Deriy, A. L.; Sun, T.; Lebensohn, R. A.; Patterson, B. M.; Clarke, A. J.

    2017-11-01

    We follow an Al-12 at. pct Cu alloy sample from the liquid state to mechanical failure, using in situ X-ray radiography during directional solidification and tensile testing, as well as three-dimensional computed tomography of the microstructure before and after mechanical testing. The solidification processing stage is simulated with a multi-scale dendritic needle network model, and the micromechanical behavior of the solidified microstructure is simulated using voxelized tomography data and an elasto-viscoplastic fast Fourier transform model. This study demonstrates the feasibility of direct in situ monitoring of a metal alloy microstructure from the liquid processing stage up to its mechanical failure, supported by quantitative simulations of microstructure formation and its mechanical behavior.

  13. Controlled Directional Solidification of Aluminum - 7 wt Percent Silicon Alloys: Comparison Between Samples Processed on Earth and in the Microgravity Environment Aboard the International Space Station

    Science.gov (United States)

    Grugel, Richard N.; Tewari, Surendra N.; Erdman, Robert G.; Poirier, David R.

    2012-01-01

    An overview of the international "MIcrostructure Formation in CASTing of Technical Alloys" (MICAST) program is given. Directional solidification processing of metals and alloys is described, and why experiments conducted in the microgravity environment aboard the International Space Station (ISS) are expected to promote our understanding of this commercially relevant practice. Microstructural differences observed when comparing the aluminum - 7 wt% silicon alloys directionally solidified on Earth to those aboard the ISS are presented and discussed.

  14. Investigation of Solidification in the Laser Engineered Net shaping (LENS) Process

    International Nuclear Information System (INIS)

    Ensz, Mark; Griffith, Michelle; Hofmeister, William; Philliber, Joel A.; Smugeresky, John; Wert, Melissa

    1999-01-01

    The Laser Engineered Net Shaping (LENSm) process is a laser assisted, direct metal manufacturing process under development at Sandia National Laboratories. The process incorporates features from stereo lithography and laser surfacing, using CAD file cross-sections to control the forming process. Powder metal particles (less than 150 micrometers) are delivered in a gas stream into the focus of a NdYAG laser to form a molten pool. The part is then driven on an x/y stage to generate a three-dimensional part by layer wise, additive processing. In an effort to understand the thermal behavior of the LENS process, in-situ high-speed thermal imaging has been coupled with microstructural analysis and finite element modeling. Cooling of the melt is accomplished primarily by conduction of heat through the part and substrate, and depending on the substrate temperature and laser input energy, cooling rates can be varied from 10 ampersand sup2; to 10 ampersand sup3; K s -l . This flexibility allows control of the microstructure and properties in the part. The experiments reported herein were conducted on 316 stainless steel, using two different particle size distributions with two different average particle sizes. Thermal images of the molten pool were analyzed to determine temperature gradients and cooling rates in the vicinity of the molten pool, and this information was correlated to the microstructure and properties of the part. Some preliminary finite element modeling of the LENS process is also presented

  15. Los Alamos Plutonium Facility newly generated TRU waste certification

    International Nuclear Information System (INIS)

    Gruetzmacher, K.; Montoya, A.; Sinkule, B.; Maez, M.

    1997-01-01

    This paper presents an overview of the activities being planned and implemented to certify newly generated contact handled transuranic (TRU) waste produced by Los Alamos National Laboratory's (LANL's) Plutonium Facility. Certifying waste at the point of generation is the most important cost and labor saving step in the WIPP certification process. The pedigree of a waste item is best known by the originator of the waste and frees a site from expensive characterization activities such as those associated with legacy waste. Through a cooperative agreement with LANLs Waste Management Facility and under the umbrella of LANLs WIPP-related certification and quality assurance documents, the Plutonium Facility will be certifying its own newly generated waste. Some of the challenges faced by the Plutonium Facility in preparing to certify TRU waste include the modification and addition of procedures to meet WIPP requirements, standardizing packaging for TRU waste, collecting processing documentation from operations which produce TRU waste, and developing ways to modify waste streams which are not certifiable in their present form

  16. Researches focused on structure of aluminium alloys processed by rapid solidification, used in automotive industry

    International Nuclear Information System (INIS)

    Sfat, C.; Vasile, T.; Vasilescu, M.

    2001-01-01

    The paper present some new results focused on an aluminium high temperature alloy, obtained by 'melt spinning method'. alloy composition, processing conditions, resulted structures and the influence between them are presented. There are studied the two zone structures of the alloy and the relation between processing conditions and the characteristics of the zones, with implications on mechanical behavior in real conditions. The final conclusion show that is possible to control the structure in order to improve material behavior. (author)

  17. Numerical investigation of thermal and residual stress of sapphire during c-axis vertical Bridgman growth process considering the solidification history effect

    Science.gov (United States)

    Hwang, Ji Hoon; Lee, Young Cheol; Lee, Wook Jin

    2018-01-01

    Sapphire single crystals have been highlighted for epitaxial of gallium nitride films in high-power laser and light emitting diode industries. In this study, the evolution of thermally induced stress in sapphire during the vertical Bridgman crystal growth process was investigated using a finite element model that simplified the real Bridgman process. A vertical Bridgman process of cylindrical sapphire crystal with a diameter of 50 mm was considered for the model. The solidification history effect during the growth was modeled by the quite element technique. The effects of temperature gradient, seeding interface shape and seeding position on the thermal stress during the process were discussed based on the finite element analysis results.

  18. TRU waste form and package criteria meeting

    Energy Technology Data Exchange (ETDEWEB)

    None

    1977-08-01

    The broad subject of the meeting is the overall ERDA TRU waste management program, although the discussions also cover performance criteria for the Waste Isolation Pilot Plant and their implications for the overall TRU program. Separate abstracts were prepared for all ten presentations. (DLC)

  19. Copper-base alloys processed by rapid solidification and ion implantation

    International Nuclear Information System (INIS)

    Wood, J.V.; Elvidge, C.J.; Johnson, E.; Johansen, A.; Sarholt-Kristensen, L.; Henriksen, O.

    1985-01-01

    Alloys of Cu-Sn and Cu-B have been processed by both melt spinning and ion implantation. In some instances (e.g. Cu-Sn alloys) rapidly solidified ribbons have been subjected to further implantation. This paper describes the similarities and differences in structure of materials subjected to a dynamic and contained process. For example in Cu-B alloys (up to 2wt% Boron) extended solubility is found in implanted alloys which is not present to the same degree in rapidly solidified alloys of the same composition. Likewise the range and nature of the reversible martensitic transformation is different in both cases as examined by electron microscopy and differential scanning calorimetry. (orig.)

  20. Decontamination of TRU glove boxes

    International Nuclear Information System (INIS)

    Crawford, J.H.

    1978-03-01

    Two glove boxes that had been used for work with transuranic nuclides (TRU) for about 12 years were decontaminated in a test program to collect data for developing a decontamination facility for large equipment highly contaminated with alpha emitters. A simple chemical technique consisting of a cycle of water flushes and alkaline permanganate and oxalic acid washes was used for both boxes. The test showed that glove boxes and similar equipment that are grossly contaminated with transuranic nuclides can be decontaminated to the current DIE nonretrievable disposal guide of <10 nCi TRU/g with a moderate amount of decontamination solution and manpower. Decontamination of the first box from an estimated 1.3 Ci to about 5 mCi (6 nCi/g) required 1.3 gallons of decontamination solution and 0.03 man-hour of work for each square foot of surface area. The second box was decontaminated from an estimated 3.4 Ci to about 2.8 mCi (4.2 nCi/g) using 0.9 gallon of decontamination solution and 0.02 man-hour for each square foot of surface area. Further reductions in contamination were achieved by repetitive decontamination cycles, but the effectiveness of the technique decreased sharply after the initial cycle

  1. Hanford Site Transuranic (TRU) Waste Certification Plan

    International Nuclear Information System (INIS)

    GREAGER, T.M.

    2000-01-01

    As a generator of transuranic (TRU) and TRU mixed waste destined for disposal at the Waste Isolation Pilot Plant (WIPP), the Hanford Site must ensure that its TRU waste meets the requirements of US. Department of Energy (DOE) 0 435.1, ''Radioactive Waste Management,'' and the Contact-Handled (CH) Transuranic Waste Acceptance Criteria for the Waste Isolation Pilot Plant (WIPP-WAC). WIPP-WAC requirements are derived from the WIPP Technical Safety Requirements, WIPP Safety Analysis Report, TRUPACT-II SARP, WIPP Land Withdrawal Act, WIPP Hazardous Waste Facility Permit, and Title 40 Code of Federal Regulations (CFR) 191/194 Compliance Certification Decision. The WIPP-WAC establishes the specific physical, chemical, radiological, and packaging criteria for acceptance of defense TRU waste shipments at WIPP. The WPP-WAC also requires that participating DOE TRU waste generator/treatment/storage sites produce site-specific documents, including a certification plan, that describe their program for managing TRU waste and TRU waste shipments before transferring waste to WIPP. Waste characterization activities provide much of the data upon which certification decisions are based. Waste characterization requirements for TRU waste and TRU mixed waste that contains constituents regulated under the Resource Conservation and Recovery Act (RCRA) are established in the WIPP Hazardous Waste Facility Permit Waste Analysis Plan (WAP). The Hanford Site Quality Assurance Project Plan (QAPjP) (HNF-2599) implements the applicable requirements in the WAP and includes the qualitative and quantitative criteria for making hazardous waste determinations. The Hanford Site must also ensure that its TRU waste destined for disposal at WPP meets requirements for transport in the Transuranic Package Transporter-11 (TRUPACT-11). The US. Nuclear Regulatory Commission (NRC) establishes the TRUPACT-11 requirements in the Safety Analysis Report for the TRUPACT-II Shipping Package (TRUPACT-11 SARP). In

  2. Asymptotic description of two metastable processes of solidification for the case of large relaxation time

    International Nuclear Information System (INIS)

    Omel'yanov, G.A.

    1995-07-01

    The non-isothermal Cahn-Hilliard equations in the n-dimensional case (n = 2,3) are considered. The interaction length is proportional to a small parameter, and the relaxation time is proportional to a constant. The asymptotic solutions describing two metastable processes are constructed and justified. The soliton type solution describes the first stage of separation in alloy, when a set of ''superheated liquid'' appears inside the ''solid'' part. The Van der Waals type solution describes the free interface dynamics for large time. The smoothness of temperature is established for large time and the Mullins-Sekerka problem describing the free interface is derived. (author). 46 refs

  3. Transition from a planar interface to cellular and dendritic structures during rapid solidification processing

    Science.gov (United States)

    Laxmanan, V.

    1986-01-01

    The development of theoretical models which characterize the planar-cellular and cell-dendrite transitions is described. The transitions are analyzed in terms of the Chalmers number, the solute Peclet number, and the tip stability parameter, which correlate microstructural features and processing conditions. The planar-cellular transition is examined using the constitutional supercooling theory of Chalmers et al., (1953) and it is observed that the Chalmers number is between 0 and 1 during dendritic and cellular growth. Analysis of cell-dendrite transition data reveal that the transition occurs when the solute Peclet number goes through a minimum, the primary arm spacings go through a maximum, and the Chalmers number is equal to 1/2. The relation between the tip stability parameter and the solute Peclet number is investigated and it is noted that the tip stability parameter is useful for studying dendritic growth in alloys.

  4. High-silica glass matrix process for high-level waste solidification

    International Nuclear Information System (INIS)

    Simmons, J.H.; Macedo, P.B.

    1981-01-01

    In the search for an optimum glass matrix composition, we have determined that chemical durability and thermal stability are maximized, and that stress development is minimized for glass compositions containing large concentrations of glass-forming oxides, of which silica is the major component (80 mol%). These properties and characteristics were recently demonstrated to belong to very old geological glasses known as tektites (ages of 750,000 to 34 million years.) The barrier to simulating tektite compositions for the waste glasses was the high melting temperature (1600 to 1800 0 C) needed for these glasses. Such temperatures greatly complicate furnace design and maintenance and lead to an intolerable vaporization of many of the radioisotopes into the off-gas system. Research conducted at our laboratory led to the development of a porous high-silica waste glass material with approximately 80% SiO 2 by mole and 30% waste loading by weight. The process can handle a wide variety of compositions, and yields long, elliptical, monolithic samples, which consist of a loaded high-silica core completely enveloped in a high-silica glass tube, which has collapsed upon the core and sealed it from the outside. The outer glass layer is totally free of waste isotopes and provides an integral multibarrier protection system

  5. A model on valence state evaluation of TRU nuclides in reprocessing solutions

    International Nuclear Information System (INIS)

    Uchiyama, Gunzo; Fujine, Sachio; Yoshida, Zenko; Maeda, Mitsuru; Motoyama, Satoshi.

    1998-02-01

    A mathematical model was developed to evaluate the valence state of TRU nuclides in reprocessing process solutions. The model consists of mass balance equations, Nernst equations, reaction rate equations and electrically neutrality equations. The model is applicable for the valence state evaluation of TRU nuclides in both steady state and transient state conditions in redox equilibrium. The valence state which is difficult to measure under high radiation and multi component conditions is calculated by the model using experimentally measured data for the TRU nuclide concentrations, nitric acid and redox reagent concentrations, electrode potential and solution temperature. (author)

  6. The Through Process Simulation of Mold filling, Solidification, and Heat Treatment of the Al Alloy Bending Beam Low-pressure Casting

    International Nuclear Information System (INIS)

    Yin, Yajun; Guo, Zhao; Wang, Huan; Liao, Dunming; Chen, Tao; Zhou, Jianxin

    2015-01-01

    The research on the simulation for the through process of low-pressure casting and heat treatment is conducive to combine information technology and advanced casting technology, which will help to predict the defects and mechanical properties of the castings in the through process. In this paper, we focus on the simulation for through process of low-pressure casting and heat treatment of ZL114A Bending beam. Firstly, we analyzethe distribution of the shrinkage and porosities in filling and solidification process, and simulate the distribution of stress and strain in the late solidification of casting. Then, the numerical simulation of heat treatment process for ZL114A Bending beam is realized according to the heat treatment parameters and the corresponding simulation results of temperature field, stress, strain, and aging performance are given. Finally, we verify that simulation platform for the through process of low-pressure casting and heat treatment can serve the production practice perfectly and provide technical guidance and process optimization for the through process of low-pressure casting and heat treatment. (paper)

  7. Detachment of Tertiary Dendrite Arms during Controlled Directional Solidification in Aluminum - 7 wt Percent Silicon Alloys: Observations from Ground-based and Microgravity Processed Samples

    Science.gov (United States)

    Grugel, Richard N.; Erdman, Robert; Van Hoose, James R.; Tewari, Surendra; Poirier, David

    2012-01-01

    Electron Back Scattered Diffraction results from cross-sections of directionally solidified aluminum 7wt% silicon alloys unexpectedly revealed tertiary dendrite arms that were detached and mis-oriented from their parent arm. More surprisingly, the same phenomenon was observed in a sample similarly processed in the quiescent microgravity environment aboard the International Space Station (ISS) in support of the joint US-European MICAST investigation. The work presented here includes a brief introduction to MICAST and the directional solidification facilities, and their capabilities, available aboard the ISS. Results from the ground-based and microgravity processed samples are compared and possible mechanisms for the observed tertiary arm detachment are suggested.

  8. A microencapsulation process of liquid mercury by sulfur polymer stabilization/solidification technology. Part II: Durability of materials

    Directory of Open Access Journals (Sweden)

    López-Delgado, A.

    2012-02-01

    Full Text Available Under the European LIFE Program a microencapsulation process was developed for liquid mercury using Sulfur Polymer Stabilization/Solidification (SPSS technology, obtaining a stable concrete-like sulfur matrix that allows the immobilization of mercury for long-term storage. The process description and characterization of the materials obtained were detailed in Part I. The present document, Part II, reports the results of different tests carried out to determine the durability of Hg-S concrete samples with very high mercury content (up to 30 % w/w. Different UNE and RILEM standard test methods were applied, such as capillary water absorption, low pressure water permeability, alkali/acid resistance, salt mist aging, freeze-thaw resistance and fire performance. The samples exhibited no capillarity and their resistance in both alkaline and acid media was very high. They also showed good resistance to very aggressive environments such as spray salt mist, freeze-thaw and dry-wet. The fire hazard of samples at low heat output was negligible.

    Dentro del Programa Europeo LIFE, se ha desarrollado un proceso de microencapsulación de mercurio liquido, utilizando la tecnología de estabilización/solidificación con azufre polimérico (SPSS. Como resultado se ha obtenido un material estable tipo concreto que permite la inmovilización de mercurio y su almacenamiento a largo plazo. La descripción del proceso y la caracterización de los materiales obtenidos, denominados concretos Hg-S, se detallan en la Parte I. El presente trabajo, Parte II, incluye los resultados de los diferentes ensayos realizados para determinar la durabilidad de las muestras de concreto Hg-S con un contenido de mercurio de hasta el 30 %. Se han utilizado diferentes métodos de ensayo estándar, UNE y RILEM, para determinar propiedades como la absorción de agua por capilaridad, la permeabilidad de agua a baja presión, la resistencia a álcali y ácido, el comportamiento en

  9. UNCONSTRAINED MELTING AND SOLIDIFICATION INSIDE ...

    African Journals Online (AJOL)

    2015-09-01

    Sep 1, 2015 ... There is a large number of experimental and numerical works on melting and solidification of PCM[6-10], and also its usage as thermal management in building [11-14], electronic devices [15-16] and solar energy. [17-20].Most investigated geometries in melting and freezing process are sphere (spherical.

  10. The Production of Material with Ultrafine Grain Structure in Al-Zn Alloy in the Process of Rapid Solidification

    Directory of Open Access Journals (Sweden)

    Szymaneka M.

    2014-06-01

    Full Text Available In the aluminium alloy family, Al-Zn materials with non-standard chemical composition containing Mg and Cu are a new group of alloys, mainly owing to their high strength properties. Proper choice of alloying elements, and of the method of molten metal treatment and casting enable further shaping of the properties. One of the modern methods to produce materials with submicron structure is a method of Rapid Solidification. The ribbon cast in a melt spinning device is an intermediate product for further plastic working. Using the technique of Rapid Solidification it is not possible to directly produce a solid structural material of the required shape and length. Therefore, the ribbon of an ultrafine grain or nanometric structure must be subjected to the operations of fragmentation, compaction, consolidation and hot extrusion.

  11. Three-dimensional cellular automaton-finite element modeling of solidification grain structures for arc-welding processes

    International Nuclear Information System (INIS)

    Chen, Shijia; Guillemot, Gildas; Gandin, Charles-André

    2016-01-01

    Solidification grain structure has significant impact on the final properties of welded parts using fusion welding processes. Direct simulation of grain structure at industrial scale is yet rarely reported in the literature and remains a challenge. A three-dimensional (3D) coupled Cellular Automaton (CA) – Finite Element (FE) model is presented that predicts the grain structure formation during multiple passes Gas Tungsten Arc Welding (GTAW) and Gas Metal Arc Welding (GMAW). The FE model is established in a level set (LS) approach that tracks the evolution of the metal-shielding gas interface due to the addition of metal. The FE method solves the mass, energy and momentum conservation equations for the metal plus shielding gas system based on an adaptive mesh (FE mesh). Fields are projected in a second FE mesh, named CA mesh. A CA grid made of a regular lattice of cubic cells is created to overlay the fixed CA mesh. The CA model based on the CA grid simulates the melting and growth of the grain boundaries in the liquid pool. In order to handle large computational domains while keeping reasonable computational costs, parallel computations and dynamic strategies for the allocation/deallocation of the CA grid are introduced. These strategies correspond to significant optimizations of the computer memories that are demonstrated. The 3D CAFE model is first applied to the simple configuration of single linear passes by GTAW of a duplex stainless steel URANUS 2202. It is then applied to a more persuasive example considering GMAW in spray transfer mode during multiple passes to fill a V-groove chamfer. Simulations reveal the possibility to handle domains with millions of grains in representative domain sizes while following the formation of textures that result from the growth competition among columnar grains. -- Graphical abstract: Simulated 3D grain structure (3D CAFE model) for GTAW multiple linear passes at the surface of a duplex stainless steel (URANUS 22002

  12. Hanford Site Transuranic (TRU) Waste Certification Plan

    International Nuclear Information System (INIS)

    GREAGER, T.M.

    1999-01-01

    The Hanford Site Transuranic Waste Certification Plan establishes the programmatic framework and criteria with in which the Hanford Site ensures that contract-handled TRU wastes can be certified as compliant with the WIPP WAC and TRUPACT-II SARP

  13. Hanford Site Transuranic (TRU) Waste Certification Plan

    Energy Technology Data Exchange (ETDEWEB)

    GREAGER, T.M.

    1999-09-09

    The Hanford Site Transuranic Waste Certification Plan establishes the programmatic framework and criteria within which the Hanford Site ensures that contract-handled TRU wastes can be certified as compliant with the WIPP WAC and TRUPACT-II SARP.

  14. Low level waste solidification practice in Japan

    International Nuclear Information System (INIS)

    Sakata, S.; Kuribayashi, H.; Kono, Y.

    1981-01-01

    Both sea dumping and land isolation are planned to be accomplished for low level waste disposal in Japan. The conceptual design of land isolation facilities has been completed, and site selection will presently get underway. With respect to ocean dumping, safety surveys are being performed along the lines of the London Dumping Convention and the Revised Definitions and Recommendations of the IAEA, and the review of Japanese regulations and applicable criteria is being expedited. This paper discusses the present approach to waste solidification practices in Japan. It reports that the bitumen solidification process and the plastic solidification process are being increasingly used in Japan. Despite higher investment costs, both processes have advantages in operating cost, and are comparable to the cement solidification process in overall costs

  15. Transuranic (TRU) Waste Phase I Retrieval Plan

    CERN Document Server

    McDonald, K M

    2000-01-01

    From 1970 to 1987, TRU and suspect TRU wastes at Hanford were placed in the SWBG. At the time of placement in the SWBG these wastes were not regulated under existing Resource Conservation and Recovery Act (RCRA) regulations, since they were generated and disposed of prior to the effective date of RCRA at the Hanford Site (1987). From the standpoint of DOE Order 5820.2A1, the TRU wastes are considered retrievably stored, and current plans are to retrieve these wastes for shipment to WIPP for disposal. This plan provides a strategy for the Phase I retrieval that meets the intent of TPA milestone M-91 and Project W-113, and incorporates the lessons learned during TRU retrieval campaigns at Hanford, LANL, and SRS. As in the original Project W-113 plans, the current plan calls for examination of approximately 10,000 suspect-TRU drums located in the 218-W-4C burial ground followed by the retrieval of those drums verified to contain TRU waste. Unlike the older plan, however, this plan proposes an open-air retrieval ...

  16. Small hazardous waste generators in developing countries: use of stabilization/solidification process as an economic tool for metal wastewater treatment and appropriate sludge disposal.

    Science.gov (United States)

    Silva, Marcos A R; Mater, Luciana; Souza-Sierra, Maria M; Corrêa, Albertina X R; Sperb, Rafael; Radetski, Claudemir M

    2007-08-25

    The aim of this study was to propose a profitable destination for an industrial sludge that can cover the wastewater treatment costs of small waste generators. Optimized stabilization/solidification technology was used to treat hazardous waste from an electroplating industry that is currently released untreated to the environment. The stabilized/solidified (S/S) waste product was used as a raw material to build concrete blocks, to be sold as pavement blocks or used in roadbeds and/or parking lots. The quality of the blocks containing a mixture of cement, lime, clay and waste was evaluated by means of leaching and solubility tests according to the current Brazilian waste regulations. Results showed very low metal leachability and solubility of the block constituents, indicating a low environmental impact. Concerning economic benefits from the S/S process and reuse of the resultant product, the cost of untreated heavy metal-containing sludge disposal to landfill is usually on the order of US$ 150-200 per tonne of waste, while 1tonne of concrete roadbed blocks (with 25% of S/S waste constitution) has a value of around US$ 100. The results of this work showed that the cement, clay and lime-based process of stabilization/solidification of hazardous waste sludge is sufficiently effective and economically viable to stimulate the treatment of wastewater from small industrial waste generators.

  17. TRU waste-sampling program

    International Nuclear Information System (INIS)

    Warren, J.L.; Zerwekh, A.

    1985-08-01

    As part of a TRU waste-sampling program, Los Alamos National Laboratory retrieved and examined 44 drums of 238 Pu- and 239 Pu-contaminated waste. The drums ranged in age from 8 months to 9 years. The majority of drums were tested for pressure, and gas samples withdrawn from the drums were analyzed by a mass spectrometer. Real-time radiography and visual examination were used to determine both void volumes and waste content. Drum walls were measured for deterioration, and selected drum contents were reassayed for comparison with original assays and WIPP criteria. Each drum tested at atmospheric pressure. Mass spectrometry revealed no problem with 239 Pu-contaminated waste, but three 8-month-old drums of 238 Pu-contaminated waste contained a potentially hazardous gas mixture. Void volumes fell within the 81 to 97% range. Measurements of drum walls showed no significant corrosion or deterioration. All reassayed contents were within WIPP waste acceptance criteria. Five of the drums opened and examined (15%) could not be certified as packaged. Three contained free liquids, one had corrosive materials, and one had too much unstabilized particulate. Eleven drums had the wrong (or not the most appropriate) waste code. In many cases, disposal volumes had been inefficiently used. 2 refs., 23 figs., 7 tabs

  18. TRU waste from the Superblock

    International Nuclear Information System (INIS)

    Coburn, T.T.

    1997-01-01

    This data analysis is to show that weapons grade plutonium is of uniform composition to the standards set by the Waste-Isolation Pilot Plant (WIPP) Transuranic Waste Characterization Quality Assurance Program Plan (TRUW Characterization QAPP, Rev. 2, DOE, Carlsbad Area Office, November 15, 1996). The major portion of Superblock transuranic (TRU) waste is glove-box trash contaminated with weapons grade plutonium. This waste originates in the Building 332 (B332) radioactive-materials area (RMA). Because each plutonium batch brought into the B332 RMA is well characterized with regard to nature and quantity of transuranic nuclides present, waste also will be well characterized without further analytical work, provided the batches are quite similar. A sample data set was created by examining the 41 incoming samples analyzed by Ken Raschke (using a γ-ray spectrometer) for isotopic distribution and by Ted Midtaune (using a calorimeter) for mass of radionuclides. The 41 samples were from separate batches analyzed May 1993 through January 1997. All available weapons grade plutonium data in Midtaune's files were used. Alloys having greater than 50% transuranic material were included. The intention of this study is to use this sample data set to judge ''similarity.''

  19. Solidification observations and sliding wear behavior of vacuum arc melting processed Ni–Al–TiC composites

    International Nuclear Information System (INIS)

    Karantzalis, A.E.; Lekatou, A.; Tsirka, K.

    2012-01-01

    Monolithic Ni 3 Al and Ni–25 at.%Al intermetallic matrix TiC-reinforced composites were successfully produced by vacuum arc melting. TiC crystals were formed through a dissolution–reprecipitation mechanism and their final morphology is explained by means of a) Jackson's classical nucleation and growth phenomena and b) solidification rate considerations. The TiC presence altered the matrix microconstituents most likely due to specific melt–particle interactions and crystal plane epitaxial matching. TiC particles caused a significant decrease on the specific wear rate of the monolithic Ni 3 Al alloy and the possible wear mechanisms are approached by means of a) surface oxidation, b) crack/flaws formation, c) material detachment and d) debris–counter surfaces interactions. - Highlights: ► Vacuum arc melting (VAM) of Ni-Al based intermetallic matrix composite materials. ► Solidification phenomena examination. ► TiC crystal formation and growth mechanisms. ► Sliding wear examination.

  20. Fundamental Metallurgy of Solidification

    DEFF Research Database (Denmark)

    Tiedje, Niels

    2004-01-01

    The text takes the reader through some fundamental aspects of solidification, with focus on understanding the basic physics that govern solidification in casting and welding. It is described how the first solid is formed and which factors affect nucleation. It is described how crystals grow from...

  1. Evaluation of a TRU fundamental criterion and reference TRU waste units

    International Nuclear Information System (INIS)

    Klett, R.

    1993-01-01

    The comparison of two options for regulating transuranic (TRU) waste disposal is explained in this paper. The two options are (1) fundamental and derived standards developed specifically for the TRU waste and (2) a family of procedures that use a reference to the TRU waste unit with procedures that use a reference to the TRU waste unit with commercial high-level waste (HLW) criteria. Background information pertaining to both options is covered. A section on criteria specifically for TRUE waste suggests a methodology for developing or adapting fundamental and derived criteria that are consistent with all other aspects of the standards. The section on references TRU waste units covers all the parameter variations that have been suggested for this option. The technical bases of each approach is reviewed, implementation is discussed and their relative attributes and deficiencies are evaluated

  2. RH-TRU Waste Content Codes

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions

    2007-07-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: • A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. • A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is “3.” The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR

  3. Transuranic (TRU) Waste Phase I Retrieval Plan

    International Nuclear Information System (INIS)

    MCDONALD, K.M.

    1999-01-01

    From 1970 to 1987, TRU and suspect TRU wastes at Hanford were placed in the SWBG. At the time of placement in the SWBG these wastes were not regulated under existing Resource Conservation and Recovery Act (RCRA) regulations, since they were generated and disposed of prior to the effective date of RCRA at the Hanford Site (1987). From the standpoint of DOE Order 5820.2A', the TRU wastes are considered retrievably stored, and current plans are to retrieve these wastes for shipment to WIPP for disposal. This plan provides a strategy for the Phase I retrieval that meets the intent of TPA milestone M-91 and Project W-113, and incorporates the lessons learned during TRU retrieval campaigns at Hanford, LANL, and SRS. As in the original Project W-I13 plans, the current plan calls for examination of approximately 10,000 suspect-TRU drums located in the 218-W-4C burial ground followed by the retrieval of those drums verified to contain TRU waste. Unlike the older plan, however, this plan proposes an open-air retrieval scenario similar to those used for TRU drum retrieval at LANL and SRS. Phase I retrieval consists of the activities associated with the assessment of approximately 10,000 55-gallon drums of suspect TRU-waste in burial ground 218-W-4C and the retrieval of those drums verified to contain TRU waste. Four of the trenches in 218-W-4C (Trenches 1,4,20, and 29) are prime candidates for Phase I retrieval because they contain large numbers of suspect TRU drums, stacked from 2 to 5 drums high, on an asphalt pad. In fact, three of the trenches (Trenches 1,20, and 29) contain waste that has not been covered with soil, and about 1500 drums can be retrieved without excavation. The other three trenches in 218-W-4C (Trenches 7, 19, and 24) are not candidates for Phase I retrieval because they contain significant numbers of boxes. Drums will be retrieved from the four candidate trenches, checked for structural integrity, overpacked, if necessary, and assayed at the burial

  4. Transuranic (TRU) Waste Phase I Retrieval Plan

    International Nuclear Information System (INIS)

    MCDONALD, K.M.

    2000-01-01

    From 1970 to 1987, TRU and suspect TRU wastes at Hanford were placed in the SWBG. At the time of placement in the SWBG these wastes were not regulated under existing Resource Conservation and Recovery Act (RCRA) regulations, since they were generated and disposed of prior to the effective date of RCRA at the Hanford Site (1987). From the standpoint of DOE Order 5820.2A1, the TRU wastes are considered retrievably stored, and current plans are to retrieve these wastes for shipment to WIPP for disposal. This plan provides a strategy for the Phase I retrieval that meets the intent of TPA milestone M-91 and Project W-113, and incorporates the lessons learned during TRU retrieval campaigns at Hanford, LANL, and SRS. As in the original Project W-113 plans, the current plan calls for examination of approximately 10,000 suspect-TRU drums located in the 218-W-4C burial ground followed by the retrieval of those drums verified to contain TRU waste. Unlike the older plan, however, this plan proposes an open-air retrieval scenario similar to those used for TRU drum retrieval at LANL and SRS. Phase I retrieval consists of the activities associated with the assessment of approximately 10,000 55-gallon drums of suspect TRU-waste in burial ground 218-W-4C and the retrieval of those drums verified to contain TRU waste. Four of the trenches in 218-W-4C (Trenches 1, 4, 20, and 29) are prime candidates for Phase I retrieval because they contain large numbers of suspect TRU drums, stacked from 2 to 5 drums high, on an asphalt pad. In fact, three of the trenches (Trenches 1 , 20, and 29) contain waste that has not been covered with soil, and about 1500 drums can be retrieved without excavation. The other three trenches in 218-W-4C (Trenches 7, 19, and 24) are not candidates for Phase I retrieval because they contain significant numbers of boxes. Drums will be retrieved from the four candidate trenches, checked for structural integrity, overpacked, if necessary, and assayed at the burial

  5. Effect of directional solidification rate on the microstructure and properties of deformation-processed Cu–7Cr–0.1Ag in situ composites

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Keming [Jiangxi Key Laboratory for Advanced Copper and Tungsten Materials, Jiangxi Academy of Sciences, Nanchang 330029 (China); School of Mechanical, Materials and Mechatronic Engineering, University of Wollongong, NSW 2522 (Australia); Jiang, Zhengyi; Zhao, Jingwei [School of Mechanical, Materials and Mechatronic Engineering, University of Wollongong, NSW 2522 (Australia); Zou, Jin; Chen, Zhibao [Jiangxi Key Laboratory for Advanced Copper and Tungsten Materials, Jiangxi Academy of Sciences, Nanchang 330029 (China); Lu, Deping, E-mail: llludp@163.com [Jiangxi Key Laboratory for Advanced Copper and Tungsten Materials, Jiangxi Academy of Sciences, Nanchang 330029 (China)

    2014-11-05

    Highlights: • Effect of directional solidification (DS) rate on a Cu–Cr–Ag in situ composite. • The microstructure and properties of the DS in situ composite were investigated. • The second-phase Cr grains were parallel to drawing direction, and were finer. • The tensile strength was higher and the combination of properties was better. - Abstract: The influence of directional solidification rate on the microstructure, mechanical properties and conductivity of deformation-processed Cu–7Cr–0.1Ag in situ composites produced by thermo-mechanical processing was systematically investigated. The microstructure was analyzed by optical microscopy and scanning electronic microscopy. The mechanical properties and conductivity were evaluated by tensile-testing machine and micro-ohmmeter, respectively. The results indicate that the size, shape and distribution of second-phase Cr grains are significantly different in the Cu–7Cr–0.1Ag alloys with different growth rates. At a growth rate of 200 μm s{sup −1}, the Cr grains transform into fine Cr fiber-like grains parallel to the pulling direction from the Cr dendrites. The tensile strength of the Cu–7Cr–0.1Ag in situ composites from the directional solidification (DS) alloys is significantly higher than that from the as-cast alloy, while the conductivity of the in situ composites from the DS alloys is slightly lower than that from the as-cast alloy. The following combinations of tensile strength, elongation to fracture and conductivity of the Cu–7Cr–0.1Ag in situ composites from the DS alloy with a growth rate of 200 μm s{sup −1} and a cumulative cold deformation strain of 8 after isochronic aging treatment for 1 h can be obtained respectively as: (i) 1067 MPa, 2.9% and 74.9% IACS; or (ii) 1018 MPa, 3.0%, and 76.0% IACS or (iii) 906 MPa, 3.3% and 77.6% IACS.

  6. Solid waste and chemical sludges: Stabilization/solidification processes and qualification of related products. Rifiuti solidi e fanghi: Processi di stabilizzazione/solidificazione e qualificazione dei prodotti ottenuti

    Energy Technology Data Exchange (ETDEWEB)

    Balzamo, S.; De Angelis, G. (ENEA, Casaccia (Italy). Area Energia Ambiente e Salute)

    A wide programme on cementation of radioactive and/or toxic wastes is being conducted at ENEA (Italian Agency for Energy, New Technologies and the Environment) laboratories. The main goal of the research work is to achieve solidified products which are reliable for transport and final disposal, as well as, to study possible reuse for civil purposes. Several characterization tests are made aiming at the optimization of process parameters and the verification of the quality of the final waste forms. Particular attention is being devoted to the problems related to the waste-matrix interaction, because no waste can be considered 'inert' from this point of view. It is therefore necessary to investigate the nature and the amount of such interactions through an accurate study of the chemical behaviour of the main waste components. That should allow researchers to get useful information to prevent the embedded wastes from causing deleterious effects to the solidification matrix.

  7. Solidification and casting

    CERN Document Server

    Cantor, Brian

    2002-01-01

    INDUSTRIAL PERSPECTIVEDirect chillcasting of aluminium alloysContinuous casting of aluminium alloysContinuous casting of steelsCastings in the automotive industryCast aluminium-silicon piston alloysMODELLING AND SIMULATIONModelling direct chill castingMold filling simulation of die castingThe ten casting rulesGrain selection in single crystal superalloy castingsDefects in aluminium shape castingPattern formation during solidificationPeritectic solidificationSTRUCTURE AND DEFECTSHetergeneous nucleation in aluminium alloysCo

  8. Method of plastic solidification of radioactive wastes

    International Nuclear Information System (INIS)

    Oikawa, Yasuo; Tokimitsu, Fujio.

    1986-01-01

    Purpose: To prevent occurrence of deleterious cracks to the inside and the surface of solidification products, as well as eliminate gaps between the products and the vessel inner wall upon plastic solidification processing for powdery or granular radioactive wastes. Method: An appropriate amount of thermoplastic resins such as styrenic polymer or vinyl acetate type polymer as a low shrinking agent is added and mixed with unsaturated polyester resins to be mixed with radioactive wastes so as to reduce the shrinkage-ratio to 0 % upon curing reaction. Thus, a great shrinkage upon hardening the mixture is suppressed to prevent the occurrence of cracks to the surface and the inside of the solidification products, as well as prevent the gaps between the inner walls of a drum can vessel and the products upon forming solidification products to the inside of the drum can. The resultant solidification products have a large compression strength and can sufficiently satisfy the evaluation standards as the plastic solidification products of radioactive wastes. (Horiuchi, T.)

  9. Optimisation of the microstructure of YBa2Cu3O7-δ superconducting ceramics textured by horizontal solidification. Effects of a magnetic field applied vertically during the solidification process

    International Nuclear Information System (INIS)

    Durand, L.

    1995-01-01

    We have shown that il was possible to disperse in a very homogeneous way, into the 123 matrix, 211 particles regular in size around 1μm, by directional solidification of a mixture of 123 plus 211 sol-gel powders. The addition of 0.5 wt% of Pt lo this precursor mixture has permitted us to refine further the size of the 211 particles. Moreover, we have established that the finer the 123 powder and the faster the heating rate, the smaller the 211 particles. From another hand, we have found that the size and the microstructure of the solidified single domain was dependant upon numerous parameters. In particular, we have noted that if the cold pressed green body was not sintered before being melted, grains of solid CuO appeared into the liquid which next hardly recombined with it, giving rise to a comb like solidification front, witness of a local un-stability. Moreover, in these particular conditions, the 123 formed was found to be understoichiometric in Cu and the single domains severely limited in extension. Finally, we have observed that the application of a 4 T magnetic field during the solidification of a bar the longest axis of which was horizontal tended to align the c axis of the 123 grains parallel to the field. Incidentally, we have discovered that in the same conditions, the 211 particles trapped into the 123 matrix were also preferentially oriented. (author)

  10. TruStore: Implementing a Trusted Store for Android

    OpenAIRE

    Yury, Zhauniarovich; Olga, Gadyatskaya; Bruno, Crispo

    2013-01-01

    In the Android ecosystem, the process of verifying the integrity of downloaded apps is left to the user. Different from other systems, e.g., Apple, App Store, Google does not provide any certified vetting process for the Android apps. This choice has a lot of advantages but it is also the open door to possible attacks as the recent one shown by Bluebox. To address this issue, we present how to enable the deployment of application certification service, we called TruStores, for the Android pla...

  11. Los Alamos controlled air incinerator upgrade for TRU/mixed waste operations

    International Nuclear Information System (INIS)

    Vavruska, J.S.; Borduin, L.C.; Hutchins, D.A.; Warner, C.L.; Thompson, T.K.

    1989-01-01

    The Los Alamos Controlled Air Incinerator (CAI) is undergoing a major process upgrade to accept Laboratory-generated transuranic (TRU) and TRU mixed wastes on a production basis. In the interim,prior to the scheduled 1992 operation of a new on-site LLW/mixed waste incinerator, the CAI will also be accepting solid and liquid low-level mixed wastes. This paper describes major modifications that have been made to the process to enhance safety and ensure reliability for long-term, routine waste incineration operations. The regulatory requirements leading to operational status of the system are also briefly described. The CAI was developed in the mid-1970s as a demonstration system for volume reduction of TRU combustible solid wastes. It continues as a successful R and D system well into the 1980s during which incineration tests on a wide variety of radioactive and chemical waste forms were performed. In 1985, a DOE directive required Los Alamos to reduce the volume of its TRU waste prior to ultimate placement in the geological repository at the Waste Isolation Pilot Project (WIPP). With only minor modifications to the original process flowsheet, the Los Alamos CAI was judged capable of conversion to a TRU waste operations mode. 9 refs., 1 fig

  12. High-Speed Synchrotron X-ray Imaging Studies of the Ultrasound Shockwave and Enhanced Flow during Metal Solidification Processes

    Science.gov (United States)

    Tan, Dongyue; Lee, Tung Lik; Khong, Jia Chuan; Connolley, Thomas; Fezzaa, Kamel; Mi, Jiawei

    2015-07-01

    The highly dynamic behavior of ultrasonic bubble implosion in liquid metal, the multiphase liquid metal flow containing bubbles and particles, and the interaction between ultrasonic waves and semisolid phases during solidification of metal were studied in situ using the complementary ultrafast and high-speed synchrotron X-ray imaging facilities housed, respectively, at the Advanced Photon Source, Argonne National Laboratory, US, and Diamond Light Source, UK. Real-time ultrafast X-ray imaging of 135,780 frames per second revealed that ultrasonic bubble implosion in a liquid Bi-8 wt pctZn alloy can occur in a single wave period (30 kHz), and the effective region affected by the shockwave at implosion was 3.5 times the original bubble diameter. Furthermore, ultrasound bubbles in liquid metal move faster than the primary particles, and the velocity of bubbles is 70 ~ 100 pct higher than that of the primary particles present in the same locations close to the sonotrode. Ultrasound waves can very effectively create a strong swirling flow in a semisolid melt in less than one second. The energetic flow can detach solid particles from the liquid-solid interface and redistribute them back into the bulk liquid very effectively.

  13. Potential Flammable Gas Explosion in the TRU Vent and Purge Machine

    International Nuclear Information System (INIS)

    Vincent, A

    2006-01-01

    The objective of the analysis was to determine the failure of the Vent and Purge (V and P) Machine due to potential explosion in the Transuranic (TRU) drum during its venting and/or subsequent explosion in the V and P machine from the flammable gases (e.g., hydrogen and Volatile Organic Compounds [VOCs]) vented into the V and P machine from the TRU drum. The analysis considers: (a) increase in the pressure in the V and P cabinet from the original deflagration in the TRU drum including lid ejection, (b) pressure wave impact from TRU drum failure, and (c) secondary burns or deflagrations resulting from excess, unburned gases in the cabinet area. A variety of cases were considered that maximized the pressure produced in the V and P cabinet. Also, cases were analyzed that maximized the shock wave pressure in the cabinet from TRU drum failure. The calculations were performed for various initial drum pressures (e.g., 1.5 and 6 psig) for 55 gallon TRU drum. The calculated peak cabinet pressures ranged from 16 psig to 50 psig for various flammable gas compositions. The blast on top of cabinet and in outlet duct ranged from 50 psig to 63 psig and 12 psig to 16 psig, respectively, for various flammable gas compositions. The failure pressures of the cabinet and the ducts calculated by structural analysis were higher than the pressure calculated from potential flammable gas deflagrations, thus, assuring that V and P cabinet would not fail during this event. National Fire Protection Association (NFPA) 68 calculations showed that for a failure pressure of 20 psig, the available vent area in the V and P cabinet is 1.7 to 2.6 times the required vent area depending on whether hydrogen or VOCs burn in the V and P cabinet. This analysis methodology could be used to design the process equipment needed for venting TRU waste containers at other sites across the Department of Energy (DOE) Complex

  14. Nuclear-waste-management technical support in the development of nuclear-waste-form criteria for the NRC. Task 2. Alternative TRU technologies

    International Nuclear Information System (INIS)

    Bida, G.; MacKenzie, D.R.

    1982-02-01

    Three main areas of transuranic (TRU) waste management are addressed: immobilization processes and waste forms for ultimate geologic disposal of TRU waste; decontamination as a method for TRU waste management; and potential problems associated with gas generation by certain TRU wastes. Waste forms are considered in terms of the regulations and criteria proposed in 10 CFR 60. Evaluation of the waste forms is based principally on ability to meet the release rate criterion of 10 -5 /year given in the Performance Objectives of Section 111, but also on the general requirements of Section 133. The two classes of metallic waste which are candidates for decontamination treatment are Zircaloy cladding hulls from light water reactor fuel elements, and failed facilities and equipment. Decontamination methods are addressed with regard to their ability to remove contamination to a level below the 10 nCi/g TRU limit. Other important factors are the volume reduction achieved, and compatibility of the secondary waste streams with acceptable waste forms. Gas generation by combustible TRU wastes and cast concretes containing TRU isotopes is discussed, and its potential for damage to a geologic repository is considered. Exclusion of combustible TRU waste from repositories is recommended. Conclusions are drawn about the suitability of various waste forms and recommendations are made regarding further work needed in the development of specific TRU waste forms

  15. Development of a safe TRU transportation system (STRUTS) for DOE's TRU waste

    International Nuclear Information System (INIS)

    Edling, D.A.; Hopkins, D.R.; Walls, H.C.

    1978-01-01

    Transportation, the link between TRU waste generation and WIPP (Waste Isolation Pilot Project) and a vital link in the overall TRU waste management program, must be addressed. The program must have many facets: ensuring public and carrier acceptance, formation of a functional and current transportation data base, systems integration, maximum utilization of existing technology, and effective implementation and integration of the transport system into current and planned operational systems

  16. Solidification and vitrification life-cycle economics study

    International Nuclear Information System (INIS)

    Gimpel, R.F.

    1992-01-01

    Solidification (making concrete) and vitrification (making glass) are frequently the treatment methods recommended for treating inorganic or radioactive wastes. Ex-situ solidification and vitrification are the competing methods for treating in excess of 450 000 cm 3 of low-level radioactive and mixed wastes at the Fernald Environmental Management Project (FEMP) located near Cincinnati, Ohio. This paper summarizes a detailed study done to: (1) compare the economics of the solidification and vitrification processes, (2) determine if the stigma assigned to vitrification is warranted and, (3) determine if investing millions of dollars into vitrification development, along with solidification development, at Fernald is warranted

  17. Study on characteristics of spent PWR cladding hull for categorizing into Non-TRU waste

    International Nuclear Information System (INIS)

    Jung, In Ha; Kim, Jong Ho; Park, Jang Jin; Shin, Jin Myeong; Lee, Ho Hee; Yang, Myung Seung

    2005-01-01

    AFCI and GEN-IV programs aim for decreasing the high level radioactive wastes to be disposed. They also try to get valuable materials to recycle as resources such as uranium and plutonium. On the other hand, cladding hull expected to be one-thirds in volume of spent fuel assembly has not studied so much in the point view of recycling to reuse. Since traditional process of reprocessing was wet process, cladding hull generating through the reprocessing process was unavoidably contaminated with TRU by acid solvent during the process. Therefore, cladding hull has been classified into TRU wastes or high level wastes. According to the strategy for TRU high level radioactive wastes of USA as well as Korea, it regulates in two respects. One is activity and the other is heat generation. In respect of activity, TRU waste contains more than 100 nCi/kg of alpha emits with longer half life than 20 years and higher than 92 in atomic number. Also, wastes are categorized into TRU waste when it generates higher than 2kW/m3, in the respect of heat generation. Our results as well as literatures, almost all of TRU nuclides in the cladding hull are responsible for remained uranium and plutonium owing to pellet-cladding interaction. In addition, recoiled fission products on the surface of the cladding hull serve as heat generator. Up to now, decontamination of the cladding hull generating from the reprocessing of wet process is regarded as valueless and un-economic works owing to the amount of second waste produced

  18. Physical and chemical feasibility of fueling molten salt reactors with TRU's trifluorides

    International Nuclear Information System (INIS)

    Ignatiev, V.; Feinberg, O.; Konakov, S.; Subbotine, S.; Surenkov, A.; Zakirov, R.

    2001-01-01

    The molten salt reactor (MSR) concept is very important for consideration as an element of future nuclear energy systems. These reactor systems are unique in many ways. Particularly, the MSRs appear to have substantial promise not only as advanced TRU free system operating in U-Th cycle, but also as transmuter of TRU. Physical and chemical feasibility of fueling MSR with TRU trifluorides is examined. Solvent compositions with and without U-Th as fissile / fertile addition are considered. The principle reactor and fuel cycle variables available for optimizing the performance of MSR as TRU transmuting system are discussed. These efforts led to the definition in minimal TRU mass flow rate, reduced total losses to waste and maximum possible burn up rate for the molten salt transmuter. The current status of technology and prospects for revisited interest are summarized. Significant chemical problems are remain to be resolved at the end of prior MSRs programs, notably, graphite life durability, tritium control, fate of noble metal fission products. Questions arising from plutonium and minor actinide fueling include: corrosion and container chemistry, new redox buffer for systems without uranium, analytical chemistry instrumentation, adequate constituent solubilities, suitable fuel processing and waste form development. However these problems appear to be soluble. (author)

  19. Solidification with back-diffusion of irregular eutectics

    Directory of Open Access Journals (Sweden)

    M. Trepczyńska-Łent

    2008-10-01

    Full Text Available The definition of the α - parameter back-diffusion has been introduced in the work. The alternative models of solidification were describedtaking into consideration back-diffusion process. The possibility of using those models for eutectic alloys solidification is worthyof interest.

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

    International Nuclear Information System (INIS)

    1988-06-01

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

  1. ''New ' technology of solidification of liquid radioactive waste'

    International Nuclear Information System (INIS)

    Sytyl, V.A.; Svistova, L.M.; Spiridonova, V.P.

    1998-01-01

    It is generally accepted that the best method of processing of radioactive waste is its solidification and then storage. At present time, three methods of solidification of radioactive waste are widely used in the world: cementation, bituminous grouting and vitrification. But they do not solve the problem of ecologically processing of waste because of different disadvantages. General disadvantages are: low state of filling, difficulties in solidification of the crystalline hydrated forms of radioactive waste; particular sphere of application and economical difficulties while processing the great volume of waste. In connection with it the urgent necessity is emerging: to develop less expensive and ecologically more reliable technology of solidification of radioactive waste. A new method of solidification is presented with its technical schema. (N.C.)

  2. Conceptual design study and evaluation of an advanced treatment process applying a submerged combustion technique for spent solvents

    International Nuclear Information System (INIS)

    Uchiyama, Gunzo; Maeda, Mitsuru; Fijine, Sachio; Chida, Mitsuhisa; Kirishima, Kenji.

    1993-10-01

    An advanced treatment process based on a submerged combustion technique was proposed for spent solvents and the distillation residues containing transuranium (TRU) nuclides. A conceptual design study and the preliminary cost estimation of the treatment facility applying the process were conducted. Based on the results of the study, the process evaluation on the technical features, such as safety, volume reduction of TRU waste and economics was carried out. The key requirements for practical use were also summarized. It was shown that the process had the features as follows: the simplified treatment and solidification steps will not generate secondary aqueous wastes, the volume of TRU solid waste will be reduced less than one tenth of that of a reference technique (pyrolysis process), and the facility construction cost is less than 1 % of the total construction cost of a future large scale reprocessing plant. As for the low level wastes of calcium phosphate, it was shown that the further removal of β · γ nuclides with TRU nuclides from the wastes would be required for the safety in interim storage and transportation and for the load of shielding. (author)

  3. Documentation of TRU biological transport model (BIOTRAN)

    Energy Technology Data Exchange (ETDEWEB)

    Gallegos, A.F.; Garcia, B.J.; Sutton, C.M.

    1980-01-01

    Inclusive of Appendices, this document describes the purpose, rationale, construction, and operation of a biological transport model (BIOTRAN). This model is used to predict the flow of transuranic elements (TRU) through specified plant and animal environments using biomass as a vector. The appendices are: (A) Flows of moisture, biomass, and TRU; (B) Intermediate variables affecting flows; (C) Mnemonic equivalents (code) for variables; (D) Variable library (code); (E) BIOTRAN code (Fortran); (F) Plants simulated; (G) BIOTRAN code documentation; (H) Operating instructions for BIOTRAN code. The main text is presented with a specific format which uses a minimum of space, yet is adequate for tracking most relationships from their first appearance to their formulation in the code. Because relationships are treated individually in this manner, and rely heavily on Appendix material for understanding, it is advised that the reader familiarize himself with these materials before proceeding with the main text.

  4. Documentation of TRU biological transport model (BIOTRAN)

    International Nuclear Information System (INIS)

    Gallegos, A.F.; Garcia, B.J.; Sutton, C.M.

    1980-01-01

    Inclusive of Appendices, this document describes the purpose, rationale, construction, and operation of a biological transport model (BIOTRAN). This model is used to predict the flow of transuranic elements (TRU) through specified plant and animal environments using biomass as a vector. The appendices are: (A) Flows of moisture, biomass, and TRU; (B) Intermediate variables affecting flows; (C) Mnemonic equivalents (code) for variables; (D) Variable library (code); (E) BIOTRAN code (Fortran); (F) Plants simulated; (G) BIOTRAN code documentation; (H) Operating instructions for BIOTRAN code. The main text is presented with a specific format which uses a minimum of space, yet is adequate for tracking most relationships from their first appearance to their formulation in the code. Because relationships are treated individually in this manner, and rely heavily on Appendix material for understanding, it is advised that the reader familiarize himself with these materials before proceeding with the main text

  5. TRU partnership-Working smarter-Not harder

    International Nuclear Information System (INIS)

    Armstrong, D.W.; Briggs, S.R.; Martin, M.R.; Turner, D.R.

    1994-01-01

    The open-quotes TRU Partnershipclose quotes was initiated and continues to function under the catch phrase philosophy of open-quotes work smarter, not harderclose quotes. The parntership participants have realized that DOE no longer has the funding available to reinvent the wheel at each site. Information and experiences from each site need to accurately and timely provided to the other sites for their use. The project teams from the different TRU waste handling sites benefit enormously from the strong network that has developed between TRU partnership participants. The partnership working interface places design manager in touch with design manager, project manager with project manager, etc. across site boundaries, and equally important, across corporate boundaries. The TRU Partnership has created a team atmosphere for the participants. The team focus is on the common challenge of managing TRU waste projects to support site needs and the needs of the national TRU waste program. Although consistency of approach for all projects at any given site is important, the TRU Partnership provides an intersite forum to establish consistency and understanding across all DOE projects managing TRU waste. The TRU Partnership has adopted the Westinghouse Electric Corporation open-quotes Savings Through Sharingclose quotes philosophy as an integral part of its organizational objectives. As applied by the group, the approach concentrates on information and experiences that can enhance development and reduce costs for (TRU) waste projects

  6. Polymer Solidification Technology - Technical Issues and Challenges

    International Nuclear Information System (INIS)

    Jensen, Charles; Kim, Juyoul

    2010-01-01

    Many factors come into play, most of which are discovered and resolved only during full-scale solidification testing of each of the media commonly used in nuclear power plants. Each waste stream is unique, and must be addressed accordingly. This testing process is so difficult that Diversified's Vinyl Ester Styrene and Advanced Polymer Solidification are the only two approved processes in the United States today. This paper summarizes a few of the key obstacles that must be overcome to achieve a reliable, repeatable process for producing an approved Stable Class B and C waste form. Before other solidification and encapsulation technologies can be considered compliant with the requirements of a Stable waste form, the tests, calculations and reporting discussed above must be conducted for both the waste form and solidification process used to produce the waste form. Diversified's VERI TM and APS TM processes have gained acceptance in the UK. These processes have also been approved and gained acceptance in the U. S. because we have consistently overcome technical hurdles to produce a complaint product. Diversified Technologies processes are protected intellectual property. In specific instances, we have patents pending on key parts of our process technology

  7. Investigation of columnar-to-equiaxed transition in solidification processing of AlSi alloys in microgravity – The CETSOL project

    International Nuclear Information System (INIS)

    Zimmermann, G; Sturz, L; Billia, B; Mangelinck-Noël, N; Thi, H Nguyen; Gandin, Ch- A; Browne, D J; Mirihanage, W U

    2011-01-01

    Grain structures observed in most casting processes of metallic alloys are the result of a competition between the growth of several arrays of dendrites that develop under constrained and unconstrained conditions. Often this leads to a transition from columnar to equiaxed grain growth during solidification (CET). A microgravity environment results in suppression of buoyancy-driven melt flow and so enables growth of equiaxed grains free of sedimentation and buoyancy effects. This contribution presents first results obtained in experiments on-board the International Space Station (ISS), which were performed in the frame of the ESA-MAP programme CETSOL. Hypoeutectic aluminium-silicon alloys with and without grain refiners were processed successfully in a low gradient furnace (MSL-LGF). First analysis shows that in the non grain refined samples columnar dendritic growth exists, whereas CET is observed in the grain refined samples. From analysis of the thermal data and the grain structure the critical parameters for the temperature gradient and the cooling rate describing CET are determined. These data are used for initial numerical simulations to predict the position of the columnar-to-equiaxed transition and will form a unique database for calibration and further development of numerical CET-modeling.

  8. Sensitivity of Transmutation Capability to Recycling Scenarios in KALIMER-600 TRU Burner

    International Nuclear Information System (INIS)

    Lee, Yong Kyo; Kim, Myung Hyun

    2013-01-01

    The purpose of this study is to test transmutation and design feasibility of KALIMER burner caused from many limitations in recycling options; such as low recovery factors and external feed. Design impact from many recycling options will be tested as a sensitivity to various recycling process parameters under many recycling scenarios. Through this study, possibilities when Pyro-processing is realized with SFR can be expected in the recycling scenarios. For the development of sodium-cooled fast reactor(SFR) technology, prototype KALIMER plant is now under R and D stage in Korea. For the future application of SFR for waste transmutation, KALIMER core was designed for TRU burner by KAERI. Feasibility of TRU burner cannot be evaluated exactly because overall functional parameters in pyro-processing recycling process has not been verified yet. There is great possibility to accept undesirable process functions in pyro-processing. Only TRU nuclides composition a little differs between PWR SF and CANDU SF so first scenario has no problem operating SFR. In second scenario, the radiotoxicity of waste at 99% of TRU RF have to be confirmed whether it is proper level to reposit as Low and Intermediate Level Wastes or not. And the reactor safety at high RF of RE must be inspected. Not only third scenario but also several scenarios for good measure are being calculated and will be evaluated

  9. Development of an Alternative Treatment Scheme for Sr/TRU Removal: Permanganate Treatment of AN-107 Waste

    Energy Technology Data Exchange (ETDEWEB)

    RT Hallen; SA Bryan; FV Hoopes

    2000-08-04

    A number of Hanford tanks received waste containing organic complexants, which increase the volubility of Sr-90 and transuranic (TRU) elements. Wastes from these tanks require additional pretreatment to remove Sr-90 and TRU for immobilization as low activity waste (Waste Envelope C). The baseline pretreatment process for Sr/TRU removal was isotopic exchange and precipitation with added strontium and iron. However, studies at both Battelle and Savannah River Technology Center (SRTC) have shown that the Sr/Fe precipitates were very difficult to filter. This was a result of the formation of poor filtering iron solids. An alternate treatment technology was needed for Sr/TRU removal. Battelle had demonstrated that permanganate treatment was effective for decontaminating waste samples from Hanford Tank SY-101 and proposed that permanganate be examined as an alternative Sr/TRU removal scheme for complexant-containing tank wastes such as AW107. Battelle conducted preliminary small-scale experiments to determine the effectiveness of permanganate treatment with AN-107 waste samples that had been archived at Battelle from earlier studies. Three series of experiments were performed to evaluate conditions that provided adequate Sr/TRU decontamination using permanganate treatment. The final series included experiments with actual AN-107 diluted feed that had been obtained specifically for BNFL process testing. Conditions that provided adequate Sr/TRU decontamination were identified. A free hydroxide concentration of 0.5M provided adequate decontamination with added Sr of 0.05M and permanganate of 0.03M for archived AN-107. The best results were obtained when reagents were added in the sequence Sr followed by permanganate with the waste at ambient temperature. The reaction conditions for Sr/TRU removal will be further evaluated with a 1-L batch of archived AN-107, which will provide a large enough volume of waste to conduct crossflow filtration studies (Hallen et al. 2000a).

  10. Development of an Alternative Treatment Scheme for Sr/TRU Removal: Permanganate Treatment of AN-107 Waste

    International Nuclear Information System (INIS)

    Hallen, R.T.; Bryan, S.A.; Hoopes, F.V.

    2000-01-01

    A number of Hanford tanks received waste containing organic complexants, which increase the volubility of Sr-90 and transuranic (TRU) elements. Wastes from these tanks require additional pretreatment to remove Sr-90 and TRU for immobilization as low activity waste (Waste Envelope C). The baseline pretreatment process for Sr/TRU removal was isotopic exchange and precipitation with added strontium and iron. However, studies at both Battelle and Savannah River Technology Center (SRTC) have shown that the Sr/Fe precipitates were very difficult to filter. This was a result of the formation of poor filtering iron solids. An alternate treatment technology was needed for Sr/TRU removal. Battelle had demonstrated that permanganate treatment was effective for decontaminating waste samples from Hanford Tank SY-101 and proposed that permanganate be examined as an alternative Sr/TRU removal scheme for complexant-containing tank wastes such as AW107. Battelle conducted preliminary small-scale experiments to determine the effectiveness of permanganate treatment with AN-107 waste samples that had been archived at Battelle from earlier studies. Three series of experiments were performed to evaluate conditions that provided adequate Sr/TRU decontamination using permanganate treatment. The final series included experiments with actual AN-107 diluted feed that had been obtained specifically for BNFL process testing. Conditions that provided adequate Sr/TRU decontamination were identified. A free hydroxide concentration of 0.5M provided adequate decontamination with added Sr of 0.05M and permanganate of 0.03M for archived AN-107. The best results were obtained when reagents were added in the sequence Sr followed by permanganate with the waste at ambient temperature. The reaction conditions for Sr/TRU removal will be further evaluated with a 1-L batch of archived AN-107, which will provide a large enough volume of waste to conduct crossflow filtration studies (Hallen et al. 2000a)

  11. Radioactive gas solidification treatment device

    International Nuclear Information System (INIS)

    Igarashi, Ryokichi; Watanabe, Yu; Seki, Eiji.

    1992-01-01

    In a radioactive gas solidification treatment device by using sputtering, spiral pipelines are disposed with a gap therebetween for cooling an ion injection electrode by passing cooling water during operation of the solidification treatment. During the operation of the solidification treatment, cooling water is passed in the pipelines to cool the ion injection electrode. During storage, a solidification vessel is cooled by natural heat dissipation from an exposed portion at the surface of the solidification vessel. Accordingly, after-heat of radioactive gas solidified in a metal accumulation layer can be removed efficiently, safely and economically to improve the reliability. (N.H.)

  12. CH-TRU Waste Content Codes

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2008-01-16

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  13. Characterization optimization for the National TRU waste system

    International Nuclear Information System (INIS)

    Basabilvazo, George T.; Countiss, S.; Moody, D.C.; Jennings, S.G.; Lott, S.A.

    2002-01-01

    On March 26, 1999, the Waste Isolation Pilot Plant (WIPP) received its first shipment of transuranic (TRU) waste. On November 26, 1999, the Hazardous Waste Facility Permit (HWFP) to receive mixed TRU waste at WIPP became effective. Having achieved these two milestones, facilitating and supporting the characterization, transportation, and disposal of TRU waste became the major challenges for the National TRU Waste Program. Significant challenges still remain in the scientific, engineering, regulatory, and political areas that need to be addressed. The National TRU Waste System Optimization Project has been established to identify, develop, and implement cost-effective system optimization strategies that address those significant challenges. Fundamental to these challenges is the balancing and prioritization of potential regulatory changes with potential technological solutions. This paper describes some of the efforts to optimize (to make as functional as possible) characterization activities for TRU waste.

  14. Examination of solidified and stabilized matrices as a result of solidification and stabilization process of arseniccontaining sludge with portland cement and lime

    Directory of Open Access Journals (Sweden)

    Tanapon Phenrat

    2004-02-01

    Full Text Available By solidification and stabilization (S/S with Portland cement and lime, it is possible to reduce arsenic concentration in leachate of the arsenic-containing sludge from arsenic removal process by coagulation with ferric chloride. From the initial arsenic concentration in leachate of unsolidified /unstabilized sludge which was around 20.75 mg/L, the arsenic concentrations in leachate of solidified/stabilized waste were reduced to 0.3, 0.58, 1.09, and 1.85 mg/L for the waste-to-binder ratios of 0.15, 0.25, 0.5, and 1, respectively, due tothe formation of insoluble calcium-arsenic compounds. To be more cost effective for the future, alternative uses of these S/S products were also assessed by measurement of compressive strength of the mortar specimens. It was found that the compressive strengths of these matrices were from 28 ksc to 461 ksc. In conclusion, considering compressive strength and leachability of the solidified matrices, some of these solidified/ stabilized products have potential to serve as an interlocking concrete paving block.

  15. Defect generation during solidification of aluminium foams

    International Nuclear Information System (INIS)

    Mukherjee, M.; Garcia-Moreno, F.; Banhart, J.

    2010-01-01

    The reason for the frequent occurrence of cell wall defects in metal foams was investigated. Aluminium foams often expand during solidification, a process which is referred as solidification expansion (SE). The effect of SE on the structure of aluminium foams was studied in situ by X-ray radioscopy and ex situ by X-ray tomography. A direct correlation between the magnitude of SE and the number of cell wall ruptures during SE and finally the number of defects in the solidified foams was found.

  16. Incorporating interfacial phenomena in solidification models

    Science.gov (United States)

    Beckermann, Christoph; Wang, Chao Yang

    1994-01-01

    A general methodology is available for the incorporation of microscopic interfacial phenomena in macroscopic solidification models that include diffusion and convection. The method is derived from a formal averaging procedure and a multiphase approach, and relies on the presence of interfacial integrals in the macroscopic transport equations. In a wider engineering context, these techniques are not new, but their application in the analysis and modeling of solidification processes has largely been overlooked. This article describes the techniques and demonstrates their utility in two examples in which microscopic interfacial phenomena are of great importance.

  17. Enhancing TRU burning and Am transmutation in Advanced Recycling Reactor

    International Nuclear Information System (INIS)

    Ikeda, Kazumi; Kochendarfer, Richard A.; Moriwaki, Hiroyuki; Kunishima, Shigeru

    2011-01-01

    Research highlights: → This ARR is an oxide fueled sodium cooled reactor based on innovative technologies to destruct TRU. → TRU burning core is designed to burn TRU at 28 kg/TW th h, adding moderator pins of B 4 C (Enriched B-11). → Am transmutation core can transmute Am at 34 kg/TW th h, adding uranium free AmN blanket to TRU burning core. → The TRU burning core improves TRU burning by 40-50% than the previous core. → The Am transmutation core can transmute Am effectively, keeping the void reactivity acceptable. - Abstract: This paper presents about conceptual designs of Advanced Recycling Reactor (ARR) focusing on enhancement in transuranics (TRU) burning and americium (Am) transmutation. The design has been conducted in the context of the Global Nuclear Energy Partnership (GNEP) seeking to close nuclear fuel cycle in ways that reduce proliferation risks, reduce the nuclear waste in the US and further improve global energy security. This study strives to enhance the TRU burning and the Am transmutation, assuming the development of related technologies in this study, while the ARR based on mature technologies was designed in the previous study. It has followed that the provided TRU burning core is designed to burn TRU at 28 kg/TW th h, by adding moderator pins of B 4 C (Enriched B-11) and the Am transmutation core will be able to transmute Am at 34 kg/TW th h, by locating Am blanket of AmN around the TRU burning core. It indicates that these concepts improve TRU burning by 40-50% than the previous core and can transmute Am effectively, keeping the void reactivity acceptable.

  18. Solidification of highly active liquid waste

    International Nuclear Information System (INIS)

    Morris, J.B.

    1985-03-01

    This document contains the annual progress reports on the following subjects: Joule ceramic melter; microwave vitrification; glass technology; identification, evaluation and review of potential alternative solidification processes; rotary kiln calcination; alternative glass feedstocks; volatile ruthenium trapping by solid adsorbents; irrigated baffle column dust scrubber. (author)

  19. 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

  20. The effect of the melt spinning processing parameters on the solidification structures in Ti-30 at.% Ni-20 at.% Cu shape memory alloys

    International Nuclear Information System (INIS)

    Kim, Yeon-wook; Yun, Young-mok; Nam, Tae-hyun

    2006-01-01

    Solidification structures and shape memory characteristics of Ti-30 at.% Ni-20 at.% Cu alloy ribbons prepared by melt spinning were investigated by means of differential scanning calorimetry and X-ray diffraction. In these experiments particular attention has been paid to change the ejection temperature of the melt from 1350 to 1500 deg. C and the velocity of cooling wheel from 33 to 55 m/s. Then the cooling rates of ribbons were controlled. The effect of this cooling rate on solidification structures and martensitic transformation behaviors is discussed

  1. FY 1997 report on the study on solidification process of high-temperature melt of heat resistant metals; 1997 nendo chosa hokokusho (tainetsu kinzoku koon yueki no gyoko katei no kenkyu)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    Study was made on a solidification process of metal melt under micro-gravity condition in an underground non-gravity experiment center, considering that improvement of the heat resistance of turbine blades for jet engines and power generation gas turbines contributes to prevention of global warming through improvement of thermal engine efficiencies and consumption reduction of precious fossil fuel. Study was made on a simulation program and precise measurement of thermal properties for precision casting of heat-resistant alloy members. Study was also made on Al and Zn alloys and their welding for production and evaluation technologies of new metal textures by supercooling solidification. Some issues for strongly desired improvement of a simulation program for precision casting were clarified. In addition, since thermal property data of practical heat-resistant polyalloy members are poor, data and measurement method for precision casting were clarified. It was also suggested that basic elucidation of the solidification process under micro- gravity condition is possible. 34 refs., 41 figs., 5 tabs.

  2. Solidification of oils and organic liquids

    International Nuclear Information System (INIS)

    Clark, D.E.; Colombo, P.; Neilson, R.M. Jr.

    1982-07-01

    The suitability of selected solidification media for application in the disposal of low-level oil and other organic liquid wastes has been investigated. In the past, these low-level wastes (LLWs) have commonly been immobilized by sorption onto solid absorbents such as vermiculite or diatomaceous earth. Evolving regulations regarding the disposal of these materials encourage solidification. Solidification media which were studied include Portland type I cement; vermiculite plus Portland type I cement; Nuclear Technology Corporation's Nutek 380-cement process; emulsifier, Portland type I cement-sodium silicate; Delaware Custom Materiel's cement process; and the US Gypsum Company's Envirostone process. Waste forms have been evaluated as to their ability to reliably produce free standing monolithic solids which are homogeneous (macroscopically), contain < 1% free standing liquids by volume and pass a water immersion test. Solidified waste form specimens were also subjected to vibratory shock testing and flame testing. Simulated oil wastes can be solidified to acceptable solid specimens having volumetric waste loadings of less than 40 volume-%. However, simulated organic liquid wastes could not be solidified into acceptable waste forms above a volumetric loading factor of about 10 volume-% using the solidification agents studied

  3. 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.

  4. 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

  5. Microwave solidification project overview

    Energy Technology Data Exchange (ETDEWEB)

    Sprenger, G.

    1993-01-01

    The Rocky Flats Plant Microwave Solidification Project has application potential to the Mixed Waste Treatment Project and the The Mixed Waste Integrated Program. The technical areas being addressed include (1) waste destruction and stabilization; (2) final waste form; and (3) front-end waste handling and feed preparation. This document covers need for such a program; technology description; significance; regulatory requirements; and accomplishments to date. A list of significant reports published under this project is included.

  6. Microwave solidification project overview

    International Nuclear Information System (INIS)

    Sprenger, G.

    1993-01-01

    The Rocky Flats Plant Microwave Solidification Project has application potential to the Mixed Waste Treatment Project and the The Mixed Waste Integrated Program. The technical areas being addressed include (1) waste destruction and stabilization; (2) final waste form; and (3) front-end waste handling and feed preparation. This document covers need for such a program; technology description; significance; regulatory requirements; and accomplishments to date. A list of significant reports published under this project is included

  7. Thermosolutal convection during dendritic solidification

    Science.gov (United States)

    Heinrich, J. C.; Nandapurkar, P.; Poirier, D. R.; Felicelli, S.

    1989-01-01

    This paper presents a mathematical model for directional solidification of a binary alloy including a dendritic region underlying an all-liquid region. It is assumed initially that there exists a nonconvecting state with planar isotherms and isoconcentrates solidifying at a constant velocity. The stability of this system has been analyzed and nonlinear calculations are performed that show the effect of convection in the solidification process when the system is unstable. Results of calculations for various cases defined by the initial temperature gradient at the dendrite tips and varying strength of the gravitational field are presented for systems involving lead-tin alloys. The results show that the systems are stable for a gravitational constant of 0.0001 g(0) and that convection can be suppressed by appropriate choice of the container's size for higher values of the gravitational constant. It is also concluded that for the lead-tin systems considered, convection in the mushy zone is not significant below the upper 20 percent of the dendritic zone, if al all.

  8. Software documentation for TRU certification program

    International Nuclear Information System (INIS)

    CLINTON, R.

    1999-01-01

    The document provides validation information for software used to support TRU operational activities. Calculations were performed using a spreadsheet application. This document provides information about the usage of the software application, Microsoft(reg s ign) Excel. Microsoft(reg s ign) Excel spreadsheets were used to perform specific calculations to determine the amount of containers to visually examine and to perform analyses on container head-gas data. Contained in this document are definitions of formulas and variables with relation to the Excel codes used. Also, a demonstration is provided using predetermined values to obtain predetermined results

  9. Advanced modeling of solidification

    International Nuclear Information System (INIS)

    Bousquet-Melou, P.; Fichot, F.; Goyeau, B.; Gobin, D.; Quintard, M.

    2001-01-01

    A theoretical and numerical macroscopic modeling of the solidification of binary mixtures is presented. The growth of a solid-liquid region (mushy zone), represented by a non-homogeneous porous medium, is considered. A macroscopic model for momentum, heat and mass transfer during solidification is derived using the volume averaging method, and the effective transport properties (permeability, effective diffusivities, mass exchange coefficients) are defined by associated closure problems (set of microscopic balance equations). Consequently, the effects of the dendritic geometry (tortuosity) and of microscopic transfer phenomena (dispersion, interfacial exchange) are introduced in the averaged balance equations and in the representation of the effective transport coefficients. This closure method provides an original approach of solidification modeling. The resulting macroscopic model is based on the local thermal equilibrium assumption (one-temperature model) while a two-phase description of macroscopic species transfer is introduced using solid and liquid mass exchange coefficients. The phase diagram is used to predict the solid and liquid equilibrium concentrations at the solid-liquid interface. This two-phase approach extends the classical limiting cases that correspond to the lever-rule and Scheil descriptions. (authors)

  10. Experimental investigation of molten salt droplet quenching and solidification processes of heat recovery in thermochemical hydrogen production

    International Nuclear Information System (INIS)

    Ghandehariun, S.; Wang, Z.; Naterer, G.F.; Rosen, M.A.

    2015-01-01

    Highlights: • Thermal efficiency of a thermochemical cycle of hydrogen production is improved. • Direct contact heat recovery from molten salt is analyzed. • Falling droplets quenched into water are investigated experimentally. - Abstract: This paper investigates the heat transfer and X-ray diffraction patterns of solidified molten salt droplets in heat recovery processes of a thermochemical Cu–Cl cycle of hydrogen production. It is essential to recover the heat of the molten salt to enhance the overall thermal efficiency of the copper–chlorine cycle. A major portion of heat recovery within the cycle can be achieved by cooling and solidifying the molten salt exiting an oxygen reactor. Heat recovery from the molten salt is achieved by dispersing the molten stream into droplets. In this paper, an analytical study and experimental investigation of the thermal phenomena of a falling droplet quenched into water is presented, involving the droplet surface temperature during descent and resulting composition change in the quench process. The results show that it is feasible to quench the molten salt droplets for an efficient heat recovery process without introducing any material imbalance for the overall cycle integration.

  11. TRU waste transportation -- The flammable gas generation problem

    International Nuclear Information System (INIS)

    Connolly, M.J.; Kosiewicz, S.T.

    1997-01-01

    The Nuclear Regulatory Commission (NRC) has imposed a flammable gas (i.e., hydrogen) concentration limit of 5% by volume on transuranic (TRU) waste containers to be shipped using the TRUPACT-II transporter. This concentration is the lower explosive limit (LEL) in air. This was done to minimize the potential for loss of containment during a hypothetical 60 day period. The amount of transuranic radionuclide that is permissible for shipment in TRU waste containers has been tabulated in the TRUPACT-II Safety Analysis Report for Packaging (SARP, 1) to conservatively prevent accumulation of hydrogen above this 5% limit. Based on the SARP limitations, approximately 35% of the TRU waste stored at the Idaho National Engineering and Environmental Lab (INEEL), Los Alamos National Lab (LANL), and Rocky Flats Environmental Technology Site (RFETS) cannot be shipped in the TRUPACT-II. An even larger percentage of the TRU waste drums at the Savannah River Site (SRS) cannot be shipped because of the much higher wattage loadings of TRU waste drums in that site's inventory. This paper presents an overview of an integrated, experimental program that has been initiated to increase the shippable portion of the Department of Energy (DOE) TRU waste inventory. In addition, the authors will estimate the anticipated expansion of the shippable portion of the inventory and associated cost savings. Such projection should provide the TRU waste generating sites a basis for developing their TRU waste workoff strategies within their Ten Year Plan budget horizons

  12. DOE's plan for buried transuranic (TRU) contaminated waste

    International Nuclear Information System (INIS)

    Mathur, J.; D'Ambrosia, J.; Sease, J.

    1987-01-01

    Prior to 1970, TRU-contaminated waste was buried as low-level radioactive waste. In the Defense Waste Management Plan issued in 1983, the plan for this buried TRU-contaminated waste was to monitor the buried waste, take remedial actions, and to periodically evaluate the safety of the waste. In March 1986, the General Accounting Office (GAO) recommended that the Department of Energy (DOE) provide specific plans and cost estimates related to buried TRU-contaminated waste. This plan is in direct response to the GAO request. Buried TRU-contaminated waste and TRU-contaminated soil are located in numerous inactive disposal units at five DOE sites. The total volume of this material is estimated to be about 300,000 to 500,000 m 3 . The DOE plan for TRU-contaminated buried waste and TRU-contaminated soil is to characterize the disposal units; assess the potential impacts from the waste on workers, the surrounding population, and the environment; evaluate the need for remedial actions; assess the remedial action alternatives; and implement and verify the remedial actions as appropriate. Cost estimates for remedial actions for the buried TRU-contaminated waste are highly uncertain, but they range from several hundred million to the order of $10 billion

  13. Repackaging SRS Black Box TRU Waste

    International Nuclear Information System (INIS)

    Swale, D. J.; Stone, K.A.; Milner, T. N.

    2006-01-01

    Historically, large items of TRU Waste, which were too large to be packaged in drums for disposal have been packaged in various sizes of custom made plywood boxes at the Savannah River Site (SRS), for many years. These boxes were subsequently packaged into large steel ''Black Boxes'' for storage at SRS, pending availability of Characterization and Certification capability, to facilitate disposal of larger items of TRU Waste. There are approximately 107 Black Boxes in inventory at SRS, each measuring some 18' x 12' x 7', and weighing up to 45,000 lbs. These Black Boxes have been stored since the early 1980s. The project to repackage this waste into Standard Large Boxes (SLBs), Standard Waste Boxes (SWB) and Ten Drum Overpacks (TDOP), for subsequent characterization and WIPP disposal, commenced in FY04. To date, 10 Black Boxes have been repackaged, resulting in 40 SLB-2's, and 37 B25 overpack boxes, these B25's will be overpacked in SLB-2's prior to shipping to WIPP. This paper will describe experience to date from this project

  14. FIPS: a process for the solidification of fission product solutions using a drum drier. [HTGR fuel reprocessing

    Energy Technology Data Exchange (ETDEWEB)

    Halaszovich, St.; Laser, M.; Merz, E.; Thiele, D.

    1976-08-01

    A new process consisting of the steps concentration of the fission product solution, denitration of the solution by addition of formaldehyde, addition of glass-forming additives, drying of the slurry using a drum drier, melting of the dry product in the crucible by rising level in-pot-melting, and off-gas treatment and recovery of nitric acid is under development. A small plant with a capacity of 1 kg glass per hour has been tested in hot cells with fission product solutions from LWR fuel element reprocessing since December 1974. The equipment is very simple to operate and to control. No serious problems arose during operation.

  15. A microencapsulation process of liquid mercury by sulfur polymer stabilization/solidification technology. Part I: Characterization of materials

    Energy Technology Data Exchange (ETDEWEB)

    Lopez-Delgado, A.; Lopez, F. A.; Alguacil, F. J.; Padilla, I.; Guerrero, A.

    2012-11-01

    European Directives consider mercury a priority hazardous substance due to its adverse effects on human health and the environment. In response to environmental concerns, a microencapsulation process has been developed within the European LIFE program as a long-term storage option for mercury. This process leads to the obtainment of a stable concrete-like sulfur matrix that allows the immobilization of mercury. The final product, in the form of a solid block containing up to 30 % Hg, exhibits excellent mechanical properties (compressive strength 53-61MPa and flexural strength 7-10 MPa), low porosity (0.57 % PHe), very low total pore volume (0.63x10-2 cm{sup 3} g{sup -}1), and extremely low permeability (coefficient of water absorption by capillarity 0.07 g cm{sup -}2). Toxicity characteristic leaching tests reveal a mercury concentration in leachates well below the 0.2 mg L{sup -}1 set out in US EPA Land Disposal Restrictions (LDRs). The values of mercury vapor emissions of final products were lower than those of cinnabar and meta cinnabar. (Author)

  16. Major Components of the National TRU Waste System Optimization Project

    International Nuclear Information System (INIS)

    Moody, D.C.; Bennington, B.; Sharif, F.

    2002-01-01

    The National Transuranic (TRU) Program (NTP) is being optimized to allow for disposing of the legacy TRU waste at least 10 years earlier than originally planned. This acceleration will save the nation an estimated $713. The Department of Energy's (DOE'S) Carlsbad Field Office (CBFO) has initiated the National TRU Waste System Optimization Project to propose, and upon approvaI, implement activities that produce significant cost saving by improving efficiency, thereby accelerating the rate of TRU waste disposal without compromising safety. In its role as NTP agent of change, the National TRU Waste System Optimization Project (the Project) (1) interacts closely with all NTP activities. Three of the major components of the Project are the Central Characterization Project (CCP), the Central Confirmation Facility (CCF), and the MobiIe/Modular Deployment Program.

  17. Gravitational influences on the liquid-state homogenization and solidification of aluminum antimonide. [space processing of solar cell material

    Science.gov (United States)

    Ang, C.-Y.; Lacy, L. L.

    1979-01-01

    Typical commercial or laboratory-prepared samples of polycrystalline AlSb contain microstructural inhomogeneities of Al- or Sb-rich phases in addition to the primary AlSb grains. The paper reports on gravitational influences, such as density-driven convection or sedimentation, that cause microscopic phase separation and nonequilibrium conditions to exist in earth-based melts of AlSb. A triple-cavity electric furnace is used to homogenize the multiphase AlSb samples in space and on earth. A comparative characterization of identically processed low- and one-gravity samples of commercial AlSb reveals major improvements in the homogeneity of the low-gravity homogenized material.

  18. Investigation on the asymmetry of thermal condition and grain defect formation in the customary directional solidification process

    International Nuclear Information System (INIS)

    Ma, D; Wu, Q; Hollad, S; Bührig-Polaczek, A

    2012-01-01

    In the present study, the non-uniformity of the thermal condition and the corresponding grain defect formation in the customary Bridgman process were investigated. The casting clusters in radial alignment were directionally solidified in a Bridgman furnace. It was found that in the casting cluster, the shadow side facing the central rod was ineffectively heated in the hot zone and ineffectively cooled in the cooling zone during withdrawal, compared with the heater side facing the furnace heater. The metallographic examination of the simplified turbine blades exhibited that the platforms on the shadow side are very prone to stray grain formation, while the heater side reveals a markedly lower tendency for that. The asymmetric thermal condition causes the asymmetrical formation of these grain defects. This non-uniformity of the thermal condition should be minimized as far as possible, in order to effectively optimize the quality of the SC superalloy components.

  19. 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.

  20. 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

  1. Multi-scale Modeling of Dendritic Alloy Solidification

    OpenAIRE

    Dagner, Johannes

    2009-01-01

    Solidification of metallic melts is one of the most important processes in material science. The microstructure, which is formed during freezing, determines the mechanical properties of the final product largely. Many physical phenomena influence the solidification process and hence the resulting microstructure. One important parameter is influence of melt flow, which may modify heat and species transport on a large range of length- and time-scales. On the micro-scale, it influences the conce...

  2. Actinide recycling by pyro process for future nuclear fuel cycle system

    International Nuclear Information System (INIS)

    Inoue, T.

    2001-01-01

    Pyrometallurgical technology is one of the potential devices for the future nuclear fuel cycle. Not only economic advantage but also environmental safety and strong resistance for proliferation are required. So as to satisfy the requirements, actinide recycling applicable to LWR and FBR cycles by pyro-process has been developed over a ten-year period at the CRIEPI. The main technology is electrorefining for U and Pu separation and reductive extraction for TRU separation, which can be applied on oxide fuels through reduction process as well as metal fuels. The application of this technology for separation of TRU in HLLW through chlorination could contribute to the improvement of public acceptance with regard to geologic disposal. The main achievements are summarised as follows: - Elemental technologies such as electrorefining, reductive extraction, injection casting and salt waste treatment and solidification have been successfully developed with lots of experiments. - Fuel dissolution into molten salt and uranium recovery on solid cathode for electrorefining has been demonstrated at an engineering scale facility in Argonne National Laboratory using spent fuels and at the CRIEPI through uranium tests. - Single element tests using actinides showed Li reduction to be technically feasible; the subjects of technical feasibility on multi-element systems and on effective recycle of Li by electrolysis of Li 2 O remain to be addressed. - Concerning the treatment of HLLW for actinide separation, the conversion to chlorides through oxides has also been established through uranium tests. - It is confirmed that more than 99% of TRU nuclides can be recovered from high-level liquid waste by TRU tests. - Through these studies, the process flowsheets for reprocessing of metal and oxide fuels and for partitioning of TRU separation have been established. The subjects to be emphasised for further development are classified into three categories: process development (demonstration

  3. A microencapsulation process of liquid mercury by sulfur polymer stabilization/solidification technology. Part I: Characterization of materials

    Directory of Open Access Journals (Sweden)

    López-Delgado, A.

    2012-02-01

    Full Text Available European Directives consider mercury a priority hazardous substance due to its adverse effects on human health and the environment. In response to environmental concerns, a microencapsulation process has been developed within the European LIFE program as a long-term storage option for mercury. This process leads to the obtainment of a stable concrete-like sulfur matrix that allows the immobilization of mercury. The final product, in the form of a solid block containing up to 30 % Hg, exhibits excellent mechanical properties (compressive strength 53-61MPa and flexural strength 7-10 MPa, low porosity (0.57 % PHe, very low total pore volume (0.63x10-2 cm3 g-1, and extremely low permeability (coefficient of water absorption by capillarity 0.07 g cm-2. Toxicity characteristic leaching tests reveal a mercury concentration in leachates well below the 0.2 mg L-1 set out in US EPA Land Disposal Restrictions (LDRs. The values of mercury vapor emissions of final products were lower than those of cinnabar and metacinnabar.

    Resumen Las Directivas Europeas consideran al mercurio una sustancia de peligrosidad prioritaria debido a sus efectos adversos sobre la salud humana y sobre el medio ambiente. En respuesta a estas preocupaciones ambientales, y dentro del Programa Europeo LIFE, se ha desarrollado un proceso de microencapsulación como una opción al almacenamiento a largo plazo del mercurio. Con este proceso se obtiene un material estable, tipo concreto, de matriz de azufre que permite la inmovilización del mercurio. El producto final, en forma de un bloque sólido, contiene hasta un 30 % de Hg, presenta excelentes propiedades mecánicas (resistencia a la compresión 53-61 MPa, y a la flexión 7-10 MPa, baja porosidad (0,57 % PHe, muy bajo volumen total de poro (0,63 x 10-2 cm3 g-1 y una permeabilidad extremadamente baja (coeficiente de absorción de

  4. Utilization of the MPI Process for in-tank solidification of heel material in large-diameter cylindrical tanks

    Energy Technology Data Exchange (ETDEWEB)

    Kauschinger, J.L.; Lewis, B.E.

    2000-01-01

    A major problem faced by the US Department of Energy is remediation of sludge and supernatant waste in underground storage tanks. Exhumation of the waste is currently the preferred remediation method. However, exhumation cannot completely remove all of the contaminated materials from the tanks. For large-diameter tanks, amounts of highly contaminated ``heel'' material approaching 20,000 gal can remain. Often sludge containing zeolite particles leaves ``sand bars'' of locally contaminated material across the floor of the tank. The best management practices for in-tank treatment (stabilization and immobilization) of wastes require an integrated approach to develop appropriate treatment agents that can be safely delivered and mixed uniformly with sludge. Ground Environmental Services has developed and demonstrated a remotely controlled, high-velocity jet delivery system termed, Multi-Point-Injection (MPI). This robust jet delivery system has been field-deployed to create homogeneous monoliths containing shallow buried miscellaneous waste in trenches [fiscal year (FY) 1995] and surrogate sludge in cylindrical (FY 1998) and long, horizontal tanks (FY 1999). During the FY 1998 demonstration, the MPI process successfully formed a 32-ton uniform monolith of grout and waste surrogates in about 8 min. Analytical data indicated that 10 tons of zeolite-type physical surrogate were uniformly mixed within a 40-in.-thick monolith without lifting the MPI jetting tools off the tank floor. Over 1,000 lb of cohesive surrogates, with consistencies similar to Gunite and Associated Tank (GAAT) TH-4 and Hanford tank sludges, were easily intermixed into the monolith without exceeding a core temperature of 100 F during curing.

  5. Effect Of Natural Convection On Directional Solidification Of Pure Metal

    Directory of Open Access Journals (Sweden)

    Skrzypczak T.

    2015-06-01

    Full Text Available The paper is focused on the modeling of the directional solidification process of pure metal. During the process the solidification front is sharp in the shape of the surface separating liquid from solid in three dimensional space or a curve in 2D. The position and shape of the solid-liquid interface change according to time. The local velocity of the interface depends on the values of heat fluxes on the solid and liquid sides. Sharp interface solidification belongs to the phase transition problems which occur due to temperature changes, pressure, etc. Transition from one state to another is discontinuous from the mathematical point of view. Such process can be identified during water freezing, evaporation, melting and solidification of metals and alloys, etc.

  6. Simulations of rapid pressure-induced solidification in molten metals

    International Nuclear Information System (INIS)

    Patel, Mehul V.; Streitz, Frederick H.

    2004-01-01

    The process of interest in this study is the solidification of a molten metal subjected to rapid pressurization. Most details about solidification occurring when the liquid-solid coexistence line is suddenly transversed along the pressure axis remain unknown. We present preliminary results from an ongoing study of this process for both simple models of metals (Cu) and more sophisticated material models (MGPT potentials for Ta). Atomistic (molecular dynamics) simulations are used to extract details such as the time and length scales that govern these processes. Starting with relatively simple potential models, we demonstrate how molecular dynamics can be used to study solidification. Local and global order parameters that aid in characterizing the phase have been identified, and the dependence of the solidification time on the phase space distance between the final (P,T) state and the coexistence line has been characterized

  7. Assessment of the Mechanisms for Sr-90 and TRU Removal from Complexant-Containing Tank Wastes at Hanford

    International Nuclear Information System (INIS)

    Hallen, Richard T.; Geeting, John GH; Lilga, Michael A.; Hart, Todd R.; Hoopes, Francis V.

    2005-01-01

    Small-scale tests (∼20 mL) were conducted with samples from Hanford underground storage tanks AN-102 and AN-107 to assess the mechanisms for removing Sr-90 and transuranics (TRU) from the liquid (supernatant) portion of the waste. The Sr-90 and TRU must be removed (decontaminated), in addition to Cs-137 and the entrained solids, before the supernatant can be disposed of as low-activity waste. Experiments were conducted with various reagents and modified Sr/TRU removal process conditions to more fully understand the reaction mechanisms. The optimized treatment conditions--no added hydroxide, addition of Sr (0.02M target concentration) followed by sodium permanganate (0.02M target concentration) with mixing at ambient temperature--were used as a reference for comparison. The waste was initially two orders of magnitude undersaturated with Sr; the addition of nonradioactive Sr(NO?) ? saturated the supernatant, resulting in isotopic dilution and precipitation of Sr-90 as SrCO?. The reaction chemistry of Mn species relevant to the mechanism of TRU removal by permanganate treatment was evaluated, along with the importance of various mechanisms for decontamination, such as precipitation, absorption, ligand exchange, and oxidation of organic complexants. For TRU removal, permanganate addition generally gave the highest DF. The addition of Mn of lower oxidation states (II, IV, and VI) also resulted in good TRU removal, as did complexant oxidation with periodate and addition of Zr(IV) for ligand exchange. These results suggest that permanganate treatment leads to TRU removal by multiple routes

  8. Application of insoluble tannin to recovery of uranium, TRU and heavy metals elements form radioactive liquid waste

    International Nuclear Information System (INIS)

    Hamaguchi, Kazuhiko; Shirato, Wataru; Nakamura, Yasuo; Matsumura, Tatsuro; Takeshita, Kenji; Nakano, Yoshio

    1999-01-01

    Mitsubishi Nuclear Fuel Co., Ltd. (MNF) has developed a new adsorbent, TANNIX (tread mark), for the recovery of uranium, TRU and heavy metal elements in the liquid waste, in which TANNIX derived from a natural tannin polymer. TANNIX has same advantages that handling is easier than that of standard IX-resin, and that the volume of secondary waste is reduced by burning the used TANNIX. We have replaced its radioactive liquid waste treatment system from the conventional co-precipitation process to adsorption process by using TANNIX. TANNIX was founded to be more effective for the recovery of Pu, TRU, and hexavalent chromium Cr-(VI) as well as Uranium. (author)

  9. Polyethylene solidification of low-level wastes

    International Nuclear Information System (INIS)

    Kalb, P.D.; Colombo, P.

    1985-02-01

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

  10. Transuranic (TRU) Waste Repackaging at the Nevada Test Site

    International Nuclear Information System (INIS)

    Di Sanza, E.F.; Pyles, G.; Ciucci, J.; Arnold, P.

    2009-01-01

    This paper describes the activities required to modify a facility and the process of characterizing, repackaging, and preparing for shipment the Nevada Test Site's (NTS) legacy transuranic (TRU) waste in 58 oversize boxes (OSB). The waste, generated at other U.S. Department of Energy (DOE) sites and shipped to the NTS between 1974 and 1990, requires size-reduction for off-site shipment and disposal. The waste processing approach was tailored to reduce the volume of TRU waste by employing decontamination and non-destructive assay. As a result, the low-level waste (LLW) generated by this process was packaged, with minimal size reduction, in large sea-land containers for disposal at the NTS Area 5 Radioactive Waste Management Complex (RWMC). The remaining TRU waste was repackaged and sent to the Idaho National Laboratory Consolidation Site for additional characterization in preparation for disposal at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. The DOE National Nuclear Security Administration Nevada Site Office and the NTS Management and Operating (M and O) contractor, NSTec, successfully partnered to modify and upgrade an existing facility, the Visual Examination and Repackaging Building (VERB). The VERB modifications, including a new ventilation system and modified containment structure, required an approved Preliminary Documented Safety Analysis prior to project procurement and construction. Upgrade of the VERB from a radiological facility to a Hazard Category 3 Nuclear Facility required new rigor in the design and construction areas and was executed on an aggressive schedule. The facility Documented Safety Analysis required that OSBs be vented prior to introduction into the VERB. Box venting was safely completed after developing and implementing two types of custom venting systems for the heavy gauge box construction. A remotely operated punching process was used on boxes with wall thickness of up to 3.05 mm (0.120 in) to insert aluminum

  11. Rock solidification method

    International Nuclear Information System (INIS)

    Nakaya, Iwao; Murakami, Tadashi; Miyake, Takafumi; Funakoshi, Toshio; Inagaki, Yuzo; Hashimoto, Yasuhide.

    1985-01-01

    Purpose: To convert radioactive wastes into the final state for storage (artificial rocks) in a short period of time. Method: Radioactive burnable wastes such as spent papers, cloths and oils and activated carbons are burnt into ashes in a burning furnace, while radioactive liquid wastes such as liquid wastes of boric acid, exhausted cleaning water and decontaminating liquid wastes are powderized in a drying furnace or calcining furnace. These powders are joined with silicates as such as white clay, silica and glass powder and a liquid alkali such as NaOH or Ca(OH) 2 and transferred to a solidifying vessel. Then, the vessel is set to a hydrothermal reactor, heated and pressurized, then taken out about 20 min after and tightly sealed. In this way, radioactive wastes are converted through the hydrothermal reactions into aqueous rock stable for a long period of time to obtain solidification products insoluble to water and with an extremely low leaching rate. (Ikeda, J.)

  12. Terminating Safeguards on Excess Special Nuclear Material: Defense TRU Waste Clean-up and Nonproliferation - 12426

    Energy Technology Data Exchange (ETDEWEB)

    Hayes, Timothy [Los Alamos National Laboratory, Carlsbad Operations Group (United States); Nelson, Roger [Department Of Energy, Carlsbad Operations Office (United States)

    2012-07-01

    The Department of Energy (DOE) and the National Nuclear Security Administration (NNSA) manages defense nuclear material that has been determined to be excess to programmatic needs and declared waste. When these wastes contain plutonium, they almost always meet the definition of defense transuranic (TRU) waste and are thus eligible for disposal at the Waste Isolation Pilot Plant (WIPP). The DOE operates the WIPP in a manner that physical protections for attractiveness level D or higher special nuclear material (SNM) are not the normal operating condition. Therefore, there is currently a requirement to terminate safeguards before disposal of these wastes at the WIPP. Presented are the processes used to terminate safeguards, lessons learned during the termination process, and how these approaches might be useful for future defense TRU waste needing safeguards termination prior to shipment and disposal at the WIPP. Also described is a new criticality control container, which will increase the amount of fissile material that can be loaded per container, and how it will save significant taxpayer dollars. Retrieval, compliant packaging and shipment of retrievably stored legacy TRU waste has dominated disposal operations at WIPP since it began operations 12 years ago. But because most of this legacy waste has successfully been emplaced in WIPP, the TRU waste clean-up focus is turning to newly-generated TRU materials. A major component will be transuranic SNM, currently managed in safeguards-protected vaults around the weapons complex. As DOE and NNSA continue to consolidate and shrink the weapons complex footprint, it is expected that significant quantities of transuranic SNM will be declared surplus to the nation's needs. Safeguards termination of SNM varies due to the wide range of attractiveness level of the potential material that may be directly discarded as waste. To enhance the efficiency of shipping waste with high TRU fissile content to WIPP, DOE designed an

  13. A task for laser cutting of lamellae with TruLaser 1030

    OpenAIRE

    Lazov, Lyubomir; Deneva, Hristina; Narica, Pavels

    2015-01-01

    The growing development of manufacturing, automotive, aerospace and other sectors in the industry generates the necessity to continuously expanding on modifications of electrical machinery and equipments which are used in them, as well as to improve their performance and reliability. The report presents some results from a study to the process of laser cutting through melting on lamellae for rotor and stator packages by using the laser system TruLaser 1030. Some functional dependencies are of...

  14. Pyrolysis/Steam Reforming Technology for Treatment of TRU Orphan Wastes

    International Nuclear Information System (INIS)

    Mason, J. B.; McKibbin, J.; Schmoker, D.; Bacala, P.

    2003-01-01

    Certain transuranic (TRU) waste streams within the Department of Energy (DOE) complex cannot be disposed of at the Waste Isolation Pilot Plant (WIPP) because they do not meet the shipping requirements of the TRUPACT-II or the disposal requirements of the Waste Analysis Plan (WAP) in the WIPP RCRA Part B Permit. These waste streams, referred to as orphan wastes, cannot be shipped or disposed of because they contain one or more prohibited items, such as liquids, volatile organic compounds (VOCs), hydrogen gas, corrosive acids or bases, reactive metals, or high concentrations of polychlorinated biphenyl (PCB), etc. The patented, non-incineration, pyrolysis and steam reforming processes marketed by THOR Treatment Technologies LLC removes all of these prohibited items from drums of TRU waste and produces a dry, inert, inorganic waste material that meets the existing TRUPACT-II requirements for shipping, as well as the existing WAP requirements for disposal of TRU waste at WIPP. THOR Treatment Technologies is a joint venture formed in June 2002 by Studsvik, Inc. (Studsvik) and Westinghouse Government Environmental Services Company LLC (WGES) to further develop and deploy Studsvik's patented THORSM technology within the DOE and Department of Defense (DoD) markets. The THORSM treatment process is a commercially proven system that has treated over 100,000 cu. ft. of nuclear waste from commercial power plants since 1999. Some of this waste has had contact dose rates of up to 400 R/hr. A distinguishing characteristic of the THORSM process for TRU waste treatment is the ability to treat drums of waste without removing the waste contents from the drum. This feature greatly minimizes criticality and contamination issues for processing of plutonium-containing wastes. The novel features described herein are protected by issued and pending patents

  15. Statistical analysis of radiochemical measurements of TRU radionuclides in REDC waste

    International Nuclear Information System (INIS)

    Beauchamp, J.; Downing, D.; Chapman, J.; Fedorov, V.; Nguyen, L.; Parks, C.; Schultz, F.; Yong, L.

    1996-10-01

    This report summarizes results of the study on the isotopic ratios of transuranium elements in waste from the Radiochemical Engineering Development Center actinide-processing streams. The knowledge of the isotopic ratios when combined with results of nondestructive assays, in particular with results of Active-Passive Neutron Examination Assay and Gamma Active Segmented Passive Assay, may lead to significant increase in precision of the determination of TRU elements contained in ORNL generated waste streams

  16. A strategy for analysis of TRU waste characterization needs

    International Nuclear Information System (INIS)

    Leigh, C.D.; Chu, M.S.Y.; Arvizu, J.S.; Marcinkiewicz, C.J.

    1994-01-01

    Regulatory compliance and effective management of the nation's TRU waste requires knowledge about the constituents present in the waste. With limited resources, the DOE needs a cost-effective characterization program. In addition, the DOE needs a method for predicting the present and future analytical requirements for waste characterization. Thus, a strategy for predicting the present and future waste characterization needs that uses current knowledge of the TRU inventory and prioritization of the data needs is presented

  17. Plans for Managing Hanford Remote Handled Transuranic (TRU) Waste

    International Nuclear Information System (INIS)

    MCKENNEY, D.E.

    2001-01-01

    The current Hanford Site baseline and life-cycle waste forecast predicts that approximately 1,000 cubic meters of remote-handled transuranic (RH-TRU) waste will be generated by waste management and environmental restoration activities at Hanford. These 1,000 cubic meters, comprised of both transuranic and mixed transuranic (TRUM) waste, represent a significant portion of the total estimated inventory of RH-TRU to be disposed of at the Waste Isolation Pilot Plant (WIPP). A systems engineering approach is being followed to develop a disposition plan for each RH-TRU/TRUM waste stream at Hanford. A number of significant decision-making efforts are underway to develop and finalize these disposition plans, including: development and approval of a RH-TRU/TRUM Waste Project Management Plan, revision of the Hanford Waste Management Strategic Plan, the Hanford Site Options Study (''Vision 2012''), the Canyon Disposal Initiative Record-of-Decision, and the Hanford Site Solid (Radioactive and Hazardous) Waste Program Environmental Impact Statement (SW-EIS). Disposition plans may include variations of several options, including (1) sending most RH-TRU/TRUM wastes to WIPP, (2) deferrals of waste disposal decisions in the interest of both efficiency and integration with other planned decision dates and (3) disposition of some materials in place consistent with Department of Energy Orders and the regulations in the interest of safety, risk minimization, and cost. Although finalization of disposition paths must await completion of the aforementioned decision documents, significant activities in support of RH-TRU/TRUM waste disposition are proceeding, including Hanford participation in development of the RH TRU WIPP waste acceptance criteria, preparation of T Plant for interim storage of spent nuclear fuel sludge, sharing of technology information and development activities in cooperation with the Mixed Waste Focus Area, RH-TRU technology demonstrations and deployments, and

  18. Transuranic (Tru) waste volume reduction operations at a plutonium facility

    Energy Technology Data Exchange (ETDEWEB)

    Cournoyer, Michael E [Los Alamos National Laboratory; Nixon, Archie E [Los Alamos National Laboratory; Dodge, Robert L [Los Alamos National Laboratory; Fife, Keith W [Los Alamos National Laboratory; Sandoval, Arnold M [Los Alamos National Laboratory; Garcia, Vincent E [Los Alamos National Laboratory

    2010-01-01

    Programmatic operations at the Los Alamos National Laboratory Plutonium Facility (TA 55) involve working with various amounts of plutonium and other highly toxic, alpha-emitting materials. The spread of radiological contamination on surfaces, airborne contamination, and excursions of contaminants into the operator's breathing zone are prevented through use of a variety of gloveboxes (the glovebox, coupled with an adequate negative pressure gradient, provides primary confinement). Size-reduction operations on glovebox equipment are a common activity when a process has been discontinued and the room is being modified to support a new customer. The Actin ide Processing Group at TA-55 uses one-meter-long glass columns to process plutonium. Disposal of used columns is a challenge, since they must be size-reduced to get them out of the glovebox. The task is a high-risk operation because the glass shards that are generated can puncture the bag-out bags, leather protectors, glovebox gloves, and the worker's skin when completing the task. One of the Lessons Learned from these operations is that Laboratory management should critically evaluate each hazard and provide more effective measures to prevent personnel injury. A bag made of puncture-resistant material was one of these enhanced controls. We have investigated the effectiveness of these bags and have found that they safely and effectively permit glass objects to be reduced to small pieces with a plastic or rubber mallet; the waste can then be easily poured into a container for removal from the glove box as non-compactable transuranic (TRU) waste. This size-reduction operation reduces solid TRU waste generation by almost 2% times. Replacing one-time-use bag-out bags with multiple-use glass crushing bags also contributes to reducing generated waste. In addition, significant costs from contamination, cleanup, and preparation of incident documentation are avoided. This effort contributes to the Los Alamos

  19. Transuranic (Tru) waste volume reduction operations at a plutonium facility

    International Nuclear Information System (INIS)

    Cournoyer, Michael E.; Nixon, Archie E.; Dodge, Robert L.; Fife, Keith W.; Sandoval, Arnold M.; Garcia, Vincent E.

    2010-01-01

    Programmatic operations at the Los Alamos National Laboratory Plutonium Facility (TA 55) involve working with various amounts of plutonium and other highly toxic, alpha-emitting materials. The spread of radiological contamination on surfaces, airborne contamination, and excursions of contaminants into the operator's breathing zone are prevented through use of a variety of gloveboxes (the glovebox, coupled with an adequate negative pressure gradient, provides primary confinement). Size-reduction operations on glovebox equipment are a common activity when a process has been discontinued and the room is being modified to support a new customer. The Actin ide Processing Group at TA-55 uses one-meter-long glass columns to process plutonium. Disposal of used columns is a challenge, since they must be size-reduced to get them out of the glovebox. The task is a high-risk operation because the glass shards that are generated can puncture the bag-out bags, leather protectors, glovebox gloves, and the worker's skin when completing the task. One of the Lessons Learned from these operations is that Laboratory management should critically evaluate each hazard and provide more effective measures to prevent personnel injury. A bag made of puncture-resistant material was one of these enhanced controls. We have investigated the effectiveness of these bags and have found that they safely and effectively permit glass objects to be reduced to small pieces with a plastic or rubber mallet; the waste can then be easily poured into a container for removal from the glove box as non-compactable transuranic (TRU) waste. This size-reduction operation reduces solid TRU waste generation by almost 2% times. Replacing one-time-use bag-out bags with multiple-use glass crushing bags also contributes to reducing generated waste. In addition, significant costs from contamination, cleanup, and preparation of incident documentation are avoided. This effort contributes to the Los Alamos National

  20. Transuranic (TRU) waste volume reduction operations at a plutonium facility

    International Nuclear Information System (INIS)

    Cournoyer, Michael E.; Nixon, Archie E.; Fife, Keith W.; Sandoval, Arnold M.; Garcia, Vincent E.; Dodge, Robert L.

    2011-01-01

    Programmatic operations at the Los Alamos National Laboratory Plutonium Facility (TA-55) involve working with various amounts of plutonium and other highly toxic, alpha-emitting materials. The spread of radiological contamination on surfaces, airborne contamination, and excursions of contaminants into the operator's breathing zone are prevented through use of a variety of gloveboxes (the glovebox, coupled with an adequate negative pressure gradient, provides primary confinement). Size-reduction operations on glovebox equipment are a common activity when a process has been discontinued and the room is being modified to support a new customer. The Actinide Processing Group at TA-55 uses one-meter or longer glass columns to process plutonium. Disposal of used columns is a challenge, since they must be size-reduced to get them out of the glovebox. The task is a high-risk operation because the glass shards that are generated can puncture the bag-out bags, leather protectors, glovebox gloves, and the worker's skin when completing the task. One of the Lessons Learned from these operations is that Laboratory management should critically evaluate each hazard and provide more effective measures to prevent personnel injury. A bag made of puncture-resistant material was one of these enhanced controls. We have investigated the effectiveness of these bags and have found that they safely and effectively permit glass objects to be reduced to small pieces with a plastic or rubber mallet; the waste can then be easily poured into a container for removal from the glovebox as non-compactable transuranic (TRU) waste. This size-reduction operation reduces solid TRU waste volume generation by almost 2½ times. Replacing one-time-use bag-out bags with multiple-use glass crushing bags also contributes to reducing generated waste. In addition, significant costs from contamination, cleanup, and preparation of incident documentation are avoided. This effort contributes to the Los Alamos

  1. TRU waste transport economics: an overview

    International Nuclear Information System (INIS)

    Edling, D.A.; Hopkins, D.R.; Walls, H.C.

    1978-01-01

    There are currently three predominant methods used to transport transuranium contaminated waste. These are: (1) ATMX Railcars--500 and 600 series, (2) Super Tigers, and (3) Poly Panthers. Both the ATMX-500 and 600 series railcars are massive doubly walled steel railcars which provide the equivalent protection of a Type B package. In ATMX-600 the rapid loading and unloading of the 9 x 9 x 50 feet cargo space is achieved by prepackaging the TRU waste into standard 20-foot steel cargo containers. The ATMX-500 railcars are divided into three inside bays, having dimensions of 16 (l) x 9.25 (w) x 6.25 (h) feet. A typical load consists of 128 55-gallon drums (however, space can accommodate 192 drums), 12 fiberglass boxes (4 x 4 x 7), or a combination of palletized drums and boxes. A Super Tiger is an overpack authorized for Type A, Type B, and large quantities of radioactive materials having outside dimensions of 8 x 8 x 20 feet. Maximum payload is approximately 28,700 lb with a gross weight of 45,000 lb. The primary factors influencing transport costs are examined including freight rates of transport mode, effective cargo (weight and volume) management, effective utilization of available space (package design), transport mileage, and rental fees or initial capital outlay. Miscellaneous factors are also examined

  2. Solidification and vitrification life-cycle economics study

    International Nuclear Information System (INIS)

    Gimpel, R.F.

    1992-01-01

    Solidification (making concrete) and vitrification (making glass) are frequently the treatment methods recommended for treating inorganic or radioactive wastes. Solidification is generally perceived as the most economical treatment method, whereas vitrification is considered (by many) as the most effective of all treatment methods. Unfortunately, vitrification has acquired the stigma that it is too expensive to receive further consideration as an alternative to solidification in high volume treatment applications. Ex situ solidification and vitrification are the competing methods for treating in excess of 450,000 m 3 of low-level radioactive and mixed waste at the Fernald Environmental Management Project (FEMP or simply, Fernald) located near Cincinnati, Ohio. This paper s a detailed study done to: compare the economics of the solidification and vitrification processes; determine if the stigma assigned to vitrification is warranted; determine if investing millions of dollars into vitrification development, along with solidification development, at Fernald is warranted. Common parameters were determined and detailed life-cycle cost estimates were made. Incorporating the unit costs into a computer spreadsheet allowed 'what if' scenarios to be performed. Some scenarios investigated included variation of: remediation times, amount of wastes treated, treatment efficiencies, electrical and material costs and escalation

  3. Weld solidification cracking in 304 to 304L stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Hochanadel, Patrick W [Los Alamos National Laboratory; Lienert, Thomas J [Los Alamos National Laboratory; Martinez, Jesse N [Los Alamos National Laboratory; Martinez, Raymond J [Los Alamos National Laboratory; Johnson, Matthew Q [Los Alamos National Laboratory

    2010-01-01

    A series of annulus welds were made between 304 and 304L stainless steel coaxial tubes using both pulsed laser beam welding (LBW) and pulsed gas tungsten arc welding (GTAW). In this application, a change in process from pulsed LBW to pulsed gas tungsten arc welding was proposed to limit the possibility of weld solidification cracking since weldability diagrams developed for GTAW display a greater range of compositions that are not crack susceptible relative to those developed for pulsed LBW. Contrary to the predictions of the GTAW weldability diagram, cracking was found. This result was rationalized in terms of the more rapid solidification rate of the pulsed gas tungsten arc welds. In addition, for the pulsed LBW conditions, the material compositions were predicted to be, by themselves, 'weldable' according to the pulsed LBW weldability diagram. However, the composition range along the tie line connecting the two compositions passed through the crack susceptible range. Microstructurally, the primary solidification mode (PSM) of the material processed with higher power LBW was determined to be austenite (A), while solidification mode of the materials processed with lower power LBW apparently exhibited a dual PSM of both austenite (A) and ferrite-austenite (FA) within the same weld. The materials processed by pulsed GT A W showed mostly primary austenite solidification, with some regions of either primary austenite-second phase ferrite (AF) solidification or primary ferrite-second phase austenite (FA) solidification. This work demonstrates that variations in crack susceptibility may be realized when welding different heats of 'weldable' materials together, and that slight variations in processing can also contribute to crack susceptibility.

  4. Weld solidification cracking in 304 to 204L stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Hochanadel, Patrick W [Los Alamos National Laboratory; Lienert, Thomas J [Los Alamos National Laboratory; Martinez, Jesse N [Los Alamos National Laboratory; Johnson, Matthew Q [Los Alamos National Laboratory

    2010-09-15

    A series of annulus welds were made between 304 and 304L stainless steel coaxial tubes using both pulsed laser beam welding (LBW) and pulsed gas tungsten arc welding (GTAW). In this application, a change in process from pulsed LBW to pulsed gas tungsten arc welding was proposed to limit the possibility of weld solidification cracking since weldability diagrams developed for GTAW display a greater range of compositions that are not crack susceptible relative to those developed for pulsed LBW. Contrary to the predictions of the GTAW weldability diagram, cracking was found.This result was rationalized in terms of the more rapid solidification rate of the pulsed gas tungsten arc welds. In addition, for the pulsed LBW conditions, the material compositions were predicted to be, by themselves, 'weldable' according to the pulsed LBW weldability diagram. However, the composition range along the tie line connecting the two compositions passed through the crack susceptible range. Microstructurally, the primary solidification mode (PSM) of the material processed with higher power LBW was determined to be austenite (A), while solidification mode of the materials processed with lower power LBW apparently exhibited a dual PSM of both austenite (A) and ferrite-austenite (FA) within the same weld. The materials processed by pulsed GTAW showed mostly primary austenite solidification, with some regions of either primary austenite-second phase ferrite (AF) solidification or primary ferrite-second phase austenite (FA) solidification. This work demonstrates that variations in crack susceptibility may be realized when welding different heats of 'weldable' materials together, and that slight variations in processing can also contribute to crack susceptibility.

  5. Solidification of radioactive aqueous solution

    Energy Technology Data Exchange (ETDEWEB)

    Aikawa, Hideaki; Kato, Kiyoshi; Wadachi, Yoshiki

    1970-09-07

    A process for solidifying a radioactive waste solution is provided, using as a solidifying agent a mixture of calcined gypsum and burnt vermiculite. The quantity ratio of the mixture is preferred to be 1:1 by volume. The quantity of impregnation is 1/2 of the volume of the total quantity of the solidifying agent. In embodiments, 10 liters of plutonium waste solution was mixed with a mixture of 1:1 calcined gypsum and burnt vermiculite contained in a 20-liter cylindrical steel container lined with asphalt. The plutonium waste solution from the laboratory was neutralized with a caustic soda aqueous solution to prevent explosion due to the nitration of organic compounds. The neutralization is not always necessary. A market available dental gypsum was calcined at 400 to 500/sup 0/C and a vermiculite from Illinois was burnt at 1,100/sup 0/C to prepare the agents. The time required for the impregnation with 10 liters of plutonium solution was four minutes. After impregnation, the temperature rose to 40/sup 0/C within 30 minutes to one hour. Next, it was cooled to room temperature by standing for 3-4 hours. Solidification time was about 1 hour. The Japan Atomic Energy Research Insitute had treated and disposed about 1,000 tons of plutonium waste by this process as of August 19, 1970.

  6. RH-TRU Waste Content Codes (RH-TRUCON)

    International Nuclear Information System (INIS)

    2007-01-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is '3.' The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR limits based

  7. RH-TRU Waste Content Codes (RH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2007-08-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: • A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. • A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is “3.” The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR

  8. RH-TRU Waste Content Codes (RH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions

    2007-05-30

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: • A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. • A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is “3.” The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR

  9. RH-TRU Waste Content Codes (RH-Trucon)

    International Nuclear Information System (INIS)

    2007-01-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is '3.' The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR limits based

  10. A study of the alterability of terminal residues stabilized by a mineral-binder-based solidification/stabilization process; Etude de l`alterabilite de residus ultimes stabilises par un procede de solidification, stabilisation a base de liants mineraux

    Energy Technology Data Exchange (ETDEWEB)

    Humez, N.; Prost, R. [Institut National de Recherches Agronomiques (INRA), 78 - Versailles (France). Unite de Science du Sol; Piketty, L.; Schnuriger, B.; Benchara, A.; Pichat, P. [Centre de Solidification, Stabilisation et Stockage, Sarp Industries, 78 - Limay (France)

    1997-12-31

    A new laboratory approach has been developed for the study of the long term behaviour of stabilized terminal residues, and is based on the experimental method for the study of rock alteration. The method consists in examining the migration of elements that are present in the solution contained in the pores, and studying the structural modifications which result from the dissolution of certain constituents, the transformation and the neo-formation of other ones in the material. The method allows for the evaluation, in a short time, of the alterability and perenniality of stabilized terminal wastes. Results show that the long term alterability of solidified/stabilized samples that have been processed using the ECOFIX method, is low

  11. Pyrometallurgical process of actinide metal

    International Nuclear Information System (INIS)

    Yoo, Jae Hyung; Kang, Young Ho; Woo, Mun Sik; Hwang, Sung Chan

    1999-06-01

    Major subject on pyrometallurgical partitioning technology is to separate transmutation elements (TRU) from rare earth elements(RE). Distribution coefficients of TRU and RE between molten chloride and liquid cadmium were measured for reductive extraction, and TRU were separated from RE in simplified molten chloride system by electrorefining. And separation efficiency between TRU and RE were estimated by using thermodynamics data. The results indicate that uranium, neptunium and plutonium are easy to separate from RE but some amount of RE accompany americium, and that processes have to be optimized to attain good separation efficiency of TRU. (author)

  12. Modular radwaste volume reduction and solidification systems

    International Nuclear Information System (INIS)

    Miller, E.L.

    1986-01-01

    This paper describes both the modular transportable and the modular mobile liquid radwaste volume reduction and solidification units based on a General Electric Company developed and patented process called AZTECH (a trademark of GE). An AZTECH system removes all water by azeotropic distillation and encapsulates the remaining solids in a polyester compound. The resulting monolith is suitable for either long term above ground storage or shallow land burial. Pilot and demonstration plant testing has confirmed the design parameters. The three processing modules are covered together with data which resulted in Nuclear Regulatory Commission approval on Dec. 30, 1985

  13. Linear Stability of Binary Alloy Solidification for Unsteady Growth Rates

    Science.gov (United States)

    Mazuruk, K.; Volz, M. P.

    2010-01-01

    An extension of the Mullins and Sekerka (MS) linear stability analysis to the unsteady growth rate case is considered for dilute binary alloys. In particular, the stability of the planar interface during the initial solidification transient is studied in detail numerically. The rapid solidification case, when the system is traversing through the unstable region defined by the MS criterion, has also been treated. It has been observed that the onset of instability is quite accurately defined by the "quasi-stationary MS criterion", when the growth rate and other process parameters are taken as constants at a particular time of the growth process. A singular behavior of the governing equations for the perturbed quantities at the constitutional supercooling demarcation line has been observed. However, when the solidification process, during its transient, crosses this demarcation line, a planar interface is stable according to the linear analysis performed.

  14. Development of waste packages for TRU-disposal. 5. Development of cylindrical metal package for TRU wastes

    International Nuclear Information System (INIS)

    Mine, Tatsuya; Mizubayashi, Hiroshi; Asano, Hidekazu; Owada, Hitoshi; Otsuki, Akiyoshi

    2005-01-01

    Development of the TRU waste package for hulls and endpieces compression canisters, which include long-lived and low sorption nuclides like C-14 is essential and will contribute a lot to a reasonable enhancement of safety and economy of the TRU-disposal system. The cylindrical metal package made of carbon steel for canisters to enhance the efficiency of the TRU-disposal system and to economically improve their stacking conditions was developed. The package is a welded cylindrical construction with a deep drawn upper cover and a disc plate for a bottom cover. Since the welding is mainly made only for an upper cover and a bottom disc plate, this package has a better containment performance for radioactive nuclide and can reduce the cost for construction and manufacturing including its welding control. Furthermore, this package can be laid down in pile for stacking in the circular cross section disposal tunnel for the sedimentary rock, which can drastically minimize the space for disposal tunnel as mentioned previously in TRU report. This paper reports the results of the study for application of newly developed metal package into the previous TRU-disposal system and for the stacking equipment for the package. (author)

  15. W-026, transuranic waste (TRU) glovebox acceptance test report

    International Nuclear Information System (INIS)

    Leist, K.J.

    1998-01-01

    On July 18, 1997, the Transuranic (TRU) glovebox was tested using glovebox acceptance test procedure 13021A-86. The primary focus of the glovebox acceptance test was to examine control system interlocks, display menus, alarms, and operator messages. Limited mechanical testing involving the drum ports, hoists, drum lifter, compacted drum lifter, drum tipper, transfer car, conveyors, sorting table, lidder/delidder device and the TRU empty drum compactor were also conducted. As of February 25, 1998, 10 of the 102 test exceptions that affect the TRU glovebox remain open. These items will be tracked and closed via the WRAP Master Test Exception Database. As part of Test Exception resolution/closure the responsible individual closing the Test Exception performs a retest of the affected item(s) to ensure the identified deficiency is corrected, and, or to test items not previously available to support testing. Test exceptions are provided as appendices to this report

  16. An investigation of TRU recycling with various neutron spectrums

    International Nuclear Information System (INIS)

    Yong-Nam, Kim; Hong-Chul, Kim; Chi-Young, Han; Jong-Kyung, Kim; Won-Seok Park

    2003-01-01

    This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single batch fuel loading, the burn-up calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analysed in terms of burn-up reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behaviour between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction. (author)

  17. Los Alamos National Laboratory TRU waste sampling projects

    International Nuclear Information System (INIS)

    Yeamans, D.; Rogers, P.; Mroz, E.

    1997-01-01

    The Los Alamos National Laboratory (LANL) has begun characterizing transuranic (TRU) waste in order to comply with New Mexico regulations, and to prepare the waste for shipment and disposal at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico. Sampling consists of removing some head space gas from each drum, removing a core from a few drums of each homogeneous waste stream, and visually characterizing a few drums from each heterogeneous waste stream. The gases are analyzed by GC/MS, and the cores are analyzed for VOC's and SVOC's by GC/MS and for metals by AA or AE spectroscopy. The sampling and examination projects are conducted in accordance with the ''DOE TRU Waste Quality Assurance Program Plan'' (QAPP) and the ''LANL TRU Waste Quality Assurance Project Plan,'' (QAPjP), guaranteeing that the data meet the needs of both the Carlsbad Area Office (CAO) of DOE and the ''WIPP Waste Acceptance Criteria, Rev. 5,'' (WAC)

  18. RH-TRU Waste Content Codes (RH TRUCON)

    International Nuclear Information System (INIS)

    2007-01-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: (1) A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. (2) A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is ''3''. The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR

  19. RH-TRU Waste Content Codes (RH TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions

    2007-05-01

    The Remote-Handled Transuranic (RH-TRU) Content Codes (RH-TRUCON) document describes the inventory of RH-TRU waste within the transportation parameters specified by the Remote-Handled Transuranic Waste Authorized Methods for Payload Control (RH-TRAMPAC).1 The RH-TRAMPAC defines the allowable payload for the RH-TRU 72-B. This document is a catalog of RH-TRU 72-B authorized contents by site. A content code is defined by the following components: • A two-letter site abbreviation that designates the physical location of the generated/stored waste (e.g., ID for Idaho National Laboratory [INL]). The site-specific letter designations for each of the sites are provided in Table 1. • A three-digit code that designates the physical and chemical form of the waste (e.g., content code 317 denotes TRU Metal Waste). For RH-TRU waste to be transported in the RH-TRU 72-B, the first number of this three-digit code is “3.” The second and third numbers of the three-digit code describe the physical and chemical form of the waste. Table 2 provides a brief description of each generic code. Content codes are further defined as subcodes by an alpha trailer after the three-digit code to allow segregation of wastes that differ in one or more parameter(s). For example, the alpha trailers of the subcodes ID 322A and ID 322B may be used to differentiate between waste packaging configurations. As detailed in the RH-TRAMPAC, compliance with flammable gas limits may be demonstrated through the evaluation of compliance with either a decay heat limit or flammable gas generation rate (FGGR) limit per container specified in approved content codes. As applicable, if a container meets the watt*year criteria specified by the RH-TRAMPAC, the decay heat limits based on the dose-dependent G value may be used as specified in an approved content code. If a site implements the administrative controls outlined in the RH-TRAMPAC and Appendix 2.4 of the RH-TRU Payload Appendices, the decay heat or FGGR

  20. Stabilization/Solidification Remediation Method for Contaminated Soil: A Review

    Science.gov (United States)

    Tajudin, S. A. A.; Azmi, M. A. M.; Nabila, A. T. A.

    2016-07-01

    Stabilization/Solidification (S/S) is typically a process that involves a mixing of waste with binders to reduce the volume of contaminant leachability by means of physical and chemical characteristics to convert waste in the environment that goes to landfill or others possibly channels. Stabilization is attempts to reduce the solubility or chemical reactivity of the waste by changing the physical and chemical properties. While, solidification attempt to convert the waste into easily handled solids with low hazardous level. These two processes are often discussed together since they have a similar purpose of improvement than containment of potential pollutants in treated wastes. The primary objective of this review is to investigate the materials used as a binder in Stabilization/Solidification (S/S) method as well as the ability of these binders to remediate the contaminated soils especially by heavy metals.

  1. Solidification Mapping of a Nickel Alloy 718 Laboratory VAR Ingot

    Science.gov (United States)

    Watt, Trevor J.; Taleff, Eric M.; Lopez, Felipe; Beaman, Joe; Williamson, Rodney

    The solidification microstructure of a laboratory-scale Nickel alloy 718 vacuum arc remelted (VAR) ingot was analyzed. The cylindrical, 210-mm-diameter ingot was sectioned along a plane bisecting it length-wise, and this mid-plane surface was ground and etched using Canada's reagent to reveal segregation contrast. Over 350 photographs were taken of the etched mid-plane surface and stitched together to form a single mosaic image. Image data in the resulting mosaic were processed using a variety of algorithms to extract quantities such as primary dendrite orientation, primary dendrite arm spacing (PDAS), and secondary dendrite arm spacing (SDAS) as a function of location. These quantities were used to calculate pool shape and solidification rate during solidification using existing empirical relationships for Nickel Alloy 718. The details and outcomes of this approach, along with the resulting comparison to experimental processing conditions and computational models, are presented.

  2. TRU waste-assay instrumentation and application in nuclear-facility decommissioning

    International Nuclear Information System (INIS)

    Umbarger, C.J.

    1982-01-01

    The Los Alamos TRU waste assay program is developing measurement techniques for TRU and other radioactive waste materials generated by the nuclear industry, including decommissioning programs. Systems are now being fielded for test and evaluation purposes at DOE TRU waste generators. The transfer of this technology to other facilities and the commercial instrumentation sector is well in progress. 6 figures

  3. Comparison between TRU burning reactors and commercial fast reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

    2001-03-01

    Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

  4. MICROBIAL TRANSFORMATIONS OF TRU AND MIXED WASTES: ACTINIDE SPECIATION AND WASTE VOLUME REDUCTION.

    Energy Technology Data Exchange (ETDEWEB)

    FRANCIS, A.J.; DODGE, C.J.

    2006-11-16

    The overall goals of this research project are to determine the mechanism of microbial dissolution and stabilization of actinides in Department of Energy's (DOE) TRU wastes, contaminated sludges, soils, and sediments. This includes (1) investigations on the fundamental aspects of microbially catalyzed radionuclide and metal transformations (oxidation/reduction reactions, dissolution, precipitation, chelation); (2) understanding of the microbiological processes that control speciation and alter the chemical forms of complex inorganic/organic contaminant mixtures; and (3) development of new and improved microbially catalyzed processes resulting in immobilization of metals and radionuclides in the waste with concomitant waste volume reduction.

  5. MICROBIAL TRANSFORMATIONS OF TRU AND MIXED WASTES: ACTINIDE SPECIATION AND WASTE VOLUME REDUCTION

    Energy Technology Data Exchange (ETDEWEB)

    Francis, A.J.; Dodge, C.J.

    2006-06-01

    The overall goals of this research project are to determine the mechanism of microbial dissolution and stabilization of actinides in Department of Energy’s (DOE) TRU wastes, contaminated sludges, soils, and sediments. This includes (i) investigations on the fundamental aspects of microbially catalyzed radionuclide and metal transformations (oxidation/reduction reactions, dissolution, precipitation, chelation); (ii) understanding of the microbiological processes that control speciation and alter the chemical forms of complex inorganic/organic contaminant mixtures; and (iii) development of new and improved microbially catalyzed processes resulting in immobilization of metals and radionuclides in the waste with concomitant waste volume reduction.

  6. MICROBIAL TRANSFORMATIONS OF TRU AND MIXED WASTES: ACTINIDE SPECIATION AND WASTE VOLUME REDUCTION

    Energy Technology Data Exchange (ETDEWEB)

    Francis, A.J.; Dodge, C.J.

    2006-06-01

    The overall goals of this research project are to determine the mechanism of microbial dissolution and stabilization of actinides in Department of Energy's (DOE) TRU wastes, contaminated sludges, soils, and sediments. This includes (1) investigations on the fundamental aspects of microbially catalyzed radionuclide and metal transformations (oxidation/reduction reactions, dissolution, precipitation, chelation); (2) understanding of the microbiological processes that control speciation and alter the chemical forms of complex inorganic/organic contaminant mixtures; and (3) development of new and improved microbially catalyzed processes resulting in immobilization of metals and radionuclides in the waste with concomitant waste volume reduction.

  7. Development of the chemical stabilization and solidification process for the treatment of radioactive raffinate sludges at the DOE Weldon Spring Site Remedial Action Project

    International Nuclear Information System (INIS)

    Cole, P.M.; Kakaria, V.; Enger, J.

    1996-01-01

    Chemical Solidification and Stabilization (CSS) is the mixing of chemical reagents with waste to solidify and chemically stabilize the contaminated material. The resulting product is resistant to leaching of certain contaminants. CSS treatment using Class C fly ash and Portland cement was chosen as the most feasible method for treatment of the chemically and radioactively contaminated sludge (raffinate) contained in raffinate pits on the Weldon Spring Site Remedial Action Project (WSSRAP) located outside of St. Louis, Missouri. Due to the uniqueness of the material, substantial bench-scale testing was performed on the raffinate to better understand its properties. Similarly, due to mixed results in the application of CSS treatment to radioactive materials, a pilot-scale testing facility was built to verify bench testing results and to establish and quantify design parameters for the full-scale CSS production facility. This paper discusses the development of the pilot-scale testing facility, the testing plan, and the results of the testing activities. Particular attention has been given to the applicability of the CSS treatment method and to the value of pilot-scale testing

  8. Microstructural investigation of D2 tool steel during rapid solidification

    Science.gov (United States)

    Delshad Khatibi, Pooya

    Solidification is considered as a key processing step in developing the microstructure of most metallic materials. It is, therefore, important that the solidification process can be designed and controlled in such a way so as to obtain the desirable properties in the final product. Rapid solidification refers to the system's high undercooling and high cooling rate, which can yield a microstructure with unique chemical composition and mechanical properties. An area of interest in rapid solidification application is high-chromium, high-carbon tool steels which experience considerable segregation of alloying elements during their solidification in a casting process. In this dissertation, the effect of rapid solidification (undercooling and cooling rate) of D2 tool steel on the microstructure and carbide precipitation during annealing was explored. A methodology is described to estimate the eutectic and primary phase undercooling of solidifying droplets. The estimate of primary phase undercooling was confirmed using an online measurement device that measured the radiation energy of the droplets. The results showed that with increasing primary phase and eutectic undercooling and higher cooling rate, the amount of supersaturation of alloying element in metastable retained austenite phase also increases. In the case of powders, the optimum hardness after heat treatment is achieved at different temperatures for constant periods of time. Higher supersaturation of austenite results in obtaining secondary hardness at higher annealing temperature. D2 steel ingots generated using spray deposition have high eutectic undercooling and, as a result, high supersaturation of alloying elements. This can yield near net shape D2 tool steel components with good mechanical properties (specifically hardness). The data developed in this work would assist in better understanding and development of near net shape D2 steel spray deposit products with good mechanical properties.

  9. Matemathical description of solidification cooling curves of pure metals

    Directory of Open Access Journals (Sweden)

    Arno Müller

    1998-10-01

    Full Text Available The introduction of an "incubation time" to the Schwarz classical mathematical description of metals solidification, resulted in a new model called Modified Schwarz Model. By doing so it was possible to identify and quantify the "delay time" that separates the two heat waves traveling independently in a casting during the solidification: the Supercooled / Superheated Liquid and the Solid / Liquid. The thermal shock produced in the initial stage of the undercooling generation process, can be used as an important parameter in the forecasting of the solidification's behavior of pure metals and alloys, when changing mold's materials, pouring and ambient temperatures. The hypercooling proneness degree of metals and alloys, can also be calculated.

  10. Systematic evaluation of options to avoid generation of noncertifiable transuranic (TRU) waste at Los Alamos National Laboratory

    International Nuclear Information System (INIS)

    Boak, J.M.; Kosiewicz, S.T.; Triay, I.; Gruetzmacher, K.; Montoya, A.

    1998-03-01

    At present, >35% of the volume of newly generated transuranic (TRU) waste at Los Alamos National Laboratory is not certifiable for transport to the Waste Isolation Pilot Plant (WIPP). Noncertifiable waste would constitute 900--1,000 m 3 of the 2,600 m 3 of waste projected during the period of the Environmental Management (EM) Accelerated Cleanup: Focus on 2006 plan (DOE, 1997). Volume expansion of this waste to meet thermal limits would increase the shipped volume to ∼5,400 m 3 . This paper presents the results of efforts to define which TRU waste streams are noncertifiable at Los Alamos, and to prioritize site-specific options to reduce the volume of certifiable waste over the period of the EM Accelerated Cleanup Plan. A team of Los Alamos TRU waste generators and waste managers reviewed historic generation rates and thermal loads and current practices to estimate the projected volume and thermal load of TRU waste streams for Fiscal Years 1999--2006. These data defined four major problem TRU waste streams. Estimates were also made of the volume expansion that would be required to meet the permissible wattages for all waste. The four waste streams defined were: (1) 238 Pu-contaminated combustible waste from production of Radioactive Thermoelectric Generators (RTGs) with 238 Pu activity which exceeds allowable shipping limits by 10--100X. (2) 241 Am-contaminated cement waste from plutonium recovery processes (nitric and hydrochloric acid recovery) are estimated to exceed thermal limits by ∼3X. (3) 239 Pu-contaminated combustible waste, mainly organic waste materials contaminated with 239 Pu and 241 Am, is estimated to exceed thermal load requirements by a factor of ∼2X. (4) Oversized metal waste objects, (especially gloveboxes), cannot be shipped as is to WIPP because they will not fit in a standard waste box or drum

  11. Progress report on disposal concept for TRU waste in Japan

    International Nuclear Information System (INIS)

    2000-03-01

    The object of this report is to contribute towards establishing a national TRU waste disposal program by integrating the results of research and development work carried out by JNC and the electricity utilities and summarizing the findings concerning safe methods for TRU waste disposal. The report consists of 5 chapters: the first describes the boundary conditions for the review of the TRU waste disposal concept (including geological conditions) and the basic concept adopted; the second describes the generation and characteristics of TRU waste and the third outlines the disposal technology; the fourth gives the key of the safety assessment and the fifth presents the conclusions of the report and lists issues for future consideration. The geological environment of Japan is simply classified into crystalline and sedimentary rock types (in terms of groundwater flow properties and rock strength) and a set of target conditions/properties for each rock type is then established. Based on this, a case which represents the basis for performance assessment (the reference case) will be defined. Alternatives to the reference case are studied to investigate the flexibility of the disposal concept. Under the conditions assumed in this study, the perturbing events considered showed no significant effects on the dose at the 100 meter evaluation point, owing to the relatively high efficiency of the natural barrier. However, the significant effect of these events on nuclide from the EBS shows that, in the case of a less efficient natural barrier, their effects could influence resulting dose. (S.Y.)

  12. Solidification of low-level wastes by inorganic binder

    International Nuclear Information System (INIS)

    Sasaki, M.T.; Shimojo, M.; Suzuki, K.; Kajikawa, A.; Karasawa, Y.

    1995-01-01

    The use of an alkali activated slag binder has been studied for solidification and stabilization of low-level wastes in nuclear power stations and spent fuel processing facilities. The activated slag effectively formed waste products having good physical properties with high waste loading for sodium sulfate, sodium nitrate, calcium pyrophosphate/phosphate and spent ion-exchange resins. Moreover, the results of the study suggest the slag has the ability to become a common inorganic binder for the solidification of various radioactive wastes. This paper also describes the fixation of radionuclides by the activated slag binder

  13. Some techniques for the solidification of radioactive wastes in concrete

    International Nuclear Information System (INIS)

    Colombo, P.; Neilson, R. Jr.

    1976-06-01

    Some techniques for the solidification of radioactive wastes in concrete are discussed. The sources, storage, volume reduction, and solidification of liquid wastes at Brookhaven National Laboratory (BNL) using the cement-vermiculite process is described. Solid waste treatment, shipping containers, and off-site shipments of solid wastes at BNL are also considered. The properties of low-heat-generating, high-level wastes, simulating those in storage at the Savannah River Plant (SRP), solidified in concrete were determined. Polymer impregnation was found to further decrease the leachability and improve the durability of these concrete waste forms

  14. Modified sulfur cement solidification of low-level wastes

    Energy Technology Data Exchange (ETDEWEB)

    1985-10-01

    This topical report describes the results of an investigation on the solidification of low-level radioactive wastes in modified sulfur cement. The work was performed as part of the Waste Form Evaluation Program, sponsored by the US Department of Energy's Low-Level Waste Management Program. Modified sulfur cement is a thermoplastic material developed by the US Bureau of Mines. Processing of waste and binder was accomplished by means of both a single-screw extruder and a dual-action mixing vessel. Waste types selected for this study included those resulting from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste type and method of processing. Property evaluation testing was carried out on laboratory scale specimens in order to compare with waste form performance for other potential matrix materials. Waste form property testing included compressive strength, water immersion, thermal cycling and radionuclide leachability. Recommended waste loadings of 40 wt. % sodium sulfate and boric acid salts and 43 wt. % incinerator ash, which are based on processing and performance considerations, are reported. Solidification efficiencies for these waste types represent significant improvements over those of hydraulic cements. Due to poor waste form performance, incorporation of ion exchange resin waste in modified sulfur cement is not recommended.

  15. Microstructural Development in Al-Si Powder During Rapid Solidification

    Energy Technology Data Exchange (ETDEWEB)

    Genau, Amber Lynn [Iowa State Univ., Ames, IA (United States)

    2004-01-01

    Powder metallurgy has become an increasingly important form of metal processing because of its ability to produce materials with superior mechanical properties. These properties are due in part to the unique and often desirable microstructures which arise as a result of the extreme levels of undercooling achieved, especially in the finest size powder, and the subsequent rapid solidification which occurs. A better understanding of the fundamental processes of nucleation and growth is required to further exploit the potential of rapid solidification processing. Aluminum-silicon, an alloy of significant industrial importance, was chosen as a model for simple eutectic systems displaying an unfaceted/faceted interface and skewed coupled eutectic growth zone, Al-Si powder produced by high pressure gas atomization was studied to determine the relationship between microstructure and alloy composition as a function of powder size and atomization gas. Critical experimental measurements of hypereutectic (Si-rich) compositions were used to determine undercooling and interface velocity, based on the theoretical models which are available. Solidification conditions were analyzed as a function of particle diameter and distance from nucleation site. A revised microstructural map is proposed which allows the prediction of particle morphology based on temperature and composition. It is hoped that this work, by providing enhanced understanding of the processes which govern the development of the solidification morphology of gas atomized powder, will eventually allow for better control of processing conditions so that particle microstructures can be optimized for specific applications.

  16. Modified sulfur cement solidification of low-level wastes

    International Nuclear Information System (INIS)

    1985-10-01

    This topical report describes the results of an investigation on the solidification of low-level radioactive wastes in modified sulfur cement. The work was performed as part of the Waste Form Evaluation Program, sponsored by the US Department of Energy's Low-Level Waste Management Program. Modified sulfur cement is a thermoplastic material developed by the US Bureau of Mines. Processing of waste and binder was accomplished by means of both a single-screw extruder and a dual-action mixing vessel. Waste types selected for this study included those resulting from advanced volume reduction technologies (dry evaporator concentrate salts and incinerator ash) and those which remain problematic for solidification using contemporary agents (ion exchange resins). Process development studies were conducted to ascertain optimal process control parameters for successful solidification. Maximum waste loadings were determined for each waste type and method of processing. Property evaluation testing was carried out on laboratory scale specimens in order to compare with waste form performance for other potential matrix materials. Waste form property testing included compressive strength, water immersion, thermal cycling and radionuclide leachability. Recommended waste loadings of 40 wt. % sodium sulfate and boric acid salts and 43 wt. % incinerator ash, which are based on processing and performance considerations, are reported. Solidification efficiencies for these waste types represent significant improvements over those of hydraulic cements. Due to poor waste form performance, incorporation of ion exchange resin waste in modified sulfur cement is not recommended

  17. Optimization of TRU burnup in modular helium reactor

    International Nuclear Information System (INIS)

    Yonghee, Kim; Venneri, F.

    2007-01-01

    An optimization study of a single-pass TRU (transuranic) deep-burn (DB) has been performed for a block-type MHR (Modular Helium Reactor) proposed by General Atomics. Assuming a future equilibrium scenario of advanced LWRs, a high-burnup TRU vector is considered: 50 GWD/MTU and 5-year cooling. For 3-D equilibrium cores, the performance analysis is done by using a continuous energy Monte Carlo depletion code MCCARD. The core optimization is performed from the viewpoints of the core configuration, fuel management, TRISO fuel specification, and neutron spectrum. With regard to core configuration, two annular cores are investigated in terms of the neutron economy. A conventional radial shuffling scheme of fuel blocks is compared with an axial block shuffling strategy in terms of the fuel burnup and core power distributions. The impact of the kernel size of TRISO fuel is evaluated and a diluted kernel, instead of a conventional concentrated kernel, is introduced to maximize the TRU burnup by reducing the self-shielding effects of TRISO fuels. A higher graphite density is evaluated in terms of the fuel burnup. In addition, it is shown that the core power distribution can be effectively controlled by zoning of the packing fraction of TRISO fuels. We also have shown that a long-cycle DB-MHR core can be designed by using a small batch size for fuel reloading, at the expense of a marginal decrease of the TRU discharge burnup. Depending on the fuel management scheme, fuel specifications, and core parameters, the TRU burnup in an optimized DB-MHR core is over 60% in a single-pass irradiation campaign. (authors)

  18. Los Alamos National Laboratory accelerated tru waste workoff strategies

    International Nuclear Information System (INIS)

    Kosiewicz, S.T.; Triay, I.R.; Rogers, P.Z.; Christensen, D.V.

    1997-01-01

    During 1996, the Los Alamos National Laboratory (LANL) developed two transuranic (TRU) waste workoff strategies that were estimated to save $270 - 340M through accelerated waste workoff and the elimination of a facility. The planning effort included a strategy to assure that LANL would have a significant quantity (3000+ drums) of TRU waste certified for shipment to the Waste Isolation Pilot Plant (WIPP) beginning in April of 1998, when WIPP was projected to open. One of the accelerated strategies can be completed in less than ten years through a Total Optimization of Parameters Scenario (open-quotes TOPSclose quotes). open-quotes TOPSclose quotes fully utilizes existing LANL facilities and capabilities. For this scenario, funding was estimated to be unconstrained at $23M annually to certify and ship the legacy inventory of TRU waste at LANL. With open-quotes TOPSclose quotes the inventory is worked off in about 8.5 years while shipping 5,000 drums per year at a total cost of $196M. This workoff includes retrieval from earthen cover and interim storage costs. The other scenario envisioned funding at the current level with some increase for TRUPACT II loading costs, which total $16M annually. At this funding level, LANL estimates it will require about 17 years to work off the LANL TRU legacy waste while shipping 2,500 drums per year to WIPP. The total cost will be $277M. This latter scenario decreases the time for workoff by about 19 years from previous estimates and saves an estimated $190M. In addition, the planning showed that a $70M facility for TRU waste characterization was not needed. After the first draft of the LANL strategies was written, Congress amended the WIPP Land Withdrawal Act (LWA) to accelerate the opening of WIPP to November 1997. Further, the No Migration Variance requirement for the WIPP was removed. This paper discusses the LANL strategies as they were originally developed. 1 ref., 3 figs., 2 tabs

  19. Development and application of new parameters for TRU transmutation effectiveness

    International Nuclear Information System (INIS)

    Han, Chi Young

    2005-02-01

    Four new parameters (incineration branching ratio, incineration rate, incineration time, and incineration buckling) have been developed to evaluate quantitatively the TRU transmutation effectiveness and applied to transmutation of uranium and TRU. From the incineration branching ratio, it is possible to analyze the main contributors to fission reaction for transmutation of a target nuclide. From the incineration rate, it is available to evaluate the transmutation effectiveness in the viewpoint of a relative incineration rate to incineration potential of a target nuclide and its family. This parameter is also used to calculate the incineration time and incineration buckling together with the incineration branching ratio. The incineration time makes it possible to discuss more practically the transmutation speed instead of the existing other parameters. The incineration buckling can be used to evaluate the time behavior of the incineration rate and also employed to support the results from the incineration time. Taking into account the transmutation effectiveness and potential of uranium and TRU derived by using the parameters and an existing neutron economy parameter, it was noted that the thermal neutron energy is very preferable from the transmutation effectiveness point of view, on the other hand the fast neutron energy is effective from the transmutation potential. Applying them to the typical critical and subcritical TRU burners, it is indicated that the critical reactor containing fertile uranium undergoes effectively the selective TRU transmutation on the present fast spectrum. It was also noted that the uranium-free subcritical reactor could be operated effectively on a little softer spectrum due to the larger neutron excess in the present spectrum. It is expected that the new parameters developed in this study and the results are directly applicable to practical transmutation reactor design, in particular accelerator-driven transmutation reactor

  20. The melting and solidification of nanowires

    International Nuclear Information System (INIS)

    Florio, B. J.; Myers, T. G.

    2016-01-01

    A mathematical model is developed to describe the melting of nanowires. The first section of the paper deals with a standard theoretical situation, where the wire melts due to a fixed boundary temperature. This analysis allows us to compare with existing results for the phase change of nanospheres. The equivalent solidification problem is also examined. This shows that solidification is a faster process than melting; this is because the energy transfer occurs primarily through the solid rather than the liquid which is a poorer conductor of heat. This effect competes with the energy required to create new solid surface which acts to slow down the process, but overall conduction dominates. In the second section, we consider a more physically realistic boundary condition, where the phase change occurs due to a heat flux from surrounding material. This removes the singularity in initial melt velocity predicted in previous models of nanoparticle melting. It is shown that even with the highest possible flux the melting time is significantly slower than with a fixed boundary temperature condition.

  1. The melting and solidification of nanowires

    Science.gov (United States)

    Florio, B. J.; Myers, T. G.

    2016-06-01

    A mathematical model is developed to describe the melting of nanowires. The first section of the paper deals with a standard theoretical situation, where the wire melts due to a fixed boundary temperature. This analysis allows us to compare with existing results for the phase change of nanospheres. The equivalent solidification problem is also examined. This shows that solidification is a faster process than melting; this is because the energy transfer occurs primarily through the solid rather than the liquid which is a poorer conductor of heat. This effect competes with the energy required to create new solid surface which acts to slow down the process, but overall conduction dominates. In the second section, we consider a more physically realistic boundary condition, where the phase change occurs due to a heat flux from surrounding material. This removes the singularity in initial melt velocity predicted in previous models of nanoparticle melting. It is shown that even with the highest possible flux the melting time is significantly slower than with a fixed boundary temperature condition.

  2. The melting and solidification of nanowires

    Energy Technology Data Exchange (ETDEWEB)

    Florio, B. J., E-mail: brendan.florio@ul.ie [University of Limerick, Mathematics Applications Consortium for Science and Industry (MACSI), Department of Mathematics and Statistics (Ireland); Myers, T. G., E-mail: tmyers@crm.cat [Centre de Recerca Matemàtica (Spain)

    2016-06-15

    A mathematical model is developed to describe the melting of nanowires. The first section of the paper deals with a standard theoretical situation, where the wire melts due to a fixed boundary temperature. This analysis allows us to compare with existing results for the phase change of nanospheres. The equivalent solidification problem is also examined. This shows that solidification is a faster process than melting; this is because the energy transfer occurs primarily through the solid rather than the liquid which is a poorer conductor of heat. This effect competes with the energy required to create new solid surface which acts to slow down the process, but overall conduction dominates. In the second section, we consider a more physically realistic boundary condition, where the phase change occurs due to a heat flux from surrounding material. This removes the singularity in initial melt velocity predicted in previous models of nanoparticle melting. It is shown that even with the highest possible flux the melting time is significantly slower than with a fixed boundary temperature condition.

  3. Solidification analysis of a centrifugal atomizer using the Al-32.7wt.% Cu alloy

    Energy Technology Data Exchange (ETDEWEB)

    Osborne, Matthew G. [Iowa State Univ., Ames, IA (United States)

    1998-02-23

    A centrifugal atomizer (spinning disk variety) was designed and constructed for the production of spherical metal powders, 100-1,000 microns in diameter in an inert atmosphere. Initial atomization experiments revealed the need for a better understanding of how the liquid metal was atomized and how the liquid droplets solidified. To investigate particle atomization, Ag was atomized in air and the process recorded on high-speed film. To investigate particle solidification, Al-32.7 wt.% Cu was atomized under inert atmosphere and the subsequent particles were examined microscopically to determine solidification structure and rate. This dissertation details the experimental procedures used in producing the Al-Cu eutectic alloy particles, examination of the particle microstructures, and determination of the solidification characteristics (e.g., solidification rate) of various phases. Finally, correlations are proposed between the operation of the centrifugal atomizer and the observed solidification spacings.

  4. Effects Disposal Condition and Ground Water to Leaching Rate of Radionuclides from Solidification Products

    International Nuclear Information System (INIS)

    Herlan Martono; Wati

    2008-01-01

    Effects disposal condition and ground water to leaching rate of radionuclides from solidification products have been studied. The aims of leaching test at laboratory to get the best composition of solidified products for continuous process or handling. The leaching rate of radionuclides from the many kinds of matrix from smallest to bigger are glass, thermosetting plastic, urea formaldehyde, asphalt, and cement. Glass for solidification of high level waste, thermosetting plastic and urea formaldehyde for solidification of low and intermediate waste, asphalt and cement for solidification of low and intermediate level waste. In shallow land burial, ground water rate is fast, debit is high, and high permeability, so the probability contact between solidification products and ground water is occur. The pH of ground water increasing leaching rate, but cation in the ground water retard leaching rate. Effects temperature radiation and radiolysis to solidification products is not occur. In the deep repository, ground water rate is slow, debit is small, and low permeability, so the probability contact between solidification products and ground water is very small. There are effect cooling time and distance between pits to rock temperature. Alfa radiation effects can be occur, but there is no contact between solidification products and ground water, so that there is not radiolysis. (author)

  5. Optimisation of the microstructure of YBa{sub 2}Cu{sub 3}O{sub 7-{delta}} superconducting ceramics textured by horizontal solidification. Effects of a magnetic field applied vertically during the solidification process; Optimisation de la microstructure de ceramiques supraconductrices texturees YBa{sub 2}Cu{sub 3}O{sub 7-{delta}} par solidification horizontale. Effets de l`application d`un champ magnetique vertical au cours de la croissance

    Energy Technology Data Exchange (ETDEWEB)

    Durand, L

    1995-12-14

    We have shown that il was possible to disperse in a very homogeneous way, into the 123 matrix, 211 particles regular in size around 1{mu}m, by directional solidification of a mixture of 123 plus 211 sol-gel powders. The addition of 0.5 wt% of Pt lo this precursor mixture has permitted us to refine further the size of the 211 particles. Moreover, we have established that the finer the 123 powder and the faster the heating rate, the smaller the 211 particles. From another hand, we have found that the size and the microstructure of the solidified single domain was dependant upon numerous parameters. In particular, we have noted that if the cold pressed green body was not sintered before being melted, grains of solid CuO appeared into the liquid which next hardly recombined with it, giving rise to a comb like solidification front, witness of a local un-stability. Moreover, in these particular conditions, the 123 formed was found to be understoichiometric in Cu and the single domains severely limited in extension. Finally, we have observed that the application of a 4 T magnetic field during the solidification of a bar the longest axis of which was horizontal tended to align the c axis of the 123 grains parallel to the field. Incidentally, we have discovered that in the same conditions, the 211 particles trapped into the 123 matrix were also preferentially oriented. (author).

  6. Solidification microstructure of centrifugally cast Inconel 625

    Directory of Open Access Journals (Sweden)

    Silvia Barella

    2017-07-01

    Full Text Available Centrifugal casting is a foundry process allowing the production of near net-shaped axially symmetrical components. The present study focuses on the microstructural characterization of centrifugally cast alloys featuring different chemical compositions for the construction of spheres applied in valves made of alloy IN625 for operation at high pressure. Control of the solidification microstructure is needed to assure the reliability of the castings. Actually, a Ni-base superalloy such as this one should have an outstanding combination of mechanical properties, high temperature stability and corrosion resistance. Alloys such as IN625 are characterised by a large amount of alloying elements and a wide solidification range, so they can be affected by micro-porosity defects, related to the shrinkage difference between the matrix and the secondary reinforcing phases (Nb-rich carbides and Laves phase. In this study, the microstructure characterization was performed as a function of the applied heat treatments and it was coupled with a calorimetric analysis in order to understand the mechanism ruling the formation of micro-porosities that can assure alloy soundness. The obtained results show that the presence of micro-porosities is governed by morphology and by the size of the secondary phases, and the presence of the observed secondary phases is detrimental to corrosion resistance.

  7. Comparative assessment of disposal of TRU waste in a greater-confinement disposal facility

    International Nuclear Information System (INIS)

    Cohn, J.J.; Smith, C.F.; Ciminesi, F.J.; Dickman, P.T.; O'Neal, D.A.

    1982-11-01

    This study reviewed previous work that established generic limits for shallow land burial of TRU contaminated wastes and extended previous methodology to estimate approximate appropriate burial limits for TRU wastes in an arid zone greater confinement disposal facility (GCDF). An erosion scenario provided the limiting pathway in the previous determination of generic shallow land burial limits. Erosion removed the cover soil, exposing the waste mass to habitation and agriculture. For the deep burial concept (that is, burial at a depth greater than 10 m [33 ft]), the aquifer transport scenario was controlling. In both cases, the assumed site conditions were characteristic of a humid zone in which groundwater flows immediately below the waste deposit. In deriving limits for an arid site GCDF, either the erosion/reclaimer or the aquifer transport scenario could provide the controlling pathway, depending on the nuclide and the assumed burial depth. The derived limits were higher for the arid sited GCDF than those of the generic humid study. The physical processes that increase limits relative to the generic study include increased time during which radioactive decay occurs prior to release and increased dilution. Some nuclides were effectively unlimited in an arid zone GCDF, while others (notably Pu-239) were affected on a much smaller scale, primarily due to very long half-lives. As a final comment, the limit values derived in this report represent adjustments to the calculations of the Healy and Rodgers report (LA-UR-79-100). Those original calculations were very conservative, utilizing a worst case approach, but nevertheless involving significant levels of uncertainty in key assumptions. Consequently, the results are assumption dependent. Other approaches to such an analysis could, and should be used to develop site specific concentration limits for TRU wastes

  8. Solidification Sequence of Spray-Formed Steels

    Science.gov (United States)

    Zepon, Guilherme; Ellendt, Nils; Uhlenwinkel, Volker; Bolfarini, Claudemiro

    2016-02-01

    Solidification in spray-forming is still an open discussion in the atomization and deposition area. This paper proposes a solidification model based on the equilibrium solidification path of alloys. The main assumptions of the model are that the deposition zone temperature must be above the alloy's solidus temperature and that the equilibrium liquid fraction at this temperature is reached, which involves partial remelting and/or redissolution of completely solidified droplets. When the deposition zone is cooled, solidification of the remaining liquid takes place under near equilibrium conditions. Scanning electron microscopy (SEM) and optical microscopy (OM) were used to analyze the microstructures of two different spray-formed steel grades: (1) boron modified supermartensitic stainless steel (SMSS) and (2) D2 tool steel. The microstructures were analyzed to determine the sequence of phase formation during solidification. In both cases, the solidification model proposed was validated.

  9. Study on integrated TRU multi-recycling in sodium cooled fast reactor CDFR

    International Nuclear Information System (INIS)

    Hu Yun; Xu Mi; Wang Kan

    2010-01-01

    In view of recently proposed closed fuel cycle strategy which would recycle the integrated transuranics (TRU) from PWR spent fuel in the fast reactors, the neutronics characteristics of TRU recycled in China Demonstration Fast Reactor (CDFR) are studied in this paper. The results show that loading integrated TRU to substitute pure Pu as driver fuel will mainly make the influence on sodium void worth and negligible effects on other parameters, and hence TRU recycling in CDFR is feasible from viewpoint of core neutronics. If TRU is multi-recycled, the variation of TRU composition depends on fuel types and the ratio of TRU and U when recycling. It is indicated that, when TRU is multi-recycled in CDFR with MOX fuel, the minor actinides (MA) fraction in TRU will firstly decrease to ∼7.24% (minimum) within 8 TRU recycle times and then slowly increase to ∼7.7% after 20 TRU recycle times; while when TRU is multi-recycled in CDFR with metal fuel (TRU-U-10Zr), the MA fraction in TRU will gradually approach to an equilibrium state with the MA fraction of ∼3.8%, demonstrating better MA transmutation effect in metal fuel core. No matter 7.7 or 3.8%, they are both lower than ∼10% in PWR spent fuel with burnup of 45 GWd/tU, which presents satisfying effect of MA amount controlling for TRU multi-recycling strategy. On the other hand, the corresponding recycling parameters such as TRU heat release and neutron emission rate are also much lower in metal fuel than those in MOX fuel. Moreover, TRU recycled in metal fuel will bring greater fissile Pu isotopes equilibrium fraction due to better breeding capability of metal fuel. Finally, it could be summarized that integrated TRU multi-recycling in fast reactor can make contributions to both breeding and transmutation, and such strategy is a prospective closed fuel cycle manner to achieve the object of effective control of cumulated MA amount and sustainable development of nuclear energy.

  10. Origin of grain orientation during solidification of an aluminum alloy

    International Nuclear Information System (INIS)

    Wei, H.L.; Elmer, J.W.; DebRoy, T.

    2016-01-01

    The evolution of grain morphology during solidification of a moving aluminum alloy pool is simulated by considering heat transfer, flow of liquid metal in the molten pool and solidification parameters. The computationally efficient model consists of a 3D coupled heat transfer and fluid flow simulation to predict the molten pool shape and temperature field, and a 2D model of grain formation in the molten pool. The results demonstrate that columnar grains grow in a curved pattern rather than along straight lines from the fusion boundary towards the center of the molten pool. The calculated results are validated with independent experimental data. The computed ratio of local temperature gradient to solidification rate, G/R, is used to model the columnar to equiaxed transition during solidification. The simulated results show that only curved columnar grains are formed when the scanning speed is low (2.0 mm/s). In contrast, a transition from curved columnar to equiaxed morphologies occurs at the higher scanning speeds of 8.0 mm/s and 11.5 mm/s, with higher equiaxed grain fraction at higher speed. The similarities between the physical processes governing fusion welding and additive manufacturing (AM) make the model capable of predicting grain orientation in both processes.

  11. Gamma-ray spectrometry combined with acceptable knowledge (GSAK). A technique for characterization of certain remote-handled transuranic (RH-TRU) wastes. Part 1. Methodology and techniques

    International Nuclear Information System (INIS)

    Hartwell, J.K.; McIlwain, M.E.

    2005-01-01

    Gamma-ray spectrometry combined with acceptable knowledge (GSAK) is a technique for the characterization of certain remote-handled transuranic (RH-TRU) wastes. GSAK uses gamma-ray spectrometry to quantify a portion of the fission product inventory of RH-TRU wastes. These fission product results are then coupled with calculated inventories derived from acceptable process knowledge to characterize the radionuclide content of the assayed wastes. GSAK has been evaluated and tested through several test exercises. GSAK approach is described, while test results are presented in Part II. (author)

  12. Gamma-ray spectrometry combined with acceptable knowledge (GSAK). A technique for characterization of certain remote-handled transuranic (RH-TRU) wastes. Part 2. Testing and results

    International Nuclear Information System (INIS)

    Hartwell, J.K.; McIlwain, M.E.

    2005-01-01

    Gamma-ray spectrometry combined with acceptable knowledge (GSAK) is a technique for the characterization of certain remote-handled transuranic (RH-TRU) wastes. GSAK uses gamma-ray spectrometry to quantify a portion of the fission product inventory of RH-TRU wastes. These fission product results are then coupled with calculated inventories derived from acceptable process knowledge to characterize the radionuclide content of the assayed wastes. GSAK has been evaluated and tested through several test exercises. These tests and their results are described; while the former paper in this issue presents the methodology, equipment and techniques. (author)

  13. Characterizing cemented TRU waste for RCRA hazardous constituents

    International Nuclear Information System (INIS)

    Yeamans, D.R.; Betts, S.E.; Bodenstein, S.A.

    1996-01-01

    Los Alamos National Laboratory (LANL) has characterized drums of solidified transuranic (TRU) waste from four major waste streams. The data will help the State of New Mexico determine whether or not to issue a no-migration variance of the Waste Isolation Pilot Plant (WIPP) so that WIPP can receive and dispose of waste. The need to characterize TRU waste stored at LANL is driven by two additional factors: (1) the LANL RCRA Waste Analysis Plan for EPA compliant safe storage of hazardous waste; (2) the WIPP Waste Acceptance Criteria (WAC) The LANL characterization program includes headspace gas analysis, radioassay and radiography for all drums and solids sampling on a random selection of drums from each waste stream. Data are presented showing that the only identified non-metal RCRA hazardous component of the waste is methanol

  14. Research on safety evaluation for TRU waste disposal

    International Nuclear Information System (INIS)

    Senoo, M.; Shirahashi, K.; Sakamoto, Y.; Moriyama, N.; Konishi, M.

    1989-01-01

    Studies on adsorption behavior of transuranic (TRU) elements have been performed from the view point of validating the data for safety assessment and investigating adsorption behavior of TRU elements. Distribution coefficient (Kd value) of plutonium between groundwater and soils sampled at the planning site of low level waste disposal facility were measured for safety assessment. Kd values measured were the order of 10 3 ml/g. For investigating adsorption behavior, pH dependency of Kd value of neptunium and Am for soils were studied. It was concluded that pH dependency of Kd value of neptunium was mainly owing to amount of surface charge of soils, on the other hand that of Am was due to chemical form of Am. Influence of carbonation of cement for adsorption behavior of neptunium and plutonium was also investigated and it was concluded that Kd value of carbonated cement was lower than that of fresh cement

  15. Assessment of Hanford burial grounds and interim TRU storage

    International Nuclear Information System (INIS)

    Geiger, J.F.; Brown, D.J.; Isaacson, R.E.

    1977-08-01

    A review and assessment is made of the Hanford low level solid radioactive waste management sites and facilities. Site factors considered favorable for waste storage and disposal are (1) limited precipitation, (2) a high deficiency of moisture in the underlying sediments (3) great depth to water table, all of which minimize radionuclide migration by water transport, and (4) high sorbtive capacity of the sediments. Facilities are in place for 20 year retrievable storage of transuranic (TRU) wastes and for disposal of nontransuranic radioactive wastes. Auxiliary facilities and services (utilities, roads, fire protection, shops, etc.) are considered adequate. Support staffs such as engineering, radiation monitoring, personnel services, etc., are available and are shared with other operational programs. The site and associated facilities are considered well suited for solid radioactive waste storage operations. However, recommendations are made for study programs to improve containment, waste package storage life, land use economy, retrievability and security of TRU wastes

  16. Savannah River Site Operating Experience with Transuranic (TRU) Waste Retrieval

    International Nuclear Information System (INIS)

    Stone, K.A.; Milner, T.N.

    2006-01-01

    Drums of TRU Waste have been stored at the Savannah River Site (SRS) on concrete pads from the 1970's through the 1980's. These drums were subsequently covered with tarpaulins and then mounded over with dirt. Between 1996 and 2000 SRS ran a successful retrieval campaign and removed some 8,800 drums, which were then available for venting and characterization for WIPP disposal. Additionally, a number of TRU Waste drums, which were higher in activity, were stored in concrete culverts, as required by the Safety Analysis for the Facility. Retrieval of drums from these culverts has been ongoing since 2002. This paper will describe the operating experience and lessons learned from the SRS retrieval activities. (authors)

  17. The new Japanese policy for TRU-waste management

    International Nuclear Information System (INIS)

    Yamamoto, M.

    1992-01-01

    In July 1991, the Advisory Committee on Radioactive Waste of the Japan Atomic Energy Commission announced its report on a new Japanese policy for TRU-waste management. The total volume of radioactive wastes which contain TRU nuclides has reached the equivalent of about 40,000,200-liter drums, and is expected to grow to about 300,000 drums by the year 2010. Further development is required to reduce the volume of the existing waste and to decrease the amount of waste being generated. Wastes with concentration levels exceeding a threshold limit of 1 Giga-Becquerel per ton will be disposed in an underground facility. Those wastes with lower activities will be sent to a shallow-land burial facility. The goal of research and development is the completion of the disposal system by the late 1990's. (author)

  18. Parametric Criticality Safety Calculations for Arrays of TRU Waste Containers

    Energy Technology Data Exchange (ETDEWEB)

    Gough, Sean T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-10-26

    The Nuclear Criticality Safety Division (NCSD) has performed criticality safety calculations for finite and infinite arrays of transuranic (TRU) waste containers. The results of these analyses may be applied in any technical area onsite (e.g., TA-54, TA-55, etc.), as long as the assumptions herein are met. These calculations are designed to update the existing reference calculations for waste arrays documented in Reference 1, in order to meet current guidance on calculational methodology.

  19. Fluid flow solidification simulation of molten alloys

    International Nuclear Information System (INIS)

    Kaschnitz, E.

    1997-01-01

    In an effort to minimize costs and to obtain optimum designs, computer simulation of shape casting processes is more and more used as a development tool. Accurate predictions are possible by means of three dimensional fluid flow and solidification modelling. The bases of the model are the transient laminar Navier-Stokes-equations for a Newtonian fluid including the tracking of the free surface. They are describing the melt flow pattern during the mold filling sequence. Simultaneously, the temperature development in the alloy and mold is calculated using Fourier's heat transfer equation. At OEGI, a commercial software package (MAGMAsoft) with a finite difference equation solver is used for improvement of casting processes. Different examples of industrial applications will be shown. (author)

  20. Recent Advances in Study of Solid-Liquid Interfaces and Solidification of Metals

    Directory of Open Access Journals (Sweden)

    Mohsen Asle Zaeem

    2018-02-01

    Full Text Available Solidification occurs in several material processing methods, such as in casting, welding, and laser additive manufacturing of metals, and it controls the nano- and microstructures, as well as the overall properties of the products[...

  1. SILICATE TECHNOLOGY CORPORATION'S SOLIDIFICATION/ STABILIZATION TECHNOLOGY FOR ORGANIC AND INORGANIC CONTAMINANTS IN SOILS - APPLICATIONS ANALYSIS REPORT

    Science.gov (United States)

    This Applications Analysis Report evaluates the solidification/stabilization treatment process of Silicate Technology Corporation (STC) for the on-site treatment of hazardous waste. The STC immobilization technology utilizes a proprietary product (FMS Silicate) to chemically stab...

  2. Solidification at the High and Low Rate Extreme

    Energy Technology Data Exchange (ETDEWEB)

    Meco, Halim [Iowa State Univ., Ames, IA (United States)

    2004-12-19

    The microstructures formed upon solidification are strongly influenced by the imposed growth rates on an alloy system. Depending on the characteristics of the solidification process, a wide range of growth rates is accessible. The prevailing solidification mechanisms, and thus the final microstructure of the alloy, are governed by these imposed growth rates. At the high rate extreme, for instance, one can have access to novel microstructures that are unattainable at low growth rates. While the low growth rates can be utilized for the study of the intrinsic growth behavior of a certain phase growing from the melt. Although the length scales associated with certain processes, such as capillarity, and the diffusion of heat and solute, are different at low and high rate extremes, the phenomena that govern the selection of a certain microstructural length scale or a growth mode are the same. Consequently, one can analyze the solidification phenomena at both high and low rates by using the same governing principles. In this study, we examined the microstructural control at both low and high extremes. For the high rate extreme, the formation of crystalline products and factors that control the microstructure during rapid solidification by free-jet melt spinning are examined in Fe-Si-B system. Particular attention was given to the behavior of the melt pool at different quench-wheel speeds. Since the solidification process takes place within the melt-pool that forms on the rotating quench-wheel, we examined the influence of melt-pool dynamics on nucleation and growth of crystalline solidification products and glass formation. High-speed imaging of the melt-pool, analysis of ribbon microstructure, and measurement of ribbon geometry and surface character all indicate upper and lower limits for melt-spinning rates for which nucleation can be avoided, and fully amorphous ribbons can be achieved. Comparison of the relevant time scales reveals that surface-controlled melt

  3. Observed TRU data from nuclear utility waste streams

    International Nuclear Information System (INIS)

    Wessman, R.A.; Floyd, J.G.; Leventhal, L.

    1990-01-01

    TMA/Norcal has performed 10CFR61 analysis of radioactive waste streams from BWR's and PWR's since 1983. Many standard and non-routine sample types have been received for analysis from nuclear power plants nation-wide. In addition to the 10CFR61 Tables I and II analyses, we also have analyzed for many of the supplementary isotopes. As part of this program TRU analyses are required. As a result, have accumulated a significant amount of data for plutonium, americium, and curium in radioactive waste for many different sample matrices from many different waste streams. This paper will present our analytical program for 10CFR61 TRU. The laboratory methodology including chemical and radiometric procedures is discussed. The sensitivity of our measurements and ability to meet the lower limits of detection is also discussed. Secondly, a review of TRU data is presented. Scaling factors and their ranges from selected PWR stations are included. We discuss some features of, and limits to, interpretation of these data. 8 refs., 3 tabs

  4. Gas generation and migration analysis for TRU waste disposal system

    International Nuclear Information System (INIS)

    Ando, Kenichi; Noda, Masaru; Yamamoto, Mikihiko; Mihara, Morihiro

    2005-09-01

    In TRU waste disposal system, significant quantities of gases may be generated due to metal corrosion, radiolysis effect and microorganism activities. It is therefore recommended that the potential impact of gas generation and migration on TRU waste repository should be evaluated. In this study, gas generation rates were calculated in the repository and gas migration analysis in the disposal system were carried out using two phase flow model with results of gas generation rates. First, the time dependencies of gas generation rate in each TRU waste repositories were evaluated based on amounts of metal, organic matter and radioactivity. Next, the accumulation pressure of gases and expelled pore water volume nuclides in the repository were calculated by TOUGH2 code. After that, the results showed that the increase of gas pressure was the range of 1.3 to 1.4 MPa. In the repository with and without buffer, the rate of expelled pore water was 0.006 - 0.009 m 3 /y and 0.018 - 0.24m 3 /y, respectively. In addition, the radioactive gas migration through the repository and geosphere are evaluated. And re-saturation analysis is also performed to evaluate the initial condition of the system. (author)

  5. Test Plan: WIPP bin-scale CH TRU waste tests

    International Nuclear Information System (INIS)

    Molecke, M.A.

    1990-08-01

    This WIPP Bin-Scale CH TRU Waste Test program described herein will provide relevant composition and kinetic rate data on gas generation and consumption resulting from TRU waste degradation, as impacted by synergistic interactions due to multiple degradation modes, waste form preparation, long-term repository environmental effects, engineered barrier materials, and, possibly, engineered modifications to be developed. Similar data on waste-brine leachate compositions and potentially hazardous volatile organic compounds released by the wastes will also be provided. The quantitative data output from these tests and associated technical expertise are required by the WIPP Performance Assessment (PA) program studies, and for the scientific benefit of the overall WIPP project. This Test Plan describes the necessary scientific and technical aspects, justifications, and rational for successfully initiating and conducting the WIPP Bin-Scale CH TRU Waste Test program. This Test Plan is the controlling scientific design definition and overall requirements document for this WIPP in situ test, as defined by Sandia National Laboratories (SNL), scientific advisor to the US Department of Energy, WIPP Project Office (DOE/WPO). 55 refs., 16 figs., 19 tabs

  6. TRU waste certification and TRUPACT-2 payload verification

    International Nuclear Information System (INIS)

    Hunter, E.K.; Johnson, J.E.

    1990-01-01

    The Waste Isolation Pilot Plant (WIPP) established a policy that requires each waste shipper to verify that all waste shipments meet the requirements of the Waste Acceptance Criteria (WAC) prior to being shipped. This verification provides assurance that transuranic (TRU) wastes meet the criteria while still retained in a facility where discrepancies can be immediately corrected. Each Department of Energy (DOE) TRU waste facility planning to ship waste to the Waste Isolation Pilot Plant (WIPP) is required to develop and implement a specific program including Quality Assurance (QA) provisions to verify that waste is in full compliance with WIPP's WAC. This program is audited by a composite DOE and contractor audit team prior to granting the facility permission to certify waste. During interaction with the Nuclear Regulatory Commission (NRC) on payload verification for shipping in TRUPACT-II, a similar system was established by DOE. The TRUPACT-II Safety Analysis Report (SAR) contains the technical requirements and physical and chemical limits that payloads must meet (like the WAC). All shippers must plan and implement a payload control program including independent QA provisions. A similar composite audit team will conduct preshipment audits, frequent subsequent audits, and operations inspections to verify that all TRU waste shipments in TRUPACT-II meet the requirements of the Certificate of Compliance issued by the NRC which invokes the SAR requirements. 1 fig

  7. MSFR TRU-burning potential and comparison with an SFR

    Energy Technology Data Exchange (ETDEWEB)

    Fiorina, C.; Cammi, A. [Politecnico di Milano: Via La Masa 34, 20136 Milan (Italy); Franceschini, F. [Westinghouse Electric Company LL: 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States); Krepel, J. [Paul Scherrer Institut - PSI WEST, 5234 Villigen (Switzerland)

    2013-07-01

    The objective of this work is to evaluate the Molten Salt Fast Reactor (MSFR) potential benefits in terms of transuranics (TRU) burning through a comparative analysis with a sodium-cooled FR. The comparison is based on TRU- and MA-burning rates, as well as on the in-core evolution of radiotoxicity and decay heat. Solubility issues limit the TRU-burning rate to 1/3 that achievable in traditional low-CR FRs (low-Conversion-Ratio Fast Reactors). The softer spectrum also determines notable radiotoxicity and decay heat of the equilibrium actinide inventory. On the other hand, the liquid fuel suggests the possibility of using a Pu-free feed composed only of Th and MA (Minor Actinides), thus maximizing the MA burning rate. This is generally not possible in traditional low-CR FRs due to safety deterioration and decay heat of reprocessed fuel. In addition, the high specific power and the lack of out-of-core cooling times foster a quick transition toward equilibrium, which improves the MSFR capability to burn an initial fissile loading, and makes the MSFR a promising system for a quick (i.e., in a reactor lifetime) transition from the current U-based fuel cycle to a novel closed Th cycle. (authors)

  8. TRU Waste Management Program. Cost/schedule optimization analysis

    International Nuclear Information System (INIS)

    Detamore, J.A.; Raudenbush, M.H.; Wolaver, R.W.; Hastings, G.A.

    1985-10-01

    This Current Year Work Plan presents in detail a description of the activities to be performed by the Joint Integration Office Rockwell International (JIO/RI) during FY86. It breaks down the activities into two major work areas: Program Management and Program Analysis. Program Management is performed by the JIO/RI by providing technical planning and guidance for the development of advanced TRU waste management capabilities. This includes equipment/facility design, engineering, construction, and operations. These functions are integrated to allow transition from interim storage to final disposition. JIO/RI tasks include program requirements identification, long-range technical planning, budget development, program planning document preparation, task guidance development, task monitoring, task progress information gathering and reporting to DOE, interfacing with other agencies and DOE lead programs, integrating public involvement with program efforts, and preparation of reports for DOE detailing program status. Program Analysis is performed by the JIO/RI to support identification and assessment of alternatives, and development of long-term TRU waste program capabilities. These analyses include short-term analyses in response to DOE information requests, along with performing an RH Cost/Schedule Optimization report. Systems models will be developed, updated, and upgraded as needed to enhance JIO/RI's capability to evaluate the adequacy of program efforts in various fields. A TRU program data base will be maintained and updated to provide DOE with timely responses to inventory related questions

  9. MANAGEMENT OF TRANSURANIC (TRU) WASTE RETRIEVAL PROJECT RISKS SUCCESSES IN THE STARTUP OF THE HANFORD 200 AREA TRU WASTE RETRIEVAL PROJECT

    International Nuclear Information System (INIS)

    GREENWLL, R.D.

    2005-01-01

    A risk identification and mitigation method applied to the Transuranic (TRU) Waste Retrieval Project performed at the Hanford 200 Area burial grounds is described. Retrieval operations are analyzed using process flow diagramming. and the anticipated project contingencies are included in the Authorization Basis and operational plans. Examples of uncertainties assessed include degraded container integrity, bulged drums, unknown containers, and releases to the environment. Identification and mitigation of project risks contributed to the safe retrieval of over 1700 cubic meters of waste without significant work stoppage and below the targeted cost per cubic meter retrieved. This paper will be of interest to managers, project engineers, regulators, and others who are responsible for successful performance of waste retrieval and other projects with high safety and performance risks

  10. Decontamination impacts on solidification

    International Nuclear Information System (INIS)

    Piciulo, P.L.; Davis, M.S.

    1985-01-01

    The increased occupational exposure resulting from the accumulation of activated corrosion products in the primary system of LWRs has led to the development of chemical methods to remove the contamination. In the past, the problem of enhanced migration of radionuclides away from trenches used to dispose of low-level radioactive waste, has been linked to the presence, at the disposal unit, of chelating or complexing agents such as those used in decontamination processes. These agents have further been found to reduce the normal sorptive capacity of soils for radionuclides. The degree to which these agents inhibit the normal sorptive processes is dependent on the type of complexing agent, the radionuclide of concern, the soil properties and whether the nuclide is present as a complex or is already sorbed to the soil. Since the quantity of reagent employed in a full system decontamination is large (200 to 25,000 kg), the potential for enhanced migration of radionuclides from a site used to dispose of the decontamination wastes should be addressed and guidelines established for the safe disposal of these wastes

  11. CH-TRU Waste Content Codes (CH-TRUCON)

    International Nuclear Information System (INIS)

    2005-01-01

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes 'shipping categories' that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the 'General Case,' which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for 'Close-Proximity Shipments' (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for 'Controlled Shipments

  12. MCNP Modeling Results for Location of Buried TRU Waste Drums

    International Nuclear Information System (INIS)

    Steinman, D K; Schweitzer, J S

    2006-01-01

    In the 1960's, fifty-five gallon drums of TRU waste were buried in shallow pits on remote U.S. Government facilities such as the Idaho National Engineering Laboratory (now split into the Idaho National Laboratory and the Idaho Completion Project [ICP]). Subsequently, it was decided to remove the drums and the material that was in them from the burial pits and send the material to the Waste Isolation Pilot Plant in New Mexico. Several technologies have been tried to locate the drums non-intrusively with enough precision to minimize the chance for material to be spread into the environment. One of these technologies is the placement of steel probe holes in the pits into which wireline logging probes can be lowered to measure properties and concentrations of material surrounding the probe holes for evidence of TRU material. There is also a concern that large quantities of volatile organic compounds (VOC) are also present that would contaminate the environment during removal. In 2001, the Idaho National Engineering and Environmental Laboratory (INEEL) built two pulsed neutron wireline logging tools to measure TRU and VOC around the probe holes. The tools are the Prompt Fission Neutron (PFN) and the Pulsed Neutron Gamma (PNG), respectively. They were tested experimentally in surrogate test holes in 2003. The work reported here estimates the performance of the tools using Monte-Carlo modelling prior to field deployment. A MCNP model was constructed by INEEL personnel. It was modified by the authors to assess the ability of the tools to predict quantitatively the position and concentration of TRU and VOC materials disposed around the probe holes. The model was used to simulate the tools scanning the probe holes vertically in five centimetre increments. A drum was included in the model that could be placed near the probe hole and at other locations out to forty-five centimetres from the probe-hole in five centimetre increments. Scans were performed with no chlorine in the

  13. CH-TRU Waste Content Codes (CH-TRUCON)

    International Nuclear Information System (INIS)

    2007-01-01

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes 'shipping categories' that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the 'General Case,' which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for 'Close-Proximity Shipments' (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for 'Controlled Shipments

  14. CH-TRU Waste Content Codes (CH-TRUCON)

    International Nuclear Information System (INIS)

    2006-01-01

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes 'shipping categories' that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the 'General Case,' which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for 'Close-Proximity Shipments' (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for 'Controlled Shipments

  15. CH-TRU Waste Content Codes (CH-TRUCON)

    International Nuclear Information System (INIS)

    2008-01-01

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes 'shipping categories' that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the 'General Case,' which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for 'Close-Proximity Shipments' (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for 'Controlled Shipments

  16. CH-TRU Waste Content Codes (CH-TRUCON)

    International Nuclear Information System (INIS)

    2004-01-01

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes 'shipping categories' that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the 'General Case,' which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for 'Close-Proximity Shipments' (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for 'Controlled Shipments

  17. CH-TRU Content Codes (CH-TRUCON)

    International Nuclear Information System (INIS)

    2005-01-01

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes 'shipping categories' that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the 'General Case,' which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for 'Close-Proximity Shipments' (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for 'Controlled Shipments

  18. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2006-09-15

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  19. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-05-01

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  20. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2007-02-15

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  1. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-06-20

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  2. CH-TRU Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-10-15

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  3. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2006-06-20

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  4. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-01-15

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codesand corresponding shipping categories for "Controlled Shipments

  5. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2006-12-20

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  6. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2006-08-15

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  7. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2006-01-18

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  8. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2004-10-01

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  9. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-03-15

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  10. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2007-09-20

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  11. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2007-08-15

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  12. CH-TRU Waste Content Codes (CH TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2004-12-01

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  13. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-11-20

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  14. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-12-15

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  15. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-01-30

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  16. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2005-08-15

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  17. CH-TRU Waste Content Codes (CH-TRUCON)

    Energy Technology Data Exchange (ETDEWEB)

    Washington TRU Solutions LLC

    2007-06-15

    The CH-TRU Waste Content Codes (CH-TRUCON) document describes the inventory of the U.S. Department of Energy (DOE) CH-TRU waste within the transportation parameters specified by the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC). The CH-TRAMPAC defines the allowable payload for the Transuranic Package Transporter-II (TRUPACT-II) and HalfPACT packagings. This document is a catalog of TRUPACT-II and HalfPACT authorized contents and a description of the methods utilized to demonstrate compliance with the CH-TRAMPAC. A summary of currently approved content codes by site is presented in Table 1. The CH-TRAMPAC describes "shipping categories" that are assigned to each payload container. Multiple shipping categories may be assigned to a single content code. A summary of approved content codes and corresponding shipping categories is provided in Table 2, which consists of Tables 2A, 2B, and 2C. Table 2A provides a summary of approved content codes and corresponding shipping categories for the "General Case," which reflects the assumption of a 60-day shipping period as described in the CH-TRAMPAC and Appendix 3.4 of the CH-TRU Payload Appendices. For shipments to be completed within an approximately 1,000-mile radius, a shorter shipping period of 20 days is applicable as described in the CH-TRAMPAC and Appendix 3.5 of the CH-TRU Payload Appendices. For shipments to WIPP from Los Alamos National Laboratory (LANL), Nevada Test Site, and Rocky Flats Environmental Technology Site, a 20-day shipping period is applicable. Table 2B provides a summary of approved content codes and corresponding shipping categories for "Close-Proximity Shipments" (20-day shipping period). For shipments implementing the controls specified in the CH-TRAMPAC and Appendix 3.6 of the CH-TRU Payload Appendices, a 10-day shipping period is applicable. Table 2C provides a summary of approved content codes and corresponding shipping categories for "Controlled Shipments

  18. Grout and glass performance in support of stabilization/solidification of ORNL tank sludges

    International Nuclear Information System (INIS)

    Spence, R.D.; Mattus, C.H.; Mattus, A.J.

    1998-09-01

    Wastewater at Oak Ridge National Laboratory (ORNL) is collected, evaporated, and stored in the Melton Valley Storage Tanks (MVST) and Bethel Valley Evaporator Storage Tanks (BVEST) pending treatment for disposal. In addition, some sludges and supernatants also requiring treatment remain in two inactive tank systems: the gunite and associated tanks (GAAT) and the old hydrofracture (OHF) tank. The waste consists of two phases: sludge and supernatant. The sludges contain a high amount of radioactivity, and some are classified as TRU sludges. Some Resource Conservation and Recovery Act (RCRA) metal concentrations are high enough to be defined as RCRA hazardous; therefore, these sludges are presumed to be mixed TRU waste. Grouting and vitrification are currently two likely stabilization/solidification alternatives for mixed wastes. Grouting has been used to stabilize/solidify hazardous and low-level radioactive waste for decades. Vitrification has been developed as a high-level radioactive alternative for decades and has been under development recently as an alternative disposal technology for mixed waste. The objective of this project is to define an envelope, or operating window, for grout and glass formulations for ORNL tank sludges. Formulations will be defined for the average composition of each of the major tank farms (BVEST/MVST, GAAT, and OHF) and for an overall average composition of all tank farms. This objective is to be accomplished using surrogates of the tank sludges with hot testing of actual tank sludges to check the efficacy of the surrogates

  19. Evaluation of stabilization-solidification techniques

    International Nuclear Information System (INIS)

    Goubier, R.

    1989-01-01

    This paper reports that among the techniques applied to treat polluting residue in France for the past ten years has been the mixing of pollutants with reactive agents in order to fix the contaminants and to give them a solid consistency. The first applications of these stabilization/solidification processes occurred in 1978 in the treatment of oil residues from the AMOCO CADIZ spill. They have also been used for the treatment of a mayor dump site for petroleum residues, for the disposal of mineral sludges of a detoxication plant, and for the rehabilitation of sites contaminated by various industrial residues, specially acid tars generated by oil refining plants. Although from the beginning these techniques appeared to be able to transform filthy lagoons into solid and apparently safe areas, it was necessary to evaluate their efficiency and to determine the conditions and limits of application

  20. Solidification control in continuous casting of steel

    Indian Academy of Sciences (India)

    Unknown

    Solidification in continuous casting (CC) technology is initiated in a water- ..... to fully austenitic solidification, and FP between 0 and 1 indicates mixed mode. ... the temperature interval (LIT – TSA) corresponding to fs = 0⋅9 → 1, is in reality the.

  1. General characteristics of eutectic alloy solidification mechanisms

    International Nuclear Information System (INIS)

    Lemaignan, Clement.

    1977-01-01

    The eutectic alloy sodification was studied in binary systems: solidification of non facetted - non facetted eutectic alloy (theoretical aspects, variation of the lamellar spacing, crystallographic relation between the various phases); solidification of facetted - non facetted eutectic alloy; coupled growth out of eutectic alloy; eutectic nucleation [fr

  2. TRU Waste Inventory Collection and Work-Off Plans for the Centralization of TRU Waste Characterization/Certification at INL - On Your Mark - Get Set

    International Nuclear Information System (INIS)

    McTaggart, J.; Lott, S.

    2009-01-01

    The U.S. Department of Energy (DOE) amended the Record of Decision (ROD) for the Waste Management Program: Treatment and Storage of Transuranic Waste to centralize transuranic (TRU) waste characterization/certification from fourteen TRU waste sites. This centralization will allow for treatment, characterization and certification of TRU waste from the fourteen sites, thirteen of which are sites with small quantities of TRU waste, at the Idaho National Laboratory (INL) prior to shipping the waste to the Waste Isolation Pilot Plant (WIPP) for disposal. Centralization of this TRU waste will avoid the cost of building treatment, characterization, certification, and shipping capabilities at each of the small quantity sites that currently do not have existing facilities. Advanced Mixed Waste Treatment Project (AMWTP) and Idaho Nuclear Technology and Engineering Center (INTEC) will provide centralized shipping facilities, to WIPP, for all of the small quantity sites. Hanford, the one large quantity site identified in the ROD, has a large number of waste in containers that are over-packed into larger containers which are inefficient for shipment to and disposal at WIPP. The AMWTP at the INL will reduce the volume of much of the CH waste and make it much more efficient to ship and dispose of at WIPP. In addition, the INTEC has a certified remote handled (RH) TRU waste characterization/certification program at INL to disposition TRU waste from the sites identified in the ROD. (authors)

  3. TRU waste inventory collection and work-off plans for the centralization of TRU waste characterization at INL - on your mark - get set - 9410

    International Nuclear Information System (INIS)

    Mctaggert, Jerri Lynne; Lott, Sheila; Gadbury, Casey

    2009-01-01

    The U.S. Department of Energy (DOE) amended the Record of Decision (ROD) for the Waste Management Program: Treatment and Storage ofTransuranic Waste to centralize transuranic (TRU) waste characterization/certification from fourteen TRU waste sites. This centralization will allow for treatment, characterization and certification ofTRU waste from the fourteen sites, thirteen of which are sites with small quantities ofTRU waste, at the Idaho National Laboratory (INL) prior to shipping the waste to the Waste Isolation Pilot Plant (WIPP) for disposal. Centralization ofthis TRU waste will avoid the cost ofbuilding treatment, characterization, certification, and shipping capabilities at each ofthe small quantity sites that currently do not have existing facilities. Advanced Mixed Waste Treatment Project (AMWTP) and Idaho Nuclear Technology and Engineering Center (INTEC) will provide centralized shipping facilities, to WIPP, for all ofthe small quantity sites. Hanford, the one large quantity site identified in the ROD, has a large number ofwaste in containers that are overpacked into larger containers which are inefficient for shipment to and disposal at WIPP. The AMWTP at the INL will reduce the volume ofmuch of the CH waste and make it much more efficient to ship and dispose of at WIPP. In addition, the INTEC has a certified remote handled (RH) TRU waste characterization/certification program at INL to disposition TRU waste from the sites identified in the ROD.

  4. Mobile concrete solidification systems for power reactor waste

    International Nuclear Information System (INIS)

    Tchemitcheff, E.; Bordas, Y.

    1990-01-01

    In late 1988 SGN received an order from Electricite de France (EDF) for the construction of a mobile concrete solidification system to process secondary system resins generated by the P'4 and N4 series PWR power plants in France. This order was placed in view of SGN's experience with low- and medium-level radioactive waste treatment and conditioning over a period of almost 20 years. In addition to the construction of fixed waste processing facilities using more conventional technologies, SGN has been involved in application of the mobile system concept to the bituminization process in the United States, which led to the construction and commissioning of two transportable systems in collaboration with its American licensee US Ecology. It has also conducted large-scale R ampersand D on LLW/MLW concrete solidification, particularly for ion exchange resins. 5 figs

  5. Techniques for the solidification of high-level wastes

    International Nuclear Information System (INIS)

    1977-01-01

    The problem of the long-term management of the high-level wastes from the reprocessing of irradiated nuclear fuel is receiving world-wide attention. While the majority of the waste solutions from the reprocessing of commercial fuels are currently being stored in stainless-steel tanks, increasing effort is being devoted to developing technology for the conversion of these wastes into solids. A number of full-scale solidification facilities are expected to come into operation in the next decade. The object of this report is to survey and compare all the work currently in progress on the techniques available for the solidification of high-level wastes. It will examine the high-level liquid wastes arising from the various processes currently under development or in operation, the advantages and disadvantages of each process for different types and quantities of waste solutions, the stages of development, the scale-up potential and flexibility of the processes

  6. TRU waste certification and TRUPACT-II payload verification

    International Nuclear Information System (INIS)

    Hunter, E.K.; Johnson, J.E.

    1990-01-01

    The Waste Isolation Pilot Plant (WIPP) established a policy (subsequently confirmed and required by DOE Order 5820.2A, Radioactive Waste Management, September 1988) that requires each waste shipper to verify that all waste shipments meet the requirements of the Waste Acceptance Criteria (WAC) prior to being shipped. This verification provides assurance that transuranic (TRU) wastes meet the criteria while still retained in a facility where discrepancies can be immediately corrected. In this manner, problems that would arise if WAC violations were discovered at the receiver, where corrective facilities are not available, are avoided. Each Department of Energy (DOE) TRU waste facility planning to ship waste to the Waste Isolation Pilot Plant (WIPP) is required to develop and implement a specific program including Quality Assurance (QA) provisions to verify that waste is in full compliance with WIPP's WAC. This program is audited by a composite DOE and contractor audit team prior to granting the facility permission to certify waste. During interaction with the Nuclear Regulatory Commission (NRC) on payload verification for shipping in TRUPACT-II, a similar system was established by DOE. The TRUPACT-II Safety Analysis Report (SAR) contains the technical requirements and physical and chemical limits that payloads must meet (like the WAC). All shippers must plan and implement a payload control program including independent QA provisions. A similar composite audit team will conduct preshipment audits, frequent subsequent audits, and operations inspections to verify that all TRU waste shipments in TRUPACT-II meet the requirements of the Certificate of Compliance (C of C) issued by the NRC which invokes the SAR requirements. 1 fig

  7. Radiolytic decomposition of organic C-14 released from TRU waste

    International Nuclear Information System (INIS)

    Kani, Yuko; Noshita, Kenji; Kawasaki, Toru; Nishimura, Tsutomu; Sakuragi, Tomofumi; Asano, Hidekazu

    2007-01-01

    It has been found that metallic TRU waste releases considerable portions of C-14 in the form of organic molecules such as lower molecular weight organic acids, alcohols and aldehydes. Due to the low sorption ability of organic C-14, it is important to clarify the long-term behavior of organic forms under waste disposal conditions. From investigations on radiolytic decomposition of organic carbon molecules into inorganic carbonic acid, it is expected that radiation from TRU waste will decompose organic C-14 into inorganic carbonic acid that has higher adsorption ability into the engineering barriers. Hence we have studied the decomposition behavior of organic C-14 by gamma irradiation experiments under simulated disposal conditions. The results showed that organic C-14 reacted with OH radicals formed by radiolysis of water, to produce inorganic carbonic acid. We introduced the concept of 'decomposition efficiency' which expresses the percentage of OH radicals consumed for the decomposition reaction of organic molecules in order to analyze the experimental results. We estimated the effect of radiolytic decomposition on the concentration of organic C-14 in the simulated conditions of the TRU disposal system using the decomposition efficiency, and found that the concentration of organic C-14 in the waste package will be lowered when the decomposition of organic C-14 by radiolysis was taken into account, in comparison with the concentration of organic C-14 without radiolysis. Our prediction suggested that some amount of organic C-14 can be expected to be transformed into the inorganic form in the waste package in an actual system. (authors)

  8. CsIX/TRU Grout Feasibility Study

    Energy Technology Data Exchange (ETDEWEB)

    S. J. Losinski; C. M. Barnes; B. K. Grover

    1998-11-01

    A settlement agreement between the Department of Energy (DOE) and the State of Idaho mandates that liquid waste now stored at the Idaho Nuclear Technology Engineering Center (INTEC - formerly the Idaho Chemical Processing Plant, ICPP) will be calcined by the end of year 2012. This study investigates an alternative treatment of the liquid waste that removes undissolved solids (UDS) by filtration and removes cesium by ion exchange followed by cement-based grouting of the remaining liquid into 55-gal drums. Operations are assumed to be from January 2008 through December 2012. The grouted waste will be contact-handled and will be shipped to the Waste Isolation Pilot Plant (WIPP) in New Mexico for disposal. The small volume of secondary wastes such as the filtered solids and cesium sorbent (resin) would remain in storage at the Idaho National Engineering and Environmental Laboratory for treatment and disposal under another project, with an option to dispose of the filtered solids as a r emote-handled waste at WIPP.

  9. CsIX/TRU Grout Feasibility Study

    International Nuclear Information System (INIS)

    Losinski, S. J.; Barnes, C. M.; Grover, B. K.

    1998-01-01

    A settlement agreement between the Department of Energy (DOE) and the State of Idaho mandates that liquid waste now stored at the Idaho Nuclear Technology Engineering Center (INTEC - formerly the Idaho Chemical Processing Plant, ICPP) will be calcined by the end of year 2012. This study investigates an alternative treatment of the liquid waste that removes undissolved solids (UDS) by filtration and removes cesium by ion exchange followed by cement-based grouting of the remaining liquid into 55-gal drums. Operations are assumed to be FR-om January 2008 through December 2012. The grouted waste will be contact-handled and will be shipped to the Waste Isolation Pilot Plant (WIPP) in New Mexico for disposal. The small volume of secondary wastes such as the filtered solids and cesium sorbent (resin) would remain in storage at the Idaho National Engineering and Environmental Laboratory for treatment and disposal under another project, with an option to dispose of the filtered solids as a r emote-handled waste at WIPP

  10. TRU Waste Sampling Program: Volume I. Waste characterization

    International Nuclear Information System (INIS)

    Clements, T.L. Jr.; Kudera, D.E.

    1985-09-01

    Volume I of the TRU Waste Sampling Program report presents the waste characterization information obtained from sampling and characterizing various aged transuranic waste retrieved from storage at the Idaho National Engineering Laboratory and the Los Alamos National Laboratory. The data contained in this report include the results of gas sampling and gas generation, radiographic examinations, waste visual examination results, and waste compliance with the Waste Isolation Pilot Plant-Waste Acceptance Criteria (WIPP-WAC). A separate report, Volume II, contains data from the gas generation studies

  11. Remotely operated facility for in situ solidification of fissile uranium

    International Nuclear Information System (INIS)

    McGinnis, C.P.; Collins, E.D.; Patton, B.D.

    1986-01-01

    A heavily shielded, remotely operated facility, located within the Radiochemical processing Plant at Oak Ridge National Laboratory (ORNL), has been designed and is being operated to convert approx.1000 kg of fissile uranium (containing approx.75% 235 U, approx.10% 233 U, and approx.140 ppM 232 U) from a nitrate solution (130 g of uranium per L) to a solid oxide form. This project, the Consolidated Edison Uranium Solidification Program (CEUSP), is being carried out in order to prepare a stable uranium form for longterm storage. This paper describes the solidification process selected, the equipment and facilities required, the experimental work performed to ensure successful operation, some problems that were solved, and the initial operations

  12. Metastable and unstable cellular solidification of colloidal suspensions

    Science.gov (United States)

    Deville, Sylvain; Maire, Eric; Bernard-Granger, Guillaume; Lasalle, Audrey; Bogner, Agnès; Gauthier, Catherine; Leloup, Jérôme; Guizard, Christian

    2009-12-01

    Colloidal particles are often seen as big atoms that can be directly observed in real space. They are therefore becoming increasingly important as model systems to study processes of interest in condensed-matter physics such as melting, freezing and glass transitions. The solidification of colloidal suspensions has long been a puzzling phenomenon with many unexplained features. Here, we demonstrate and rationalize the existence of instability and metastability domains in cellular solidification of colloidal suspensions, by direct in situ high-resolution X-ray radiography and tomography observations. We explain such interface instabilities by a partial Brownian diffusion of the particles leading to constitutional supercooling situations. Processing under unstable conditions leads to localized and global kinetic instabilities of the solid/liquid interface, affecting the crystal morphology and particle redistribution behaviour.

  13. Numerical model for dendritic solidification of binary alloys

    Science.gov (United States)

    Felicelli, S. D.; Heinrich, J. C.; Poirier, D. R.

    1993-01-01

    A finite element model capable of simulating solidification of binary alloys and the formation of freckles is presented. It uses a single system of equations to deal with the all-liquid region, the dendritic region, and the all-solid region. The dendritic region is treated as an anisotropic porous medium. The algorithm uses the bilinear isoparametric element, with a penalty function approximation and a Petrov-Galerkin formulation. Numerical simulations are shown in which an NH4Cl-H2O mixture and a Pb-Sn alloy melt are cooled. The solidification process is followed in time. Instabilities in the process can be clearly observed and the final compositions obtained.

  14. Dynamics of liquid solidification thermal resistance of contact layer

    CERN Document Server

    Lipnicki, Zygmunt

    2017-01-01

    This monograph comprehensively describes phenomena of heat flow during phase change as well as the dynamics of liquid solidification, i.e. the development of a solidified layer. The book provides the reader with basic knowledge for practical designs, as well as with equations which describe processes of energy transformation. The target audience primarily comprises researchers and experts in the field of heat flow, but the book may also be beneficial for both practicing engineers and graduate students.

  15. Characteristics of Cement Solidification of Metal Hydroxide Waste

    Directory of Open Access Journals (Sweden)

    Dae-Seo Koo

    2017-02-01

    Full Text Available To perform the permanent disposal of metal hydroxide waste from electro-kinetic decontamination, it is necessary to secure the technology for its solidification. The integrity tests on the fabricated solidification should also meet the criteria of the Korea Radioactive Waste Agency. We carried out the solidification of metal hydroxide waste using cement solidification. The integrity tests such as the compressive strength, immersion, leach, and irradiation tests on the fabricated cement solidifications were performed. It was also confirmed that these requirements of the criteria of Korea Radioactive Waste Agency on these cement solidifications were met. The microstructures of all the cement solidifications were analyzed and discussed.

  16. Characteristics of cement solidification of metal hydroxide waste

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Dae Seo; Sung, Hyun Hee; Kim, Seung Soo; Kim, Gye Nam; Choi, Jong Won [Dept. of Decontemination Decommission Technology Development, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-02-15

    To perform the permanent disposal of metal hydroxide waste from electro-kinetic decontamination, it is necessary to secure the technology for its solidification. The integrity tests on the fabricated solidification should also meet the criteria of the Korea Radioactive Waste Agency. We carried out the solidification of metal hydroxide waste using cement solidification. The integrity tests such as the compressive strength, immersion, leach, and irradiation tests on the fabricated cement solidifications were performed. It was also confirmed that these requirements of the criteria of Korea Radioactive Waste Agency on these cement solidifications were met. The microstructures of all the cement solidifications were analyzed and discussed.

  17. Non-Equilibrium Solidification of Undercooled Metallic Melts

    Directory of Open Access Journals (Sweden)

    Dieter M. Herlach

    2014-06-01

    Full Text Available If a liquid is undercooled below its equilibrium melting temperature an excess Gibbs free energy is created. This gives access to solidification of metastable solids under non-equilibrium conditions. In the present work, techniques of containerless processing are applied. Electromagnetic and electrostatic levitation enable to freely suspend a liquid drop of a few millimeters in diameter. Heterogeneous nucleation on container walls is completely avoided leading to large undercoolings. The freely suspended drop is accessible for direct observation of rapid solidification under conditions far away from equilibrium by applying proper diagnostic means. Nucleation of metastable crystalline phases is monitored by X-ray diffraction using synchrotron radiation during non-equilibrium solidification. While nucleation preselects the crystallographic phase, subsequent crystal growth controls the microstructure evolution. Metastable microstructures are obtained from deeply undercooled melts as supersaturated solid solutions, disordered superlattice structures of intermetallics. Nucleation and crystal growth take place by heat and mass transport. Comparative experiments in reduced gravity allow for investigations on how forced convection can be used to alter the transport processes and design materials by using undercooling and convection as process parameters.

  18. Transuranic (TRU) waste management at Savannah River - past, present and future

    International Nuclear Information System (INIS)

    D'Ambrosia, J.T.

    1985-01-01

    Defense TRU waste at Savannah River (SR) results from the Department of Energy's (DOE) national defense activities, including the operation of production reactors and fuel reprocessing plants and research and development activities. TRU waste is material declared as having negligible economic value, contaminated with alpha-emitting radionuclides of atomic number greater than 92, and half-lives longer than 20 years, in concentrations greater than 100 nCi/g. TRU waste has been retrievably stored at SR since 1974 awaiting disposal. The Waste Isolation Pilot Plant (WIPP), now under construction in New Mexico, is a research and development facility for demonstrating the safe disposal of defense TRU waste, including that in storage at SR. The major objective of the TRU program at SR is to support the TRU National Program, which is dedicated to preparing waste for, and emplacing waste in, the WIPP. Thus, the SR Program also supports WIPP operations. The SR Site specific goals are to phase out the indefinite storage of TRU waste, which has been the mode of waste management since 1974, and to dispose of SR's Defense TRU waste

  19. Global cooperation and conceptual design toward GNEP. Enhanced TRU burning fast reactor

    International Nuclear Information System (INIS)

    Ikeda, Kazumi; Maddox, James W.; Nakazato, Wataru; Kunishima, Shigeru

    2008-01-01

    In support of the GNEP (Global Nuclear Energy Partnership) program, AREVA and Mitsubishi Heavy Industries, Ltd. (MHI) seek to develop an ARR (Advanced Recycling Reactor) in concern with a CFTC (Consolidated Fuel Treatment Facility). This report presents the examination of more effective transuranics (TRU) burning core. Therefore some innovative technologies have been examined under the safety requirements; MA bearing fuel with 50% TRU fraction, moderator pin, fuel of high Am fraction, and Am blanket. The function of moderator is to enhance TRU burning capability, while increasing the Doppler effect and reducing the positive sodium void effect. The aim of 50% TRU fraction is to increase TRU burning capability by curbing plutonium production. Both high Am fraction of fuel and Am blanket can promote Am transmutation. According to the detailed calculation of high TRU (MA 15%, Pu 35% average) contained oxide fueled core with moderator pins of 12% arranged driver fuel assemblies, TRU conversion ratio decreases down to 0.33 and TRU burning capability is improved to 67kg/TWeh. Deploying Am blanket which is oxide fuel with Am 50% and U 50%, the total of Am transmutation capability becomes 69 kg/TWeh. (author)

  20. Utilization of coal fly ash in solidification of liquid radioactive waste from research reactor.

    Science.gov (United States)

    Osmanlioglu, Ahmet Erdal

    2014-05-01

    In this study, the potential utilization of fly ash was investigated as an additive in solidification process of radioactive waste sludge from research reactor. Coal formations include various percentages of natural radioactive elements; therefore, coal fly ash includes various levels of radioactivity. For this reason, fly ashes have to be evaluated for potential environmental implications in case of further usage in any construction material. But for use in solidification of radioactive sludge, the radiological effects of fly ash are in the range of radioactive waste management limits. The results show that fly ash has a strong fixing capacity for radioactive isotopes. Specimens with addition of 5-15% fly ash to concrete was observed to be sufficient to achieve the target compressive strength of 20 MPa required for near-surface disposal. An optimum mixture comprising 15% fly ash, 35% cement, and 50% radioactive waste sludge could provide the solidification required for long-term storage and disposal. The codisposal of radioactive fly ash with radioactive sludge by solidification decreases the usage of cement in solidification process. By this method, radioactive fly ash can become a valuable additive instead of industrial waste. This study supports the utilization of fly ash in industry and the solidification of radioactive waste in the nuclear industry.

  1. Analysis of TRU waste for RCRA-listed elements

    International Nuclear Information System (INIS)

    Mahan, C.; Gerth, D.; Yoshida, T.

    1996-01-01

    Analytical methods for RCRA listed elements on Portland cement type waste have been employed using both microwave and open hot plate digestions with subsequent analysis by inductively coupled plasma atomic emission spectroscopy (ICP-AES), inductively coupled plasma mass spectrometry (ICP-AES), inductively coupled plasma mass spectrometry (ICP-MS), graphite furnace atomic absorption (GFAA) and cold vapor atomic absorption and fluorescence (CVAA/CVAFS). Four different digestion procedures were evaluated including an open hot plate nitric acid digestion, EPA SW-846 Method 3051, and 2 methods using modifications to Method 3051. The open hot plate and the modified Method 3051, which used aqua regia for dissolution, were the only methods which resulted in acceptable data quality for all 14 RCRA-listed elements. Results for the nitric acid open hot plate digestion were used to qualify the analytical methods for TRU waste characterization, and resulted in a 99% passing score. Direct chemical analysis of TRU waste is being developed at Los Alamos National Laboratory in an attempt to circumvent the problems associated with strong acid digestion methods. Technology development includes laser induced breakdown spectroscopy (LIBS), laser ablation inductively coupled plasma mass spectrometry (LA-ICPMS), dc arc CID atomic emission spectroscopy (DC-AES), and glow discharge mass spectrometry (GDMS). Analytical methods using the Portland cement matrix are currently being developed for each of the listed techniques. Upon completion of the development stage, blind samples will be distributed to each of the technology developers for RCRA metals characterization

  2. Waste Isolation Pilot Plant TruDock crane system analysis

    International Nuclear Information System (INIS)

    Morris, B.C.; Carter, M.

    1996-10-01

    The WIPP TruDock crane system located in the Waste Handling Building was identified in the WIPP Safety Analysis Report (SAR), November 1995, as a potential accident concern due to failures which could result in a dropped load. The objective of this analysis is to evaluate the frequency of failure of the TruDock crane system resulting in a dropped load and subsequent loss of primary containment, i.e. drum failure. The frequency of dropped loads was estimated to be 9.81E-03/year or approximately one every 102 years (or, for the 25% contingency, 7.36E-03/year or approximately one every 136 years). The dominant accident contributor was the failure of the cable/hook assemblies, based on failure data obtained from NUREG-0612, as analyzed by PLG, Inc. The WIPP crane system undergoes a rigorous test and maintenance program, crane operation is discontinued following any abnormality, and the crane operator and load spotter are required to be trained in safe crane operation, therefore it is felt that the WIPP crane performance will exceed the data presented in NUREG-0612 and the estimated failure frequency is felt to be conservative

  3. Deep-Burn MHR Neutronic Analysis with a SiC-Gettered TRU Kernel

    International Nuclear Information System (INIS)

    Jo, Chang Keun; Noh, Jae Man; Kim, Yong Hee; Venneric, F.

    2010-01-01

    This paper is focused on the nuclear core design of a DB-MHR (Deep Burn-Modular Helium Reactor) core loaded with a SiC-gettered TRU fuel. The SiC oxygen getter is added to reduce the CO pressure in the buffer zone of TRISO. In the paper, the cycle length, reactivity swing, discharged burnup, and the burning rate of plutonium were calculated for the DB-MHR. Also, impacts of uranium addition to the TRU kernel were investigated. Recently, the decay heat of TRU fueled DB core was found to be highly dependent on the TRU loading: the higher the loading, the higher the decay heat. The high decay heat of TRU fuel may lead to unacceptably high peak fuel temperature during an LPCC (Low Pressure Conduction Cooling) accident. Thus, we tried to minimize the decay heat of the core for a minimal peak fuel temperature during LPCC

  4. CONCRETE CONTAINERS FOR LONG TERM STORAGE AND FINAL DISPOSAL OF TRU WASTE AND LONG LIVED ILW

    International Nuclear Information System (INIS)

    Sakamoto, H.; Asano, H.; Tunaboylu, K.; Mayer, G.; Klubertanz, G.; Kobayashi, S.; Komuro, T.; Wagner, E.

    2003-01-01

    Transuranic (TRU) waste packaging development has been conducted since 1998 by the Radioactive Waste Management Funding and Research Centre (RWMC) to support the TRU waste disposal concept in Japan. In this paper, the overview of development status of the reinforced concrete package is introduced. This package has been developed in order to satisfy the Japanese TRU waste disposal concept based on current technology and to provide a low cost package. Since 1998, the basic design work (safety evaluation, manufacturing and handling procedure, economic evaluation, elemental tests etc.) have been carried out. As a result, the basic specification of the package was decided. This report presents the concept as well as the results of basic design, focused on safety analysis and handling procedure of the package. Two types of the packages exist: - Package-A: for non-heat generating TRU waste from reprocessing in 200 l drums and - Package-B: for heat generating TRU-waste from reprocessing

  5. A Comparative study of solidification of Al-Cu alloy under flow of cylindrical radial heat and the unidirectional vertically

    Directory of Open Access Journals (Sweden)

    Jean Robert P. Rodrigues

    2014-09-01

    Full Text Available In spite of technological importance of solidification of metallic alloys under radial heat flow, relatively few studies have been carried out in this area. In this work the solidification of Al 4.5 wt% Cu cylinders against a steel massive mold is analyzed and compared with unidirectional solidification against a cooled mold. Initially temperature variations at different positions in the casting and in the mold were measured during solidification using a data acquisition system. These temperature variations were introduced in a numerical method in order to determine the variation of heat transfer coefficient at metal/mold interface by inverse method. The primary and secondary dendrite arm spacing variations were measured through optical microscopy. Comparisons carried out between experimental and numerical data showed that the numerical method describes well the solidification processes under radial heat flux.

  6. A study for the safety evaluation of geological disposal of TRU waste and influence on disposal site design by change of amount of TRU waste (Joint research)

    International Nuclear Information System (INIS)

    Hasegawa, Makoto; Kondo, Hitoshi; Takahashi, Kuniaki; Funabashi, Hideaki; Kawatsuma, Shinji; Kamei, Gento; Hirano, Fumio; Mihara, Morihiro; Ueda, Hiroyoshi; Ohi, Takao; Hyodo, Hideaki

    2011-02-01

    In the safety evaluation of the geological disposal of the TRU waste, it is extremely important to share the information with the Research and development organization (JAEA: that is also the waste generator) by the waste disposal entrepreneur (NUMO). In 2009, NUMO and JAEA set up a technical commission to investigate the reasonable TRU waste disposal following a cooperation agreement between these two organizations. In this report, the calculation result of radionuclide transport for a TRU waste geological disposal system was described, by using the Tiger code and the GoldSim code at identical terms. Tiger code is developed to calculate a more realistic performance assessment by JAEA. On the other hand, GoldSim code is the general simulation software that is used for the computation modeling of NUMO TRU disposal site. Comparing the calculation result, a big difference was not seen. Therefore, the reliability of both codes was able to be confirmed. Moreover, the influence on the disposal site design (Capacity: 19,000m 3 ) was examined when 10% of the amount of TRU waste increased. As a result, it was confirmed that the influence of the site design was very little based on the concept of the Second Progress Report on Research and Development for TRU Waste Disposal in Japan. (author)

  7. Development of the remote-handled transuranic waste radioassay data quality objectives. An evaluation of RH-TRU waste inventories, characteristics, radioassay methods and capabilities

    Energy Technology Data Exchange (ETDEWEB)

    Meeks, A.M.; Chapman, J.A.

    1997-09-01

    The Waste Isolation Pilot Plant will accept remote-handled transuranic waste as early as October of 2001. Several tasks must be accomplished to meet this schedule, one of which is the development of Data Quality Objectives (DQOs) and corresponding Quality Assurance Objectives (QAOs) for the assay of radioisotopes in RH-TRU waste. Oak Ridge National Laboratory (ORNL) was assigned the task of providing to the DOE QAO, information necessary to aide in the development of DQOs for the radioassay of RH-TRU waste. Consistent with the DQO process, information needed and presented in this report includes: identification of RH-TRU generator site radionuclide data that may have potential significance to the performance of the WIPP repository or transportation requirements; evaluation of existing methods to measure the identified isotopic and quantitative radionuclide data; evaluation of existing data as a function of site waste streams using documented site information on fuel burnup, radioisotope processing and reprocessing, special research and development activities, measurement collection efforts, and acceptable knowledge; and the current status of technologies and capabilities at site facilities for the identification and assay of radionuclides in RH-TRU waste streams. This report is intended to provide guidance in developing the RH-TRU waste radioassay DQOs, first by establishing a baseline from which to work, second, by identifying needs to fill in the gaps between what is known and achievable today and that which will be required before DQOs can be formulated, and third, by recommending measures that should be taken to assure that the DQOs in fact balance risk and cost with an achievable degree of certainty.

  8. Development of the remote-handled transuranic waste radioassay data quality objectives. An evaluation of RH-TRU waste inventories, characteristics, radioassay methods and capabilities

    International Nuclear Information System (INIS)

    Meeks, A.M.; Chapman, J.A.

    1997-09-01

    The Waste Isolation Pilot Plant will accept remote-handled transuranic waste as early as October of 2001. Several tasks must be accomplished to meet this schedule, one of which is the development of Data Quality Objectives (DQOs) and corresponding Quality Assurance Objectives (QAOs) for the assay of radioisotopes in RH-TRU waste. Oak Ridge National Laboratory (ORNL) was assigned the task of providing to the DOE QAO, information necessary to aide in the development of DQOs for the radioassay of RH-TRU waste. Consistent with the DQO process, information needed and presented in this report includes: identification of RH-TRU generator site radionuclide data that may have potential significance to the performance of the WIPP repository or transportation requirements; evaluation of existing methods to measure the identified isotopic and quantitative radionuclide data; evaluation of existing data as a function of site waste streams using documented site information on fuel burnup, radioisotope processing and reprocessing, special research and development activities, measurement collection efforts, and acceptable knowledge; and the current status of technologies and capabilities at site facilities for the identification and assay of radionuclides in RH-TRU waste streams. This report is intended to provide guidance in developing the RH-TRU waste radioassay DQOs, first by establishing a baseline from which to work, second, by identifying needs to fill in the gaps between what is known and achievable today and that which will be required before DQOs can be formulated, and third, by recommending measures that should be taken to assure that the DQOs in fact balance risk and cost with an achievable degree of certainty

  9. Solidification and performance of cement doped with phenol

    International Nuclear Information System (INIS)

    Vipulanandan, C.; Krishnan, S.

    1991-01-01

    Treating mixed hazardous wastes using the solidification/stabilization technology is becoming a critical element in waste management planning. The effect of phenol, a primary constituent in many hazardous wastes, on the setting and solidification process of Type I Portland cement was evaluated. The leachability of phenol from solidified cement matrix (TCLP test) and changes in mechanical properties were studied after curing times up to 28 days. The changes in cement hydration products due to phenol were studied using the X-ray diffraction (XRD) powder technique. Results show that phenol interferes with initial cement hydration by reducing the formation of calcium hydroxide and also reduces the compressive strength of cement. A simple model has been proposed to quantify the phenol leached from the cement matrix during the leachate test

  10. 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.

  11. Maximizing DOE R and D efforts in tru waste management learning from international programs

    International Nuclear Information System (INIS)

    Saxman, P.A.; Loughead, J.S.C.

    1990-01-01

    Through the International Technology Exchange Program, Department of Energy (DOE) technical specialists maintain a formal dialogue with research and Development (R and D) specialists from nuclear programs in other countries. The objective of these exchanges is to seek innovative waste management solutions, maximize progress for ongoing R and D activities, and minimize the development time required for implementation of transuranic (TRU) waste processing technologies and waste assay developments. Based on information provided by PNC during the exchange, DOE specialists evaluated PNC's efforts to implement technologies and techniques from their R and D program activities. This paper presents several projects with particular potential for DOE operations, and suggests several ways that these concepts could be used to advantage by DOE or commercial programs

  12. Advanced dust monitoring system applied to new TRU handling facility of JAERI

    International Nuclear Information System (INIS)

    Yabuta, H.; Shigeta, Y.; Sawahata, K.; Hasegawa, K.

    1993-01-01

    In JAERI, a large, scale multipurpose facility is under construction, which consists of a TRU waste management testing installation, a solution fuel treatment installation and critical assemblies with uranium and/or plutonium solution fuel. The facility is also equipped with a lot of gloveboxes for handling and treatment of solution fuel and hot cells for research on reprocessing process. As there may be a relatively high potential of air contamination, it is important to monitor air contamination effectively and efficiently. An advanced dust monitoring system was introduced for convenience of handling and automatical measurement of filter papers, by developing a filter-holder with an IC memory and a radioactivity measuring device with an automatic filter-holder changing mechanism as a part of a centralized monitoring system with a computer

  13. A study on decontamination of TRU, Co, and Mo using plasma surface etching technique

    International Nuclear Information System (INIS)

    Seo, Y.D.; Kim, Y.S.; Paek, S.H.; Lee, K.H.; Jung, C.H.; Oh, W.Z.

    2001-01-01

    Recently dry decontamination/surface-cleaning technology using plasma etching has been focused in the nuclear industry. In this study, the applicability and the effectiveness of this new dry processing technique are experimentally investigated by examining the etching reaction of UO 2 , Co, and Mo in r.f. plasma with the etchant gas of CF 4 /O 2 mixture. UO 2 is chosen as a representing material for uranium and TRU (TRans-Uranic) compounds and metallic Co and Mo are selected because they are the principal contaminants in the spent nuclear components such as valves and pipes made of stainless steel or INCONEL. Results show that in all cases maximum etching rate is achieved when the mole fraction of O 2 to CF 4 /O 2 mixture gas is 20 %, regardless of temperature and r.f. power. (author)

  14. Micro-scale thermocapillary convection with solidification

    International Nuclear Information System (INIS)

    Yang, W.J.; Liu, J.C.; Chai, A.T.

    1991-01-01

    This paper reports on an experimental study performed on heat transfer in sessile drops of lysozyme solutions with solidification. Solidification inside the sessile drop is initiated by means of the center cooling method. The internal flow behavior and solidification front movement are observed using a microscope-video monitor system. Results are obtained for lysozyme, and buffer solutions, and water, representing media possessing surface tension coefficients. It is disclosed that the time history of the solidification front movement can be divided into two stages; initial and stable. In the stable stage, the front movement x follows the power-law behavior x = Ct n . C is an empirical constant, and t denotes time. The exponent n takes on a value close to unity in the stable stage

  15. Evolution of solidification texture during additive manufacturing

    Science.gov (United States)

    Wei, H. L.; Mazumder, J.; DebRoy, T.

    2015-01-01

    Striking differences in the solidification textures of a nickel based alloy owing to changes in laser scanning pattern during additive manufacturing are examined based on theory and experimental data. Understanding and controlling texture are important because it affects mechanical and chemical properties. Solidification texture depends on the local heat flow directions and competitive grain growth in one of the six preferred growth directions in face centered cubic alloys. Therefore, the heat flow directions are examined for various laser beam scanning patterns based on numerical modeling of heat transfer and fluid flow in three dimensions. Here we show that numerical modeling can not only provide a deeper understanding of the solidification growth patterns during the additive manufacturing, it also serves as a basis for customizing solidification textures which are important for properties and performance of components. PMID:26553246

  16. Method of reprocessing radioactive asphalt solidification products

    International Nuclear Information System (INIS)

    Nakaya, Iwao; Murakami, Tadashi; Miyake, Takafumi; Inagaki, Yuzo.

    1986-01-01

    Purpose: To obtain heat-stable solidification products and decrease the total volume thereof by modifying the solidified form by the reprocessing of existent radioactive asphalt solidification products. Method: Radioactive asphalt solidification products are heated into a fluidized state. Then, incombustible solvents such as perchloroethylene or trichloroethylene are added to a dissolving tank to gradually dissolve the radioactive asphalt solidification products. Thus, organic materials such as asphalts are transferred into the solvent layer, while inorganic materials containing radioactive materials remain as they are in the separation tank. Then, the inorganic materials containing the radioactive materials are taken out and then solidified, for example, by converting them into a rock or glass form. (Kawakami, Y.)

  17. Efficient estimation of diffusion during dendritic solidification

    Science.gov (United States)

    Yeum, K. S.; Poirier, D. R.; Laxmanan, V.

    1989-01-01

    A very efficient finite difference method has been developed to estimate the solute redistribution during solidification with diffusion in the solid. This method is validated by comparing the computed results with the results of an analytical solution derived by Kobayashi (1988) for the assumptions of a constant diffusion coefficient, a constant equilibrium partition ratio, and a parabolic rate of the advancement of the solid/liquid interface. The flexibility of the method is demonstrated by applying it to the dendritic solidification of a Pb-15 wt pct Sn alloy, for which the equilibrium partition ratio and diffusion coefficient vary substantially during solidification. The fraction eutectic at the end of solidification is also obtained by estimating the fraction solid, in greater resolution, where the concentration of solute in the interdendritic liquid reaches the eutectic composition of the alloy.

  18. Improved cement solidification of low and intermediate level radioactive wastes

    International Nuclear Information System (INIS)

    1993-01-01

    Cementation was the first and is still the most widely applied technique for the conditioning of low and intermediate level radioactive wastes. Compared with other solidification techniques, cementation is relatively simple and inexpensive. However, the quality of the final cemented waste forms depends very much on the composition of the waste and the type of cement used. Different kinds of cement are used for different kinds of waste and the compatibility of a specific waste with a specific cement type should always be carefully evaluated. Cementation technology is continuously being developed in order to improve the characteristics of cemented waste in accordance with the increasing requirements for quality of the final solidified waste. Various kinds of additives and chemicals are used to improve the cemented waste forms in order to meet all safety requirements. This report is meant mainly for engineers and designers, to provide an explanation of the chemistry of cementation systems and to facilitate the choice of solidification agents and processing equipment. It reviews recent developments in cementation technology for improving the quality of cemented waste forms and provides a brief description of the various cement solidification processes in use. Refs, figs and tabs

  19. Premature melt solidification during mold filling and its influence on the as-cast structure

    Science.gov (United States)

    Wu, M.; Ahmadein, M.; Ludwig, A.

    2018-03-01

    Premature melt solidification is the solidification of a melt during mold filling. In this study, a numerical model is used to analyze the influence of the pouring process on the premature solidification. The numerical model considers three phases, namely, air, melt, and equiaxed crystals. The crystals are assumed to have originated from the heterogeneous nucleation in the undercooled melt resulting from the first contact of the melt with the cold mold during pouring. The transport of the crystals by the melt flow, in accordance with the socalled "big bang" theory, is considered. The crystals are assumed globular in morphology and capable of growing according to the local constitutional undercooling. These crystals can also be remelted by mixing with the superheated melt. As the modeling results, the evolutionary trends of the number density of the crystals and the volume fraction of the solid crystals in the melt during pouring are presented. The calculated number density of the crystals and the volume fraction of the solid crystals in the melt at the end of pouring are used as the initial conditions for the subsequent solidification simulation of the evolution of the as-cast structure. A five-phase volume-average model for mixed columnar-equiaxed solidification is used for the solidification simulation. An improved agreement between the simulation and experimental results is achieved by considering the effect of premature melt solidification during mold filling. Finally, the influences of pouring parameters, namely, pouring temperature, initial mold temperature, and pouring rate, on the premature melt solidification are discussed.

  20. Combined transuranic-strontium extraction process

    Science.gov (United States)

    Horwitz, E. Philip; Dietz, Mark L.

    1992-01-01

    The transuranic (TRU) elements neptunium, plutonium and americium can be separated together with strontium from nitric acid waste solutions in a single process. An extractant solution of a crown ether and an alkyl(phenyl)-N,N-dialkylcarbanylmethylphosphine oxide in an appropriate diluent will extract the TRU's together with strontium, uranium and technetium. The TRU's and the strontium can then be selectively stripped from the extractant for disposal.

  1. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    Energy Technology Data Exchange (ETDEWEB)

    McDeavitt, Sean M

    2011-04-29

    Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500ºC to 600ºC) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: 1. Hot working fabrication using mechanical alloying and extrusion • Design, fabricate, and assemble extrusion equipment • Extrusion database on DU metal • Extrusion database on U-10Zr alloys • Extrusion database on U-20xx-10Zr alloys • Evaluation and testing of tube sheath metals 2. Low-temperature sintering of U alloys • Design, fabricate, and assemble equipment • Sintering database on DU metal • Sintering database on U-10Zr alloys • Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research & Development (FCR&D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich

  2. Powder Metallurgy of Uranium Alloy Fuels for TRU-Burning Reactors Final Technical Report

    International Nuclear Information System (INIS)

    McDeavitt, Sean M.

    2011-01-01

    Overview Fast reactors were evaluated to enable the transmutation of transuranic isotopes generated by nuclear energy systems. The motivation for this was that TRU isotopes have high radiotoxicity and relatively long half-lives, making them unattractive for disposal in a long-term geologic repository. Fast reactors provide an efficient means to utilize the energy content of the TRUs while destroying them. An enabling technology that requires research and development is the fabrication metallic fuel containing TRU isotopes using powder metallurgy methods. This project focused upon developing a powder metallurgical fabrication method to produce U-Zr-transuranic (TRU) alloys at relatively low processing temperatures (500 C to 600 C) using either hot extrusion or alpha-phase sintering for charecterization. Researchers quantified the fundamental aspects of both processing methods using surrogate metals to simulate the TRU elements. The process produced novel solutions to some of the issues relating to metallic fuels, such as fuel-cladding chemical interactions, fuel swelling, volatility losses during casting, and casting mold material losses. Workscope There were two primary tasks associated with this project: (1) Hot working fabrication using mechanical alloying and extrusion - Design, fabricate, and assemble extrusion equipment - Extrusion database on DU metal - Extrusion database on U-10Zr alloys - Extrusion database on U-20xx-10Zr alloys - Evaluation and testing of tube sheath metals (2) Low-temperature sintering of U alloys - Design, fabricate, and assemble equipment - Sintering database on DU metal - Sintering database on U-10Zr alloys - Liquid assisted phase sintering on U-20xx-10Zr alloys Appendices Outline Appendix A contains a Fuel Cycle Research and Development (FCR and D) poster and contact presentation where TAMU made primary contributions. Appendix B contains MSNE theses and final defense presentations by David Garnetti and Grant Helmreich outlining the

  3. A Novel and Cost Effective Approach to the Decommissioning and Decontamination of Legacy Glove Boxes - Minimizing TRU Waste and Maximizing LLW Waste - 13634

    Energy Technology Data Exchange (ETDEWEB)

    Pancake, Daniel; Rock, Cynthia M.; Creed, Richard [Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439 (United States); Donohoue, Tom; Martin, E. Ray; Mason, John A. [ANTECH Corporation 9050 Marshall Court, Westminster, CO, 80031 (United States); Norton, Christopher J.; Crosby, Daniel [Environmental Alternatives, Inc., 149 Emerald Street, Suite R, Keene, NH 03431 (United States); Nachtman, Thomas J. [InstaCote, Inc., 160 C. Lavoy Road, Erie, MI, 48133 (United States)

    2013-07-01

    This paper describes the process of decommissioning two gloveboxes at the Argonne National Laboratory (ANL) that were employed for work with plutonium and other radioactive materials. The decommissioning process involved an initial phase of clearing tools and materials from the glove boxes and disconnecting them from the laboratory infrastructure. The removed materials, assessed as Transuranic (TRU) waste, were packaged into 55 gallon (200 litre) drums and prepared for ultimate disposal at the Waste Isolation Pilot Plant (WIPP) at Carlsbad New Mexico. The boxes were then sampled to determine the radioactive contents by means of smears that were counted with alpha and beta detectors to determine the residual surface contamination, especially in terms of alpha particle emitters that are an indicator of TRU activity. Paint chip samples were also collected and sent for laboratory analysis in order to ascertain the radioactive contamination contributing to the TRU activity as a fixed contamination. The investigations predicted that it may be feasible to reduce the residual surface contamination and render the glovebox structure low level waste (LLW) for disposal. In order to reduce the TRU activity a comprehensive decontamination process was initiated using chemical compounds that are particularly effective for lifting and dissolving radionuclides that adhere to the inner surfaces of the gloveboxes. The result of the decontamination process was a reduction in the TRU surface activity on the inner surfaces of the gloveboxes by four orders of magnitude in terms of disintegrations per unit area (DPA). The next phase of the process involved a comprehensive assay of the gloveboxes using a combination of passive neutron and gamma ray scintillation detectors and a shielded and collimated high purity Germanium (HPGe) gamma ray detector. The HPGe detector was used to obtain gamma ray spectra for a variety of measurement positions within the glovebox. The spectra were used to

  4. A Novel and Cost Effective Approach to the Decommissioning and Decontamination of Legacy Glove Boxes - Minimizing TRU Waste and Maximizing LLW Waste - 13634

    International Nuclear Information System (INIS)

    Pancake, Daniel; Rock, Cynthia M.; Creed, Richard; Donohoue, Tom; Martin, E. Ray; Mason, John A.; Norton, Christopher J.; Crosby, Daniel; Nachtman, Thomas J.

    2013-01-01

    This paper describes the process of decommissioning two gloveboxes at the Argonne National Laboratory (ANL) that were employed for work with plutonium and other radioactive materials. The decommissioning process involved an initial phase of clearing tools and materials from the glove boxes and disconnecting them from the laboratory infrastructure. The removed materials, assessed as Transuranic (TRU) waste, were packaged into 55 gallon (200 litre) drums and prepared for ultimate disposal at the Waste Isolation Pilot Plant (WIPP) at Carlsbad New Mexico. The boxes were then sampled to determine the radioactive contents by means of smears that were counted with alpha and beta detectors to determine the residual surface contamination, especially in terms of alpha particle emitters that are an indicator of TRU activity. Paint chip samples were also collected and sent for laboratory analysis in order to ascertain the radioactive contamination contributing to the TRU activity as a fixed contamination. The investigations predicted that it may be feasible to reduce the residual surface contamination and render the glovebox structure low level waste (LLW) for disposal. In order to reduce the TRU activity a comprehensive decontamination process was initiated using chemical compounds that are particularly effective for lifting and dissolving radionuclides that adhere to the inner surfaces of the gloveboxes. The result of the decontamination process was a reduction in the TRU surface activity on the inner surfaces of the gloveboxes by four orders of magnitude in terms of disintegrations per unit area (DPA). The next phase of the process involved a comprehensive assay of the gloveboxes using a combination of passive neutron and gamma ray scintillation detectors and a shielded and collimated high purity Germanium (HPGe) gamma ray detector. The HPGe detector was used to obtain gamma ray spectra for a variety of measurement positions within the glovebox. The spectra were used to

  5. Electric melting furnace for waste solidification

    International Nuclear Information System (INIS)

    Masaki, Toshio.

    1990-01-01

    To avoid electric troubles or reduction of waste processing performance even when platinum group elements are contained in wastes to be applied with glass solidification. For this purpose, a side electrode is disposed to the side wall of a melting vessel and a central electrode serving as a counter electrode is disposed about at the center inside the melting vessel. With such a constitution, if conductive materials are deposited at the bottom of the furnace or the bottom of the melting vessel, heating currents flow selectively between the side electrode and the central electrode. Accordingly, no electric currents flow through the conductive deposits thereby enabling to prevent abnormal heating in the bottom of the furnace. Further, heat generated by electric supply between the side electrode and the central electrode is supplied efficiently to raw material on the surface of the molten glass liquid to improve the processing performance. Further, disposition of the bottom electrode at the bottom of the furnace enables current supply between the central electrode and the bottom electrode to facilitate the temperature control for the molten glass in the furnace than in the conventional structure. (I.S.)

  6. On the role of solidification modelling in Integrated Computational Materials Engineering “ICME”

    International Nuclear Information System (INIS)

    Schmitz, G J; Böttger, B; Apel, M

    2016-01-01

    Solidification during casting processes marks the starting point of the history of almost any component or product. Integrated Computational Materials Engineering (ICME) [1-4] recognizes the importance of further tracking the history of microstructure evolution along the subsequent process chain. Solidification during joining processes in general happens quite late during production, where the parts to be joined already have experienced a number of processing steps which affected their microstructure. Reliable modelling of melting and dissolution of these microstructures represents a key issue before eventually modelling ‘re’-solidification e.g. during welding or soldering. Some instructive examples of microstructure evolution during a joining process obtained on the basis of synthetic and simulated initial microstructures of an Al-Cu binary model system are discussed. (paper)

  7. Solidification of high-level radioactive wastes. Final report

    International Nuclear Information System (INIS)

    1979-06-01

    A panel on waste solidification was formed at the request of the Nuclear Regulatory Commission to study the scientific and technological problems associated with the conversion of liquid and semiliquid high-level radioactive wastes into a stable form suitable for transportation and disposition. Conclusions reached and recommendations made are as follows. Many solid forms described in this report could meet standards as stringent as those currently applied to the handling, storage, and transportation of spent fuel assemblies. Solid waste forms should be selected only in the context of the total radioactive waste management system. Many solid forms are likely to be satisfactory for use in an appropriately designed system, The current United States policy of deferring the reprocessing of commercial reactor fuel provides additional time for R and D solidification technology for this class of wastes. Defense wastes which are relatively low in radioactivity and thermal power density can best be solidified by low-temperature processes. For solidification of fresh commercial wastes that are high in specific activity and thermal power density, the Panel recommends that, in addition to glass, the use of fully-crystalline ceramics and metal-matrix forms be actively considered. Preliminary analysis of the characteristics of spent fuel pins indicates that they may be eligible for consideration as a waste form. Because the differences in potential health hazards to the public resulting from the use of various solid form and disposal options are likely to be small, the Panel concludes that cost, reliability, and health hazards to operating personnel will be major considerations in choosing among the options that can meet safety requiremens. The Panel recommends that responsibility for all radioactive waste management operations (including solidification R and D) should be centralized

  8. Solidification processing method for radioactive waste

    International Nuclear Information System (INIS)

    Hiraki, Akimitsu; Tanaka, Keiji; Heta, Katsutoshi.

    1991-01-01

    The pressure in a vessel containing radioactive wastes is previously reduced and cement mortar prepared by kneading cement, sand and kneading agent with water is poured under shaking substantially to the upper end of the vessel. After the lowering of the mortar level due to the deforming has been terminated, the pressure is increased gradually. Then, the cement mortar is further poured substantially to the upper end of the vessel again. With such a two step pouring method, spaces other than the radioactive wastes in the vessel can be filled substantially completely with the cement mortar. Accordingly, it is possible to avoid the problem in view of the strength due to the formation of gaps at the inside of the vessel, or leaching of radioactive materials due to the intrusion of water into the gaps. Further, if washing water is reutilized as water for kneading or washing after the precipitation of the solid contents, the amount of the secondary wastes generated can be reduced. (T.M.)

  9. Plutonium (TRU) transmutation and {sup 233}U production by single-fluid type accelerator molten-salt breeder (AMSB)

    Energy Technology Data Exchange (ETDEWEB)

    Furukaw, Kazuo [Tokai Univ., Kanagawa (Japan); Kato, Yoshio [Japan Atom. Ene. Res. Inst., Ibaraki (Japan); Chigrinov, Sergey E. [Academy of Science, Minsk (Belarus)

    1995-10-01

    For practical/industrial disposition of Pu(TRU) by accelerator facility, not only physical soundness and safety but also the following technological rationality should be required: (1) few R&D items including radiation damage, heat removal and material compatibility; (2) few operation/maintenance/processing works: (3) few reproduction of radioactivity; (4) effective energy production in parallel. This will be achieved by the new modification of Th-fertilizing Single-Fluid type Accelerator Molten-Salt Breeder (AMSB), by which a global nuclear energy strategy for next century might be prepared.

  10. X-Ray Radiographic Observation of Directional Solidification Under Microgravity: XRMON-GF Experiments on MASER12 Sounding Rocket Mission

    Science.gov (United States)

    Reinhart, G.; NguyenThi, H.; Bogno, A.; Billia, B.; Houltz, Y.; Loth, K.; Voss, D.; Verga, A.; dePascale, F.; Mathiesen, R. H.; hide

    2012-01-01

    The European Space Agency (ESA) - Microgravity Application Promotion (MAP) programme entitled XRMON (In situ X-Ray MONitoring of advanced metallurgical processes under microgravity and terrestrial conditions) aims to develop and perform in situ X-ray radiography observations of metallurgical processes in microgravity and terrestrial environments. The use of X-ray imaging methods makes it possible to study alloy solidification processes with spatio-temporal resolutions at the scales of relevance for microstructure formation. XRMON has been selected for MASER 12 sounding rocket experiment, scheduled in autumn 2011. Although the microgravity duration is typically six minutes, this short time is sufficient to investigate a solidification experiment with X-ray radiography. This communication will report on the preliminary results obtained with the experimental set-up developed by SSC (Swedish Space Corporation). Presented results dealing with directional solidification of Al-Cu confirm the great interest of performing in situ characterization to analyse dynamical phenomena during solidification processes.

  11. TECHNOLOGY EVALUATION REPORT: SILICATE TECHNOLOGY CORPORATION - SOLIDIFICATION/STABILIZATION OF PCP AND INORGANIC CONTAMINANTS IN SOILS - SELMA, CA

    Science.gov (United States)

    This Technolgy Evaluation Report evaluates the solidification/stabilization process of Silicate Technology Corporation (STC) for the on-site treatment of contaminated soil The STC immobilization technology uses a proprietary product (FMS Silicate) to chemically stabilize and ...

  12. Feasibility analysis of constant TRU feeding in waste transmutation system using accelerator-driven subcritical system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kun Jai; Cho, Nam Zin; Jo, Chang Keun; Park, Chang Je; Kim, Do Sam; Park, Jeong Hwan [Korea Advanced Institute of Science and Technology, Taejon (Korea)

    1999-03-01

    It is probable that the issue of nuclear spent fuel and high-level waste can have negative impact on the future expansion of nuclear power programs. Accelerator-driven nuclear waste transmutation with constant composition TRU feeding which satisfies non-proliferation condition will help establish the long-range nuclear waste disposal strategy. In this study, current status of accelerator-driven transmutation of waste technology, and feasibility analysis of constant composition TRU feeding system were investigated. We ascertained that solid system using constant composition TRU is feasible with the the capability of transmutation. (author). 13 refs., 53 figs., 20 tabs.

  13. Properties and solidification of decontamination wastes

    International Nuclear Information System (INIS)

    Davis, M.S.; Piciulo, P.L.; Bowerman, B.S.; Adams, J.W.; Milian, L.

    1983-01-01

    LWRs will require one or more chemical decontaminations to achieve their designed lifetimes. Primary system decontamination is designed to lower radiation fields in areas where plant maintenance personnel must work. Chemical decontamination methods are either hard (concentrated chemicals, approximately 5 to 25 weight percent) or soft (dilute chemicals less than 1 percent by weight). These methods may have different chemical reagents, some tailor-made to the crud composition and many methods are and will be proprietary. One factor common to most commercially available processes is the presence of organic acids and chelates. These types of organic reagents are known to enhance the migration of radionuclides after disposal in a shallow land burial site. The NRC sponsors two programs at Brookhaven National Laboratory that are concerned with the management of decontamination wastes which will be generated by the full system decontamination of LWRs. These two programs focus on potential methods for degrading or converting decontamination wastes to more acceptable forms prior to disposal and the impact of disposing of solidified decontamination wastes. The results of the solidification of simulated decontamination resin wastes will be presented. Recent results on combustion of simulated decontamintion wastes will be described and procedures for evaluating the release of decontamination reagents from solidified wastes will be summarized

  14. Interface Pattern Selection in Directional Solidification

    Science.gov (United States)

    Trivedi, Rohit; Tewari, Surendra N.

    2001-01-01

    The central focus of this research is to establish key scientific concepts that govern the selection of cellular and dendritic patterns during the directional solidification of alloys. Ground-based studies have established that the conditions under which cellular and dendritic microstructures form are precisely where convection effects are dominant in bulk samples. Thus, experimental data can not be obtained terrestrially under pure diffusive regime. Furthermore, reliable theoretical models are not yet possible which can quantitatively incorporate fluid flow in the pattern selection criterion. Consequently, microgravity experiments on cellular and dendritic growth are designed to obtain benchmark data under diffusive growth conditions that can be quantitatively analyzed and compared with the rigorous theoretical model to establish the fundamental principles that govern the selection of specific microstructure and its length scales. In the cellular structure, different cells in an array are strongly coupled so that the cellular pattern evolution is controlled by complex interactions between thermal diffusion, solute diffusion and interface effects. These interactions give infinity of solutions, and the system selects only a narrow band of solutions. The aim of this investigation is to obtain benchmark data and develop a rigorous theoretical model that will allow us to quantitatively establish the physics of this selection process.

  15. Method of manufacturing borosilicate glass solidification products

    International Nuclear Information System (INIS)

    Tanaka, Tsuneya.

    1986-01-01

    Purpose: To obtain glass solidification products efficiently in a dry process from medium and high level radioactive liquid wastes discharged from PWR type reactors. Method: Boric acid-containing radioactive liquid wastes generated from primary coolants of PWR type reactors are evaporated to condensate as the pre-treatment. The concentrated liquid wastes are supplied to a drum type rotary kiln. While on the other hand, usual glass frits are introduced into the kiln. The liquid wastes are dried in the rotary kiln, as well as B 2 O 3 and the glass frits in the liquid wastes are combined into glass particles. In this case, since the kiln is rotated, no glass particles are deposited on the wall of the kiln. Then, the glass particles are introduced for melting into a high frequency melting furnace made of metal. The melting temperature is set to 1100 - 1150 deg C. The molten borosilicate glass is recovered from the bottom of the melting furance, contained in a canister and cooled for several hours, and then a cover is welded to the canister. (Ikeda, J.)

  16. Divorced Eutectic Solidification of Mg-Al Alloys

    Science.gov (United States)

    Monas, Alexander; Shchyglo, Oleg; Kim, Se-Jong; Yim, Chang Dong; Höche, Daniel; Steinbach, Ingo

    2015-08-01

    We present simulations of the nucleation and equiaxed dendritic growth of the primary hexagonal close-packed -Mg phase followed by the nucleation of the -phase in interdendritic regions. A zoomed-in region of a melt channel under eutectic conditions is investigated and compared with experiments. The presented simulations allow prediction of the final properties of an alloy based on process parameters. The obtained results give insight into the solidification processes governing the microstructure formation of Mg-Al alloys, allowing their targeted design for different applications.

  17. Biópsia hepática com agulha tru-cut guiada por videolaparoscopia em caprinos Videolaparoscopic guided hepatic biopsy with tru-cut needle in goats

    Directory of Open Access Journals (Sweden)

    A.L.L. Duarte

    2009-02-01

    Full Text Available Descreve-se a técnica de biópsia hepática com agulha tru-cut guiada por videolaparoscopia em 12 caprinos machos, castrados, hígidos, sem raça definida, distribuídos em dois grupos (G: G1, com cinco animais de 12 meses e G2, com sete de seis meses de idade. O procedimento foi realizado sob anestesia geral intravenosa com o animal em decúbito lateral esquerdo. O pneumoperitônio e a laparoscopia foram procedidos no flanco direito, aproximadamente a 10cm ventral aos processos transversos das vértebras lombares. A agulha tru-cut foi introduzida no 11º espaço intercostal direito, a aproximadamente 12cm ventral à coluna vertebral, para punção e remoção de fragmento do lobo hepático direito. O tempo operatório médio foi de 23 minutos e cinco segundos. A hemorragia causada pela perfuração hepática cessou em dois minutos em 75% dos animais e em três minutos, nos 25% restantes. Nas avaliações clínicas feitas no pré-jejum e às 24, 48 e 72 horas após a biópsia, não foram observadas alterações (P>0,05 da temperatura retal, das frequências cardíaca e respiratória e dos movimentos rumenais nos dois grupos. A biópsia hepática com agulha tru-cut guiada por videolaparoscopia foi considerada eficaz para uso em caprinos, permitindo a obtenção de fragmentos hepáticos suficientes para exame histológico.The videolaparoscopic guided hepatic biopsy with tru-cut needle is described in 12 healthy, no defined breed, castrated male goats. Animals were distributed in two groups: G1 (n=5 12-month-old animals; and G2 (n=7 six-month-old animals. The procedure was performed with the animal in left lateral recumbency and under total intravenous anesthesia. Pneumoperitoneum and laparoscopy were performed in the right flank, approximately 10cm ventral to the transverse processes of the lumbar vertebrae. The tru-cut needle was inserted into the right eleventh intercostal space, around 12cm ventral to the spinal column, for punching and

  18. Cement solidification of spent ion exchange resins produced by the nuclear industry

    International Nuclear Information System (INIS)

    Jaouen, C.; Vigreux, B.

    1988-01-01

    Cement solidification technology has been applied to spent ion exchange resins for many years in countries throughout the world (at reactors, research centers and spent fuel reprocessing plants). Changing specifications for storage of radioactive waste have, however, confronted the operators of such facilities with a number of problems. Problems related both to the cement solidification process (water/cement/resin interactions and chemical interactions) and to its utilization (mixing, process control, variable feed composition, etc.) have often led waste producers to prefer other, polymer-based processes, which are very expensive and virtually incompatible with water. This paper discusses research on cement solidification of ion exchange resins since 1983 and the development of application technologies adapted to nuclear service conditions and stringent finished product quality requirements

  19. Characterization of a sodium-cooled fast reactor in an MHR-SFR synergy for TRU transmutation

    International Nuclear Information System (INIS)

    Hong, Ser Gi; Kim, Yonghee; Venneri, Francesco

    2008-01-01

    In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50-65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR-SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core

  20. Report of conceptual design for TRU solid waste facilities adjacent to 200H Area: Savannah River Plant

    International Nuclear Information System (INIS)

    1978-02-01

    Facilities for consolidating Savannah River Plant solid transuranic (TRU) waste and placing in long-term safe, retrievable storage have been designed conceptually. A venture guidance appraisal of cost for the facilities has been prepared. The proposed site of the new processing area is adjacent to existing H Area facilities. The scopes of work comprising the conceptual design describe facilities for: exhuming high-level TRU waste from buried and pad-stored locations in the plant burial ground; opening, emptying, and sorting waste containers and their contents within shielded, regulated enclosures; volume-reducing the noncombustibles by physical processes and decontaminating the metal waste; burning combustibles; fixing the consolidated waste forms in a concrete matrix within a double-walled steel container; placing product containers in a retrievable surface storage facility adjacent to the existing plant burial ground; and maintaining accountability of all special nuclear materials. Processing, administration, and auxiliary service buildings are to be located adjacent to existing H Area facilities where certain power and waste liquid services will be shared

  1. Solidification microstructures of aluminium-uranium alloys

    International Nuclear Information System (INIS)

    Ambrozio Filho, F.; Vieira, R.R.

    1976-01-01

    The solidification of microstrutures of aluminium-uranium alloys in the range of 4 to 20% uranium is investigated. The solidification was obtained both in ingot molds and under controlled directional solidification. The conditions for the presence of primary crystals and eutectic are discussed and an analysis of the influence of variables (growth rate and thermal gradient in the liquid) on the alloy structure is made. The effect of cooling rate on the alloy structures has been determined. It is found that the resulting structure can be derived from the kinectics concept, as required by the coupled-zone theory. Suggestions on the qualitative intervals of composition and temperatures with eutectic growth are presented [pt

  2. Enthalpies of a binary alloy during solidification

    Science.gov (United States)

    Poirier, D. R.; Nandapurkar, P.

    1988-01-01

    The purpose of the paper is to present a method of calculating the enthalpy of a dendritic alloy during solidification. The enthalpies of the dendritic solid and interdendritic liquid of alloys of the Pb-Sn system are evaluated, but the method could be applied to other binaries, as well. The enthalpies are consistent with a recent evaluation of the thermodynamics of Pb-Sn alloys and with the redistribution of solute in the same during dendritic solidification. Because of the heat of mixing in Pb-Sn alloys, the interdendritic liquid of hypoeutectic alloys (Pb-rich) of less than 50 wt pct Sn has enthalpies that increase as temperature decreases during solidification.

  3. A comparative neutronic analysis of 150MWe TRU burner according to the coolant alteration

    International Nuclear Information System (INIS)

    Yoo, J. W.; Kim, S. J.; Kim, Y. I.

    2000-01-01

    A comparative neutronic analysis has been conducted for the small TRU burner according to their coolant material. The use of Pb-Bi coolant gave a low burnup reactivity swing and negative or less positive coolant void coefficient with harder neutron spectrum. By a lower burnup reactivity swing and higher conversion ratio of Pb-Bi cooled core, the total amount of TRU consumption was found to be small compared with Na cooled core despite of the higher MA consumption ratio of Pb-Bi cooled core. However, Pb-Bi cooled reactor have a lager margin in the coolant void coefficient, so that a variable MA composition can be loaded in the core. Accordingly, even though the Pb-Bi cooled TRU burner has not effectiveness on TRU burning in the same geometry and material condition, a flexible MA loading is envisaged to result in 10 times larger MA burning amount, still preserving a low coolant void worth

  4. Radioactive waste processing device

    International Nuclear Information System (INIS)

    Ikeda, Takashi; Funabashi, Kiyomi; Chino, Koichi.

    1992-01-01

    In a waste processing device for solidifying, pellets formed by condensing radioactive liquid wastes generated from a nuclear power plant, by using a solidification agent, sodium chloride, sodium hydroxide or sodium nitrate is mixed upon solidification. In particular, since sodium sulfate in a resin regenerating liquid wastes absorbs water in the cement upon cement solidification, and increases the volume by expansion, there is a worry of breaking the cement solidification products. This reaction can be prevented by the addition of sodium chloride and the like. Accordingly, integrity of the solidification products can be maintained for a long period of time. (T.M.)

  5. Field tests on migration of TRU-nuclide, (1). General introduction

    International Nuclear Information System (INIS)

    Ogawa, Hiromichi; Tanaka, Tadao; Mukai, Masayuki

    2003-01-01

    The field migration test using TRU nuclide was carried out as a cooperative research project between JAERI (Japan Atomic Energy Research Institute) and CIRP (China Institute for Radiation Protection). This report introduced the out-line of the field migration test and described the outline of the series of 'Field Test on Migration of TRU-nuclide' and main results as a summary report. (author)

  6. Plasma processing of compacted drums of simulated radioactive waste

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  7. Plastic solidification system for radioactive waste

    International Nuclear Information System (INIS)

    Kani, Jiro; Irie, Hiromitsu; Obu, Etsuji; Nakayama, Yasuyuki; Matsuura, Hiroyuki.

    1979-01-01

    The establishment of a new solidification system is an important theme for recent radioactive-waste disposal systems. The conditions required of new systems are: (1) the volume of the solidified product to be reduced, and (2) the property of the solidified product to be superior to the conventional ones. In the plastic solidification system developed by Toshiba, the waste is first dried and then solidified with thermosetting resin. It has been confirmed that the property of the plastic solidified product is superior to that of the cement-or bitumen-solidified product. Investigation from various phases is being carried on for the application of this method to commercial plants. (author)

  8. Powder-Metallurgy Process And Product

    Science.gov (United States)

    Paris, Henry G.

    1988-01-01

    Rapid-solidification processing yields alloys with improved properties. Study undertaken to extend favorable property combinations of I/M 2XXX alloys through recently developed technique of rapid-solidification processing using powder metallurgy(P/M). Rapid-solidification processing involves impingement of molten metal stream onto rapidly-spinning chill block or through gas medium using gas atomization technique.

  9. Analisis Penerapan Metode Transmitter Receiver Unit (TRU Upgrading Untuk Mengatasi Traffic Congestion Jaringan GSM Pada BTS Area Purwokerto Kota

    Directory of Open Access Journals (Sweden)

    Alfin Hikmaturokhman

    2011-05-01

    Full Text Available Semakin banyaknya pengguna selular maka akan semakin banyak trafik yang akan tertampung. Trafik yang melebihi kapasitas kanal yang disediakan dapat menyebabkan kondisi Traffic Congestion. Untuk menanganinya diperlukan metode penambahan kapasitas kanal agar semua trafik dapat tertampung dengan baik. Metode ini disebut dengan TRU Upgrading. Transmitter Receiver Unit (TRU adalah hardware yang terletak pada Radio Base Station dalam BTS yang berisi slot-slot kanal sedangkan metode TRU Upgrading adalah metode dengan menambahkan/upgrade kapasitas kanal yang tersedia dari konfigurasi TRU yang telah ada sebelumnya, misalkan pada BTS Pabuaran memiliki konfigurasi 3x2x3 karena terjadi kejenuhan pelanggan maka konfigurasi TRU diupgrade menjadi 3x4x3. Perubahan konfigurasi TRU maka merubah konfigurasi BTS-nya serta menambah kapasitas kanalnya. Key Performance Indicator (KPI yang baik pada Indosat adalah menggunakan batas GoS 2%. Nilai GoS ini dikaitkan dengan tabel Erlang untuk mendapatkan sebuah nilai intensitas trafik. Jika nilai intensitas trafik konfigurasi TRU yang digunakan kurang dari nilai intensitas trafik pelanggan maka disebut traffic congestion. Sebagai akibat dari traffic congestion adalah kondisi blocking. TRU Upgrading ini dilakukan dengan harapan nilai blocking panggilan menjadi 0 %. Pada Purwokerto kota, diterapkan  TRU Upgrading untuk cell Grendeng 3, Pabuaran 2, dan Unsoed 1 karena trafik pelanggan yang terjadi melebihi nilai intensitas trafik dari konfigurasi TRU yang digunakan.   Untuk cell Unsoed 1 dan Grendeng 3 meski telah dilakukan TRU Upgrading menjadi 4 buah TRU tetap terjadi traffic congestion sebesar 8 sampai dengan 15 Erlang dikarenakan pada cell-cell ini mengcover area yang padat penduduk. Sedang untuk Pabuaran 2 penerapan TRU upgrading mencapai keefektifan sebesar 100%.

  10. Effects of conversion ratio change on the core performances in medium to large TRU burning reactors

    International Nuclear Information System (INIS)

    Song, Hoon; Kim, Sang-Ji; Yoo, Jae-Woon; Kim, Yeong-Il

    2009-01-01

    Conceptual fast reactor core designs with sodium coolant are developed at 1,500, 3,000 and 4,500 MWt which are configured to transmute recycled transuranics (TRU) elements with external feeds consisting of LWR spent fuel. Even at each pre-determined power level, the performance parameters, reactivity coefficients and their implications on the safety analysis can be different when the target TRU conversion ratio changes. In order to address this aspect of design, a study on TRU conversion ratio change was performed. The results indicate that it is feasible to design a TRU burner core to accommodate a wide range of conversion ratios by employing different fuel cladding thicknesses. The TRU consumption rate is found to be proportional to the core power without any significant deterioration in the core performance at higher power levels. A low conversion ratio core has an increased TRU consumption rate and much faster burnup reactivity loss, which calls for appropriate means for reactivity compensation. As for the reactivity coefficients related with the conversion ratio change, the core with a low conversion ratio has a less negative Doppler coefficient, a more negative axial expansion coefficient, a more negative control rod worth per rod, a more negative radial expansion coefficient, a less positive sodium density coefficient and a less positive sodium void worth. A slight decrease in the delayed neutron fraction is also noted, reflecting the fertile U-238 fraction reduction. (author)

  11. Fractal growth in impurity-controlled solidification in lipid monolayers

    DEFF Research Database (Denmark)

    Fogedby, Hans C.; Sørensen, Erik Schwartz; Mouritsen, Ole G.

    1987-01-01

    A simple two-dimensional microscopic model is proposed to describe solidifcation processes in systems with impurities which are miscible only in the fluid phase. Computer simulation of the model shows that the resulting solids are fractal over a wide range of impurity concentrations and impurity...... diffusional constants. A fractal-forming mechanism is suggested for impurity-controlled solidification which is consistent with recent experimental observations of fractal growth of solid phospholipid domains in monolayers. The Journal of Chemical Physics is copyrighted by The American Institute of Physics....

  12. Quality assurance procedures for the analysis of TRU waste samples

    International Nuclear Information System (INIS)

    Glasgow, D.C. Giaquinto, J.M.; Robinson, L.

    1995-01-01

    The Waste Isolation Pilot Plant (WIPP) project was undertaken in response to the growing need for a national repository for transuranic (TRU) waste. Guidelines for WIPP specify that any waste item to be interred must be fully characterized and analyzed to determine the presence of chemical compounds designated hazardous and certain toxic elements. The Transuranic Waste Characterization Program (TWCP) was launched to develop analysis and quality guidelines, certify laboratories, and to oversee the actual waste characterizations at the laboratories. ORNL is participating in the waste characterization phase and brings to bear a variety of analytical techniques including ICP-AES, cold vapor atomic absorption, and instrumental neutron activation analysis (INAA) to collective determine arsenic, cadmium, barium, chromium, mercury, selenium, silver, and other elements. All of the analytical techniques involved participate in a cooperative effort to meet the project objectives. One important component of any good quality assurance program is determining when an alternate method is more suitable for a given analytical problem. By bringing to bear a whole arsenal of analytical techniques working toward common objectives, few analytical problems prove to be insurmountable. INAA and ICP-AES form a powerful pair when functioning in this cooperative manner. This paper will provide details of the quality assurance protocols, typical results from quality control samples for both INAA and ICP-AES, and detail method cooperation schemes used

  13. Containerless solidification of BiFeO3 oxide under microgravity

    Science.gov (United States)

    Yu, Jianding; Arai, Yasutomo; Koshikawa, Naokiyo; Ishikawa, Takehito; Yoda, Shinichi

    1999-07-01

    Containerless solidification of BiFeO3 oxide has been carried out under microgravity with Electrostatic Levitation Furnace (ELF) aboard on the sounding rocket (TR-IA). It is a first containerless experiment using ELF under microgravity for studying the solidification of oxide insulator material. Spherical BiFeO3 sample with diameter of 5mm was heated by two lasers in oxygen and nitrogen mixing atmosphere, and the sample position by electrostatic force under pinpoint model and free drift model. In order to compare the solidification behavior in microgravity with on ground, solidification experiments of BiFeO3 in crucible and drop tube were carried out. In crucible experiment, it was very difficult to get single BiFeO3 phase, because segregation of Fe2O3 occured very fast and easily. In drop tube experiment, fine homogeneous BiFeO3 microstructure was obtained in a droplet about 300 μm. It implies that containerless processing can promote the phase selection in solidification. In microgravity experiment, because the heating temperature was lower than that of estimated, the sample was heated into Fe2O3+liquid phase region. Fe2O3 single crystal grew on the surface of the spherical sample, whose sample was clearly different from that observed in ground experiments.

  14. Functional Nanoclay Suspension for Printing-Then-Solidification of Liquid Materials.

    Science.gov (United States)

    Jin, Yifei; Compaan, Ashley; Chai, Wenxuan; Huang, Yong

    2017-06-14

    Additive manufacturing (AM) enables the freeform fabrication of complex structures from various build materials. The objective of this study is to develop a novel Laponite nanoclay-enabled "printing-then-solidification" additive manufacturing approach to extrude complex three-dimensional (3D) structures made of various liquid build materials. Laponite, a member of the smectite mineral family, is investigated to serve as a yield-stress support bath material for the extrusion printing of liquid build materials. Using the printing-then-solidification approach, the printed structure remains liquid and retains its shape with the help of the Laponite support bath. Then the completed liquid structures are solidified in situ by applying suitable cross-linking mechanisms. Finally, the solidified structures are harvested from the Laponite nanoclay support bath for any further processing as needed. Due to its chemical and physical stability, liquid build materials with different solidification/curing/gelation mechanisms can be fabricated in the Laponite bath using the printing-then-solidification approach. The feasibility of the proposed Laponite-enabled printing-then-solidification approach is demonstrated by fabricating several complicated structures made of various liquid build materials, including alginate with ionic cross-linking, gelatin with thermal cross-linking, and SU-8 with photo-cross-linking. During gelatin structure printing, living cells are included and the postfabrication cell viability is above 90%.

  15. Effect of Chemical Composition on Susceptibility to Weld Solidification Cracking in Austenitic Weld Metal

    Science.gov (United States)

    Kadoi, Kota; Shinozaki, Kenji

    2017-12-01

    The influence of the chemical composition, especially the niobium content, chromium equivalent Creq, and nickel equivalent Nieq, on the weld solidification cracking susceptibility in the austenite single-phase region in the Schaeffler diagram was investigated. Specimens were fabricated using the hot-wire laser welding process with widely different compositions of Creq, Nieq, and niobium in the region. The distributions of the susceptibility, such as the crack length and brittle temperature range (BTR), in the Schaeffler diagram revealed a region with high susceptibility to solidification cracking. Addition of niobium enhanced the susceptibility and changed the distribution of the susceptibility in the diagram. The BTR distribution was in good agreement with the distribution of the temperature range of solidification (Δ T) calculated by solidification simulation based on Scheil model. Δ T increased with increasing content of alloying elements such as niobium. The distribution of Δ T was dependent on the type of alloying element owing to the change of the partitioning behavior. Thus, the solidification cracking susceptibility in the austenite single-phase region depends on whether the alloy contains elements. The distribution of the susceptibility in the region is controlled by the change in Δ T and the segregation behavior of niobium with the chemical composition.

  16. Solidification of ash from incineration of low-level radioactive waste

    International Nuclear Information System (INIS)

    Roberson, W.A.; Albenesius, E.L.; Becker, G.W.

    1983-01-01

    The safe disposal of both high-level and low-level radioactive waste is a problem of increasing national attention. A full-scale incineration and solidification process to dispose of suspect-level and low-level beta-gamma contaminated combustible waste is being demonstrated at the Savannah River Plant (SRP) and Savannah River Laboratory (SRL). The stabilized wasteform generated by the process will meet or exceed all future anticipated requirements for improved disposal of low-level waste. The incineration process has been evaluated at SRL using nonradioactive wastes, and is presently being started up in SRP to process suspect-level radioactive wastes. A cement solidification process for incineration products is currently being evaluated by SRL, and will be included with the incineration process in SRP during the winter of 1984. The GEM alumnus author conducted research in a related disposal solidification program during the GEM-sponsored summer internship, and upon completion of the Masters program, received full-time responsibility for developing the incineration products solidification process

  17. Sufficient condition for generation of multiple solidification front in one-dimensional solidification of binary alloys

    International Nuclear Information System (INIS)

    Bobula, E.; Kalicka, Z.

    1981-10-01

    In the paper we consider the one-dimensional solidification of binary alloys in the finite system. The authors present the sufficient condition for solidification in the liquid in front of the moving solid-liquid interface. The effect may produce a fluctuating concentration distributin in the solid. The convection in the liquid and supercooling required for homogeneous nucleation are omitted. A local-equilibrium approximation at the liquid-solid interface is supposed. (author)

  18. Documentation of acceptable knowledge for Los Alamos National Laboratory Plutonium Facility TRU waste stream

    International Nuclear Information System (INIS)

    Montoya, A.J.; Gruetzmacher, K.M.; Foxx, C.L.; Rogers, P.Z.

    1998-03-01

    Characterization of transuranic waste from the LANL Plutonium Facility for certification and transportation to WIPP includes the use of acceptable knowledge as specified in the WIPP Quality Assurance Program Plan. In accordance with a site specific procedure, documentation of acceptable knowledge for retrievably stored and currently generated transuranic waste streams is in progress at LANL. A summary overview of the TRU waste inventory is complete and documented in the Sampling Plan. This document also includes projected waste generation, facility missions, waste generation processes, flow diagrams, times, and material inputs. The second part of acceptable knowledge documentation consists of assembling more detailed acceptable knowledge information into auditable records and is expected to require several years to complete. These records for each waste stream must support final assignment of waste matrix parameters, EPA hazardous waste numbers, and radionuclide characterization. They must also include a determination whether waste streams are defense waste streams for compliance with the WIPP Land Withdrawal Act. The LANL Plutonium Facility's mission is primarily plutonium processing in basic special nuclear material (SNM) research activities to support national defense and energy programs. It currently has about 100 processes ranging from SNM recovery from residues to development of plutonium 238 heat sources for space applications. Its challenge is to characterize and certify waste streams from such diverse and dynamic operations using acceptable knowledge. This paper reports the progress on the certification of the first of these waste streams to the WIPP WAC

  19. W-026, transuranic waste restricted waste management (TRU RWM) glovebox operational test report

    Energy Technology Data Exchange (ETDEWEB)

    Leist, K.J.

    1998-02-18

    The TRU Waste/Restricted Waste Management (LLW/PWNP) Glovebox 401 is designed to accept and process waste from the Transuranic Process Glovebox 302. Waste is transferred to the glovebox via the Drath and Schraeder Bagless Transfer Port (DO-07401) on a transfer stand. The stand is removed with a hoist and the operator inspects the waste (with the aid of the Sampling and Treatment Director) to determine a course of action for each item. The waste is separated into compliant and non compliant. One Trip Port DO-07402A is designated as ``Compliant``and One Trip Port DO-07402B is designated as ``Non Compliant``. As the processing (inspection, bar coding, sampling and treatment) of the transferred items takes place, residue is placed in the appropriate One Trip port. The status of the waste items is tracked by the Data Management System (DMS) via the Plant Control System (PCS) barcode interface. As an item is moved for sampling or storage or it`s state altered by treatment, the Operator will track an items location using a portable barcode reader and entry any required data on the DMS console. The Operational Test Procedure (OTP) will perform evolutions (described here) using the Plant Operating Procedures (POP) in order to verify that they are sufficient and accurate for controlled glovebox operation.

  20. W-026, transuranic waste restricted waste management (TRU RWM) glovebox operational test report

    International Nuclear Information System (INIS)

    Leist, K.J.

    1998-01-01

    The TRU Waste/Restricted Waste Management (LLW/PWNP) Glovebox 401 is designed to accept and process waste from the Transuranic Process Glovebox 302. Waste is transferred to the glovebox via the Drath and Schraeder Bagless Transfer Port (DO-07401) on a transfer stand. The stand is removed with a hoist and the operator inspects the waste (with the aid of the Sampling and Treatment Director) to determine a course of action for each item. The waste is separated into compliant and non compliant. One Trip Port DO-07402A is designated as ''Compliant''and One Trip Port DO-07402B is designated as ''Non Compliant''. As the processing (inspection, bar coding, sampling and treatment) of the transferred items takes place, residue is placed in the appropriate One Trip port. The status of the waste items is tracked by the Data Management System (DMS) via the Plant Control System (PCS) barcode interface. As an item is moved for sampling or storage or it's state altered by treatment, the Operator will track an items location using a portable barcode reader and entry any required data on the DMS console. The Operational Test Procedure (OTP) will perform evolutions (described here) using the Plant Operating Procedures (POP) in order to verify that they are sufficient and accurate for controlled glovebox operation

  1. Solidification at the micro-scale

    International Nuclear Information System (INIS)

    Howe, A.

    2003-01-01

    The experimental determination and computer simulation of the micro-segregation accompanying the solidification of alloys continues to be a subject of much academic and industrial interest. Both are subject to progressively more sophisticated analyses, and a discussion is offered regarding the development and practical use of such studies. Simple steels are particularly difficult targets for such work: solidification does not end conveniently in a eutectic, the rapid diffusion particularly in the delta-ferrite phase obscures most evidence of what had occurred at the micro-scale during solidification, and one or more subsequent solid state phase transformations further obscure such details. Also, solidification at the micro-scale is inherently variable: the usual, dendrite morphologies encountered are, after all, instabilities in growth behaviour, and therefore such variability should be expected. For questions such as the relative susceptibility of different grades to particular problems, it is the average, typical behaviour that is of interest, whereas for other questions such as the on-set of macro-segregation, the local variability is paramount. Depending on the question being asked, and indeed the accuracy with which validatory data are available, simple pseudo-analytical equations employing various limiting assumptions, or sophisticated models which remove the need for most such limitations, could be appropriate. This paper highlights the contribution to such studies of various collaborative research forums within the European Union with which the author is involved. (orig.) [de

  2. Solidification of low-level waste - a dilemma for the small user

    International Nuclear Information System (INIS)

    Harris, S.; Gilmore, A.

    1980-01-01

    The requirement that radioactive waste for sea disposal must be solidified by the originator is discussed. Attempts to solidify small quantities of radioactive waste such as contaminated oils and labelled benzyopyrene with other solvents are described. Encapsulation media tested were concrete and interior and exterior grade Polyfilla (a plaster and cellulose based filler). Problems were presented by the difficulty of mixing the materials and by the maximum uptake of solvents while still allowing solidification. In all cases a soft crumbling material resulted. It is concluded that solidification processing on a small scale does not make economic or scientific sense and that if solidification is necessary it would be better carried out as a national operation by collecting liquids from users. (U.K.)

  3. Positive segregation as a function of buoyancy force during steel ingot solidification

    International Nuclear Information System (INIS)

    Radovic, Zarko; Jaukovic, Nada; Lalovic, Milisav; Tadic, Nebojsa

    2008-01-01

    We analyze theoretically and experimentally solute redistribution in the dendritic solidification process and positive segregation during solidification of steel ingots. Positive segregation is mainly caused by liquid flow in the mushy zone. Changes in the liquid steel velocity are caused by the temperature gradient and by the increase in the solid fraction during solidification. The effects of buoyancy and of the change in the solid fraction on segregation intensity are analyzed. The relationships between the density change, liquid fraction and the steel composition are considered. Such elements as W, Ni, Mo and Cr decrease the effect of the density variations, i.e. they show smaller tendency to segregate. Based on the modeling and experimental results, coefficients are provided controlling the effects of chemical composition, secondary dendrite arm spacing and the solid fraction.

  4. Study and modeling of heat transfer during the solidification of semi-crystalline polymers

    Energy Technology Data Exchange (ETDEWEB)

    Le Goff, R.; Poutot, G.; Delaunay, D. [Laboratoire de Thermocinetique de l' ecole polytechnique de l' universite de Nantes, UMR CNRS 6607, rue Christian Pauc, BP 50609 44306 Nantes cedex 3 (France); Fulchiron, R.; Koscher, E. [Laboratoire des Materiaux Polymeres et des Biomateriaux, IMP/UMR CNRS 5627, Universite Claude Bernard, Lyon 1, 69622 Villeurbanne Cedex (France)

    2005-12-01

    Semi-crystalline polymers are materials whose behavior during their cooling is difficult to model because of the strong coupling between the crystallization, heat transfer, pressure and shear. Thanks to two original apparatus we study solidification of such a polymer without shear. Firstly the comparison between experimental results and a numerical model will permit to validate crystallization kinetic for cooling rate reachable by DSC. The second experiment makes it possible to analyze solidification for high cooling rate, corresponding to some manufacturing processes. It appears that crystallization has an influence on the thermal contact resistance. (author)

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

    International Nuclear Information System (INIS)

    2001-01-01

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

  6. Comparison of risk-dominant scenario assumptions for several TRU waste facilities in the DOE complex

    International Nuclear Information System (INIS)

    Foppe, T.L.; Marx, D.R.

    1999-01-01

    In order to gain a risk management perspective, the DOE Rocky Flats Field Office (RFFO) initiated a survey of other DOE sites regarding risks from potential accidents associated with transuranic (TRU) storage and/or processing facilities. Recently-approved authorization basis documents at the Rocky Flats Environmental Technology Site (RFETS) have been based on the DOE Standard 3011 risk assessment methodology with three qualitative estimates of frequency of occurrence and quantitative estimates of radiological consequences to the collocated worker and the public binned into three severity levels. Risk Class 1 and 2 events after application of controls to prevent or mitigate the accident are designated as risk-dominant scenarios. Accident Evaluation Guidelines for selection of Technical Safety Requirements (TSRs) are based on the frequency and consequence bin assignments to identify controls that can be credited to reduce risk to Risk Class 3 or 4, or that are credited for Risk Class 1 and 2 scenarios that cannot be further reduced. This methodology resulted in several risk-dominant scenarios for either the collocated worker or the public that warranted consideration on whether additional controls should be implemented. RFFO requested the survey because of these high estimates of risks that are primarily due to design characteristics of RFETS TRU waste facilities (i.e., Butler-type buildings without a ventilation and filtration system, and a relatively short distance to the Site boundary). Accident analysis methodologies and key assumptions are being compared for the DOE sites responding to the survey. This includes type of accidents that are risk dominant (e.g., drum explosion, material handling breach, fires, natural phenomena, external events, etc.), source term evaluation (e.g., radionuclide material-at-risk, chemical and physical form, damage ratio, airborne release fraction, respirable fraction, leakpath factors), dispersion analysis (e.g., meteorological

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

    International Nuclear Information System (INIS)

    DOE Carlsbad Field Office

    2001-01-01

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

  8. Rapid Solidification of AB{sub 5} Hydrogen Storage Alloys

    Energy Technology Data Exchange (ETDEWEB)

    Gulbrandsen-Dahl, Sverre

    2002-01-01

    This doctoral thesis is concerned with rapid solidification of AB{sub 5} materials suitable for electrochemical hydrogen storage. The primary objective of the work has been to characterise the microstructure and crystal structure of the produced AB{sub 5} materials as a function of the process parameters, e.g. the cooling rate during rapid solidification, the determination of which has been paid special attention to. The thesis is divided into 6 parts, of which Part I is a literature review, starting with a short presentation of energy storage alternatives. Then a general review of metal hydrides and their utilisation as energy carriers is presented. This part also includes more detailed descriptions of the crystal structure, the chemical composition and the hydrogen storage properties of AB{sub 5} materials. Furthermore, a description of the chill-block melt spinning process and the gas atomisation process is given. In Part II of the thesis a digital photo calorimetric technique has been developed and applied for obtaining in situ temperature measurements during chill-block melt spinning of a Mm(NiCoMnA1){sub 5} hydride forming alloy (Mm = Mischmetal of rare earths). Compared with conventional colour transmission temperature measurements, this technique offers a special advantage in terms of a high temperature resolutional and positional accuracy, which under the prevailing experimental conditions were found to be {+-}29 K and {+-} 0.1 mm, respectively. Moreover, it is shown that the cooling rate in solid state is approximately 2.5 times higher than that observed during solidification, indicating that the solid ribbon stayed in intimate contact with the wheel surface down to very low metal temperatures before the bond was broken. During this contact period the cooling regime shifted from near ideal in the melt puddle to near Newtonian towards the end, when the heat transfer from the solid ribbon to the wheel became the rate controlling step. In Part III of the

  9. Heat load limits for TRU drums on pads

    International Nuclear Information System (INIS)

    Steimke, J.L.; McKinley, M.S.

    1993-08-01

    Some of the Trans-Uranic (TRU) waste generated at SRS is packaged in 55 gallon, galvanized steel drums and stored on concrete pads that are exposed to the weather. It was necessary to compute how much heat can be generated by the waste in these drums without exceeding the temperature limits of the contents of the drum. This report documents the calculation of heat load limits for the drum, which depend on the temperature limits of the contents of the drum. The applicable temperature limits for the contents of the drum are the melting temperature of the polyethylene liner, 284 ± 8 F, the combustion temperature of paper, 450 F and the decomposition temperature of anionic resin, 190 F. One part of the analysis leading to the heat load limits was the collection of weather records on solar flux, wind speed and air temperature. Another part of the task was an experimental measurement of two important properties of the drum lid, the emittance and the absorptance. As used here, emittance is the rate at which an object emits infrared thermal radiation divided by the rate at which a perfect black body at the same temperature emits thermal radiation. Absorptance is the rate at which an object absorbs solar radiation divided by the rate at which a perfect black body absorbs radiation. For nine locations on each of eight typical weathered drum lids the measured emittance ranged from 0.73 ± 0.05 to 1.00 ± 0.07 (95% confidence level) and the average emittance for the eight lids was 0.85. For the eight drum lids the measured absorptance ranged from 0.64 ± 0.07 to 0.79 ± 0.07 with an average absorptance for the eight lids of 0.739

  10. Mechanism of flow reversal during solidification of an anomalous liquid

    Science.gov (United States)

    Kumar, Virkeshwar; Kumawat, Mitesh; Srivastava, Atul; Karagadde, Shyamprasad

    2017-12-01

    In a wide variety of fluidic systems involving thermal and compositional gradients, local density changes lead to the onset of natural convection that influences the process itself, for example, during phase-change phenomena and magmatic flows. Accurate knowledge of the flow characteristics is essential to quantify the impact of the flow of the processes. In this work, the first-ever demonstration of flow reversal during bottom-up solidification of water using full-field thermal and flow measurements and its direct impact on the solidifying interface is presented. Based on prior optical interferometric measurements of full-field temperature distribution in water during solidification, we use the particle image velocimetry technique to quantify and reveal the changing natural convection pattern arising solely due to the density anomaly of water between 0 °C and 4 °C. The independently captured thermal and flow fields show striking similarities and clearly elucidate the plausible mechanism explaining the formation of a curved interface at the stagnation point and the subsequent reversal of flow direction due to a changed interface morphology. A control volume analysis is further presented to estimate the energy invested in the formation of a perturbation and the resulting flip in the flow direction caused by this perturbation.

  11. Solidification of liquid concentrate and solid waste generated as by-products of the liquid radwaste treatment systems in light-water reactors

    International Nuclear Information System (INIS)

    Neilson, R.M. Jr.; Colombo, P.

    1977-01-01

    The treatment of liquid concentrate and solid waste produced in light-water reactors as by-products of liquid radwaste treatment systems consists of five basic operations: waste collection, waste pretreatment, solidification agent handling, mixing/packaging (solidification) and waste package handling. This paper will concern itself primarily with the solidification operation, however, the other operations enumerated as well as the types of wastes treated and their origins will be briefly described, especially with regards to their effects on solidification. During solidification, liquid concentrate and solid wastes are incorporated with a solidification agent to form a monolithic, free-standing solid. The basic solidification agent types either currently used in the United States or proposed for use include absorbants, hydraulic cement, urea-formaldehyde, other polymer systems, and bitumen. The operation, formulations and limitations of these agents as used for radwaste solidification will be discussed. Properties relevant to the evaluation of solidified waste forms will be identified and relative comparisons made for wastes solidified by various processes

  12. Manufacture history results of an investigation of the bitumen solidification object towards the check of an abandonment object

    International Nuclear Information System (INIS)

    Kogawa, Noboru; Kondo, Toshiyuki

    2001-08-01

    In order to make this book reflect in the investigation which turned the bitumen solidification object to maintenance of the abandonment object technical standard on condition of carrying out subterranean disposal in the future, it created for the purpose of utilizing as precious sources of information, such as a nuclide inventory in the living body, group-izing of the past campaign required for typical solidification object selection, and information offer at the time of disposal examination. A development operation history collected so that histories including the shift action in an institution of the formation of discharge reduction of the characteristic of solidification object manufacture outlines, such as composition of the process of an institution and a solidification object and a storage actual result, the contents of an examination of the past campaign, and the solidification object manufactured based on topics or radioactive iodine and radioactive carbon etc., such as the past contents of an examination/operation, may grasp comprehensively in creation, and it carried out as the composition stared the trend of future disposal fixedly. It was a period (for 16 years) until an bituminization demonstration facility processing institution will start a cold examination from April (Showa 57), 1982, and it starts a hot examination from May 4, it starts solidification processing technical development operation from October 6 and it results in the fire explosion accident on March 11 (Heisei 9), 1997, and low level radioactivity concentration waste fluid was processed 7,438m 3 and 29,967 bitumen solidification objects were manufactured. According to the accident, it is necessary to hand it down to future generations with processing technology while the bitumen solidification object manufactured in 15 years although the bituminization demonstration facility processing institution came to close the mission holds information precious when considering future disposal

  13. Americium product solidification and disposal

    International Nuclear Information System (INIS)

    Mailen, J.C.; Campbell, D.O.; Bell, J.T.; Collins, E.D.

    1987-01-01

    The americium product from the TRUEX processing plant needs to be converted into a form suitable for ultimate disposal. An evaluation of the disposal based on safety, number of process steps, demonstrated operability of the processes, production of low-level alpha waste streams, and simplicity of maintenance with low radiation exposures to personnel during maintenance, has been made. The best process is to load the americium on a cation exchange resin followed by calcination or oxidation of the resin after loading

  14. Validation of a 3D multi-physics model for unidirectional silicon solidification

    NARCIS (Netherlands)

    Simons, P.; Lankhorst, A.M.; Habraken, A.; Faber, A.J.; Tiuleanu, D.; Pingel, R.

    2012-01-01

    A model for transient movements of solidification fronts has been added to X-stream, an existing multi-physics simulation program for high temperature processes with flow and chemical reactions. The implementation uses an enthalpy formulation and works on fixed grids. First we show the results of a

  15. Application of sulfur concrete for solidification of radioactive wastes and building of repositories

    International Nuclear Information System (INIS)

    Cholerzynski, A.; Tomczak, W.; Switalski, J.

    2000-01-01

    The application of sulfur concrete as solidification material for radioactive wastes and as building material used in repositories have been presented. Their high shear strength, low level of leaching, and high radiation resistance decide of positive recommendation of such material for wide use in radioactive waste treatment processes and repositories building

  16. Basic study on decontamination of TRU wastes with cerium mediated electrolytic oxidation method

    International Nuclear Information System (INIS)

    Ishii, Junichi; Kobayashi, Fuyumi; Uchida, Shoji; Sumiya, Masato; Kida, Takashi; Shirahashi, Koichi; Umeda, Miki; Sakuraba, Koichi

    2010-03-01

    At Nuclear Fuel Cycle Safety Engineering Research Facility (NUCEF), the cerium mediated electrolytic oxidation method which is a decontamination technique to decrease the radioactivity of TRU wastes to the clearance-level has been developed for the effective reduction of TRU wastes generated from the decommissioning of a nuclear fuel reprocessing facility and so on. This method corrodes the oxide layer and the surface of metallic TRU metal wastes by the strong oxidation power of Ce 4+ in nitric acid. In this study, parameter tests were conducted to optimize the solution condition of Ce 3+ initial concentrations and nitric acid concentrations. The target corrosion rate of metallic TRU wastes set to be 2 - 4 μm/h for the practical use of this method. Under the optimized solution condition, a dissolution test of stainless steel simulating wastes was carried out. From the result of the dissolution test, the average corrosion rate was 3.3 μm/h during the test time of 90 hours. Based on the supposition that the corrosion depth of metallic TRU wastes was 20 μm enough to achieve the clearance-level, the treatment time for the decontamination was about 6 hours. It was confirmed from the result that the decontamination could be performed within one day and the decontamination solution could repeatedly reuse 15 times. (author)

  17. TRU Self-Recycling in a High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Jo, Chang Keun

    2013-01-01

    Conclusions: • Evaluated the core characteristics and performance for SR-HTR. • Self-recycling of self-generated TRUs is feasible and deep-burning of the self-generated TRU can be achieved in SR-HTR. • From the results, ⇒ TRU discharge burnup is over 60% and the uranium fuel can also be utilized very efficiently in the SR-HTR core. ⇒ In the case of separate TRU loading, the power fraction of the TRU fueled zone is substantially smaller (~10%) than that of the uranium fueled zone. ⇒ The transmutation of Pu-239 is nearly complete (~99%) in the SR-HTR core and that of Pu-241 is also extremely high. ⇒ The decay heat of SR-HTR core is evaluated to be similar to that of the 3-ring UO 2 -only loaded HTR core. • A TF-coupled analysis is required for a more concrete evaluation of TRU deep-burn in an SR-HTR

  18. A facility design for repackaging ORNL CH-TRU legacy waste in Building 3525

    International Nuclear Information System (INIS)

    Huxford, T.J.; Cooper, R.H. Jr.; Davis, L.E.; Fuller, A.B.; Gabbard, W.A.; Smith, R.B.; Guay, K.P.; Smith, L.C.

    1995-07-01

    For the last 25 years, the Oak Ridge National Laboratory (ORNL) has conducted operations which have generated solid, contact-handled transuranic (CH-TRU) waste. At present the CH-TRU waste inventory at ORNL is about 3400 55-gal drums retrievably stored in RCRA-permitted, aboveground facilities. Of the 3400 drums, approximately 2600 drums will need to be repackaged. The current US Department of Energy (DOE) strategy for disposal of these drums is to transport them to the Waste Isolation Pilot Plant (WIPP) in New Mexico which only accepts TRU waste that meets a very specific set of criteria documented in the WIPP-WAC (waste acceptance criteria). This report describes activities that were performed from January 1994 to May 1995 associated with the design and preparation of an existing facility for repackaging and certifying some or all of the CH-TRU drums at ORNL to meet the WIPP-WAC. For this study, the Irradiated Fuel Examination Laboratory (IFEL) in Building 3525 was selected as the reference facility for modification. These design activities were terminated in May 1995 as more attractive options for CH-TRU waste repackaging were considered to be available. As a result, this document serves as a final report of those design activities

  19. Proposed methods for treating high-level pyrochemical process wastes

    International Nuclear Information System (INIS)

    Johnson, T.R.; Miller, W.E.; Steunenberg, R.K.

    1985-01-01

    This survey illustrates the large variety and number of possible techniques available for treating pyrochemical wastes; there are undoubtedly other process types and many variations. The choice of a suitable process is complicated by the uncertainty as to what will be an acceptable waste form in the future for both TRU and non-TRU wastes

  20. Development of techniques for measuring plutonium contents in TRU wastes by NDA methods

    International Nuclear Information System (INIS)

    Matsubayashi, Toshiyuki; Kuwana, Katsumi; Morita, Tomio; Izuhara, Shigeomi; Suzuki, Masahiro

    1983-01-01

    In order to develop a technique for measuring the amount of plutonium in plutonium-contaminated (TRU) wastes, a passive gamma method was selected from many candidate methods, and examined for the suitability by applying the method to low density wastes. A segmented gamma scanner was used for the experiment. The instrument is composed mainly of a Ge(Li) detector, multichannel analyser, data processing system, turntable and transmission radiation source of (75)Se. A sham waste was prepared by adding plutonium oxide powder as a radiation source to waste matrix in a 20-1 carton box. The sham waste was put on the turntable, and the detector was set at 50 cm distance from the center of the turntable. 414 keV gamma ray emitted from (239)Pu was utilized for the assay of plutonium in the experiment. The effects of combustible (paper) waste matrix, organic chlorinated material matrix, and the distribution of plutonium source in a box on the count rate were examined, and it was concluded that 1) about 10 mg of (239)Pu contained in both matrices should be assayed by the passive gamma method, 2) 50 mg of (239) Pu was measured at 30 % confidence level with 2000 sec measuring time, 3) the effect of distribution of plutonium in a waste was able to be reduced to a value of less than 15 % by rotating the waste on the turntable. (Yoshitake, I.)

  1. Comparison of two selective separation method for 93Zr by using TRU and TEVA resins

    International Nuclear Information System (INIS)

    Oliveira, Thiago C.; Oliveira, Arno Heeren de

    2011-01-01

    The zirconium isotope 93 Zr is a long-lived pure β-particle-emitting radionuclide produced from 235 U fission and from neutron activation of the stable isotope 92 Zr and thus occurring as one of the radionuclides found in nuclear reactors. Due to its long half life, 93 Zr is one of the radionuclides of interest for the performance of assessment studies of waste storage or disposal. Measurement of 93 Zr is difficult owing to its trace level concentration and its low activity in nuclear wastes and further because its certified standards are not frequently available. The aim of this work was to compare two radiochemical procedure based on selective extraction using an anion-exchange chromatography, TRU and TEVA resins, in order to separate zirconium from the matrix and to analyze it by liquid scintillation spectrometry technique. To set up the radiochemical separation procedure for zirconium, a tracer solution of 95 Zr and its 724.19 keV γ-ray measurements by γ - spectrometry were used in order to follow the behavior of zirconium during the radiochemical separation. A tracer solution of 55 Fe, the main interference in the LSC measurements, was used in order to verify the decontamination factor during the separation process. The limit of detection of the 0.05 Bq 1 -1 was obtained for 55 Fe standard solutions by using a sample:cocktail ratio of 3:17 mL for Optiphase Hisafe 3 cocktail. (author)

  2. Solidification in Multicomponent Multiphase Systems (SIMMS)

    Science.gov (United States)

    Rex, S.; Hecht, U.

    2005-06-01

    The multiphase microstructures that evolve during the solidification of multicomponent alloys are attracting widespread interest for industrial applications and fundamental research.Thermodynamic databases are now well-established for many alloy systems. Thermodynamic calculations provide all the required information about phase equilibria, forming an integral part of both dedicated and comprehensive microstructure models. Among the latter, phase-field modelling has emerged as the method of choice. Solidification experiments are intended to trigger model development or to serve as benchmarks for model validation. For benchmarking, microgravity conditions offer a unique opportunity for avoiding buoyancy-induced convection and buoyancy forces in bulk samples. However, diffusion and the free-energy of interfaces and its anisotropy need to be determined.The measurement of chemical diffusivities in the liquid state can equally benefit from microgravity experiments.

  3. Retrofit of radwaste solidification systems in Spain

    International Nuclear Information System (INIS)

    Rorcillo, R.; Virzi, E.

    1983-01-01

    In order to meet current Spanish engineering criteria as well as to provide for likely future Spanish Regulatory requirements, utilities committed to a major policy change in the preferred radwaste solidification media. In the early 1970's Spanish utilities, following the United States experience, purchased inexpensive solidification systems which used urea formaldehyde (UF) as the binding matrix. By the late 1970's the Spanish utilities, seeing the deterioration of the UF position and slow progress toward its improvement, unilaterally changed their binding matrix to cement. This paper illustrates the implementation of this change at the ASCO Nuclear Plant. The problems of layout modifications, shortened delivery schedule and criteria unique for Spain are addressed. Also presented is the operating experience acquired during the pre-operational start-up of the ASCO I Radwaste System

  4. OPTIMASI PEMADATAM CEPAT PADA PENGAYAAN MINYAK IKAN HASIL SAMPING PENGALENGAN LEMURU DENGAN ASAM LEMAK w-3 MENGGUNAKAN METODE PERMUKAAN RESPON [Optimization of Rapid Solidification to Enrich Fish Oil from by-Product of Lemuru Canning Processing with w-3 batty Acids by Response Surface Method

    Directory of Open Access Journals (Sweden)

    Teti Estiasih1

    2005-12-01

    Full Text Available Oil from by-product of lemuru canning processing was a source of w-3 fatty acid but its characteristics had out been known. The content of w-3 fatty acids of this oil had to be increased. Various methods are available to enhance w-3 fatty acids concentration Rapid solidification was one of limited methods to enrich fish oil by w-3 fatty acids containing triglycerides.This research was conducted to optimize rapid solidification condition to enrich fish oil from by product of lemuru canning processing with w-3 fatty acids and characterize the enriched oil compared by International Association of Fish Meal and Oil Manufacturers standard. In optimization process, the content of EPA+OHA and yield .was maximized. A two-factors central composite design in Response Surface Method was used to study the effect of solvent-to-oil ratio (X1 and extraction time (X2. The response (Y is the multiplication of EPA+DHA content by yield.The results showed that under optimum conditions the maximum response were obtained at a solvent-to-oil ratio of 3,95:1(vw and extraction time of 24,93 hours. The w-3 fatty acids enriched fish oil had EPA+DHA content of 33,33% and yield of 9.40% (w/w.The produced w-3 fatty acids enriched fish oil had good quality based on food grade fish oil standard, unless Fe and Cu content. Chelation could reduce these oxidizing metals.

  5. Alternative solidification techniques for radioactive ion exchange resins and liquid concentrates

    International Nuclear Information System (INIS)

    Thegerstroem, C.

    1980-01-01

    Methods, that are used or are under development for solidification of radioactive ion exchange resins or liquid concentrates, utilize normally cement, bitumen or some polymere as matrix material. This report contains a review and a description of these solidification processes and their products, especially of relatively new techniques that are under development in different countries. It is possible that solidification in thermosetting resins will be more used in the future, especially when product quality requirements are high (for instance when solidifying medium level resins) or when special waste categories has to be solidified. However it is not probable that thermosetting resins will be extensively used in a broad application as matrix material. In that case the methods are to complicated and expensive compared to, for instance, solidification in concrete. Systems for incorporation in polyesteremulsions (Dow-process) have a potential as they are quite simple and can accept a large variation of liquid wastes. Some methods in an early stage of development (for instance Inert Carrier Radwaste Process) will have to be tested in active application before they can be further evaluated. (author)

  6. Cement radwaste solidification studies third annual report

    International Nuclear Information System (INIS)

    Brown, D.J.; James, J.M.; Lee, D.J.; Smith, D.L.; Walker, A.T.

    1982-03-01

    This report summarises cement radwaste studies carried out at AEE Winfrith during 1981 on the encapsulation of medium and low active waste in cement. During the year more emphasis has been placed on the work which is directly related to the solidification of SGHWR active sludge. Information has been obtained on the properties of 220 dm 3 drums of cemented waste. The use of cement grouts for the encapsulation of solid items has also been investigated during 1981. (U.K.)

  7. Initial stages of solidification of eutectic alloys

    International Nuclear Information System (INIS)

    Lemaignan, Clement

    1980-01-01

    The study of the various initial stages of eutectic solidification - i.e. primary nucleation, eutectic structure formation and stable growth conditions - was undertaken with various techniques including low angle neutron diffusion, in-situ electron microscopy on solidifying alloys and classical metallography. The results obtained allow to discuss the effect of metastable states during primary nucleation, of surface dendrite during eutectic nucleation and also of the crystallographic anisotropy during growth. (author) [fr

  8. Microwave solidification development for Rocky Flats waste

    Energy Technology Data Exchange (ETDEWEB)

    Dixon, D.; Erle, R.; Eschen, V. [and others

    1994-04-01

    The Microwave Engineering Team at the Rocky Flats Plant has developed a production-scale system for the treatment of hazardous, radioactive, and mixed wastes using microwave energy. The system produces a vitreous final form which meets the acceptance criteria for shipment and disposal. The technology also has potential for application on various other waste streams from the public and private sectors. Technology transfer opportunities are being identified and pursued for commercialization of the microwave solidification technology.

  9. Microwave solidification development for Rocky Flats waste

    International Nuclear Information System (INIS)

    Dixon, D.; Erle, R.; Eschen, V.

    1994-04-01

    The Microwave Engineering Team at the Rocky Flats Plant has developed a production-scale system for the treatment of hazardous, radioactive, and mixed wastes using microwave energy. The system produces a vitreous final form which meets the acceptance criteria for shipment and disposal. The technology also has potential for application on various other waste streams from the public and private sectors. Technology transfer opportunities are being identified and pursued for commercialization of the microwave solidification technology

  10. A Review of Permanent Magnet Stirring During Metal Solidification

    Science.gov (United States)

    Zeng, Jie; Chen, Weiqing; Yang, Yindong; Mclean, Alexander

    2017-12-01

    Rather than using conventional electromagnetic stirring (EMS) with three-phase alternating current, permanent magnet stirring (PMS), based on the use of sintered NdFeB material which has excellent magnetic characteristics, can be employed to generate a magnetic field for the stirring of liquid metal during solidification. Recent experience with steel casting indicates that PMS requires less than 20 pct of the total energy compared with EMS. Despite the excellent magnetic density properties and low power consumption, this relatively new technology has received comparatively little attention by the metal casting community. This paper reviews simulation modeling, experimental studies, and industrial trials of PMS conducted during recent years. With the development of magnetic simulation software, the magnetic field and associated flow patterns generated by PMS have been evaluated. Based on the results obtained from laboratory experiments, the effects of PMS on metal solidification structures and typical defects such as surface pinholes and center cavities are summarized. The significance of findings obtained from trials of PMS within the metals processing sector, including the continuous casting of steel, are discussed with the aim of providing an overview of the relevant parameters that are of importance for further development and industrial application of this innovative technology.

  11. Containerless solidification of undercooled oxide and metallic eutectic melts

    International Nuclear Information System (INIS)

    Li Mingjun; Nagashio, Kosuke; Kuribayashi, Kazuhiko

    2004-01-01

    A high-speed video was employed to monitor the in situ recalescence of undercooled oxide Al 2 O 3 -36.8 at.% ZrO 2 and metallic Ni-18.7 at.% Sn eutectics that were processed on an aero-acoustic levitator and an electromagnetic levitator, respectively. For the oxide eutectic, the entire sample becomes brighter and brighter without any clear recalescence front during spontaneous crystallization. When the sample was seeded at desired undercoolings, crystallization started from the seeding point and then spread through the entire sample. Microstructures of the oxide solidified via both the spontaneous crystallization and external seeding consist of many independent eutectic colonies at the sample surface, indicating that copious nucleation takes place regardless of melt undercooling and solidification mode. For the metallic eutectics, two kinds of recalescence are visualized. The surface and cross sectional microstructures reveal that copious nucleation is also responsible for the formation of independent eutectic colonies distributing within the entire sample. It is not possible to measure the growth velocity of a single eutectic colony using optical techniques under the usual magnification. The conventional nucleation concept derived from single-phase alloys may not be applicable to the free solidification of the undercooled double-phase oxide and metallic eutectic systems

  12. Crystallographic investigation of grain selection during initial solidification

    International Nuclear Information System (INIS)

    Esaka, H; Shinozuka, K; Kataoka, Y

    2016-01-01

    Normally, macroscopic solidified structure consists of chill, columnar and equiaxed zones. In a chill zone, many fine grains nucleate on the mold surface and grow their own preferred growth direction. Only a few of them continue to grow because of grain selection. In order to understand the grain selection process, crystallographic investigation has been carried out in the zone of initial solidification in this study. 10 g of Al-6 wt%Si alloy was melted at 850 °C and poured on the thick copper plate. Longitudinal cross section of the solidified shell was observed by a SEM and analyzed by EBSD. The result of EBSD mapping reveals that crystallographic orientation was random in the range of initial solidification. Further, some grains are elongated along their <100> direction. Columnar grains, whose growth directions are almost parallel to the heat flow direction, develop via grain selection. Here, a dendrite whose growth direction is close to the heat flow direction overgrows the other dendrite whose growth direction is far from the heat flow direction. However, sometimes we observed that dendrite, whose zenith angle is large, overgrew the other dendrite. It can be deduced that the time of nucleation on the mold surface is not constant. (paper)

  13. Neutron spectrum effects on TRU recycling in Pb-Bi cooled fast reactor core

    International Nuclear Information System (INIS)

    Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

    2003-01-01

    This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction

  14. MANAGEING THE RETRIEVAL RISK OF BURIED TRANSURANIC (TRU) WASTE WITH UNIQUE CHARACTERISTICS

    International Nuclear Information System (INIS)

    WOJTASEK, R.D.; GREENWELL, R.D.

    2005-01-01

    United States-Department of Energy (DOE) sites that store transuranic (TRU) waste are almost certain to encounter waste packages with characteristics that are so unique as to warrant special precautions for retrieval. At the Hanford Site, a subgroup of stored TRU waste (12 drums) had special considerations due to the radioactive source content of plutonium oxide (PuO 2 ), and the potential for high heat generation, pressurization, criticality, and high radiation. These characteristics bear on the approach to safely retrieve, overpack, vent, store, and transport the waste package. Because of the potential risk to personnel, contingency planning for unexpected conditions played an effective roll in work planning and in preparing workers for the field inspection activity. As a result, the integrity inspections successfully confirmed waste package configuration and waste confinement without experiencing any perturbations due to unanticipated packaging conditions. This paper discusses the engineering and field approach to managing the risk of retrieving TRU waste with unique characteristics

  15. Inspection method for solidification product of radioactive waste and method of preparing solidification product of radiation waste

    International Nuclear Information System (INIS)

    Izumida, Tatsuo; Tamada, Shin; Matsuda, Masami; Kamata, Shoji; Kikuchi, Makoto.

    1993-01-01

    A powerful X-ray generation device using an electron-ray accelerator is used for inspecting presence or absence of inner voids in solidification products of radioactive wastes during or after solidification. By installing the X-ray CT system and the radioactive waste solidifying facility together, CT imaging for solidification products is conducted in a not-yet cured state of solidifying materials during or just after the injection. If a defect that deteriorates the durability of the solidification products should be detected, the solidification products are repaired, for example, by applying vibrations to the not-yet cured solidification products. Thus, since voids or cracks in the radioactive wastes solidification products, which were difficult to be measured so far, can be measured in a short period of time accurately thereby enabling to judge adaptability to the disposal standards, inspection cost for the radioactive waste solidification product can be saved remarkably. Further, the inside of the radioactive waste solidification products can be evaluated correctly and visually, so that safety in the ground disposal storage of the radioactive solidification products can be improved remarkably. (N.H.)

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

    International Nuclear Information System (INIS)

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

    1993-04-01

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

  17. TRU-ART: A cost-effective prototypical neutron imaging technique for transuranic waste certification systems

    International Nuclear Information System (INIS)

    Horton, W.S.

    1989-01-01

    The certification of defense radioactive waste as either transuranic or low-level waste requires very sensitive and accurate assay instrumentation to determine the specific radioactivity within an individual waste package. An assay instrument that employs a new technique (TRU-ART), which can identify the location of the radioactive material within a waste package, was designed, fabricated, and tested to potentially enhance the certification of problem defense waste drums. In addition, the assay instrumentation has potential application in radioactive waste reprocessing and neutron tomography. The assay instrumentation uses optimized electronic signal responses from an array of boral- and cadmium-shielded polyethylene-moderated 3 H detector packages. Normally, thermal neutrons that are detected by 3 H detectors have very poor spatial dependency that may be used to determine the location of the radioactive material. However, these shielded-detector packages of the TRU-ART system maintain the spatial dependency of the radioactive material in that the point of fast neutron thermalization is immediately adjacent to the 3 H detector. The TRU-ART was used to determine the location of radioactive material within three mock-up drums (empty, peat moss, and concrete) and four actual waste drums. The TRU-ART technique is very analogous to emission tomography. The mock-up drum and actual waste drum data, which were collected by the TRU-ART, were directly input into a algebraic reconstruction code to produce three-dimensional isoplots. Finally, a comprehensive fabrication cost estimate of the fielded drum assay system and the TRU-ART system was determined, and, subsequently, these estimates were used in a cost-benefit analysis to compare the economic advantage of the respective systems

  18. Direct numerical simulation of solidification microstructures affected by fluid flow

    International Nuclear Information System (INIS)

    Juric, D.

    1997-12-01

    The effects of fluid flow on the solidification morphology of pure materials and solute microsegregation patterns of binary alloys are studied using a computational methodology based on a front tracking/finite difference method. A general single field formulation is presented for the full coupling of phase change, fluid flow, heat and solute transport. This formulation accounts for interfacial rejection/absorption of latent heat and solute, interfacial anisotropies, discontinuities in material properties between the liquid and solid phases, shrinkage/expansion upon solidification and motion and deformation of the solid. Numerical results are presented for the two dimensional dendritic solidification of pure succinonitrile and the solidification of globulitic grains of a plutonium-gallium alloy. For both problems, comparisons are made between solidification without fluid flow and solidification within a shear flow

  19. Nondestructive assay of TRU waste using gamma-ray active and passive computed tomography

    International Nuclear Information System (INIS)

    Roberson, G.P.; Decman, D.; Martz, H.; Keto, E.R.; Johansson, E.M.

    1995-01-01

    The authors have developed an active and passive computed tomography (A and PCT) scanner for assaying radioactive waste drums. Here they describe the hardware components of their system and the software used for data acquisition, gamma-ray spectroscopy analysis, and image reconstruction. They have measured the performance of the system using ''mock'' waste drums and calibrated radioactive sources. They also describe the results of measurements using this system to assay a real TRU waste drum with relatively low Pu content. The results are compared with X-ray NDE studies of the same TRU waste drum as well as assay results from segmented gamma scanner (SGS) measurements

  20. The effect of vibration on alpha radiolysis of transuranic (TRU) waste

    International Nuclear Information System (INIS)

    Zerwekh, A.; Kosiewicz, S.; Warren, J.

    1993-01-01

    This paper reports on previously unpublished scoping work related to the potential for vibration to redistribute radionuclides on transuranic (TRU) waste. If this were to happen, the amount of gases generated, including hydrogen, could be increased above the undisturbed levels. This could be an important consideration for transport of TRU wastes either at DOE sites or from them to a future repository, e.g., the Waste Isolation Pilot Plant (WIPP). These preliminary data on drums of real waste seem to suggest that radionuclide redistribution does not occur. However improvements in the experimental methodology are suggested to enhance safety of future experiments on real wastes as well as to provide more rigorous data

  1. On the Role of Mantle Overturn during Magma Ocean Solidification

    Science.gov (United States)

    Boukaré, C. E.; Parmentier, E.; Parman, S. W.

    2017-12-01

    Solidification of potential global magma ocean(s) (MO) early in the history of terrestrial planets may play a key role in the evolution of planetary interiors by setting initial conditions for their long-term evolution. Constraining this initial structure of solid mantles is thus crucial but remains poorly understood. MO fractional crystallization has been proposed to generate gravitationally unstable Fe-Mg chemical stratification capable of driving solid-state mantle overturn. Fractional solidification and overturn hypothesis, while only an ideal limiting case, can explain important geochemical features of both the Moon and Mars. Current overturn models consider generally post-MO overturn where the cumulate pile remains immobile until the end of MO solidification. However, if the cumulate pile overturns during MO solidification, the general picture of early planet evolution might differ significantly from the static crystallization models. We show that the timing of mantle overturn can be characterized with a dimensionless number measuring the ratio of the MO solidification time and the purely compositional overturn timescale. Syn-solidification overturn occurs if this dimensionless parameter, Rc, exceeds a critical value. Rc is mostly affected by the competition between the MO solidification time and mantle viscosity. Overturn that occurs during solidification can result in smaller scales of mantle chemical heterogeneity that could persist for long times thus influencing the whole evolution of a planetary body. We will discuss the effects of compaction/percolation on mantle viscosity. If partially molten cumulate do not have time to compact during MO solidification, viscosity of cumulates would be significantly lower as the interstitcial melt fraction would be large. Both solid mantle remelting during syn-solidification overturn and porous convection of melt retained with the cumulates are expected to reduce the degree of fractional crystallization. Syn-solidification

  2. Numerical Simulation on the Origin of Solidification Cracking in Laser Welded Thick-Walled Structures

    Directory of Open Access Journals (Sweden)

    Nasim Bakir

    2018-06-01

    Full Text Available One of the main factors affecting the use of lasers in the industry for welding thick structures is the process accompanying solidification cracks. These cracks mostly occurring along the welding direction in the welding center, and strongly affect the safety of the welded components. In the present study, to obtain a better understanding of the relation between the weld pool geometry, the stress distribution and the solidification cracking, a three-dimensional computational fluid dynamic (CFD model was combined with a thermo-mechanical model. The CFD model was employed to analyze the flow of the molten metal in the weld pool during the laser beam welding process. The weld pool geometry estimated from the CFD model was used as a heat source in the thermal model to calculate the temperature field and the stress development and distributions. The CFD results showed a bulging region in the middle depth of the weld and two narrowing areas separating the bulging region from the top and bottom surface. The thermo-mechanical simulations showed a concentration of tension stresses, transversally and vertically, directly after the solidification during cooling in the region of the solidification cracking.

  3. Containerless solidification of acoustically levitated Ni-Sn eutectic alloy

    Energy Technology Data Exchange (ETDEWEB)

    Geng, D.L.; Xie, W.J.; Wei, B. [Northwestern Polytechnical University, Department of Applied Physics, Xi' an (China)

    2012-10-15

    Containerless solidification of Ni-18.7at%Sn eutectic alloy has been achieved with a single-axis acoustic levitator. The temperature, motion, and oscillation of the sample were monitored by a high speed camera. The temperature of the sample can be determined from its image brightness, although the sample moves vertically and horizontally during levitation. The experimentally observed frequency of vertical motion is in good agreement with theoretical prediction. The sample undergoes shape oscillation before solidification finishes. The solidification microstructure of this alloy consists of a mixture of anomalous eutectic plus regular lamellar eutectic. This indicates the achievement of rapid solidification under acoustic levitation condition. (orig.)

  4. Solidification of Hypereutectic Thin Wall Ductile Cast Iron

    DEFF Research Database (Denmark)

    Pedersen, Karl Martin; Tiedje, Niels Skat

    2006-01-01

    solidification. The first stage, which was relatively short, had none or very little recalescence. Further under cooling, followed by reheating during recalescence, was necessary to initiate the second part of the eutectic solidification. Both the secondary under cooling and recalescence was larger in the 3 mm...... a higher Si content in the ferrite around the larger nodules compared to the ferrite around the rest of the nodules. This indicates that solidification took place along the following path: The solidification starts with nucleation and growth of primary graphite nodules. This probably starts during...

  5. Solidification structure and abrasion resistance of high chromium white irons

    Science.gov (United States)

    Doğan, Ö. N.; Hawk, J. A.; Laird, G.

    1997-06-01

    Superior abrasive wear resistance, combined with relatively low production costs, makes high Cr white cast irons (WCIs) particularly attractive for applications in the grinding, milling, and pumping apparatus used to process hard materials. Hypoeutectic, eutectic, and hypereutectic cast iron compositions, containing either 15 or 26 wt pct chromium, were studied with respect to the macrostructural transitions of the castings, solidification paths, and resulting microstructures when poured with varying superheats. Completely equiaxed macrostructures were produced in thick section castings with slightly hypereutectic compositions. High-stress abrasive wear tests were then performed on the various alloys to examine the influence of both macrostructure and microstructure on wear resistance. Results indicated that the alloys with a primarily austenitic matrix had a higher abrasion resistance than similar alloys with a pearlitic/bainitic matrix. Improvement in abrasion resistance was partially attributed to the ability of the austenite to transform to martensite at the wear surface during the abrasion process.

  6. Advanced powder metallurgy aluminum alloys via rapid solidification technology

    Science.gov (United States)

    Ray, R.

    1984-01-01

    Aluminum alloys containing 10 to 11.5 wt. pct. of iron and 1.5 to 3 wt. pct. of chromium using the technique of rapid solidification powder metallurgy were studied. Alloys were prepared as thin ribbons (.002 inch thick) rapidly solidified at uniform rate of 10(6) C/second by the melt spinning process. The melt spun ribbons were pulverized into powders (-60 to 400 mesh) by a rotating hammer mill. The powders were consolidated by hot extrusion at a high reduction ratio of 50:1. The powder extrusion temperature was varied to determine the range of desirable processing conditions necessary to yield useful properties. Powders and consolidated alloys were characterized by SEM and optical metallography. The consolidated alloys were evaluated for (1) thermal stability, (2) tensile properties in the range, room temperature to 450 F, and (3) notch toughness in the range, room temperature to 450 F.

  7. Radwaste volume reduction and solidification by General Electric

    International Nuclear Information System (INIS)

    Green, T.A.; Weech, M.E.; Miller, G.P.; Eberle, J.W.

    1982-01-01

    Since 1978 General Electric has been actively engaged in developing a volume reduction and solidifcation system or treatment of radwaste generated in commercial nuclear power plants. The studies have been aimed at defining an integrated system that would be directly responsive to the rapid evolving needs of the industry for the volume reduction and solidification of low-level radwaste. The resulting General Electric Volume Reduction System (GEVRS) is an integrated system based on two processes: the first uses azeotropic distillation technology and is called AZTECH, and the second is controlled-air incineration...called INCA. The AZTECH process serves to remove water from concentrated salt solutions, ion exchange resins and filter sludge slurries and then encapsulates the dried solids into a dense plastic product. The INCA unit serves to reduce combustible wastes to ashes suitable for encapsulation into the same plastic product produced by AZTECH

  8. Atomic mobility in liquid and fcc Al-Si-Mg-RE (RE = Ce, Sc) alloys and its application to the simulation of solidification processes in RE-containing A357 alloys

    International Nuclear Information System (INIS)

    Lu, Zhao; Zhang, Lijun

    2017-01-01

    This paper first provides a critical review of experimental and theoretically-predicted diffusivities in both liquid and fcc Al-Si-Mg-RE (RE = Ce, Sc) alloys as-reported by previous researchers. The modified Sutherland equation is then employed to predict self- and impurity diffusivities in Al-Si-Mg-RE melts. The self-diffusivity of metastable fcc Sc is evaluated via the first-principles computed activation energy and semi-empirical relations. Based on the critically-reviewed and presently evaluated diffusivity information, atomic mobility descriptions for liquid and fcc phases in the Al-Si-Mg-RE systems are established by means of the Diffusion-Controlled TRAnsformation (DICTRA) software package. Comprehensive comparisons show that most of the measured and theoretically-predicted diffusivities can be reasonably reproduced by the present atomic mobility descriptions. The atomic mobility descriptions for liquid and fcc Al-Si-Mg-RE alloys are further validated by comparing the model-predicted differential scanning calorimetry curves for RE-containing A357 alloys during solidification against experimental data. Detailed analysis of the curves and microstructures in RE-free and RE-containing A357 alloys indicates that both Ce and Sc can serve as the grain refiner for A357 alloys, and that the grain refinement efficiency of Sc is much higher.

  9. Atomic mobility in liquid and fcc Al-Si-Mg-RE (RE = Ce, Sc) alloys and its application to the simulation of solidification processes in RE-containing A357 alloys

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Zhao; Zhang, Lijun [Central South Univ., Changsha (China). State Key Lab of Powder Metallurgy; Tang, Ying [Thermo-Calc Software AB, Solna (Sweden)

    2017-06-15

    This paper first provides a critical review of experimental and theoretically-predicted diffusivities in both liquid and fcc Al-Si-Mg-RE (RE = Ce, Sc) alloys as-reported by previous researchers. The modified Sutherland equation is then employed to predict self- and impurity diffusivities in Al-Si-Mg-RE melts. The self-diffusivity of metastable fcc Sc is evaluated via the first-principles computed activation energy and semi-empirical relations. Based on the critically-reviewed and presently evaluated diffusivity information, atomic mobility descriptions for liquid and fcc phases in the Al-Si-Mg-RE systems are established by means of the Diffusion-Controlled TRAnsformation (DICTRA) software package. Comprehensive comparisons show that most of the measured and theoretically-predicted diffusivities can be reasonably reproduced by the present atomic mobility descriptions. The atomic mobility descriptions for liquid and fcc Al-Si-Mg-RE alloys are further validated by comparing the model-predicted differential scanning calorimetry curves for RE-containing A357 alloys during solidification against experimental data. Detailed analysis of the curves and microstructures in RE-free and RE-containing A357 alloys indicates that both Ce and Sc can serve as the grain refiner for A357 alloys, and that the grain refinement efficiency of Sc is much higher.

  10. Radiological Characterization Methodology for INEEL-Stored Remote-Handled Transuranic (RH TRU) Waste from Argonne National Laboratory-East

    International Nuclear Information System (INIS)

    Kuan, P.; Bhatt, R.N.

    2003-01-01

    An Acceptable Knowledge (AK)-based radiological characterization methodology is being developed for RH TRU waste generated from ANL-E hot cell operations performed on fuel elements irradiated in the EBR-II reactor. The methodology relies on AK for composition of the fresh fuel elements, their irradiation history, and the waste generation and collection processes. Radiological characterization of the waste involves the estimates of the quantities of significant fission products and transuranic isotopes in the waste. Methods based on reactor and physics principles are used to achieve these estimates. Because of the availability of AK and the robustness of the calculation methods, the AK-based characterization methodology offers a superior alternative to traditional waste assay techniques. Using the methodology, it is shown that the radiological parameters of a test batch of ANL-E waste is well within the proposed WIPP Waste Acceptance Criteria limits

  11. Solidification of eutectic system alloys in space (M-19)

    Science.gov (United States)

    Ohno, Atsumi

    1993-01-01

    cast by the Ohno Continuous Casting Process and they show the unidirectionally solidified structure. Each flight and ground sample was made of these same rods. The dimensions of all samples are 4.5 mm in diameter and 23.5 mm in length. Each sample is put in a graphite capsule and then vacuum sealed in a double silica ampoule. Then the ampoule is put in the tantalum cartridge and sealed by electron beam welding. For onbard experiments, a Continuous Heating Furnance (CHF) will be used for melting and solidifying samples under microgravity conditions. Six flight samples will be used. Four samples are hypo-eutectic and two are hyper-eutectic alloys. The surface of the two hypo-eutectic alloy samples are covered with aluminum oxide film to prevent Marangoni convection expected under microgravity conditions. Each sample will be heated to 700 C and held at that temperature for 5 min. After that the samples will be allowed to cool to 500 C in the furnace and they will be taken out of the furnace for He gas cooling. The heating and cooling diagrams for the flight experiments are shown. After collecting the flight samples, the solidified structures of the samples will be examined and the mechanisms of eutectic solidification under microgravity conditions will be determined. It is likely that successful flight experiment results will lead to production of high quality eutectic alloys and eutectic composite materials in space.

  12. Solidification of radioactive incinerator ash

    International Nuclear Information System (INIS)

    Schuler, T.F.; Charlesworth, D.L.

    1986-01-01

    The Ashcrete process will solidify ash generated by the Beta Gamma Incinerator (BGI) at the Savannah River Plant (SRP). The system remotely handles, adds material to, and tumbles drums of ash to produce ashcrete, a stabilized wasteform. Full-scale testing of the Ashcrete unit began at Savannah River Laboratory (SRL) in January 1984, using nonradioactive ash. Tests determined product homogeneity, temperature distribution, compressive strength, and final product formulation. Product formulations that yielded good mix homogeneity and final product compressive strength were developed. Drum pressurization and temperature rise (resulting from the cement's heat of hydration) were also studied to verify safe storage and handling characteristics. In addition to these tests, an expert system was developed to assist process troubleshooting

  13. Phase-field simulation of peritectic solidification closely coupled with directional solidification experiments in an Al-36 wt% Ni alloy

    International Nuclear Information System (INIS)

    Siquieri, R; Emmerich, H; Doernberg, E; Schmid-Fetzer, R

    2009-01-01

    In this work we present experimental and theoretical investigations of the directional solidification of Al-36 wt% Ni alloy. A phase-field approach (Folch and Plapp 2005 Phys. Rev. E 72 011602) is coupled with the CALPHAD (calculation of phase diagrams) method to be able to simulate directional solidification of Al-Ni alloy including the peritectic phase Al 3 Ni. The model approach is calibrated by systematic comparison to microstructures grown under controlled conditions in directional solidification experiments. To illustrate the efficiency of the model it is employed to investigate the effect of temperature gradient on the microstructure evolution of Al-36 wt% Ni during solidification.

  14. A coupled model on fluid flow, heat transfer and solidification in continuous casting mold

    Directory of Open Access Journals (Sweden)

    Xu-bin Zhang

    2017-11-01

    Full Text Available Fluid flow, heat transfer and solidification of steel in the mold are so complex but crucial, determining the surface quality of the continuous casting slab. In the current study, a 2D numerical model was established by Fluent software to simulate the fluid flow, heat transfer and solidification of the steel in the mold. The VOF model and k-ε model were applied to simulate the flow field of the three phases (steel, slag and air, and solidification model was used to simulate the solidification process. The phenomena at the meniscus were also explored through interfacial tension between the liquid steel and slag as well as the mold oscillation. The model included a 20 mm thick mold to clarify the heat transfer and the temperature distribution of the mold. The simulation results show that the liquid steel flows as upper backflow and lower backflow in the mold, and that a small circulation forms at the meniscus. The liquid slag flows away from the corner at the meniscus or infiltrates into the gap between the mold and the shell with the mold oscillating at the negative strip stage or at the positive strip stage. The simulated pitch and the depth of oscillation marks approximate to the theoretical pitch and measured depth on the slab.

  15. Joining of superalloy Inconel 600 by diffusion induced isothermal solidification of a liquated insert metal

    International Nuclear Information System (INIS)

    Egbewande, A.T.; Chukwukaeme, C.; Ojo, O.A.

    2008-01-01

    The effect of process variables on the microstructure of transient liquid phase bonded IN 600 using a commercial filler alloy was studied. Microstructural examination of bonded specimens showed that isothermal solidification of the liquated insert occurred during holding at the joining temperatures. In cases where the holding time was insufficient for complete isothermal solidification, the residual liquid transformed on cooling into a centerline eutectic product. The width of the eutectic decreased with increased holding time and an increase in initial gap width resulted in thicker eutectic width in specimens bonded at the same temperature and for equivalent holding times. In addition to the centerline eutectic microconstituent, precipitation of boron-rich particles was observed within the base metal region adjacent to the substrate-joint interface. Formation of these particles appeared to have influenced the rate of solidification of the liquated interlayer during bonding. In contrast to the conventional expectation of an increase in the rate of isothermal solidification with an increase in temperature, a decrease in the rate was observed with an increase in temperatures above 1160 deg. C. This could be related to a decrease in solubility of boron in nickel above the Ni-B eute