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Sample records for dwpf canistered waste

  1. Chemical compatibility of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1993-01-01

    The Waste Acceptance Preliminary Specifications (WAPS) require that the contents of the canistered waste form are compatible with one another and the stainless steel canister. The canistered waste form is a closed system comprised of a stainless steel vessel containing waste glass, air, and condensate. This system will experience a radiation field and an elevated temperature due to radionuclide decay. This report discusses possible chemical reactions, radiation interactions, and corrosive reactions within this system both under normal storage conditions and after exposure to temperatures up to the normal glass transition temperature, which for DWPF waste glass will be between 440 and 460 degrees C. Specific conclusions regarding reactions and corrosion are provided. This document is based on the assumption that the period of interim storage prior to packaging at the federal repository may be as long as 50 years

  2. Inorganic analyses of volatilized and condensed species within prototypic Defense Waste Processing Facility (DWPF) canistered waste

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1992-01-01

    The high-level radioactive waste currently stored in carbon steel tanks at the Savannah River Site (SRS) will be immobilized in a borosilicate glass in the Defense Waste Processing Facility (DWPF). The canistered waste will be sent to a geologic repository for final disposal. The Waste Acceptance Preliminary Specifications (WAPS) require the identification of any inorganic phases that may be present in the canister that may lead to internal corrosion of the canister or that could potentially adversely affect normal canister handling. During vitrification, volatilization of mixed (Na, K, Cs)Cl, (Na, K, Cs) 2 SO 4 , (Na, K, Cs)BF 4 , (Na, K) 2 B 4 O 7 and (Na,K)CrO 4 species from glass melt condensed in the melter off-gas and in the cyclone separator in the canister pour spout vacuum line. A full-scale DWPF prototypic canister filled during Campaign 10 of the SRS Scale Glass Melter was sectioned and examined. Mixed (NaK)CI, (NaK) 2 SO 4 , (NaK) borates, and a (Na,K) fluoride phase (either NaF or Na 2 BF 4 ) were identified on the interior canister walls, neck, and shoulder above the melt pour surface. Similar deposits were found on the glass melt surface and on glass fracture surfaces. Chromates were not found. Spinel crystals were found associated with the glass pour surface. Reference amounts of the halides and sulfates were found retained in the glass and the glass chemistry, including the distribution of the halides and sulfates, was homogeneous. In all cases where rust was observed, heavy metals (Zn, Ti, Sn) from the cutting blade/fluid were present indicating that the rust was a reaction product of the cutting fluid with glass and heat sensitized canister or with carbon-steel contamination on canister interior. Only minimal water vapor is present so that internal corrosion of the canister, will not occur

  3. Dew point, internal gas pressure, and chemical composition of the gas within the free volume of DWPF canistered waste forms

    International Nuclear Information System (INIS)

    Harbour, J.R.; Herman, D.T.; Crump, S.; Miller, T.J.; McIntosh, J.

    1996-01-01

    The Defense Waste Processing Facility (DWPF) produced 55 canistered waste forms containing simulated waste glass during the four Waste Qualification campaigns of the DWPF Startup Test Program. Testing of the gas within the free volume of these canisters for dew point, internal gas pressure, and chemical composition was performed as part of a continuing effort to demonstrate compliance with the Waste Acceptance Product Specifications. Results are presented for six glass-filled canisters. The dew points within the canisters met the acceptance criterion of < 20 degrees C for all six canisters. Factors influencing the magnitude of the dew point are presented. The chemical composition of the free volume gas was indistinguishable from air for all six canisters. Hence, no foreign materials were present in the gas phase of these canisters. The internal gas pressures within the sealed canisters were < 1 atm at 25 degrees C for all six canisters which readily met the acceptance criterion of an internal gas pressure of less than 1.5 atm at 25 degrees C. These results provided the evidence required to demonstrate compliance with the Waste Acceptance Product Specifications

  4. Initial results from the canistered waste forms produced during the first campaign of the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.

    1995-01-01

    As part of the Defense Waste Processing Facility (DWPF) Startup Test Program, approximately 90 canisters will be filled with glass containing simulated radioactive waste during five separate campaigns. The first campaign is a facility acceptance test to demonstrate the operability of the facility and to collect initial data on the glass and the canistered waste forms. During the next four campaigns (the waste qualification campaigns) data will be obtained which will be used to demonstrate that the DWPF product meets DOE's Waste Acceptance Product Specifications (WAPS). Currently 12 of the 16 canisters have been filled with glass during the first campaign (FA-13). This paper describes the tests that have been carried out on these 12 glass-filled canisters and presents the data with reference to the acceptance criteria of the WAPS. These tests include measurement of canister dimensions prior to and after glass filling. dew point, composition, and pressure of the gas within the free volume of the canister, fill height, free volume, weight, leak rates of welds and temporary seals, and weld parameters

  5. Canister disposition plan for the DWPF Startup Test Program

    International Nuclear Information System (INIS)

    Harbour, J.R.; Payne, C.H.

    1990-01-01

    This report details the disposition of canisters and the canistered waste forms produced during the DWPF Startup Test Program. The six melter campaigns (DWPF Startup Tests FA-13, WP-14, WP-15, WP-16, WP-17, and FA-18) will produce 126 canistered waste forms. In addition, up to 20 additional canistered waste forms may be produced from glass poured during the transition between campaigns. In particular, this canister disposition plan (1) assigns (by alpha-numeric code) a specific canister to each location in the six campaign sequences, (2) describes the method of access for glass sampling on each canistered waste form, (3) describes the nature of the specific tests which will be carried out, (4) details which tests will be carried out on each canistered waste form, (5) provides the sequence of these tests for each canistered waste form, and (6) assigns a storage location for each canistered waste form. The tests are designed to provide evidence, as detailed in the Waste Form Compliance Plan (WCP 1 ), that the DWPF product will comply with the Waste Acceptance Product Specifications (WAPS 2 ). The WAPS must be met before the canistered waste form is accepted by DOE for ultimate disposal at the Federal Repository. The results of these tests will be included in the Waste Form Qualification Report (WQR)

  6. Can-in-canister cold demonstration in DWPF (U)

    International Nuclear Information System (INIS)

    Kuehn, N.H.

    1996-07-01

    The Department of Energy Fissile Materials Disposition Program is evaluating a number of options for disposition of weapons-usable plutonium surplus to national defense needs. One of the immobilization options is the Can-In-Canister approach. In this option small cans of a plutonium glass, which contains a neutron absorber, are placed on a support structure in a large Savannah River Site Defense Waste Processing Facility (DWPF) canister. The top is then welded onto the canister. This canister is filled with High Level Waste (HLW) glass at the DWPF. The HLW glass provides the radiation source for proliferation resistance. These canisters are to be placed in a Federal Repository. To provide information on the technical feasibility of this option prior to the Record of Decision on plutonium disposition, the Department of Energy Fissile Materials Disposition Program funded a demonstration in the DWPF. This demonstration was conducted before the start of radioactive operations. Two test canisters containing cans of surrogate (non- radioactive) plutonium glass were successfully filled with simulated HLW glass at the DWPF using standard pouring procedures. One canister had twenty cans of surrogate plutonium glass. The other had eight cans of surrogate plutonium glass. After the canisters were filled, the contents of the canisters were examined to provide data on the effect of the rack and cans on the filling of the DWPF canister, the effect of the pour on the surrogate plutonium glass and the effect of the rack and cans on the simulated HLW glass. There was no deformation of the support racks during the pour. The simulated HLW glass filled all the regions around the rack and cans and the regions between the cans and the wall of the canister. This report discusses the design of the racks and cans, the modification of the DWPF canisters to accommodate the rack and cans, the conditions during the pours and the results of the post pour analysis

  7. Computer Modeling Of High-Level Waste Glass Temperatures Within DWPF Canisters During Pouring And Cool Down

    International Nuclear Information System (INIS)

    Amoroso, J.

    2011-01-01

    This report describes the results of a computer simulation study to predict the temperature of the glass at any location inside a DWPF canister during pouring and subsequent cooling. These simulations are an integral part of a larger research focus aimed at developing methods to predict, evaluate, and ultimately suppress nepheline formation in HLW glasses. That larger research focus is centered on holistically understanding nepheline formation in HLW glass by exploring the fundamental thermal and chemical driving forces for nepheline crystallization with respect to realistic processing conditions. Through experimental work, the goal is to integrate nepheline crystallization potential in HLW glass with processing capability to ultimately optimize waste loading and throughput while maintaining an acceptable product with respect to durability. The results of this study indicated severe temperature gradients and prolonged temperature dwell times exist throughout different locations in the canister and that the time and temperatures that HLW glass is subjected to during processing is a function of pour rate. The simulations indicate that crystallization driving forces are not uniform throughout the glass volume in a DWPF (or DWPF-like) canister and illustrate the importance of considering overall kinetics (chemical and thermal driving forces) of nepheline formation when developing methods to predict and suppress its formation in HLW glasses. The intended path forward is to use the simulation data both as a driver for future experimental work and, as an investigative tool for evaluating the impact of experimental results. Simulation data will be used to develop laboratory experiments to more acutely evaluate nepheline formation in HLW glass by incorporating the simulated temperatures throughout the canister into the laboratory experiments. Concurrently, laboratory experiments will be performed to identify nepheline crystallization potential in HLW glass as a function of

  8. Characterization of projected DWPF glasses heat treated to simulate canister centerline cooling

    International Nuclear Information System (INIS)

    Marra, S.L.; Jantzen, C.M.

    1992-05-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Eventually these canistered waste forms will be sent to a geologic repository for final disposal. In order to assure acceptability by the repository, the Department of Energy has defined requirements which DWPF canistered waste forms must meet. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to identify the crystalline phases expected to be present in the final glass product. Knowledge of the thermal history of the borosilicate glass during filling and cooldown of the canister is necessary to determine the amount and type of crystalline phases present in the final glass product. Glass samples of seven projected DWPF compositions were cooled following the same temperature profile as that of glass at the centerline of the full-scale DWPF canister. The glasses were characterized by x-ray diffraction and scanning electron microscopy to identify the crystalline phases present The volume percents of each crystalline phase present were determined by quantitative x-ray diffraction. The Product Consistency Test (PCI) was used to determine the durability of the heat-treated glasses

  9. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1982-01-01

    A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant liquid, high-level radioactive waste into a solid form, such as borosilicate glass. To prevent the spread of radioactivity, the outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated by-products, which are difficult to immobilize by vitrification

  10. DWPF waste form compliance plan (Draft Revision)

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Marra, S.L.

    1991-01-01

    The Department of Energy currently has over 100 million liters of high-level radioactive waste in storage at the Savannah River Site (SRS). In the late 1970's, the Department of Energy recognized that there were significant safety and cost advantages associated with immobilizing the high-level waste in a stable solid form. Several alternative waste forms were evaluated in terms of product quality and reliability of fabrication. This evaluation led to a decision to build the Defense Waste Processing Facility (DWPF) at SRS to convert the easily dispersed liquid waste to borosilicate glass. In accordance with the NEPA (National Environmental Policy Act) process, an Environmental Impact Statement was prepared for the facility, as well as an Environmental Assessment of the alternative waste forms, and issuance of a Record of Decision (in December, 1982) on the waste form. The Department of Energy, recognizing that start-up of the DWPF would considerably precede licensing of a repository, instituted a Waste Acceptance Process to ensure that these canistered waste forms would be acceptable for eventual disposal at a federal repository. This report is a revision of the DWPF compliance plan

  11. DWPF waste glass Product Composition Control System

    International Nuclear Information System (INIS)

    Brown, K.G.; Postles, R.L.

    1992-01-01

    The Defense Waste Processing Facility (DWPF) will be used to blend aqueous radwaste (PHA) with solid radwaste (Sludge) in a waste receipt vessel (the SRAT). The resulting SRAT material is transferred to the SME an there blended with ground glass (Frit) to produce a batch of melter feed slurry. The SME material is passed to a hold tank (the MFT) which is used to continuously feed the DWPF melter. The melter. The melter produces a molten glass wasteform which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in a geologic repository. The Product Composition Control System (PCCS) is the system intended to ensure that the melt will be processible and that the glass wasteform will be acceptable. This document provides a description of this system

  12. Burst Test Qualification Analysis of DWPF Canister-Plug Weld

    International Nuclear Information System (INIS)

    Gupta, N.K.; Gong, Chung.

    1995-02-01

    The DWPF canister closure system uses resistance welding for sealing the canister nozzle and plug to ensure leak tightness. The welding group at SRTC is using the burst test to qualify this seal weld in lieu of the shear test in ASME B ampersand PV Code, Section IX, paragraph QW-196. The burst test is considered simpler and more appropriate than the shear test for this application. Although the geometry, loading and boundary conditions are quite different in the two tests, structural analyses show similarity in the failure mode of the shear test in paragraph QW-196 and the burst test on the DWPF canister nozzle Non-linear structural analyses are performed using finite element techniques to study the failure mode of the two tests. Actual test geometry and realistic stress strain data for the 304L stainless steel and the weld material are used in the analyses. The finite element models are loaded until failure strains are reached. The failure modes in both tests are shear at the failure points. Based on these observations, it is concluded that the use of a burst test in lieu of the shear test for qualifying the canister-plug weld is acceptable. The burst test analysis for the canister-plug also yields the burst pressures which compare favorably with the actual pressure found during burst tests. Thus, the analysis also provides an estimate of the safety margins in the design of these vessels

  13. Development of the DWPF canister temporary shrink-fit seal

    International Nuclear Information System (INIS)

    Kelker, J.W. Jr.

    1986-04-01

    The Defense Waste Processing Facility is being constructed at The Savannah River Plant for the containerization of high-level nuclear waste in a wasteform for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in type 304L stainless steel canisters, 2-feet in diameter x 9-feet 10-inches long, containing a flanged 6-in.-diam pipe fill-nozzle. The canisters have a minimum wall thickness of 3/8 in. Utilizing the heat from the glass filling operation, a shrink-fit seal for a plug in the end of the canister fill nozzle was developed that: will withstand the radioactive environment; will prevent the spread of contamination, and will keep moisture and water from entering the canister during storage and decontamination of the canister by wet-frit blasting to remove smearable and oxide-film fixed radioactive nuclides; is removable and can be replaced by a new oversize plug in the event the seal fails the pressure decay leakage test ( -4 atm cc/sec helium); will keep the final weld closure clean and free of nuclear contamination; will withstand being pressed into the nozzle without exposing external contamination or completely breaking the seal; is reliable; and is easily installed. The seal consists of: a removable sleeve (with a tapered bore) which is shrink-fitted into the nozzle bore during canister fabrication; and a tapered plug which is placed into the sleeved nozzle after the canister is filled with radioactive molten glass. A leak-tight shrink-fit seal is formed between the nozzle, sleeve, and plug upon temperature equilibrium. The temporarily sealed canister is transferred from the Melt cell to the Decon cell, and the surface is decontaminated. Next it is transferred to the Weld/Test cell where the temporary seal is pressed down into the nozzle, revealing a clean cavity where the canister final closure weld is made

  14. Decontamination of Savannah River Plant waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1982-01-01

    A Defense Waste Processing Facility (DWPF) is currently being designed to convert Savannah River Plant (SRP) liquid, high-level radioactive waste into a solid form, such as borosilicate glass. The outside of the canisters of waste glass must have very low levels of smearable radioactive contamination before they are removed from the DWPF to prevent the spread of radioactivity. Several techniques were considered for canister decontamination: high-pressure water spray, electropolishing, chemical dissolution, and abrasive blasting. An abrasive blasting technique using a glass frit slurry has been selected for use in the DWPF. No additional equipment is needed to process waste generated from decontamination. Frit used as the abrasive will be mixed with the waste and fed to the glass melter. In contrast, chemical and electrochemical techniques require more space in the DWPF, and produce large amounts of contaminated byproducts which are difficult to immobilize by vitrification

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  16. The DWPF waste form qualification program

    International Nuclear Information System (INIS)

    Marra, S.L.; Plodinec, M.J.

    1994-01-01

    Prior to the introduction of radioactive feed into the Defense Waste Processing Facility for immobilization in borosilicate glass an extensive waste qualification program must be completed. The DWPF must demonstrate its ability to comply with the Waste Acceptance Product Specifications. This ability is being demonstrated through laboratory and pilot scale work and will be completed after the full operation of the DWPF using various simulated feeds

  17. RECENT PROCESS AND EQUIPMENT IMPROVEMENTS TO INCREASE HIGH LEVEL WASTE THROUGHPUT AT THE DEFENSE WASTE PROCESSING FACILITY (DWPF)

    International Nuclear Information System (INIS)

    Smith, M; Allan Barnes, A; Jim Coleman, J; Robert Hopkins, R; Dan Iverson, D; Richard Odriscoll, R; David Peeler, D

    2006-01-01

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF), the world's largest operating high level waste (HLW) vitrification plant, began stabilizing about 35 million gallons of SRS liquid radioactive waste by-product in 1996. The DWPF has since filled over 2000 canisters with about 4000 pounds of radioactive glass in each canister. In the past few years there have been several process and equipment improvements at the DWPF to increase the rate at which the waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process and therefore minimized process upsets and thus downtime. These improvements, which include glass former optimization, increased waste loading of the glass, the melter glass pump, the melter heated bellows liner, and glass surge protection software, will be discussed in this paper

  18. MCC-15: waste/canister accident testing and analysis method

    International Nuclear Information System (INIS)

    Slate, S.C.; Pulsipher, B.A.; Scott, P.A.

    1985-02-01

    The Materials Characterization Center (MCC) at the Pacific Northwest Laboratory (PNL) is developing standard tests to characterize the performance of nuclear waste forms under normal and accident conditions. As part of this effort, the MCC is developing MCC-15, Waste/Canister Accident Testing and Analysis. MCC-15 is used to test canisters containing simulated waste forms to provide data on the effects of accidental impacts on the waste form particle size and on canister integrity. The data is used to support the design of transportation and handling equipment and to demonstrate compliance with repository waste acceptance specifications. This paper reviews the requirements that led to the development of MCC-15, describes the test method itself, and presents some early results from tests on canisters representative of those proposed for the Defense Waste Processing Facility (DWPF). 13 references, 6 figures

  19. The remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility (DWPF) is a complete production area being constructed at the Savannah River Plant for the immobilization of nuclear waste in glass. The remote handling of canisters filled with nuclear waste in glass is an essential part of the process of the DWPF at the Savannah River Plant. The canisters are filled with nuclear waste containing up to 235,000 curies of radioactivity. Handling and movement of these canisters must be accomplished remotely since they radiate up to 5000 R/h. Within the Vitrification Building during filling, cleaning, and sealing, canisters are moved using standard cranes and trolleys and a specially designed grapple. During transportation to the Glass Waste Storage Building, a one-of-a-kind, specially designed Shielded Canister Transporter (SCT) is used. 8 figs

  20. Effect of canister size on costs of disposal of SRP high-level wastes

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-01-01

    The current plan for managing the high-level nuclear wastes at the Savannah River Plant (SRP) calls for processing them into solid forms contained in stainless steel canisters for eventual disposal in a federal geologic repository. A new SRP facility called the Defense Waste Processing Facility (DWPF) is being designed for the onsite waste processing operations. Preliminary evaluations indicate that costs of the overall disposal operation will depend significantly on the size of the canisters, which determines the number of waste forms to be processed. The objective of this study was to evaluate the effects of canister size on costs of DWPF process operations, including canister procurement, waste solidification, and interim storage, on offsite transport, and on repository costs of disposal, including provision of suitable waste packages

  1. DISPOSABLE CANISTER WASTE ACCEPTANCE CRITERIA

    Energy Technology Data Exchange (ETDEWEB)

    R.J. Garrett

    2001-07-30

    The purpose of this calculation is to provide the bases for defining the preclosure limits on radioactive material releases from radioactive waste forms to be received in disposable canisters at the Monitored Geologic Repository (MGR) at Yucca Mountain. Specifically, this calculation will provide the basis for criteria to be included in a forthcoming revision of the Waste Acceptance System Requirements Document (WASRD) that limits releases in terms of non-isotope-specific canister release dose-equivalent source terms. These criteria will be developed for the Department of Energy spent nuclear fuel (DSNF) standard canister, the Multicanister Overpack (MCO), the naval spent fuel canister, the High-Level Waste (HLW) canister, the plutonium can-in-canister, and the large Multipurpose Canister (MPC). The shippers of such canisters will be required to demonstrate that they meet these criteria before the canisters are accepted at the MGR. The Quality Assurance program is applicable to this calculation. The work reported in this document is part of the analysis of DSNF and is performed using procedure AP-3.124, Calculations. The work done for this analysis was evaluated according to procedure QAP-2-0, Control of Activities, which has been superseded by AP-2.21Q, Quality Determinations and Planning for Scientific, Engineering, and Regulatory Compliance Activities. This evaluation determined that such activities are subject to the requirements of DOE/RW/0333P, Quality Assurance Requirements and Description (DOE 2000). This work is also prepared in accordance with the development plan titled Design Basis Event Analyses on DOE SNF and Plutonium Can-In-Canister Waste Forms (CRWMS M&O 1999a) and Technical Work Plan For: Department of Energy Spent Nuclear Fuel Work Packages (CRWMS M&O 2000d). This calculation contains no electronic data applicable to any electronic data management system.

  2. Waste canister for storage of nuclear wastes

    Science.gov (United States)

    Duffy, James B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall.

  3. Waste canister for storage of nuclear wastes

    International Nuclear Information System (INIS)

    Duffy, J.B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall. 4 claims, 4 figures

  4. Conditions for precipitation of copper phases in DWPF waste glass

    International Nuclear Information System (INIS)

    Schumacher, R.F.; Ramsey, W.G.

    1993-01-01

    The Defense Waste Processing Facility (DWPF) precipitate hydrolysis process requires the use of copper formate catalyst. The expected absorbed radiation doses to the precipitate require levels of copper formate that increase the potential for the precipitation of metallic copper in the DWPF Melter. The conditions required to avoid the precipitation of copper are described

  5. Defense Waste Processing Facility (DWPF), Modular CSSX Unit (CSSX), and Waste Transfer Line System of Salt Processing Program (U)

    International Nuclear Information System (INIS)

    CHANG, ROBERT

    2006-01-01

    All of the waste streams from ARP, MCU, and SWPF processes will be sent to DWPF for vitrification. The impact these new waste streams will have on DWPF's ability to meet its canister production goal and its ability to support the Salt Processing Program (ARP, MCU, and SWPF) throughput needed to be evaluated. DWPF Engineering and Operations requested OBU Systems Engineering to evaluate DWPF operations and determine how the process could be optimized. The ultimate goal will be to evaluate all of the Liquid Radioactive Waste (LRW) System by developing process modules to cover all facilities/projects which are relevant to the LRW Program and to link the modules together to: (1) study the interfaces issues, (2) identify bottlenecks, and (3) determine the most cost effective way to eliminate them. The results from the evaluation can be used to assist DWPF in identifying improvement opportunities, to assist CBU in LRW strategic planning/tank space management, and to determine the project completion date for the Salt Processing Program

  6. A technical basis to relax the dew point specification for the environment in the vapor space in DWPF canisters

    International Nuclear Information System (INIS)

    Louthan, M.R. Jr.

    1995-05-01

    This memorandum establishes the technical basis to conclude that relaxing, from 0 C to 20 C, the dew point specification for the atmosphere in the vapor space (free volume) of a DWPF canister will not provide an environment that will cause significant amounts of corrosion induced degradation of the canister wall. The conclusion is based on engineering analysis, experience and review of the corrosion literature. The basic assumptions underlying the conclusion are: (1) the canister was fabricated from Type 304L stainless steel; (2) the corrosion behavior of the canister material, including base metal, fusion zones and heat effected zones, is typified by literature data for, and industrial experience with, 300 series austenitic stainless steels; and (3) the glass-metal crevices created during the pouring operation will not alter the basic corrosion resistance of the steel although such crevices might serve as sites for the initiation of minor amounts of corrosion on the canister wall

  7. Characterization of the Defense Waste Processing Facility (DWPF) Environmental Assessment (EA) glass Standard Reference Material. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Crawford, C.L.; Pickett, M.A.

    1993-06-01

    Liquid high-level nuclear waste at the Savannah River Site (SRS) will be immobilized by vitrification in borosilicate glass. The glass will be produced and poured into stainless steel canisters in the Defense Waste Processing Facility (DWPF). Other waste form producers, such as West Valley Nuclear Services (WVNS) and the Hanford Waste Vitrification Project (HWVP), will also immobilize high-level radioactive waste in borosilicate glass. The canistered waste will be stored temporarily at each facility for eventual permanent disposal in a geologic repository. The Department of Energy has defined a set of requirements for the canistered waste forms, the Waste Acceptance Product Specifications (WAPS). The current Waste Acceptance Primary Specification (WAPS) 1.3, the product consistency specification, requires the waste form producers to demonstrate control of the consistency of the final waste form using a crushed glass durability test, the Product Consistency Test (PCI). In order to be acceptable, a waste glass must be more durable during PCT analysis than the waste glass identified in the DWPF Environmental Assessment (EA). In order to supply all the waste form producers with the same standard benchmark glass, 1000 pounds of the EA glass was fabricated. The chemical analyses and characterization of the benchmark EA glass are reported. This material is now available to act as a durability and/or redox Standard Reference Material (SRM) for all waste form producers.

  8. DWPF process control

    International Nuclear Information System (INIS)

    Heckendoin, F.M. II

    1983-01-01

    The Defense Waste Processing Facility (DWPF) for waste vitrification at the Savannah River Plant (SRP) is in the final design stage. Instrumentation to provide the parameter sensing required to assure the quality of the two-foot-diameter, ten-foot-high waste canister is in the final stage of development. All step of the process and instrumentation are now operating as nearly full-scale prototypes at SRP. Quality will be maintained by assuring that only the intended material enters the canisters, and by sensing the resultant condition of the filled canisters. Primary emphasis will be on instrumentation of the process

  9. Canister arrangement for storing radioactive waste

    Science.gov (United States)

    Lorenzo, D.K.; Van Cleve, J.E. Jr.

    1980-04-23

    The subject invention relates to a canister arrangement for jointly storing high level radioactive chemical waste and metallic waste resulting from the reprocessing of nuclear reactor fuel elements. A cylindrical steel canister is provided with an elongated centrally disposed billet of the metallic waste and the chemical waste in vitreous form is disposed in the annulus surrounding the billet.

  10. Defense Waste Processing Facility Canister Closure Weld Current Validation Testing

    Energy Technology Data Exchange (ETDEWEB)

    Korinko, P. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Maxwell, D. N. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2018-01-29

    Two closure welds on filled Defense Waste Processing Facility (DWPF) canisters failed to be within the acceptance criteria in the DWPF operating procedure SW4-15.80-2.3 (1). In one case, the weld heat setting was inadvertently provided to the canister at the value used for test welds (i.e., 72%) and this oversight produced a weld at a current of nominally 210 kA compared to the operating procedure range (i.e., 82%) of 240 kA to 263 kA. The second weld appeared to experience an instrumentation and data acquisition upset. The current for this weld was reported as 191 kA. Review of the data from the Data Acquisition System (DAS) indicated that three of the four current legs were reading the expected values, approximately 62 kA each, and the fourth leg read zero current. Since there is no feasible way by further examination of the process data to ascertain if this weld was actually welded at either the target current or the lower current, a test plan was executed to provide assurance that these Nonconforming Welds (NCWs) meet the requirements for strength and leak tightness. Acceptance of the welds is based on evaluation of Test Nozzle Welds (TNW) made specifically for comparison. The TNW were nondestructively and destructively evaluated for plug height, heat tint, ultrasonic testing (UT) for bond length and ultrasonic volumetric examination for weld defects, burst pressure, fractography, and metallography. The testing was conducted in agreement with a Task Technical and Quality Assurance Plan (TTQAP) (2) and applicable procedures.

  11. Technical bases for the DWPF testing program

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1990-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will be the first production facility in the United States for the immobilization of high-level nuclear waste. Production of DWPF canistered wasteforms will begin prior to repository licensing, so decisions on facility startup will have to be made before the final decisions on repository design are made. The Department of Energy's Office of Civilian Radioactive Waste Management (RW) has addressed this discrepancy by defining a Waste Acceptance Process. This process provides assurance that the borosilicate-glass wasteform, in a stainless-steel canister, produced by the DWPF will be acceptable for permanent storage in a federal repository. As part of this process, detailed technical specifications have been developed for the DWPF product. SRS has developed detailed strategies for demonstrating compliance with each of the Waste Acceptance Process specifications. An important part of the compliance is the testing which will be carried out in the DWPF. In this paper, the bases for each of the tests to be performed in the DWPF to establish compliance with the specifications are described, and the tests are detailed. The results of initial tests relating to characterization of sealed canisters are reported

  12. Defense Waste Processing Facility (DWPF) Viscosity Model: Revisions for Processing High TiO2 Containing Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-30

    Radioactive high-level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to guarantee, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. The DWPF SPC system is known as the Product Composition Control System (PCCS). The DWPF will soon be receiving wastes from the Salt Waste Processing Facility (SWPF) containing increased concentrations of TiO2, Na2O, and Cs2O . The SWPF is being built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to process TiO2 concentrations >2.0 wt% in the DWPF, new viscosity data were developed over the range of 1.90 to 6.09 wt% TiO2 and evaluated against the 2005 viscosity model. An alternate viscosity model is also derived for potential future use, should the DWPF ever need to process other titanate-containing ion exchange materials. The ultimate limit on the amount of TiO2 that can be accommodated from SWPF will be determined by the three PCCS models, the waste composition of a given sludge

  13. The DWPF strategy for producing an acceptable product

    International Nuclear Information System (INIS)

    Goldston, W.T.; Plodinec, M.J.

    1991-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will convert the 130 million liters of high-level nuclear waste at SRS into stable borosilicate glass. Production of canistered waste forms by the DWPF is scheduled to begin well before submission of the license application for the first repository. The Department of Energy has defined waste acceptance specifications to ensure that DWPF canistered waste forms will be acceptable for eventual disposal. To ensure that canistered waste forms meet those specifications, a program is being carried out to qualify the waste form and those aspects of the production process which affect product quality. This program includes: Pre-production qualification testing of simulated and actual waste forms; Disciplined demonstrations of the ability to produce an acceptable product during startup testing; and Application of a rigorous product control program during production

  14. DWPF glass transition temperatures: What they are and why they are important

    International Nuclear Information System (INIS)

    Marra, S.L.; Jantzen, C.M.; Ramsey, A.A.

    1991-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site will immobilize high-level radioactive liquid waste in borosilicate glass. The glass will be poured into stainless steel canisters for eventual disposal in a geologic repository. The Department of Energy has defined a set of requirements for the DWPF canistered waste form which must be met in order to assure compatibility with, and acceptance by, the repository. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to report glass transition temperatures for the projected range of compositions. This information will be used by the repository to establish waste package design limits

  15. Studies of waste-canister compatibility

    International Nuclear Information System (INIS)

    McCoy, H.E.

    1983-01-01

    Compatibility studies were conducted between 7 waste forms and 15 potential canister structural materials. The waste forms were Al-Si and Pb-Sn matrix alloys, FUETAP, glass, Synroc D, and waste particles coated with carbon or carbon plus silicon carbide. The canister materials included carbon steel (bare and with chromium or nickel coatings), copper, Monel, Cu-35% Ni, titanium (grades 2 and 12), several Inconels, aluminum alloy 5052, and two stainless steels. Tests of either 6888 or 8821 h were conducted at 100 and 300 0 C, which bracket the low and high limits expected during storage. Glass and FUETAP evolved sulfur, which reacted preferentially with copper, nickel, and alloys of these metals. The Pb-Sn matrix alloy stuck to all samples and the carbon-coated particles to most samples at 300 0 C, but the extent of chemical reaction was not determined. Testing for 0.5 h at 800 0 C was included because it is representative of a transportation accident and is required of casks containing nuclear materials. During these tests (1) glass and FUETAP evolved sulfur, (2) FUETAP evolved large amounts of gas, (3) Synroc stuck to titanium alloys, (4) glass was molten, and (5) both matrix alloys were molten with considerable chemical interactions with many of the canister samples. If this test condition were imposed on waste canisters, it would be design limiting in many waste storage concepts

  16. Friction welded closures of waste canisters

    International Nuclear Information System (INIS)

    Klein, R.F.

    1987-01-01

    Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it into a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive-seal weld, the properties and thickness of which must be at least equal to those of the canister material. All studies and tests performed in the work discussed in this paper have the inertia friction welding concept to be highly feasible in this application. This paper describes the decision to investigate the inertia friction welding process, the inertia friction welding process itself, and a proposed equipment design concept. This system would provide a positive, reliable, inspectable, and full-thickness seal weld while utilizing easily maintainable equipment. This high-quality weld can be achieved even in highly contaminated hot cell

  17. SRL canister impact tests

    International Nuclear Information System (INIS)

    Kelker, J.W. Jr.

    1986-05-01

    The Defense Waste Processing Facility (DWPF) is being constructed at the SRP for the containerization of high-level nuclear waste as a waste form for eventual permanent disposal. The waste will be incorporated in molten glass and solidified in Type 304L stainless steel canisters 2 feet in diameter x 9 feet 10 inches long. The canisters have a minimum wall thickness of 3/8 inch. Over a three-year period, nineteen drop-tests of nine canisters, filled with simulated waste glass, were made in support of the DWPF containerization program. Eight of the canister evaluation tests were of Type 304L stainless steel material and one was of commercially pure titanium. Three different length (9.44, 5.06, and 7.88 inch) nozzle configurations containing final closure upset welds were evaluated for the stainless steel canisters. All impact tests of the stainless steel canisters, which included bottom-, side-, and top-drops, were acceptable. The bottom-drop test of the titanium canister, which contained a final closure upset weld, was acceptable; however, the top-drop resulted in a breaching of the top head where it joins the nozzle. The final closure titanium upset weld was acceptable. The titanium canister wall thickness was 1/4 inch

  18. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1981-06-01

    The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

  19. Decontamination processes for waste glass canisters

    International Nuclear Information System (INIS)

    Rankin, W.N.

    1981-01-01

    The process which will be used to decontaminate waste glass canisters at the Savannah River Plant consists of: decontamination (slurry blasting); rinse (high-pressure water); and spot decontamination (high-pressure water plus slurry). No additional waste will be produced by this process because glass frit used in decontamination will be mixed with the radioactive waste and fed into the glass melter. Decontamination of waste glass canisters with chemical and abrasive blasting techniques was investigated. The ability of a chemical technique with HNO 3 -HF and H 2 C 2 O 4 to remove baked-on contamination was demonstrated. A correlation between oxide removal and decontamination was observed. Oxide removal and, thus, decontamination by abrasive blasting techniques with glass frit as the abrasive was proposed and demonstrated

  20. Decontamination of high-level waste canisters

    International Nuclear Information System (INIS)

    Nesbitt, J.F.; Slate, S.C.; Fetrow, L.K.

    1980-12-01

    This report presents evaluations of several methods for the in-process decontamination of metallic canisters containing any one of a number of solidified high-level waste (HLW) forms. The use of steam-water, steam, abrasive blasting, electropolishing, liquid honing, vibratory finishing and soaking have been tested or evaluated as potential techniques to decontaminate the outer surfaces of HLW canisters. Either these techniques have been tested or available literature has been examined to assess their applicability to the decontamination of HLW canisters. Electropolishing has been found to be the most thorough method to remove radionuclides and other foreign material that may be deposited on or in the outer surface of a canister during any of the HLW processes. Steam or steam-water spraying techniques may be adequate for some applications but fail to remove all contaminated forms that could be present in some of the HLW processes. Liquid honing and abrasive blasting remove contamination and foreign material very quickly and effectively from small areas and components although these blasting techniques tend to disperse the material removed from the cleaned surfaces. Vibratory finishing is very capable of removing the bulk of contamination and foreign matter from a variety of materials. However, special vibratory finishing equipment would have to be designed and adapted for a remote process. Soaking techniques take long periods of time and may not remove all of the smearable contamination. If soaking involves pickling baths that use corrosive agents, these agents may cause erosion of grain boundaries that results in rough surfaces

  1. High level waste vitrification at the SRP [Savannah River Plant] (DWPF [Defense Waste Processing Facility] summary)

    International Nuclear Information System (INIS)

    Weisman, A.F.; Knight, J.R.; McIntosh, D.L.; Papouchado, L.M.

    1988-01-01

    The Savannah River Plant has been operating a nuclear fuel cycle since the early 1950's. Fuel and target elements are fabricated and irradiated to produce nuclear materials. After removal from the reactors, the fuel elements are processed to extract the products, and waste is stored. During the thirty years of operation including evaporation, about 30 million gallons of high level radioactive waste has accumulated. The Defense Waste Processing Facility (DWPF) under construction at Savannah River will process this waste into a borosilicate glass for long-term geologic disposal. The construction of the DWPF is about 70% complete; this paper will describe the status of the project, including design demonstrations, with an emphasis on the melter system. 9 figs

  2. Thermal Predictions of the Cooling of Waste Glass Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2014-11-01

    Radioactive liquid waste from five decades of weapons production is slated for vitrification at the Hanford site. The waste will be mixed with glass forming additives and heated to a high temperature, then poured into canisters within a pour cave where the glass will cool and solidify into a stable waste form for disposal. Computer simulations were performed to predict the heat rejected from the canisters and the temperatures within the glass during cooling. Four different waste glass compositions with different thermophysical properties were evaluated. Canister centerline temperatures and the total amount of heat transfer from the canisters to the surrounding air are reported.

  3. RECENT PROCESS IMPROVEMENTS TO INCREASE HLW THROUGHPUT AT THE DWPF

    International Nuclear Information System (INIS)

    Herman, C

    2007-01-01

    The Savannah River Site's (SRS) Defense Waste Processing Facility (DWPF), the world's largest operating high level waste (HLW) vitrification plant, began stabilizing about 35 million gallons of SRS liquid radioactive waste by-product in 1996. The DWPF has since filled over 2000 canisters with about 4000 pounds of radioactive glass in each canister. In the past few years there have been several process and equipment improvements at the DWPF to increase the rate at which the waste can be stabilized. These improvements have either directly increased waste processing rates or have desensitized the process and therefore minimized process upsets and thus downtime. These improvements, which include glass former optimization, increased waste loading of the glass, the melter heated bellows liner, and glass surge protection software, will be discussed in this paper

  4. Demonstration of a Solution Film Leak Test Technique and Equipment for the S00645 Canister Closure

    International Nuclear Information System (INIS)

    Cannell, G.R.

    1999-01-01

    The purpose of this effort was to demonstrate that the SFT technique, when adapted to a DWPF canister nozzle, is capable of detecting leaks not meeting the Waste Acceptance Product Specifications (WAPS) acceptance criterion

  5. The DWPF: Results of full scale qualification runs leading to radioactive operations

    International Nuclear Information System (INIS)

    Marra, S.L.; Elder, H.H.; Occhipinti, J.H.; Snyder, D.E.

    1996-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site in Aiken, SC will immobilize high-level radioactive liquid waste, currently stored in underground carbon steel tanks, in borosilicate glass. The radioactive waste is transferred to the DWPF in two forms: precipitate slurry and sludge slurry. The radioactive waste is pretreated and then combined with a borosilicate glass frit in the DWPF. This homogeneous slurry is fed to a Joule-heated melter which operates at approximately 1150 degrees C. The glass is poured into stainless steel canisters for eventual disposal in a geologic repository. The DWPF product (i.e. the canistered waste form) must comply with the Waste Acceptance Product Specifications (WAPS) in order to be acceptable for disposal. The DWPF has completed Waste Qualification Runs which demonstrate the facility's ability to comply with the waste acceptance specifications. During the Waste Qualification Runs seventy-one canisters of simulated waste glass were produced in preparation for Radioactive Operations. These canisters of simulated waste glass were produced during five production campaigns which also exercised the facility prior to beginning Radioactive Operations. The results of the Waste Qualification Runs are presented

  6. DWPF glass transition temperatures - What they are and why they are important

    International Nuclear Information System (INIS)

    Marra, S.L.; Applewhite-Ramsey, A.L.; Jantzen, C.M.

    1991-01-01

    The Department of Energy has defined a set of requirements for the DWPF canistered waste form which must be met in order to assure compatibility with, and acceptance by, the first geologic repository. These requirements are the Waste Acceptance Preliminary Specifications (WAPS). The WAPS require DWPF to report glass transition temperatures for the projected range of compositions. This information will be used by the repository to establish waste package design limits

  7. Fracturing of simulated high-level waste glass in canisters

    International Nuclear Information System (INIS)

    Peters, R.D.; Slate, S.C.

    1981-09-01

    Waste-glass castings generated from engineering-scale developmental processes at the Pacific Northwest Laboratory are generally found to have significant levels of cracks. The causes and extent of fracturing in full-scale canisters of waste glass as a result of cooling and accidental impact are discussed. Although the effects of cracking on waste-form performance in a repository are not well understood, cracks in waste forms can potentially increase leaching surface area. If cracks are minimized or absent in the waste-glass canisters, the potential for radionuclide release from the canister package can be reduced. Additional work on the effects of cracks on leaching of glass is needed. In addition to investigating the extent of fracturing of glass in waste-glass canisters, methods to reduce cracking by controlling cooling conditions were explored. Overall, the study shows that the extent of glass cracking in full-scale, passively-cooled, continuous melting-produced canisters is strongly dependent on the cooling rate. This observation agrees with results of previously reported Pacific Northwest Laboratory experiments on bench-scale annealed canisters. Thus, the cause of cracking is principally bulk thermal stresses. Fracture damage resulting from shearing at the glass/metal interface also contributes to cracking, more so in stainless steel canisters than in carbon steel canisters. This effect can be reduced or eliminated with a graphite coating applied to the inside of the canister. Thermal fracturing can be controlled by using a fixed amount of insulation for filling and cooling of canisters. In order to maintain production rates, a small amount of additional facility space is needed to accomodate slow-cooling canisters. Alternatively, faster cooling can be achieved using the multi-staged approach. Additional development is needed before this approach can be used on full-scale (60-cm) canisters

  8. Projected radionuclide inventories of DWPF glass from current waste at time of production

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1993-01-01

    The Waste Acceptance Preliminary Specifications (WAPS) require that the DWPF estimate the inventory of long-lived radionuclides present in the waste glass, and report the values in the Waste Form Qualification Report. In this report, conservative (biased high) estimates of the radionuclide inventory of glass produced from waste currently in the Tank Farm are provided. In most cases, these calculated values compare favorably with actual data. In those cases where the agreement is not good, the values reported here are conservative

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

    International Nuclear Information System (INIS)

    Shanahan, K.L.

    1992-02-01

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

  10. Storage and disposal of radioactive waste as glass in canisters

    International Nuclear Information System (INIS)

    Mendel, J.E.

    1978-12-01

    A review of the use of waste glass for the immobilization of high-level radioactive waste glass is presented. Typical properties of the canisters used to contain the glass, and the waste glass, are described. Those properties are used to project the stability of canisterized waste glass through interim storage, transportation, and geologic disposal

  11. High-level radioactive waste glass and storage canister design

    International Nuclear Information System (INIS)

    Slate, S.C.; Ross, W.A.

    1979-01-01

    Management of high-level radioactive wastes is a primary concern in nuclear operations today. The main objective in managing these wastes is to convert them into a solid, durable form which is then isolated from man. A description is given of the design and evaluation of this waste form. The waste form has two main components: the solidified waste and the storage canister. The solid waste form discussed in this study is glass. Waste glasses have been designed to be inert to water attack, physically rugged, low in volatility, and stable over time. Two glass-making processes are under development at PNL. The storage canister is being designed to provide high-integrity containment for solidified wastes from processing to terminal storage. An outline is given of the steps in canister design: material selection, stress and thermal analyses, quality verification, and postfill processing. Examples are given of results obtained from actual nonradioactive demonstration tests. 14 refs

  12. Evaluation Of The Impact Of The Defense Waste Processing Facility (DWPF) Laboratory Germanium Oxide Use On Recycle Transfers To The H-Tank Farm

    International Nuclear Information System (INIS)

    Jantzen, C.; Laurinat, J.

    2011-01-01

    When processing High Level Waste (HLW) glass, the Defense Waste Processing Facility (DWPF) cannot wait until the melt or waste glass has been made to assess its acceptability, since by then no further changes to the glass composition and acceptability are possible. Therefore, the acceptability decision is made on the upstream feed stream, rather than on the downstream melt or glass product. This strategy is known as 'feed forward statistical process control.' The DWPF depends on chemical analysis of the feed streams from the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mix Evaporator (SME) where the frit plus adjusted sludge from the SRAT are mixed. The SME is the last vessel in which any chemical adjustments or frit additions can be made. Once the analyses of the SME product are deemed acceptable, the SME product is transferred to the Melter Feed Tank (MFT) and onto the melter. The SRAT and SME analyses have been analyzed by the DWPF laboratory using a 'Cold Chemical' method but this dissolution did not adequately dissolve all the elemental components. A new dissolution method which fuses the SRAT or SME product with cesium nitrate (CsNO 3 ), germanium (IV) oxide (GeO 2 ) and cesium carbonate (Cs 2 CO 3 ) into a cesium germanate glass at 1050 C in platinum crucibles has been developed. Once the germanium glass is formed in that fusion, it is readily dissolved by concentrated nitric acid (about 1M) to solubilize all the elements in the SRAT and/or SME product for elemental analysis. When the chemical analyses are completed the acidic cesium-germanate solution is transferred from the DWPF analytic laboratory to the Recycle Collection Tank (RCT) where the pH is increased to ∼12 M to be released back to the tank farm and the 2H evaporator. Therefore, about 2.5 kg/yr of GeO 2 /year will be diluted into 1.4 million gallons of recycle. This 2.5 kg/yr of GeO 2 may increase to 4 kg/yr when improvements are implemented to attain an annual canister production

  13. Copper solubility in DWPF, Batch 1 waste glass: Update report

    International Nuclear Information System (INIS)

    Schumacker, R.F.

    1992-01-01

    The ''Late Washing'' Step in the processing of precipitate will require the use of additional copper formate in the Precipitate Reactor to catalyze the hydrolysis reaction. The increased copper concentration in the melter feed increases the potential for metal precipitation during the vitrification of the melter feed. This report describes recent results with a conservative glass selected from the DWPF acceptable region in the Batch 1 Variability Study

  14. Characterization of materials for waste-canister compatibility studies

    International Nuclear Information System (INIS)

    McCoy, H.E.; Mack, J.E.

    1981-10-01

    Sample materials of 7 waste forms and 15 potential canister materials were procured for compatibility tests. These materials were characterized before being placed in test, and the results are the main topic of this report. A test capsule was designed for the tests in which disks of a single waste form were contacted with duplicate samples of canister materials. The capsules are undergoing short-term tests at 800 0 C and long-term tests at 100 and 300 0 C

  15. DWPF MATERIALS EVALUATION SUMMARY REPORT

    Energy Technology Data Exchange (ETDEWEB)

    Gee, T.; Chandler, G.; Daugherty, W.; Imrich, K.; Jankins, C.

    1996-09-12

    To better ensure the reliability of the Defense Waste Processing Facility (DWPF) remote canyon process equipment, a materials evaluation program was performed as part of the overall startup test program. Specific test programs included FA-04 ('Process Vessels Erosion/Corrosion Studies') and FA-05 (melter inspection). At the conclusion of field testing, Test Results Reports were issued to cover the various test phases. While these reports completed the startup test requirements, DWPF-Engineering agreed to compile a more detailed report which would include essentially all of the materials testing programs performed at DWPF. The scope of the materials evaouation programs included selected equipment from the Salt Process Cell (SPC), Chemical Process Cell (CPC), Melt Cell, Canister Decon Cell (CDC), and supporting facilities. The program consisted of performing pre-service baseline inspections (work completed in 1992) and follow-up inspections after completion of the DWPF cold chemical runs. Process equipment inspected included: process vessels, pumps, agitators, coils, jumpers, and melter top head components. Various NDE (non-destructive examination) techniques were used during the inspection program, including: ultrasonic testing (UT), visual (direct or video probe), radiography, penetrant testing (PT), and dimensional analyses. Finally, coupon racks were placed in selected tanks in 1992 for subsequent removal and corrosion evaluation after chemical runs.

  16. High level waste canister emplacement and retrieval concepts study

    International Nuclear Information System (INIS)

    1975-09-01

    Several concepts are described for the interim (20 to 30 years) storage of canisters containing high level waste, cladding waste, and intermediate level-TRU wastes. It includes requirements, ground rules and assumptions for the entire storage pilot plant. Concepts are generally evaluated and the most promising are selected for additional work. Follow-on recommendations are made

  17. Corrosion Testing of Monofrax K-3 Refractory in Defense Waste Processing Facility (DWPF) Alternate Reductant Feeds

    Energy Technology Data Exchange (ETDEWEB)

    Williams, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jantzen, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Burket, P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-04-06

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) uses a combination of reductants and oxidants while converting high level waste (HLW) to a borosilicate waste form. A reducing flowsheet is maintained to retain radionuclides in their reduced oxidation states which promotes their incorporation into borosilicate glass. For the last 20 years of processing, the DWPF has used formic acid as the main reductant and nitric acid as the main oxidant. During reaction in the Chemical Process Cell (CPC), formate and formic acid release measurably significant H2 gas which requires monitoring of certain vessel’s vapor spaces. A switch to a nitric acid-glycolic acid (NG) flowsheet from the nitric-formic (NF) flowsheet is desired as the NG flowsheet releases considerably less H2 gas upon decomposition. This would greatly simplify DWPF processing from a safety standpoint as close monitoring of the H2 gas concentration could become less critical. In terms of the waste glass melter vapor space flammability, the switch from the NF flowsheet to the NG flowsheet showed a reduction of H2 gas production from the vitrification process as well. Due to the positive impact of the switch to glycolic acid determined on the flammability issues, evaluation of the other impacts of glycolic acid on the facility must be examined.

  18. Product/Process (P/P) Models For The Defense Waste Processing Facility (DWPF): Model Ranges And Validation Ranges For Future Processing

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-25

    Radioactive high level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it is poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to guarantee, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository.

  19. High-level waste canister envelope study: structural analysis

    International Nuclear Information System (INIS)

    1977-11-01

    The structural integrity of waste canisters, fabricated from standard weight Type 304L stainless steel pipe, was analyzed for sizes ranging from 8 to 24 in. diameter and 10 to 16 feet long under normal, abnormal, and improbable life cycle loading conditions. The canisters are assumed to be filled with vitrified high-level nuclear waste, stored temporarily at a fuel reprocessing plant, and then transported for storage in an underground salt bed or other geologic storage. In each of the three impact conditions studies, the resulting impact force is far greater than the elastic limit capacity of the material. Recommendations are made for further study

  20. Estimation of total error in DWPF reported radionuclide inventories. Revision 1

    International Nuclear Information System (INIS)

    Edwards, T.B.

    1995-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site is required to determine and report the radionuclide inventory of its glass product. For each macro-batch, the DWPF will report both the total amount (in curies) of each reportable radionuclide and the average concentration (in curies/gram of glass) of each reportable radionuclide. The DWPF is to provide the estimated error of these reported values of its radionuclide inventory as well. The objective of this document is to provide a framework for determining the estimated error in DWPF's reporting of these radionuclide inventories. This report investigates the impact of random errors due to measurement and sampling on the total amount of each reportable radionuclide in a given macro-batch. In addition, the impact of these measurement and sampling errors and process variation are evaluated to determine the uncertainty in the reported average concentrations of radionuclides in DWPF's filled canister inventory resulting from each macro-batch

  1. Electropolishing decontamination system for high-level waste canisters

    International Nuclear Information System (INIS)

    Larson, D.E.; Berger, D.N.; Allen, R.P.; Bryan, G.H.; Place, B.G.

    1988-10-01

    As part of a US Department of Energy (DOE) project agreement with the Federal Ministry for Research and Technology (BMFT) in the Federal Republic of Germany (FRG). The Nuclear Waste Treatment Program at the Pacific Northwest Laboratory (PNL) is preparing 30 radioactive canisters containing borosilicate glass for use in high-level waste repository related tests at the Asse Salt Mine. After filling, the canisters will be welded closed and decontaminated in preparation for shipping to the FRG. Electropolishing was selected as the primary decontamination approach, and an electropolishing system with associated canister inspection equipment has been designed and fabricated for installation in a large hot cell. This remote electropolishing system, which is currently undergoing preliminary testing, is described in this report. 3 refs., 3 figs., 1 tab

  2. Waste canister closure welding using the inertia friction welding process

    International Nuclear Information System (INIS)

    Klein, R.F.; Siemens, D.H.; Kuruzar, D.L.

    1986-02-01

    Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it in a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive seal weld, the properties and thickness of which are at least equal to those of the canister material. This paper describes the inertia friction welding process and a proposed equipment design concept that will provide a positive, reliable, inspectable, and full thickness seal weld while providing easily maintainable equipment, even though the weld is made in a highly contaminated hot cell. All studies and tests performed have shown the concept to be highly feasible. 2 refs., 6 figs

  3. Transportation considerations related to waste forms and canisters for Defense TRU wastes

    International Nuclear Information System (INIS)

    Schneider, K.J.; Andrews, W.B.; Schreiber, A.M.; Rosenthal, L.J.; Odle, C.J.

    1981-09-01

    This report identifies and discusses the considerations imposed by transportation on waste forms and canisters for contact-handled, solid transuranic wastes from the US Department of Energy (DOE) activities. The report reviews (1) the existing raw waste forms and potential immobilized waste forms, (2) the existing and potential future DOE waste canisters and shipping containers, (3) regulations and regulatory trends for transporting commercial transuranic wastes on the ISA, (4) truck and rail carrier requirements and preferences for transporting the wastes, and (5) current and proposed Type B external packagings for transporting wastes

  4. Hanford Waste Vitrification Plant: Preliminary description of waste form and canister

    International Nuclear Information System (INIS)

    Mitchell, D.E.

    1986-01-01

    In July 1985, the US Department of Energy's Office of Civilian Radioactive Waste Management established the Waste Acceptance Process as the means by which defense high-level waste producers, such as the Hanford Waste Vitrification Plant, will develop waste acceptance requirements with the candidate geologic repositories. A complete description of the Waste Acceptance Process is contained in the Preliminary Hanford Waste Vitrification Plant Waste Form Qualification Plan. The Waste Acceptance Process defines three documents that high-level waste producers must prepare as a part of the process of assuming that a high-level waste product will be acceptable for disposal in a geologic repository. These documents are the Description of Waste Form and Canister, Waste Compliance Plan, and Waste Qualification Report. This document is the Hanford Waste Vitrification Plant Preliminary Description of Waste Form and Canister for disposal of Neutralized Current Acid Waste. The Waste Acceptance Specifications for the Hanford Waste Vitrification Plant have not yet been developed, therefore, this document has been structured to corresponds to the Waste Acceptance Preliminary Specifications for the Defense Waste Processing Facility High-Level Waste Form. Not all of the information required by these specifications is appropriate for inclusion in this Preliminary Description of Waste Form and Canister. Rather, this description is limited to information that describes the physical and chemical characteristics of the expected high-level waste form. The content of the document covers three major areas: waste form characteristics, canister characteristics, and canistered waste form characteristics. This information will be used by the candidate geologic repository projects as the basis for preliminary repository design activities and waste form testing. Periodic revisions are expected as the Waste Acceptance Process progresses

  5. Task plan: Temperatures in DWPF Glass Waste Storage Building

    International Nuclear Information System (INIS)

    Hardy, B.J.

    1993-01-01

    The Bechtel National, Inc. Detailed Design Instructions for Structural Design (DDI-02) requires that concrete components of the GWSB not exceed 150 degrees F for structural elements and 200 degrees F locally over a 24 hour period. In addition, the Waste Acceptance Product Specifications (WAPS) sets the maximum post cooldown temperature of the glass waste-form at 400 degrees C. Various scenarios can be postulated which result in elevated glass and concrete temperatures in the GWSB. Therefore, it is important to determine the concrete and glass temperatures during both normal and off-normal conditions. This document details specific tasks required to develop a technically defensible and verifiable methodology for determining maximum temperatures for the waste-forms and the GWSB concrete structures. All models used in this analysis will satisfy Quality Assurance requirements and be defensible to review and oversight committees

  6. Quality Assurance Program description, Defense Waste Processing Facility (DWPF)

    International Nuclear Information System (INIS)

    Maslar, S.R.

    1992-01-01

    This document describes the Westinghouse Savannah River Company's (WSRC) Quality Assurance Program for Defense Waste Processing at the Savannah River Site (SRS). WSRC is the operating contractor for the US Department of Energy (DOE) at the SRS. The following objectives are achieved through developing and implementing the Quality Assurance Program: (1) Ensure that the attainment of quality (in accomplishing defense high-level waste processing objectives at the SRS) is at a level commensurate with the government's responsibility for protecting public health and safety, the environment, the public investment, and for efficiently and effectively using national resources. (2) Ensure that high-level waste from qualification and production activities conform to requirements defined by OCRWM. These activities include production processes, equipment, and services; and products that are planned, designed, procured, fabricated, installed, tested, operated, maintained, modified, or produced

  7. Sediment mechanical response due to emplacement of a waste canister

    International Nuclear Information System (INIS)

    Karnes, C.H.; Dawson, P.R.; Silva, A.J.; Brown, W.T.

    1980-01-01

    Preliminary studies have been conducted to determine the interaction between a waste canister and seabed sediment during and after emplacement. Empirical and approximate methods for determining the depth reached by a freefall penetrator indicate that a boosted penetrator emplacement method may be necessary. Hole closure is necessary, but has not been verified because calculations and laboratory experiments show sensitivity to boundary conditions which control the degree of dynamic hole closure. Laboratory studies show that closure will take place by creep deformation but closure times in seabed environments are uncertain. For assumed thermomechanical properties of sediments, it is shown that a heat generating waste canister will probably not move a significant distancce during the heat generation period

  8. Control of DWPF [Defense Waste Processing Facility] melter feed composition

    International Nuclear Information System (INIS)

    Edwards, R.E. Jr.; Brown, K.G.; Postles, R.L.

    1990-01-01

    The Defense Waste Processing Facility will be used to immobilize Savannah River Site high-level waste into a stable borosilicate glass for disposal in a geologic repository. Proper control of the melter feed composition in this facility is essential to the production of glass which meets product durability constraints dictated by repository regulations and facility processing constraints dictated by melter design. A technique has been developed which utilizes glass property models to determine acceptable processing regions based on the multiple constraints imposed on the glass product and to display these regions graphically. This system along with the batch simulation of the process is being used to form the basis for the statistical process control system for the facility. 13 refs., 3 figs., 1 tab

  9. Description of a ceramic waste form and canister for Savannah River Plant high-level waste

    International Nuclear Information System (INIS)

    Butler, J.L.; Allender, J.S.; Gould, T.H. Jr.

    1982-04-01

    A canistered ceramic waste form for possible immobilization of Savannah River Plant (SRP) high-level radioactive wastes is described. Characteristics reported for the form include waste loading, chemical composition, heat content, isotope inventory, mechanical and thermal properties, and leach rates. A conceptual design of a potential production process for making this canistered form are also described. The ceramic form was selected in November 1981 as the primary alternative to the reference waste form, borosilicate glass, for making a final waste form decision for SRP waste by FY-1983. 11 tables

  10. Evaluation of ISDP Batch 2 Qualification Compliance to 512-S, DWPF, Tank Farm, and Saltstone Waste Acceptance Criteria

    Energy Technology Data Exchange (ETDEWEB)

    Shafer, A.

    2010-05-05

    The purpose of this report is to document the acceptability of the second macrobatch (Salt Batch 2) of Tank 49H waste to H Tank Farm, DWPF, and Saltstone for operation of the Interim Salt Disposition Project (ISDP). Tank 49 feed meets the Waste Acceptance Criteria (WAC) requirements specified by References 11, 12, and 13. Salt Batch 2 material is qualified and ready to be processed through ARP/MCU to the final disposal facilities.

  11. Defense Waste Processing Facility -- Radioactive operations -- Part 3 -- Remote operations

    International Nuclear Information System (INIS)

    Barnes, W.M.; Kerley, W.D.; Hughes, P.D.

    1997-01-01

    The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, South Carolina is the nation's first and world's largest vitrification facility. Following a ten year construction period and nearly three years of non-radioactive testing, the DWPF began radioactive operations in March 1996. Radioactive glass is poured from the joule heated melter into the stainless steel canisters. The canisters are then temporarily sealed, decontaminated, resistance welded for final closure, and transported to an interim storage facility. All of these operations are conducted remotely with equipment specially designed for these processes. This paper reviews canister processing during the first nine months of radioactive operations at DWPF. The fundamental design consideration for DWPF remote canister processing and handling equipment are discussed as well as interim canister storage

  12. Acoustic monitoring techniques for corrosion degradation in cemented waste canisters

    International Nuclear Information System (INIS)

    Naish, C.C.; Buttle, D.; Wallace-Sims, R.; O'Brien, T.M.

    1991-01-01

    This report describes work carried out to investigate acoustic emission as a monitor of corrosion and degradation of wasteforms where the waste is potentially reactive metal. Electronic monitoring equipment has been designed, built and tested to allow long-term monitoring of a number of waste packages simultaneously. Acoustic monitoring experiments were made on a range of 1 litre cemented Magnox and aluminium samples cast into canisters comparing the acoustic events with hydrogen gas evolution rates and electrochemical corrosion rates. The attenuation of the acoustic signals by the cement grout under a range of conditions has been studied to determine the volume of wasteform that can be satisfactorily monitored by one transducer. The final phase of the programme monitored the acoustic events from full size (200 litre) cemented, inactive, simulated aluminium swarf wastepackages prepared at the AEA waste cementation plant at Winfrith. (Author)

  13. Interim glycol flowsheet reduction/oxidation (redox) model for the Defense Waste Processing Facility (DWPF)

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Zamecnik, J. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Missimer, D. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-03-08

    Control of the REDuction/OXidation (REDOX) state of glasses containing high concentrations of transition metals, such as High Level Waste (HLW) glasses, is critical in order to eliminate processing difficulties caused by overly reduced or overly oxidized melts. Operation of a HLW melter at Fe+2/ΣFe ratios of between 0.09 and 0.33, a range which is not overly oxidizing or overly reducing, helps retain radionuclides in the melt, i.e. long-lived radioactive 99Tc species in the less volatile reduced Tc4+ state, 104Ru in the melt as reduced Ru+4 state as insoluble RuO2, and hazardous volatile Cr6+ in the less soluble and less volatile Cr+3 state in the glass. The melter REDOX control balances the oxidants and reductants from the feed and from processing additives such as antifoam. Currently, the Defense Waste Processing Facility (DWPF) is running a formic acid-nitric acid (FN) flowsheet where formic acid is the main reductant and nitric acid is the main oxidant. During decomposition formate and formic acid releases H2 gas which requires close control of the melter vapor space flammability. A switch to a nitric acid-glycolic acid (GN) flowsheet is desired as the glycolic acid flowsheet releases considerably less H2 gas upon decomposition. This would greatly simplify DWPF processing. Development of an EE term for glycolic acid in the GN flowsheet is documented in this study.

  14. Friction Stir Welding of Copper Canisters for Nuclear Waste

    Energy Technology Data Exchange (ETDEWEB)

    Kaellgren, Therese

    2005-07-01

    The Swedish model for final disposal of nuclear fuel waste is based on copper canisters as a corrosion barrier with an inner pressure holding insert of cast iron. One of the methods to seal the copper canister is to use the Friction Stir Welding (FSW), a method invented by The Welding Institute (TWI). This work has been focused on characterisation of the FSW joints, and modelling of the process, both analytically and numerically. The first simulations were based on Rosenthal's analytical medium plate model. The model is simple to use, but has limitations. Finite element models were developed, initially with a two-dimensional geometry. Due to the requirements of describing both the heat flow and the tool movement, three-dimensional models were developed. These models take into account heat transfer, material flow, and continuum mechanics. The geometries of the models are based on the simulation experiments carried out at TWI and at Swedish Nuclear Fuel Waste and Management Co (SKB). Temperature distribution, material flow and their effects on the thermal expansion were predicted for a full-scale canister and lid. The steady state solutions have been compared with temperature measurements, showing good agreement. Microstructure and hardness profiles have been investigated by optical microscope, Scanning Electron Microscope (SEM), Electron Back Scatter Diffraction (EBSD) and Rockwell hardness measurements. EBSD visualisation has been used to determine the grain size distribution and the appearance of twins and misorientation within grains. The orientation maps show a fine uniform equiaxed grain structure. The root of the weld exhibits the smallest grains and many annealing twins. This may be due to deformation after recrystallisation. The appearance of the nugget and the grain size depends on the position of the weld. A large difference can be seen both in hardness and grain size between the start of the weld and when the steady state is reached.

  15. Annular air space effects on nuclear waste canister temperatures in a deep geologic waste repository

    International Nuclear Information System (INIS)

    Lowry, W.E.; Cheung, H.; Davis, B.W.

    1980-01-01

    Air spaces in a deep geologic repository for nuclear high level waste will have an important effect on the long-term performance of the waste package. The important temperature effects of an annular air gap surrounding a high level waste canister are determined through 3-D numerical modeling. Air gap properties and parameters specifically analyzed and presented are the air gap size, surfaces emissivity, presence of a sleeve, and initial thermal power generation rate; particular emphasis was placed on determining the effect of these variables have on the canister surface temperature. Finally a discussion based on modeling results is presented which specifically relates the results to NRC regulatory considerations

  16. Reevaluation of Vitrified High-Level Waste Form Criteria for Potential Cost Savings at the Defense Waste Processing Facility - 13598

    Energy Technology Data Exchange (ETDEWEB)

    Ray, J.W. [Savannah River Remediation (United States); Marra, S.L.; Herman, C.C. [Savannah River National Laboratory, Savannah River Site, Aiken, SC 29808 (United States)

    2013-07-01

    At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form. (authors)

  17. Reevaluation Of Vitrified High-Level Waste Form Criteria For Potential Cost Savings At The Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Ray, J. W.; Marra, S. L.; Herman, C. C.

    2013-01-01

    At the Savannah River Site (SRS) the Defense Waste Processing Facility (DWPF) has been immobilizing SRS's radioactive high level waste (HLW) sludge into a durable borosilicate glass since 1996. Currently the DWPF has poured over 3,500 canisters, all of which are compliant with the U. S. Department of Energy's (DOE) Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms (WAPS) and therefore ready to be shipped to a federal geologic repository for permanent disposal. Due to DOE petitioning to withdraw the Yucca Mountain License Application (LA) from the Nuclear Regulatory Commission (NRC) in 2010 and thus no clear disposal path for SRS canistered waste forms, there are opportunities for cost savings with future canister production at DWPF and other DOE producer sites by reevaluating high-level waste form requirements and compliance strategies and reducing/eliminating those that will not negatively impact the quality of the canistered waste form

  18. Corrosion of canister materials for radioactive waste disposal

    Energy Technology Data Exchange (ETDEWEB)

    Kienzler, Bernhard [KIT Karlsruhe (Germany). Institut fuer Nukleare Entsorgung (INE)

    2017-08-15

    In the period between 1980 and 2004, corrosion studies on various metallic materials have been performed at the Research Center Karlsruhe. The objectives of these experimental studies addressed mainly the performance of canister materials for heat producing, high-level wastes and spent nuclear fuels for a repository in a German salt dome. Additional studies covered the performance of steels for packaging wastes with negligible heat production under conditions to be expected in rocksalt and in the Konrad iron ore mine. The results of the investigations have been published in journals and conference proceedings but also in ''grey literature''. This paper presents a summary of the results of corrosion experiments with fine-grained steels and nodular cast steel.

  19. Corrosion of canister materials for radioactive waste disposal

    International Nuclear Information System (INIS)

    Kienzler, Bernhard

    2017-01-01

    In the period between 1980 and 2004, corrosion studies on various metallic materials have been performed at the Research Center Karlsruhe. The objectives of these experimental studies addressed mainly the performance of canister materials for heat producing, high-level wastes and spent nuclear fuels for a repository in a German salt dome. Additional studies covered the performance of steels for packaging wastes with negligible heat production under conditions to be expected in rocksalt and in the Konrad iron ore mine. The results of the investigations have been published in journals and conference proceedings but also in ''grey literature''. This paper presents a summary of the results of corrosion experiments with fine-grained steels and nodular cast steel.

  20. CHARACTERIZATION OF A PRECIPITATE REACTOR FEED TANK (PRFT) SAMPLE FROM THE DEFENSE WASTE PROCESSING FACILITY (DWPF)

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, C.; Bannochie, C.

    2014-05-12

    A sample of from the Defense Waste Processing Facility (DWPF) Precipitate Reactor Feed Tank (PRFT) was pulled and sent to the Savannah River National Laboratory (SRNL) in June of 2013. The PRFT in DWPF receives Actinide Removal Process (ARP)/ Monosodium Titanate (MST) material from the 512-S Facility via the 511-S Facility. This 2.2 L sample was to be used in small-scale DWPF chemical process cell testing in the Shielded Cells Facility of SRNL. A 1L sub-sample portion was characterized to determine the physical properties such as weight percent solids, density, particle size distribution and crystalline phase identification. Further chemical analysis of the PRFT filtrate and dissolved slurry included metals and anions as well as carbon and base analysis. This technical report describes the characterization and analysis of the PRFT sample from DWPF. At SRNL, the 2.2 L PRFT sample was composited from eleven separate samples received from DWPF. The visible solids were observed to be relatively quick settling which allowed for the rinsing of the original shipping vials with PRFT supernate on the same day as compositing. Most analyses were performed in triplicate except for particle size distribution (PSD), X-ray diffraction (XRD), Scanning Electron Microscopy (SEM) and thermogravimetric analysis (TGA). PRFT slurry samples were dissolved using a mixed HNO3/HF acid for subsequent Inductively Coupled Plasma Atomic Emission Spectroscopy (ICPAES) and Inductively Coupled Plasma Mass Spectroscopy (ICP-MS) analyses performed by SRNL Analytical Development (AD). Per the task request for this work, analysis of the PRFT slurry and filtrate for metals, anions, carbon and base were primarily performed to support the planned chemical process cell testing and to provide additional component concentrations in addition to the limited data available from DWPF. Analysis of the insoluble solids portion of the PRFT slurry was aimed at detailed characterization of these solids (TGA, PSD

  1. Copper canisters for nuclear high level waste disposal. Corrosion aspects

    International Nuclear Information System (INIS)

    Werme, L.; Sellin, P.; Kjellbert, N.

    1992-10-01

    A corrosion analysis of a thick-walled copper canister for spent fuel disposal is discussed. The analysis has shown that there are no rapid mechanisms that may lead to canister failure, indicating an anticipated corrosion service life of several millions years. If further analysis of the copper canister is considered, it should be concentrated on identifying and evaluating processes other than corrosion, which may have a potential for leading to canister failure. (au)

  2. Translating DWPF design criteria into an engineered facility design

    International Nuclear Information System (INIS)

    Kemp, J.B.

    1986-01-01

    The Defense Waste Processing Facility (DWPF) takes radioactive defense waste sludge and the radioactive nuclides, cesium and strontium, from the salt solution, and incorporates them in borosilicate glass in stainless steel canisters, for subsequent disposal in a deep geologic repository. The facility was designed by Bechtel National, Inc. under a subcontract from E.I. DuPont de Nemurs and Co., the prime contractor for the Department of Energy, for the design, construction and commissioning of the plant. The design criteria were specified by the DuPont Company, based upon their extensive experience as designer, and operator since the early 1950's, of the existing Savannah River Plant facilities. Some of the design criteria imposed unusual or new requirements on the detailed design of the facilities. This paper describes some of these criteria, encompassing several engineering disciplines, and discusses the solutions and designs which were developed for the DWPF

  3. Value Engineering Study for Closing Waste Packages Containing TAD Canisters

    International Nuclear Information System (INIS)

    Colleen Shelton-Davis

    2005-01-01

    The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system

  4. Value Engineering Study for Closing Waste Packages Containing TAD Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Colleen Shelton-Davis

    2005-11-01

    The Office of Civilian Radioactive Waste Management announced their intention to have the commercial utilities package spent nuclear fuel in shielded, transportable, ageable, and disposable containers prior to shipment to the Yucca Mountain repository. This will change the conditions used as a basis for the design of the waste package closure system. The environment is now expected to be a low radiation, low contamination area. A value engineering study was completed to evaluate possible modifications to the existing closure system using the revised requirements. Four alternatives were identified and evaluated against a set of weighted criteria. The alternatives are (1) a radiation-hardened, remote automated system (the current baseline design); (2) a nonradiation-hardened, remote automated system (with personnel intervention if necessary); (3) a nonradiation-hardened, semi-automated system with personnel access for routine manual operations; and (4) a nonradiation-hardened, fully manual system with full-time personnel access. Based on the study, the recommended design is Alternative 2, a nonradiation-hardened, remote automated system. It is less expensive and less complex than the current baseline system, because nonradiation-hardened equipment can be used and some contamination control equipment is no longer needed. In addition, the inclusion of remote automation ensures throughput requirements are met, provides a more reliable process, and provides greater protection for employees from industrial accidents and radiation exposure than the semi-automated or manual systems. Other items addressed during the value engineering study as requested by OCRWM include a comparison to industry canister closure systems and corresponding lessons learned; consideration of closing a transportable, ageable, and disposable canister; and an estimate of the time required to perform a demonstration of the recommended closure system.

  5. Criticality safety calculations for the nuclear waste disposal canisters

    International Nuclear Information System (INIS)

    Anttila, M.

    1996-12-01

    The criticality safety of the copper/iron canisters developed for the final disposal of the Finnish spent fuel has been studied with the MCNP4A code based on the Monte Carlo technique and with the fuel assembly burnup programs CASMO-HEX and CASMO-4. Two rather similar types of spent fuel disposal canisters have been studied. One canister type has been designed for hexagonal VVER-440 fuel assemblies used at the Loviisa nuclear power plant (IVO canister) and the other one for square BWR fuel bundles used at the Olkiluoto nuclear power plant (TVO canister). (10 refs.)

  6. Preliminary analysis of species partitioning in the DWPF melter. Sludge batch 7A

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Smith III, F. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-01-01

    The work described in this report is preliminary in nature since its goal was to demonstrate the feasibility of estimating the off-gas carryover from the Defense Waste Processing Facility (DWPF) melter based on a simple mass balance using measured feed and glass pour stream (PS) compositions and time-averaged melter operating data over the duration of one canister-filling cycle. The DWPF has been in radioactive operation for over 20 years processing a wide range of high-level waste (HLW) feed compositions under varying conditions such as bubbled vs. non-bubbled and feeding vs. idling. So it is desirable to find out how the varying feed compositions and operating parameters would have impacted the off-gas entrainment. However, the DWPF melter is not equipped with off-gas sampling or monitoring capabilities, so it is not feasible to measure off-gas entrainment rates directly. The proposed method provides an indirect way of doing so.

  7. The design analysis of ACP-canister for nuclear waste disposal

    International Nuclear Information System (INIS)

    Raiko, H.

    1992-05-01

    The design basis, dimensioning and some manufacturing aspects of the Advanced Cold Process Canister (ACPC) for the nuclear waste disposal is summarized in the report. The strength of the canister has been evaluated in normal design load condition and in extreme high hydrostatic pressure load condition possibly caused by ice age (orig.)

  8. NAC gets OK for waste canister, looks for buyers

    International Nuclear Information System (INIS)

    Newman, P.

    1994-01-01

    The Nuclear Regulatory Commission has given design approval to the first dual-purpose waste cansiter suitable for storing and transporting irradiated nuclear fuel. The cask could be commercially available by January 1995. NRC issued a transportation certificate for the canister, which was developed by Atlanta-based NAC Services Inc., a subsidiary of NAC Holding Inc. That certificate, which says the cask is a suitable vessel for transporting radioactive wastes by rail and truck, is the first credential of a two-part licensing process the design must acquire. Testing of the cask has been extensive, including drop tests and pin-puncture tests. Roughly 19 feet long and eight feet in diameter, the cask is designed to hold 26 pressurized water reactor fuel assemblies. NAC officials say the cask design will soon be adapted to accomodate larger boiling water reactor fuel assemblies. Utilities will need some convincing that the dual-purpose $1.5 million cask is worth the money, particularly since companies currently have no use for the cask's transportation capabilities

  9. Remote handling of canisters containing nuclear waste in glass at the Savannah River Plant

    International Nuclear Information System (INIS)

    Callan, J.E.

    1986-01-01

    The Defense Waste Processing Facility is being constructed at the Savannah River Plant at a cost of $870 million to immobilize the defense high-level radioactive waste. This radioactive waste is being added to borosilicate glass for later disposal in a federal repository. The borosilicate glass is poured into stainless steel canisters for storage. These canisters must be handled remotely because of their high radioactivity, up to 5000 R/h. After the glass has been poured into the canister which will be temporarily sealed, it is transferred to a decontamination cell and decontaminated. The canister is then transferred to the weld cell where a permanent cap is welded into place. The canisters must then be transported from the processing building to a storage vault on the plant until the federal repository is available. A shielded canister transporter (SCT) has been designed and constructed for this purpose. The design of the SCT vehicle allows the safe transport of a highly radioactive canister containing borosilicate glass weighing 2300 kg with a radiation level up to 5000 R/h from one building to another. The design provides shielding for the operator in the cab of the vehicle to be below 0.5 rem/h. The SCT may also be used to load the final shipping cask when the federal repository is ready to receive the canisters

  10. Commercial radioactive waste management system feasibility with the universal canister concept. Volume 1

    International Nuclear Information System (INIS)

    Morissette, R.P.; Schneringer, P.E.; Lane, R.K.; Moore, R.L.; Young, K.A.

    1986-01-01

    A Program Research and Development Announcement (PRDA) was initiated by DOE to solicit from industry new and novel ideas for improvements in the nuclear waste management system. GA Technologies Inc. was contracted to study a system utilizing a universal canister which could be loaded at the reactor and used throughout the waste management system. The proposed canister was developed with the objective of meeting the mission requirements with maximum flexibility and at minimum cost. Canister criteria were selected from a thorough analysis of the spent fuel inventory, and canister concepts were evaluated along with the shipping and storage casks to determine the maximum payload. Engineering analyses were performed on various cask/canister combinations. One important criterion was the interchangeability of the canisters between truck and rail cask systems. A canister was selected which could hold three PWR intact fuel elements or up to eight consolidated PWR fuel elements. One canister could be shipped in an overweight truck cask or six in a rail cask. Economic analysis showed a cost savings of the reference system under consideration at that time

  11. High-level waste canister storage final design, installation, and testing. Topical report

    International Nuclear Information System (INIS)

    Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M.; Vlad, P.M.

    1998-04-01

    This report is a description of the West Valley Demonstration Project's radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project's vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access

  12. High-level waste canister storage final design, installation, and testing. Topical report

    Energy Technology Data Exchange (ETDEWEB)

    Connors, B.J.; Meigs, R.A.; Pezzimenti, D.M.; Vlad, P.M.

    1998-04-01

    This report is a description of the West Valley Demonstration Project`s radioactive waste storage facility, the Chemical Process Cell (CPC). This facility is currently being used to temporarily store vitrified waste in stainless steel canisters. These canisters are stacked two-high in a seismically designed rack system within the cell. Approximately 300 canisters will be produced during the Project`s vitrification campaign which began in June 1996. Following the completion of waste vitrification and solidification, these canisters will be transferred via rail or truck to a federal repository (when available) for permanent storage. All operations in the CPC are conducted remotely using various handling systems and equipment. Areas adjacent to or surrounding the cell provide capabilities for viewing, ventilation, and equipment/component access.

  13. Assessment of combustion and related issues in the DWPF and ITP waste tanks

    International Nuclear Information System (INIS)

    Ginsberg, T.

    1994-04-01

    This report presents a review of the safety analyses described in the DWPF Safety Analysis Report, the combustion analysis of the ITP Tanks 48 and 49, and presents conclusions drawn from interviews staff on issues related to accident analysis, in particular on issues related to combustion phenomena. The major objectives of this report are to clarify the issues related to the modes of combustion and expected loads on process vessels and structures and, in addition, to offer recommendations which would improve the defense-in-depth posture of the DWPF

  14. Heat transfer analysis of the geologic disposal of spent fuel and high-level waste storage canisters

    International Nuclear Information System (INIS)

    Allen, G.K.

    1980-08-01

    Near-field temperatures resulting from the storage of high-level waste canisters and spent unreprocessed fuel assembly canisters in geologic formations were determined. Preliminary design of the repository was modeled for a heat transfer computer code, HEATING5, which used the Crank-Nicolson finite difference method to evaluate transient heat transfer. The heat transfer system was evaluated with several two- and three-dimensional models which transfer heat by a combination of conduction, natural convention, and radiation. Physical properties of the materials in the model were based upon experimental values for the various geologic formations. The effects of canister spacing, fuel age, and use of an overpack were studied for the analysis of the spent fuel canisters; salt, granite, and basalt were considered as the storage media for spent fuel canisters. The effects of canister diameter and use of an overpack were studied for the analysis of the high-level waste canisters; salt was considered as the only storage media for high-level waste canisters. Results of the studies on spent fuel assembly canisters showed that the canisters could be stored in salt formations with a maximum heat loading of 134 kw/acre without exceeding the temperature limits set for salt stability. The use of an overpack had little effect on the peak canister temperatures. When the total heat load per acre decreased, the peak temperatures reached in the geologic formations decreased; however, the time to reach the peak temperatures increased. Results of the studies on high-level waste canisters showed that an increased canister diameter will increase the canister interior temperatures considerably; at a constant areal heat loading, a 381 mm diameter canister reached almost a 50 0 C higher temperature than a 305 mm diameter canister. An overpacked canister caused almost a 30 0 C temperature rise in either case

  15. ALTERNATIVE ANALYTICAL DIGESTION SCHEME FOR THE DEFENSE WASTE PROCESSING FACILITY (DWPF) SLURRY RECEIPT AND ADJUSTMENT TANK (SRAT) ANALYSES

    International Nuclear Information System (INIS)

    Click, D; Charles02 Coleman, C; Frank Pennebaker, F; Kristine Zeigler, K; Tommy Edwards, T

    2007-01-01

    As part of the radioactive sludge batch qualification, Savannah River National Laboratory (SRNL) performs a verification of the digestion methods to be used by the Defense Waste Processing Facility (DWPF) Lab for elemental analysis of Sludge Receipt and Adjustment Tank (SRAT) receipt process control samples and SRAT product process control samples. Verification of these methods on Sludge Batch 4 (SB4) radioactive sludge slurry indicated SB4 contains a higher concentration of aluminum (Al) than previous sludge batches. Aluminum plays a direct role in vitrification chemistry. At moderate levels, Al assists in glass forming, but at elevated levels Al can increase the viscosity of the molten glass which can adversely impact glass production rate and the volume of glass produced via limiting waste loading.3 Most of the Al present in SB4 is in the form of Al hydroxide as a mixture of gibbsite [α-aluminum trihydroxide, α-Al(OH) 3 ] and boehmite (α-aluminum oxyhydroxide, α-AlOOH) in an unknown ratio. Testing done at SRNL indicates Gibbsite is soluble at low pH but boehmite has limited solubility in the acid mixture (DWPF Cold Chem Method (CC), 25 mL nitric acid (HNO 3 ) and 25 mL hydrofluoric acid (HF)) used by DWPF to digest process control samples. Because Al plays such an important part in vitrification chemistry, it is necessary to have a robust digestion method that will dissolve all forms of Al present in the radioactive sludge while not increasing the analytical lab turnaround time. SRNL initially suggested that the DWPF lab use the sodium peroxide/hydroxide fusion (PF) digestion method4 to digest SRAT receipt and SRAT product radioactive sludge as an alternative to the acid digestion method to ensure complete digestion based on results obtained from digesting a SB4 radioactive sample.2 However, this change may have a significant impact on the DWPF lab analytical turnaround time due to the inefficiency in drying the radioactive sludge contained in a peanut

  16. Defense Waste Processing Facility (DWPF) Durability-Composition Models and the Applicability of the Associated Reduction of Constraints (ROC) Criteria for High TiO2 Containing Glasses

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Trivelpiece, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-08-30

    Radioactive high-level waste (HLW) at the Savannah River Site (SRS) has successfully been vitrified into borosilicate glass in the DWPF since 1996. Vitrification requires stringent product/process (P/P) constraints since the glass cannot be reworked once it has been poured into ten foot tall by two foot diameter canisters. A unique “feed forward” statistical process control (SPC) was developed for this control rather than relying on statistical quality control (SQC). In SPC, the feed composition to the DWPF melter is controlled prior to vitrification. In SQC, the glass product would be sampled after it is vitrified. Individual glass property-composition models form the basis for the “feed forward” SPC. The models transform constraints on the melt and glass properties into constraints on the feed composition going to the melter in order to determine, at the 95% confidence level, that the feed will be processable and that the durability of the resulting waste form will be acceptable to a geologic repository. The DWPF SPC system is known as the Product Composition Control System (PCCS). One of the process models within PCCS is known as the Thermodynamic Hydration Energy Reaction MOdel (THERMO™). The DWPF will soon be receiving increased concentrations of TiO2-, Na2O-, and Cs2O-enriched wastes from the Salt Waste Processing Facility (SWPF). The SWPF has been built to pretreat the high-curie fraction of the salt waste to be removed from the HLW tanks in the F- and H-Area Tank Farms at the SRS. In order to validate the existing TiO2 term in THERMO™ beyond 2.0 wt% in the DWPF, new durability data were developed over the target range of 2.00 to 6.00 wt% TiO2 and evaluated against the 1995 durability model. The durability was measured by the 7-day Product Consistency Test. This study documents the adequacy of the existing THERMO™ terms. It is recommended that the modified THERMO™ durability models and

  17. Development of a facility for fabricating nuclear waste canisters from radioactively contaminated steel

    International Nuclear Information System (INIS)

    Logan, J.A.; Larsen, M.M.

    1986-01-01

    This paper describes design of a facility and processes capable of using radioactively contaminated waste steel as the principal raw material for fabricating stainless steel canisters to be used for disposal of nuclear high-level waste. By such action, expenditure (i.e., permanent loss to society) of thousands of tons of uncontaminated chromium and nickel to fabricate such canisters can be avoided. Moreover, the cost and risks involved in disposing of large accumulations of radioactively contaminated steel as low-level radioactive waste (LLRW), that would otherwise be necessary, can also be avoided. The canister fabrication processes (involving centrifugal casting) described herein have been tested and proven for this application. The performance characteristics of stainless steel canisters so fabricated have been tested and agreed to by the organizations that have been involved in this development work (Battelle Memorial Institute, DuPont, EGandG and the Savannah River Laboratory) as equivalent to the performance characteristics of canisters fabricated of uncontaminated wrought stainless steel. It is estimated that the production cost for fabricating canisters by the methods described will not differ greatly from the production cost using uncontaminated wrought steel, and the other costs avoided by not having to dispose of the contaminated steel as LLRW could cause this method to produce the lowest ultimate overall costs

  18. Defense Waste Processing Facility radioactive operations -- Part 2, Glass making

    International Nuclear Information System (INIS)

    Carter, J.T.; Rueter, K.J.; Ray, J.W.; Hodoh, O.

    1996-01-01

    The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation's first and world's largest vitrification facility. Following a ten year construction period and nearly 3 year non-radioactive test program, the DWPF began radioactive operations in March, 1996. The results of the first 8 months of radioactive operations are presented. Topics include facility production from waste preparation batching to canister filling

  19. Examining the role of canister cooling conditions on the formation of nepheline from nuclear waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Christian, J. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-01

    Nepheline (NaAlSiO₄) crystals can form during slow cooling of high-level waste (HLW) glass after it has been poured into a waste canister. Formation of these crystals can adversely affect the chemical durability of the glass. The tendency for nepheline crystallization to form in a HLW glass increases with increasing concentrations of Al₂O₃ and Na₂O.

  20. Effects of stabilizers on the heat transfer characteristics of a nuclear waste canister

    International Nuclear Information System (INIS)

    Vafai, K.; Ettefagh, J.

    1986-07-01

    This report summarizes the feasibility and the effectiveness of using stabilizers (internal metal structural components) to augment the heat transfer characteristics of a nuclear waste canister. The problem was modeled as a transient two-dimensional heat transfer in two physical domains - the stabilizer and the wedge (a 30-degree-angle canister segment), which includes the heat-producing spent-fuel rods. This problem is solved by a simultaneous and interrelated numerical investigation of the two domains in cartesian and polar coordinate systems. The numerical investigations were performed for three cases. In the first case, conduction was assumed to be the dominant mechanism for heat transfer. The second case assumed that radiation was the dominant mechanism, and in the third case both radiation and conduction were considered as mechanisms of heat transfer. The results show that for typical conditions in a waste package design, the stabilizers are quite effective in reducing the overall temperature in a waste canister. Furthermore, the results show that increasing the stabilizer thickness over the thickness specified in the present design has a negligible effect on the temperature distribution in the canister. Finally, the presence of the stabilizers was found to shift the location of the peak temperature areas in the waste canister

  1. Canister materials proposed for final disposal of high level nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Mattson, E; Odoj, R; Merz, E [eds.

    1981-06-01

    The nuclear waste will be enclosed in corrosion resistant canisters. These will be deposited in repositories in geological formations, such as granite, basalt, clay, bedded or domed salt, or the sediments beneath the deep ocean floor. There the canisters will be exposed to groundwater, brine or seawater at an elevated temperature. Species formed by radiolysis may effect the corrosivity of the agent. The corrosion resistance of candidate canister materials is evaluated by corrosion tests and by thermodynamic and mass transport calculations. Examinations of ancient metal objects after long exposure in nature may give additional information. On the basis of the work carried out so far, the principal candidate canister materials are titanium materials, copper, and highpurity alumina.

  2. Iron-nickel alloys as canister material for radioactive waste disposal in underground repositories

    International Nuclear Information System (INIS)

    Apps, J.A.

    1982-01-01

    Canisters containing high-level radioactive waste must retain their integrity in an underground waste repository for at least one thousand years after burial (Nuclear Regulatory Commission, 1981). Since no direct means of verifying canister integrity is plausible over such a long period, indirect methods must be chosen. A persuasive approach is to examine the natural environment and find a suitable material which is thermodynamically compatible with the host rock under the environmental conditions with the host rock under the environmental conditions expected in a waste repository. Several candidates have been proposed, among them being iron-nickel alloys that are known to occur naturally in altered ultramafic rocks. The following review of stability relations among iron-nickel alloys below 350 0 C is the initial phase of a more detailed evaluation of these alloys as suitable canister materials

  3. DWPF Development Plan. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Holtzscheiter, E.W.

    1994-05-09

    The DWPF Development Plan is based on an evaluation process flowsheet and related waste management systems. The scope is shown in Figure 1 entitled ``DWPF Process Development Systems.`` To identify the critical development efforts, each system has been analyzed to determine: The identification of unresolved technology issues. A technology issue (TI) is one that requires basic development to resolve a previously unknown process or equipment problem and is managed via the Technology Assurance Program co-chaired by DWPF and SRTC. Areas that require further work to sufficiently define the process basis or technical operating envelop for DWPF. This activity involves the application of sound engineering and development principles to define the scope of work required to complete the technical data. The identification of the level of effort and expertise required to provide process technical consultation during the start-up and demonstration of this first of a kind plant.

  4. DWPF Development Plan. Revision 1

    International Nuclear Information System (INIS)

    Holtzscheiter, E.W.

    1994-01-01

    The DWPF Development Plan is based on an evaluation process flowsheet and related waste management systems. The scope is shown in Figure 1 entitled ''DWPF Process Development Systems.'' To identify the critical development efforts, each system has been analyzed to determine: The identification of unresolved technology issues. A technology issue (TI) is one that requires basic development to resolve a previously unknown process or equipment problem and is managed via the Technology Assurance Program co-chaired by DWPF and SRTC. Areas that require further work to sufficiently define the process basis or technical operating envelop for DWPF. This activity involves the application of sound engineering and development principles to define the scope of work required to complete the technical data. The identification of the level of effort and expertise required to provide process technical consultation during the start-up and demonstration of this first of a kind plant

  5. The Characteristics of Welding Joint on Stainless Steel as a Candidate of High Level Waste Canister

    International Nuclear Information System (INIS)

    Aisyah; Herlan-Martono

    2000-01-01

    High level waste is the waste generated from reprocessing of the spent fuels. This type of waste is vitrified with borosilicate glass to become waste-glass. This waste glass is contained in a canister made of austenitic stainless steel. The canister material is subjected to be welded during fabrication and utilization. The character of the welding joint that is the function of the electrical current used in the welding process have been studied. The strength of the joint is tested mechanically i.e.: the tensile strength and hardness test. The result shows that the higher the current used in welding process, the better the strength of the joint and as well the tensile strength. The optimum current is 110 A. From the hardness test, it was figured that the length of the HAZ area is 14 mm. The material in HAZ area is the hardest compared to the others, it is due to the appearance of the chrome-carbide. The welding of the canister with such a condition, during fabrication as well as during the utilization of the canister for the container of the high level waste with the PWHT process gives better result. (author)

  6. Decontamination of stainless steel canisters that contain high-level waste

    International Nuclear Information System (INIS)

    Bray, L.A.

    1987-01-01

    At the West Valley Nuclear Services Company (WVNSC) in West Valley, New York, high-level radioactive waste (HLW) will be vitrified into a borosilicate glass form and poured into large, stainless steel canisters. During the filling process, volatile fission products, principally 137 Cs, condense on the exterior of the canisters. The smearable contamination that remains on the canisters after they are filled and partially cooled must be removed from the canisters' exterior surfaces prior to their storage and ultimate shipment to a US Department of Energy (DOE) repository for disposal. A simple and effective method was developed for decontamination of HLW canisters. This method of chemical decontamination is applicable to a wide variety of contaminated equipment found in the nuclear industry. The process employs a reduction-oxidation system [Ce(III)/Ce(IV)] in nitric acid solution to chemically mill the surface of stainless steel, similar to the electropolishing process, but without the need for an applied electrical current. Contaminated canisters are simply immersed in the solution at controlled temperature and Ce(IV) concentration levels

  7. A review of the possible effects of hydrogen on lifetime of carbon steel nuclear waste canisters

    International Nuclear Information System (INIS)

    Turnbull, A.

    2009-07-01

    In Switzerland, the National Cooperative for the Disposal of Radioactive Waste (Nagra) is responsible for developing an effective method for the safe disposal of vitrified high level waste (HLW) and spent fuel. One of the options for disposal canisters is thick-walled carbon steel. The canisters, which would have a diameter of about 1 m and a length of about 3 m (HLW) or about 5 m (spent fuel), will be embedded in horizontal tunnels and surrounded with bentonite clay. The regulatory requirement for the minimum canister lifetime is 1000 years but demonstration of a minimum lifetime of 10,000 years would be desirable. The pore-water to which the canister will be exposed is of marine origin with about 0.1-0.3 M Cl-. Since hydrogen is generated during the corrosion process, it is necessary to assess the probability of hydrogen assisted cracking modes and to make recommendations to eliminate that probability. To that aim, key reports detailing projections for the local environment and associated corrosion rate of the waste canister have been evaluated with the focus on the implication for the absorbed hydrogen concentration in the steel. Simple calculations of hydrogen diffusion and accumulation in the inner compartment of the sealed canister indicate that a pressure equivalent to that for gas pockets external to the canister (envisaged to be about 10 MPa) may be attained in the proposed exposure time, an important consideration since it is not possible to modify the internal surface of the closure weld. Current ideas on mechanisms of hydrogen assisted cracking are assessed from which it is concluded that the mechanistic understanding and associated models of hydrogen assisted cracking are insufficient to provide a framework for quantitative prediction for this application. The emphasis then was to identify threshold conditions for cracking and to evaluate the likelihood that these may be exceeded over the lifetime of the containment. Based on an analysis of data in the

  8. Canister design concepts for disposal of spent fuel and high level waste

    Energy Technology Data Exchange (ETDEWEB)

    Patel, R.; Punshon, C.; Nicholas, J.; Bastid, P.; Zhou, R.; Schneider, C.; Bagshaw, N.; Howse, D.; Hutchinson, E. [TWI Ltd, Cambridge, (United Kingdom); Asano, R. [Hitachi Zosen Corporation, Osaka (Japan); King, S. [Integrity Corrosion Consulting Ltd, Calgary, Alberta (Canada)

    2012-10-15

    As part of its long-term plans for development of a repository for spent fuel (SF) and high level waste (HLW), Nagra is exploring various options for the selection of materials and design concepts for disposal canisters. The selection of suitable canister options is driven by a series of requirements, one of the most important of which is providing a minimum 1000 year lifetime without breach of containment. One candidate material is carbon steel, because of its relatively low corrosion rate under repository conditions and because of the advanced state of overall technical maturity related to construction and fabrication. Other materials and design options are being pursued in parallel studies. The objective of the present study was to develop conceptual designs for carbon steel SF and HLW canisters along with supporting justification. The design process and outcomes result in design concepts that deal with all key aspects of canister fabrication, welding and inspection, short-term performance (handling and emplacement) and long-term performance (corrosion and structural behaviour after disposal). A further objective of the study is to use the design process to identify the future work that is required to develop detailed designs. The development of canister designs began with the elaboration of a number of design requirements that are derived from the need to satisfy the long-term safety requirements and the operational safety requirements (robustness needed for safe handling during emplacement and potential retrieval). It has been assumed based on radiation shielding calculations that the radiation dose rate at the canister surfaces will be at a level that prohibits manual handling, and therefore a hot cell and remote handling will be needed for filling the canisters and for final welding operations. The most important canister requirements were structured hierarchically and set in the context of an overall design methodology. Conceptual designs for SF canisters

  9. Canister design concepts for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    Patel, R.; Punshon, C.; Nicholas, J.; Bastid, P.; Zhou, R.; Schneider, C.; Bagshaw, N.; Howse, D.; Hutchinson, E.; Asano, R.; King, S.

    2012-10-01

    As part of its long-term plans for development of a repository for spent fuel (SF) and high level waste (HLW), Nagra is exploring various options for the selection of materials and design concepts for disposal canisters. The selection of suitable canister options is driven by a series of requirements, one of the most important of which is providing a minimum 1000 year lifetime without breach of containment. One candidate material is carbon steel, because of its relatively low corrosion rate under repository conditions and because of the advanced state of overall technical maturity related to construction and fabrication. Other materials and design options are being pursued in parallel studies. The objective of the present study was to develop conceptual designs for carbon steel SF and HLW canisters along with supporting justification. The design process and outcomes result in design concepts that deal with all key aspects of canister fabrication, welding and inspection, short-term performance (handling and emplacement) and long-term performance (corrosion and structural behaviour after disposal). A further objective of the study is to use the design process to identify the future work that is required to develop detailed designs. The development of canister designs began with the elaboration of a number of design requirements that are derived from the need to satisfy the long-term safety requirements and the operational safety requirements (robustness needed for safe handling during emplacement and potential retrieval). It has been assumed based on radiation shielding calculations that the radiation dose rate at the canister surfaces will be at a level that prohibits manual handling, and therefore a hot cell and remote handling will be needed for filling the canisters and for final welding operations. The most important canister requirements were structured hierarchically and set in the context of an overall design methodology. Conceptual designs for SF canisters

  10. Canister Design for Deep Borehole Disposal of Nuclear Waste (CD-ROM)

    National Research Council Canada - National Science Library

    Hoag, Christopher I

    2006-01-01

    ...: 1 CD-ROM; 4 3/4 in.; 28.7 MB. ABSTRACT: The objective of this thesis was to design a canister for the disposal of spent nuclear fuel and other high-level waste in deep borehole repositories using currently available and proven oil, gas...

  11. Status of Closure Welding Technology of Canister for Transportation and Storage of High Level Radioactive Material and Waste

    International Nuclear Information System (INIS)

    Lee, H. J.; Bang, K. S.; Seo, K. S.; Seo, C. S.

    2010-10-01

    Closure seal welding is one of the key technologies in fabricating and handling the canister which is used for transportation and storage of high radioactive material and waste. Simple industrial fabrication processes are used before filling the radioactive waste into the canister. But, automatic and remote processes should be used after filling the radioactive material because the thickness of canister is not sufficient to shield the high radiation from filled material or waste. In order to simplify the welding process the closure structure of canister and the sealing method are investigated and developed properly. Two types of radioactive materials such as vitrified waste and compacted solid waste are produced in nuclear industry. Because the filling method of two types of waste is different, the shapes of closure and opening of canister and welding method is also different. The canister shape and sealing method should be standardized to standardize the handling facilities and inspection process such as leak test after closure welding. In order to improve the productivity of disposal and compatibility of the canister, the structure and shape of canister should be standardized considering the type of waste. Two kind of welding process such as arc welding and resistance welding are reported and used in the field. In the arc welding process GTAW and PAW are considered proper processes for closure welding. The closure seal welding process can be selected by considering material of canister, thickness of body, productivity, and applicable codes and rules. Because the storage time of nuclear waste in canister is very long, at least 20 years, the long-time corrosion at the weld should be estimated including mechanical integrity. Recently, the mitigation of residual stress around weld region, which causes stress corrosion cracking, is also interesting research issue

  12. Yucca Mountain project canister material corrosion studies as applied to the electrometallurgical treatment metallic waste form

    International Nuclear Information System (INIS)

    Keiser, D.D.

    1996-11-01

    Yucca Mountain, Nevada is currently being evaluated as a potential site for a geologic repository. As part of the repository assessment activities, candidate materials are being tested for possible use as construction materials for waste package containers. A large portion of this testing effort is focused on determining the long range corrosion properties, in a Yucca Mountain environment, for those materials being considered. Along similar lines, Argonne National Laboratory is testing a metallic alloy waste form that also is scheduled for disposal in a geologic repository, like Yucca Mountain. Due to the fact that Argonne's waste form will require performance testing for an environment similar to what Yucca Mountain canister materials will require, this report was constructed to focus on the types of tests that have been conducted on candidate Yucca Mountain canister materials along with some of the results from these tests. Additionally, this report will discuss testing of Argonne's metal waste form in light of the Yucca Mountain activities

  13. Effects of waste content of glass waste forms on Savannah River high-level waste disposal costs

    International Nuclear Information System (INIS)

    McDonell, W.R.; Jantzen, C.M.

    1985-01-01

    Effects of the waste content of glass waste forms of Savannah River high-level waste disposal costs are evaluated by their impact on the number of waste canisters produced. Changes in waste content affect onsite Defense Waste Processing Facility (DWPF) costs as well as offsite shipping and repository emplacement charges. A nominal 1% increase over the 28 wt % waste loading of DWPF glass would reduce disposal costs by about $50 million for Savannah River wastes generated to the year 2000. Waste form modifications under current study include adjustments of glass frit content to compensate for added salt decontamination residues and increased sludge loadings in the DWPF glass. Projected cost reductions demonstrate significant incentives for continued optimization of the glass waste loadings. 13 refs., 3 figs., 3 tabs

  14. TOUGH - a numerical model for nonisothermal unsaturated flow to study waste canister heating effects

    International Nuclear Information System (INIS)

    Pruess, K.; Wang, J.S.Y.

    1984-01-01

    The physical processes modeled and the mathematical and numerical methods employed in a simulator for non-isothermal flow of water, vapor, and air in permeable media are briefly summarized. The simulator has been applied to study thermohydrological conditions in the near vicinity of high-level nuclear waste packages emplaced in unsaturated rocks. The studies reported here specifically address the question whether or not the waste canister environment will dry up in the thermal phase. 13 references, 8 figures, 2 tables

  15. Enhanced Earthquake-Resistance on the High Level Radioactive Waste Canister

    International Nuclear Information System (INIS)

    Choi, Youngchul; Yoon, Chanhoon; Lee, Jeaowan; Kim, Jinsup; Choi, Heuijoo

    2014-01-01

    In this paper, the earthquake-resistance type buffer was developed with the method protecting safely about the earthquake. The main parameter having an effect on the earthquake-resistant performance was analyzed and the earthquake-proof type buffer material was designed. The shear analysis model was developed and the performance of the earthquake-resistance buffer material was evaluated. The dynamic behavior of the radioactive waste disposal canister was analyzed in case the earthquake was generated. In the case, the disposal canister gets the serious damage. In this paper, the earthquake-resistance buffer material was developed in order to prevent this damage. By putting the buffer in which the density is small between the canister and buffer, the earthquake-resistant performance was improved about 80%

  16. REMOTE IN-CELL SAMPLING IMPROVEMENTS PROGRAM AT THESAVANNAH RIVER SITE (SRS) DEFENSE WASTE PROCESSING FACILITY (DWPF)

    International Nuclear Information System (INIS)

    Marzolf, A

    2007-01-01

    Remote Systems Engineering (RSE) of the Savannah River National Lab (SRNL) in combination with the Defense Waste Processing Facility(DWPF) Engineering and Operations has evaluated the existing equipment and processes used in the facility sample cells for 'pulling' samples from the radioactive waste stream and performing equipment in-cell repairs/replacements. RSE has designed and tested equipment for improving remote in-cell sampling evolutions and reducing the time required for in-cell maintenance of existing equipment. The equipment within the present process tank sampling system has been in constant use since the facility start-up over 17 years ago. At present, the method for taking samples within the sample cells produces excessive maintenance and downtime due to frequent failures relative to the sampling station equipment and manipulator. Location and orientation of many sampling stations within the sample cells is not conducive to manipulator operation. The overextension of manipulators required to perform many in-cell operations is a major cause of manipulator failures. To improve sampling operations and reduce downtime due to equipment maintenance, a Portable Sampling Station (PSS), wireless in-cell cameras, and new commercially available sampling technology has been designed, developed and/or adapted and tested. The uniqueness of the design(s), the results of the scoping tests, and the benefits relative to in-cell operation and reduction of waste are presented

  17. Analysis of Welding Joint on Handling High Level Waste-Glass Canister

    International Nuclear Information System (INIS)

    Herlan Martono; Aisyah; Wati

    2007-01-01

    The analysis of welding joint of stainless steel austenitic AISI 304 for canister material has been studied. At the handling of waste-glass canister from melter below to interim storage, there is a step of welding of canister lid. Welding quality must be kept in a good condition, in order there is no gas out pass welding pores and canister be able to lift by crane. Two part of stainless steel plate in dimension (200 x 125 x 3) mm was jointed by welding. Welding was conducted by TIG machine with protection gas is argon. Electric current were conducted for welding were 70, 80, 90, 100, 110, 120, 130, and 140 A. Welded plates were cut with dimension according to JIS 3121 standard for tensile strength test. Hardness test in welding zone, HAZ, and plate were conducted by Vickers. Analysis of microstructure by optic microscope. The increasing of electric current at the welding, increasing of tensile strength of welding yields. The best quality welding yields using electric current was 110 A. At the welding with electric current more than 110 A, the electric current influence towards plate quality, so that decreasing of stainless steel plate quality and breaking at the plate. Tensile strength of stainless steel plate welding yields in requirement conditions according to application in canister transportation is 0.24 kg/mm 2 . (author)

  18. Recommendations for codes and standards to be used for design and fabrication of high level waste canister

    International Nuclear Information System (INIS)

    Bermingham, A.J.; Booker, R.J.; Booth, H.R.; Ruehle, W.G.; Shevekov, S.; Silvester, A.G.; Tagart, S.W.; Thomas, J.A.; West, R.G.

    1978-01-01

    This study identifies codes, standards, and regulatory requirements for developing design criteria for high-level waste (HLW) canisters for commercial operation. It has been determined that the canister should be designed as a pressure vessel without provision for any overpressure protection type devices. It is recommended that the HLW canister be designed and fabricated to the requirements of the ASME Section III Code, Division 1 rules, for Code Class 3 components. Identification of other applicable industry and regulatory guides and standards are provided in this report. Requirements for the Design Specification are found in the ASME Section III Code. It is recommended that design verification be conducted principally with prototype testing which will encompass normal and accident service conditions during all phases of the canister life. Adequacy of existing quality assurance and licensing standards for the canister was investigated. One of the recommendations derived from this study is a requirement that the canister be N stamped. In addition, acceptance standards for the HLW waste should be established and the waste qualified to those standards before the canister is sealed. A preliminary investigation of use of an overpack for the canister has been made, and it is concluded that the use of an overpack, as an integral part of overall canister design, is undesirable, both from a design and economics standpoint. However, use of shipping cask liners and overpack type containers at the Federal repository may make the canister and HLW management safer and more cost effective. There are several possible concepts for canister closure design. These concepts can be adapted to the canister with or without an overpack. A remote seal weld closure is considered to be one of the most suitable closure methods; however, mechanical seals should also be investigated

  19. Design and construction of the defense waste processing facility project at the Savannah River Plant

    International Nuclear Information System (INIS)

    Baxter, R.G.

    1986-01-01

    The Du Pont Company is building for the Department of Energy a facility to vitrify high-level radioactive waste at the Savannah River Plant (SRP) near Aiken, South Carolina. The Defense Waste Processing Facility (DWPF) will solidify existing and future radioactive wastes by immobilizing the waste in Processing Facility (DWPF) will solidify existing and future radioactives wastes by immobilizing the waste in borosilicate glass contained in stainless steel canisters. The canisters will be sealed, decontaminated and stored, prior to emplacement in a federal repository. At the present time, engineering and design is 90% complete, construction is 25% complete, and radioactive processing in the $870 million facility is expected to begin by late 1989. This paper describes the SRP waste characteristics, the DWPF processing, building and equipment features, and construction progress of the facility

  20. Plutonium Immobilization Project - Robotic canister loading

    International Nuclear Information System (INIS)

    Hamilton, R.L.

    2000-01-01

    The Plutonium Immobilization Program (PIP) is a joint venture between the Savannah River Site (SRS), Lawrence Livermore National Laboratory (LLNL), Argonne National Laboratory (ANL), and Pacific Northwest National Laboratory (PNNL). When operational in 2008, the PIP will fulfill the nation's nonproliferation commitment by placing surplus weapons-grade plutonium in a permanently stable ceramic form and making it unattractive for reuse. Since there are significant radiation and security concerns, the program team is developing novel and unique technology to remotely perform plutonium immobilization tasks. The remote task covered in this paper employs a jointed arm robot to load seven 3.5 inch diameter, 135-pound cylinders (magazines) through the 4 inch diameter neck of a stainless steel canister. Working through the narrow canister neck, the robot secures the magazines into a specially designed rack pre-installed in the canister. To provide the deterrent effect, the canisters are filled with a mixture of high-level waste and glass at the Defense Waste Processing Facility (DWPF)

  1. Predicted peak temperature-rises around a high-level radioactive waste canister emplaced in the deep ocean bed

    International Nuclear Information System (INIS)

    Kipp, K.L.

    1978-06-01

    A simple mathematical model of heat conduction was used to evaluate the peak temperature-rise along the wall of a canister of high-level radioactive waste buried in deep ocean sediment. Three different amounts of vitrified waste, corresponding to standard Harvest, large Harvest, and AVM canisters, and three different waste loadings were studied. Peak temperature-rise was computed for the nine cases as a function of canister geometry and storage time between reprocessing and burial. Lower waste loadings or longer storage times than initially envisaged are necessary to prevent the peak temperature-rise from exceeding 200 0 C. The use of longer, thinner cylinders only modestly reduces the storage time for a given peak temperature. Effects of stacking of waste canisters and of close-packing were also studied. (author)

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

    International Nuclear Information System (INIS)

    Michael B. Heiser; Clark B. Millet

    2005-01-01

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

  3. Research of radioactive waste storage cask/canister materials, spent nuclear fuels and various radioactive waste forms and development of their assessment methods. Final report for Stage 3

    International Nuclear Information System (INIS)

    Dobrev, D.; Balek, V.; Červinka, R.; Večerník, P.; Člupek, M.; Kouřil, M.; Novák, P.; Stoulil, J.; Silber, R.

    2013-08-01

    The main topics treated are: Research and development of methodologies for canister/cask material degradation assessment; Laboratory research of selected materials of canister/cask with radioactive waste; and Research and assessment of canister/cask materials in natural granite rocks. Two additional documents are appended: Corrosion rate determination for samples in compacted bentonite in anaerobic conditions (methodology), and Roll test for corrosion test in an occluded solution at the interface between a radioactive waste disposal canister and the bentonite cover. (P.A.)

  4. Friction stir welding - an alternative method for sealing nuclear waste storage canisters

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, R.E. [TWI Ltd, Cambridge (United Kingdom)

    2004-12-01

    When welding 50 mm thick copper a very high heat input is required to combat the high thermal diffusivity and only the Electron Beam Welding (EBW) process had this capability when this copper canister concept was conceived. Despite the encouraging results achieved using EBW with thick section copper, SKB felt that it would be prudent to assess other joining methods. This assessment concluded that friction welding, could also provide very high quality welds to satisfy the service life requirements of the SKB canister design. A friction welding variant called Friction Stir Welding (FSW) was shown to have the capability of welding 3 mm thick copper sheet with excellent integrity and reproducibility. This later provided sufficient encouragement for SKB to consider the potential of FSW as a method for joining thick section copper, using relatively simple machine tool based technology. It was thought that FSW might provide an alternative or complementary method for welding lids, or bases to canisters. In 1997 an FSW development programme started at TWI, focussed on the feasibility of welding 10 mm thick copper plate. Once this task was successfully completed, work continued to demonstrate that progressively thicker plate, up to 50 mm thick, could be joined. At this stage, with process viability established, a full size experimental FSW canister machine was designed and built. Work with this machine finished in January 2003, when it had been shown that FSW could definitely be used to weld lids to full size canisters. This report summarises the TWI development of FSW for SKB from 1997 to January 2003. It also highlights the important aspects of the process and the project milestones that will help to ensure that SKB has a welding technology that can be used with confidence for production fabrication of copper waste storage canisters in the future. The overall conclusion to this FSW development is that there is no doubt that the FSW process could be used to produce full

  5. Friction stir welding - an alternative method for sealing nuclear waste storage canisters

    International Nuclear Information System (INIS)

    Andrews, R.E.

    2004-12-01

    When welding 50 mm thick copper a very high heat input is required to combat the high thermal diffusivity and only the Electron Beam Welding (EBW) process had this capability when this copper canister concept was conceived. Despite the encouraging results achieved using EBW with thick section copper, SKB felt that it would be prudent to assess other joining methods. This assessment concluded that friction welding, could also provide very high quality welds to satisfy the service life requirements of the SKB canister design. A friction welding variant called Friction Stir Welding (FSW) was shown to have the capability of welding 3 mm thick copper sheet with excellent integrity and reproducibility. This later provided sufficient encouragement for SKB to consider the potential of FSW as a method for joining thick section copper, using relatively simple machine tool based technology. It was thought that FSW might provide an alternative or complementary method for welding lids, or bases to canisters. In 1997 an FSW development programme started at TWI, focussed on the feasibility of welding 10 mm thick copper plate. Once this task was successfully completed, work continued to demonstrate that progressively thicker plate, up to 50 mm thick, could be joined. At this stage, with process viability established, a full size experimental FSW canister machine was designed and built. Work with this machine finished in January 2003, when it had been shown that FSW could definitely be used to weld lids to full size canisters. This report summarises the TWI development of FSW for SKB from 1997 to January 2003. It also highlights the important aspects of the process and the project milestones that will help to ensure that SKB has a welding technology that can be used with confidence for production fabrication of copper waste storage canisters in the future. The overall conclusion to this FSW development is that there is no doubt that the FSW process could be used to produce full

  6. Automated waste canister docking and emplacement using a sensor-based intelligent controller

    International Nuclear Information System (INIS)

    Drotning, W.D.

    1992-08-01

    A sensor-based intelligent control system is described that utilizes a multiple degree-of-freedom robotic system for the automated remote manipulation and precision docking of large payloads such as waste canisters. Computer vision and ultrasonic proximity sensing are used to control the automated precision docking of a large object with a passive target cavity. Real-time sensor processing and model-based analysis are used to control payload position to a precision of ± 0.5 millimeter

  7. Materials evaluation programs at the Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Gee, J.T.; Iverson, D.C.; Bickford, D.F.

    1992-01-01

    The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950s to produce nuclear materials in support of the national defense effort. About 83 million gallons of high-level waste produced since operations began has been consolidated by evaporation into 33 million gallons at the waste tank farm. The Department of Energy authorized the construction of the Defense Waste Processing Facility (DWPF), the function of which is to immobilize the waste as a durable borosilicate glass contained in stainless steel canisters prior to the placement of the canisters in a federal repository. The DWPF is now mechanically complete and is undergoing commissioning and run-in activities. A brief description of the DWPF process is provided

  8. Cost analysis for application of solidified waste fission product canisters in U.S. Army steam plants

    International Nuclear Information System (INIS)

    Sande, W.E.; Bjorklund, W.J.; Brooks, N.A.

    1977-04-01

    The main objectives of the present study are to design steam plants using projected waste fission product canister characteristics, to analyze the overall impact and cost/benefit to the nuclear fuel cycle associated with these plants, and to develop plans for this application if the cost analysis so warrants it. The construction and operation of a steam plant fueled with waste fission product canisters would require the involvement and cooperation of various government agencies and private industry; thus the philosophies of these groups were studied. These philosophies are discussed, followed by a forecast of canister supply, canister characteristics, and strategies for Army canister use. Another section describes the safety and licensing of these steam plants since this affects design and capital costs. The discussion of steam plant design includes boiler concepts, boiler heat transfer, canister temperature distributions, steam plant size, and steam plant operation. Also, canister transportation is discussed since this influences operating costs. Details of economics of Army steam plants are provided including steam plant capital costs, operating costs, fuel reprocessor savings due to Army canister storage, and overall economics. Recommendations are made in the final section

  9. SCIX IMPACT ON DWPF CPC

    Energy Technology Data Exchange (ETDEWEB)

    Koopman, D.

    2011-07-14

    A program was conducted to systematically evaluate potential impacts of the proposed Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) Chemical Processing Cell (CPC). The program involved a series of interrelated tasks. Past studies of the impact of crystalline silicotitanate (CST) and monosodium titanate (MST) on DWPF were reviewed. Paper studies and material balance calculations were used to establish reasonable bounding levels of CST and MST in sludge. Following the paper studies, Sludge Batch 10 (SB10) simulant was modified to have both bounding and intermediate levels of MST and ground CST. The SCIX flow sheet includes grinding of the CST which is larger than DWPF frit when not ground. Nominal ground CST was not yet available, therefore a similar CST ground previously in Savannah River National Laboratory (SRNL) was used. It was believed that this CST was over ground and that it would bound the impact of nominal CST on sludge slurry properties. Lab-scale simulations of the DWPF CPC were conducted using SB10 simulants with no, intermediate, and bounding levels of CST and MST. Tests included both the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles. Simulations were performed at high and low acid stoichiometry. A demonstration of the extended CPC flowsheet was made that included streams from the site interim salt processing operations. A simulation using irradiated CST and MST was also completed. An extensive set of rheological measurements was made to search for potential adverse consequences of CST and MST and slurry rheology in the CPC. The SCIX CPC impact program was conducted in parallel with a program to evaluate the impact of SCIX on the final DWPF glass waste form and on the DWPF melter throughput. The studies must be considered together when evaluating the full impact of SCIX on DWPF. Due to the fact that the alternant flowsheet for DWPF has not been selected, this study did not

  10. Review of advanced techniques for waste canister labeling

    International Nuclear Information System (INIS)

    Culbreth, W.G.; Bhagi, B.K.; Kanjerla, A.

    1994-01-01

    Technology has produced several new labeling techniques that may meet the needs of a nuclear waste repository. New methods must be capable of providing permanent labels to clearly identify the contents of each package containing high-level spent nuclear fuel. Several new techniques, along with their benefits and problems, are discussed

  11. Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System

    International Nuclear Information System (INIS)

    1990-03-01

    This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs

  12. TOUGH: a numerical model for nonisothermal unsaturated flow to study waste canister heating effects

    International Nuclear Information System (INIS)

    Pruess, K.; Wang, J.S.Y.

    1983-12-01

    The physical processes modeled and the mathematical and numerical methods employed in a simulator for non-isothermal flow of water, vapor, and air in permeable media are briefly summarized. The simulator has been applied to study thermo-hydrological conditions in the near vicinity of high-level nuclear waste packages emplaced in unsaturated rocks. The studies reported here specifically address the question whether or not the waste canister environment will dry up in the thermal phase. 13 references, 8 figures, 2 tables

  13. Post-test evaluations of Waste Isolation Pilot Plant - Savannah River simulated defense HLW canisters and waste form

    International Nuclear Information System (INIS)

    Molecke, M.A.; Sorensen, N.R.; Harbour, J.R.; Ferrara, D.M.

    1993-01-01

    Eighteen nonradioactive defense high-level waste (DHLW) canisters were emplaced in and subjected to accelerated overtest thermal conditions for about three years at the bedded salt Waste Isolation Pilot Plant (WIPP) facility. Post-test laboratory corrosion results of several stainless steel 304L waste canisters, cast steel overpacks, and associated instruments ranged from negligible to moderate. We found appreciable surface corrosion and corrosion products on the cast steel overpacks. Pieces of both 304L and 316 stainless steel test apparatus underwent extensive stress-corrosion cracking failure and nonuniform attack. One of the retrieved test packages contained nonradioactive glass waste form from the Savannah River Site. We conducted post-test analyses of this glass to determine the degree of resultant glass fracturing, and whether any respirable fines were present. Linear glass fracture density ranged from about 1 to 8 fractures intersecting every 5 cm (2 inch) segment along a diameter line of the canister cross-section. Glass fines between 1 and 10 microns in diameter were detected, but were not quantified

  14. Acoustic monitoring techniques for corrosion degradation in cemented waste canisters

    International Nuclear Information System (INIS)

    Naish, C.C.; Buttle, D.; Wallace-Sims, R.; O'Brien, T.M.

    1991-01-01

    This report describes work to investigate acoustic emission as a non-intrusive monitor of corrosion and degradation of cemented wasteforms where the waste is a potentially reactive metal. The acoustic data collected shows good correlation with the corrosion rate as measured by hydrogen gas evolution rates and the electrochemically measured corrosion rates post cement hardening. The technique has been shown to be sensitive in detecting stress caused by expansive corrosion product within the cemented wasteform. The attenuation of the acoustic signal by the wasteform reduced the signal received by the monitoring equipment by a factor of 10 over a distance of approximately 150-400 mm, dependent on the water level in the cement. Full size packages were successfully monitored. It is concluded that the technique offers good potential for monitoring cemented containers of the more reactive metals, for example Magnox and aluminium. (author)

  15. Sensor-based automated docking of large waste canisters

    International Nuclear Information System (INIS)

    Drotning, W.D.

    1990-01-01

    Sensor-based programmable robots have the potential to speed up remote manipulation operations while protecting operators from exposure to radiation. Conventional master/slave manipulators have proven to be very slow in performing precision remote operations. In addition, inadvertent collisions of remotely manipulated objects with their environment increase the hazards associated with remote handling. This paper describes the development of a robotic system for the sensor-based automated remote manipulation and precision docking of large payloads. Computer vision and proximity sensing are used to control the precision docking of a large object with a passive target cavity. Specifically, a container of nuclear spent fuel on a transport vehicle is mated with an emplacement door on a vertical storage borehole at a waste repository

  16. Preliminary design of the high-level waste canister storage system: Topical report for the period of January 1, 1987--September 30, 1987

    International Nuclear Information System (INIS)

    Peters, F.E.; Leap, D.R.

    1987-11-01

    The final stage of the West Valley solidification program will be to place the high-level waste canisters in interim storage until a federal repository is ready to receive them. The waste canisters will be stored in the largest former fuel reprocessing cell at West Valley modified for this purpose. This report provides a description of the preliminary design of the Waste Canister Storage Facility. 9 refs., 14 figs., 1 tab

  17. Preliminary corrosion models for BWIP [Basalt Waste Isolation Project] canister materials

    International Nuclear Information System (INIS)

    Fish, R.L.; Anantatmula, R.P.

    1983-01-01

    Waste package development for the Basalt Waste Isolation Project (BWIP) requires the generation of materials degradation data under repository relevant conditions. These data are used to develop predictive models for the behavior of each component of waste package. The component models are exercised in performance analyses to optimize the waste package design. This document presents all repository relevant canister materials corrosion data that the BWIP and others have developed to date, describes the methodology used to develop preliminary corrosion models and provides the mathematical description of the models for both low carbon steel and Fe9Cr1Mo steel. Example environment/temperature history and model application calculations are presented to aid in understanding the models. The models are preliminary in nature and will be updated as additional corrosion data become available. 6 refs., 5 tabs

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  19. Shielded Canister Transporter

    International Nuclear Information System (INIS)

    Eidem, G.G. Jr.; Fages, R.

    1993-01-01

    The Hanford Waste Vitrification Plant (HWVP) will produce canisters filled with high-level radioactive waste immobilized in borosilicate glass. This report discusses a Shielded Canister Transporter (SCT) which will provide the means for safe transportation and handling of the canisters from the Vitrification Building to the Canister Storage Building (CSB). The stainless steel canisters are 0.61 meters in diameter, 3.0 meters tall, and weigh approximately 2,135 kilograms, with a maximum exterior surface dose rate of 90,000 R/hr. The canisters are placed into storage tubes to a maximum of three tall (two for overpack canisters) with an impact limiter placed at the tube bottom and between each canister. A floor plug seals the top of the storage tube at the operating floor level of the CSB

  20. Corrosion of high-level radioactive waste iron-canisters in contact with bentonite.

    Science.gov (United States)

    Kaufhold, Stephan; Hassel, Achim Walter; Sanders, Daniel; Dohrmann, Reiner

    2015-03-21

    Several countries favor the encapsulation of high-level radioactive waste (HLRW) in iron or steel canisters surrounded by highly compacted bentonite. In the present study the corrosion of iron in contact with different bentonites was investigated. The corrosion product was a 1:1 Fe layer silicate already described in literature (sometimes referred to as berthierine). Seven exposition test series (60 °C, 5 months) showed slightly less corrosion for the Na-bentonites compared to the Ca-bentonites. Two independent exposition tests with iron pellets and 38 different bentonites clearly proved the role of the layer charge density of the swelling clay minerals (smectites). Bentonites with high charged smectites are less corrosive than bentonites dominated by low charged ones. The type of counterion is additionally important because it determines the density of the gel and hence the solid/liquid ratio at the contact to the canister. The present study proves that the integrity of the multibarrier-system is seriously affected by the choice of the bentonite buffer encasing the metal canisters in most of the concepts. In some tests the formation of a patina was observed consisting of Fe-silicate. Up to now it is not clear why and how the patina formed. It, however, may be relevant as a corrosion inhibitor. Copyright © 2014 Elsevier B.V. All rights reserved.

  1. Selection and evaluation of inner material candidates for Spanish high level radioactive waste canisters

    International Nuclear Information System (INIS)

    Puig, Francesc; Dies, Javier; Sevilla, Manuel; Pablo, Joan de; Pueyo, Juan Jose; Miralles, Lourdes; Martinez-Esparza, Aurora

    2007-01-01

    This paper summarizes the work carried out to analyse different alternatives related to the inner material selection of the Spanish high level waste canister for long term storage. The preliminary repository design considers granitic or clay formations, compacted bentonite sealing, corrosion allowing steel canisters and glass bead filling between the fuel assemblies and canister walls. This filling material will have the primary role of avoiding the possibility of a criticality event, which becomes an issue of major importance once the container is finally breached by corrosion and flooded by groundwater. In the first place, a complete set of requirements have been devised as evaluation criteria for candidate materials examination and selection; resulting in a compilation of demands significantly deeper and more exhaustive than any other similar work found in literature, including over 20 requirements and some other general aspects that could involve improvements in repository performance. Secondly, eight materials or material families (cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, hematite, phosphates and olivine) have been chosen and examined in detail, extracting some relevant conclusions. Either cast iron, borosilicate glass, spinel or depleted uranium are considered to look quite promising for the mentioned purpose. (authors)

  2. Selection of candidate canister materials for high-level nuclear waste containment in a tuff repository

    International Nuclear Information System (INIS)

    McCright, R.D.; Weiss, H.; Juhas, M.C.; Logan, R.W.

    1983-11-01

    A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-level wastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables

  3. Defense Waste Processing Facility, Savannah River Plant

    International Nuclear Information System (INIS)

    After 10 years of research, development, and testing, the US Department of Energy is building a new facility which will prepare high-level radioactive waste for permanent disposal. The Defense Waste Processing Facility, known as the DWPF, will be the first production-scale facility of its kind in the United States. In the DWPF, high-level waste produced by defense activities at the Savannah River Plant will be processed into a solid form, borosilicate glass, suitable for permanent off-site geologic disposal. With construction beginning in the fall of 1983, the DWPT is scheduled to be operational in 1989. By 2005, the DWPF will have immobilized the backlog of high-level waste which has been accumulating in storage tanks at the Savannah River Plant since 1954. Canisters of the immobilized waste will then be ready for permanent disposal deep under the ground, safely isolated from the environment

  4. Remote process cell mercury transfer pumps for DWPF

    International Nuclear Information System (INIS)

    Nielsen, M.G.; Vaughn, V.G.

    1986-01-01

    Final design and the results of the testing performed thus far show that the water displacement of mercury to a height of 40 feet is feasible with just 6 gallons of motive water. Control of the transfer is achieved by monitoring the pump discharge pressure. An air actuated plug valve configuration successfully contained the required discharge pressure of 260 psi. The requirements of low flow and maximum separation of mercury from particulates are achieved due to the configuration of the pressure canister. The pump is capable of transferring a discrete amount of mercury with little additional slurry particulates. The success of this new pumping configuration is highlighted by the fact that it was the inspiration for other remote transfer applications tested at SRP. These application include the dual canister sample pump shown in Figure 7, as well as a successful prototype pump designed at Pacific Northwest Laboratories (PNL). The PNL pump was designed for the purpose of metering waste slurries to an electric melter. Upon completion of final pump fabrication, the Defense Waste Processing facility (DWPF) facility will have a simple and highly reliable method of remotely transferring small discrete batches of mercury as required from radioactive process vessels. 3 refs., 7 figs., 1 tab

  5. Statistical process control support during Defense Waste Processing Facility chemical runs

    International Nuclear Information System (INIS)

    Brown, K.G.

    1994-01-01

    The Product Composition Control System (PCCS) has been developed to ensure that the wasteforms produced by the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will satisfy the regulatory and processing criteria that will be imposed. The PCCS provides rigorous, statistically-defensible management of a noisy, multivariate system subject to multiple constraints. The system has been successfully tested and has been used to control the production of the first two melter feed batches during DWPF Chemical Runs. These operations will demonstrate the viability of the DWPF process. This paper provides a brief discussion of the technical foundation for the statistical process control algorithms incorporated into PCCS, and describes the results obtained and lessons learned from DWPF Cold Chemical Run operations. The DWPF will immobilize approximately 130 million liters of high-level nuclear waste currently stored at the Site in 51 carbon steel tanks. Waste handling operations separate this waste into highly radioactive sludge and precipitate streams and less radioactive water soluble salts. (In a separate facility, soluble salts are disposed of as low-level waste in a mixture of cement slag, and flyash.) In DWPF, the precipitate steam (Precipitate Hydrolysis Aqueous or PHA) is blended with the insoluble sludge and ground glass frit to produce melter feed slurry which is continuously fed to the DWPF melter. The melter produces a molten borosilicate glass which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in a geologic repository

  6. Supplemental environmental impact statement - defense waste processing facility

    International Nuclear Information System (INIS)

    1994-11-01

    This document supplements the Final Environmental Impact Statement (EIS) DOE Issued in 1982 (DOE/EIS-0082) to construct and operate the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), a major DOE installation in southwestern South Carolina. That EIS supported the decision to construct and operate the DWPF to immobilize high-level waste generated as a result of nuclear materials processing at SRS. The DWPF would use a vitrification process to incorporate the radioactive waste into borosilicate glass and seal it in stainless steel canisters for eventual disposal at a permanent geologic repository. The DWPF is now mostly constructed and nearly ready for full operation. However, DOE has made design changes to the DWPF since the 1982 EIS to improve efficiency and safety of the facility. Each of these modifications was subjected to appropriate NEPA review. The purpose of this Supplemental EIS is to assist DOE in deciding whether and how to proceed with operation of the DWPF as modified since 1982 while ensuring appropriate consideration of potential environmental effects. In this document, DOE assesses the potential environmental impacts of completing and operating the DWPF in light of these design changes, examines the impact of alternatives, and identifies potential actions to be taken to reduce adverse impacts. Evaluations of impacts on water quality, air quality, ecological systems, land use, geologic resources, cultural resources, socioeconomics, and health and safety of onsite workers and the public are included in the assessment

  7. Analysis of factors influencing the reliability of retrievable storage canisters for containment of solid high-level radioactive waste

    International Nuclear Information System (INIS)

    Mecham, W.J.; Seefeldt, W.B.; Steindler, M.J.

    1976-08-01

    The reliability of stainless steel type 304L canisters for the containment of solidified high-level radioactive wastes in the glass and calcine forms was studied. A reference system, drawn largely from information furnished by Battelle Northwest Laboratories and Atlantic Richfield Hanford Company is described. Operations include filling the canister with the appropriate waste form, interim storage at a reprocessing plant, shipment in water to a Retrievable Surface Storage Facility (RSSF), interim storage at the RSSF, and shipment to a final disposal facility. The properties of stainless steel type 304L, fission product oxides, calcine, and glass were reviewed, and mechanisms of corrosion were identified and studied. The modes of corrosion important for reliability were stress-corrosion cracking, internal pressurization of the canister by residual impurities present, intergranular attack at the waste-canister interface, and potential local effects due to migration of fission products. The key role of temperature control throughout canister lifetime is considered together with interactive effects. Methods of ameliorating adverse effects and ensuring high reliability are identified and described. Conclusions and recommendations are presented

  8. Defense waste processing facility startup progress report

    International Nuclear Information System (INIS)

    Iverson, D.C.; Elder, H.H.

    1992-01-01

    The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950's to produce nuclear materials in support of the national defense effort. About 83 million gallons of high level waste produced since operation began have been consolidated into 33 million gallons by evaporation at the waste tank farm. The Department of Energy has authorized the construction of the Defense Waste Processing Facility (DWPF) to immobilize the waste as a durable borosilicate glass contained in stainless steel canisters, prior to emplacement in a federal repository. The DWPF is now mechanically complete and undergoing commissioning and run-in activities. Cold startup testing using simulated non-radioactive feeds is scheduled to begin in November 1992 with radioactive operation scheduled to begin in May 1994. While technical issues have been identified which can potentially affect DWPF operation, they are not expected to negatively impact the start of non-radioactive startup testing

  9. Estimation of Total Error in DWPF Reported Radionuclide Inventories

    International Nuclear Information System (INIS)

    Edwards, T.B.

    1995-01-01

    This report investigates the impact of random errors due to measurement and sampling on the reported concentrations of radionuclides in DWPF's filled canister inventory resulting from each macro-batch. The objective of this investigation is to estimate the variance of the total error in reporting these radionuclide concentrations

  10. Modeling of thermal evolution of near field area around single pit mode nuclear waste canister disposal in soft rocks

    International Nuclear Information System (INIS)

    Bajpai, R.K.; Verma, A.K.; Maheshwar, Sachin

    2016-01-01

    Soft rocks like argillites/shales are under consideration worldwide as host rock for geological disposal of vitrified as well as spent fuel nuclear waste. The near field around disposed waste canister at 400-500m depth witnesses a complex heat field evolution due to varying thermal characteristics of rocks, coupling with hydraulic processes and varying intensity of heat flux from the canister. Smooth heat dissipation across the rock is desirable to avoid buildup of temperature beyond design limit (100 °C) and resultant micro fracturing due to thermal stresses in the rocks and intervening buffer clay layers. This also causes enhancement of hydraulic conductivity of the rocks, radionuclide transport and greater groundwater ingress towards the canister. Hence heat evolution modeling constitutes an important part of safety assessment of geological disposal facilities

  11. Critical review of welding technology for canisters for disposal of spent fuel and high level waste

    International Nuclear Information System (INIS)

    Pike, S.; Allen, C.; Punshon, C.; Threadgill, P.; Gallegillo, M.; Holmes, B.; Nicholas, J.

    2010-03-01

    Nagra is the Swiss national cooperative for the disposal of radioactive waste and is responsible for final disposal of all types of waste produced in Switzerland, which are partitioned into two repository types, one for spent fuel (SF), vitrified high-level waste (HLW) and long-lived intermediate level waste and one for low and intermediate level waste. In the general licences applied for these repositories, documentation has to show that long-term safety can be ensured and that factors for the construction, operation, and closure of the facility have been considered. Nagra has commissioned TWI to carry out a critical review of welding technologies for the sealing of HLW and SF canisters made of carbon steel. In conjunction with a material selection report, the information gained will be used as a preliminary step to provide input to developing design concepts for the canisters. The features to be considered are: a) Suitability of techniques for thickness of weld required; b) Suitability for remote operation, maintenance and set-up; c) Welding speed, weld quality, tolerances and cost; d) Effect of welding process on parent materials properties including microstructure corrosion resistance, distortion and residual stress; e) Potential post-weld treatments to reduce residual stress and enhance corrosion resistance; f) Suitability of inspection techniques for the weld thickness required; g) Impact of welding techniques on the canister design and material selection; h) Critique of emerging technologies which may be suitable in the future. The review of potential welding technologies began with a feasibility study carried out by TWI experts, where the unsuitable processes were rejected. For the remaining processes attention was focused on previous applications for the material and thickness suggested, and especially on safety critical applications such as applied in the nuclear and pressure vessel industry. Once the relevant information was gathered each process was

  12. Nitric-glycolic flowsheet reduction/oxidation (redox) model for the defense waste processing facility (DWPF)

    Energy Technology Data Exchange (ETDEWEB)

    Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, M. S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Trivelpiece, C. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Ramsey, W. G. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-14

    Control of the REDuction/OXidation (REDOX) state of glasses containing high concentrations of transition metals, such as High Level Waste (HLW) glasses, is critical in order to eliminate processing difficulties caused by overly reduced or overly oxidized melts. Operation of a HLW melter at Fe+2/ΣFe ratios of between 0.09 and 0.33, retains radionuclides in the melt and thus the final glass. Specifically, long-lived radioactive 99Tc species are less volatile in the reduced Tc4+ state as TcO2 than as NaTcO4 or Tc2O7, and ruthenium radionuclides in the reduced Ru4+ state are insoluble RuO2 in the melt which are not as volatile as NaRuO4 where the Ru is in the +7 oxidation state. Similarly, hazardous volatile Cr6+ occurs in oxidized melt pools as Na2CrO4 or Na2Cr2O7, while the Cr+3 state is less volatile and remains in the melt as NaCrO2 or precipitates as chrome rich spinels. The melter REDOX control balances the oxidants and reductants from the feed and from processing additives such as antifoam.

  13. DWPF recycle minimization: Brainstorming session

    International Nuclear Information System (INIS)

    Jacobs, R.A.; Poirier, M.R.

    1993-01-01

    The recycle stream from the DWPF constitutes a major source of water addition to the High Level Waste evaporator system. As now designed, the entire flow of 3.5 to 6.5 gal/min (at sign 25% and 75% attainment, respectively), or 2 gal/min during idling, flow to the 2H evaporator system (Tank 43). Substantial improvement in the HLW water balance and tank volume management is expected if the DWPF recycle to the HLW evaporator system can be significantly reduced. A task team has been appointed to study alternatives for reducing the flow to the HLW evaporator system and make recommendations for implementation and/or further study and evaluation. The brainstorming session detailed in this report was designed to produce the first cut options for the task team to further evaluate

  14. Preliminary Analysis of Species Partitioning in the DWPF Melter

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kesterson, M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Johnson, F. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McCabe, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-07-15

    The work described in this report is preliminary in nature since its goal was to demonstrate the feasibility of estimating the off-gas entrainment rates from the Defense Waste Processing Facility (DWPF) melter based on a simple mass balance using measured feed and glass pour stream compositions and timeaveraged melter operating data over the duration of one canister-filling cycle. The only case considered in this study involved the SB6 pour stream sample taken while Canister #3472 was being filled over a 20-hour period on 12/20/2010, approximately three months after the bubblers were installed. The analytical results for that pour stream sample provided the necessary glass composition data for the mass balance calculations. To estimate the “matching” feed composition, which is not necessarily the same as that of the Melter Feed Tank (MFT) batch being fed at the time of pour stream sampling, a mixing model was developed involving three preceding MFT batches as well as the one being fed at that time based on the assumption of perfect mixing in the glass pool but with an induction period to account for the process delays involved in the calcination/fusion step in the cold cap and the melter turnover.

  15. DWPF remotable television and cell lighting facilities

    International Nuclear Information System (INIS)

    Heckendorn, F.M. II.

    1984-01-01

    The Defense Waste Processing Facility (DWPF) for radioactive waste vitrification at the Savannah River Plant (SRP) is now under construction. Development of specialized low cost television (TV) viewing equipment for in-cell and within-melter applications is now complete. High resolution TV cameras not originally designed for high radiation environments have been demonstrated in crane remotable packages to be well suited to the DWPF. High intensity in-cell lighting has also been demonstrated in crane remotable assemblies. These dual 1000 W units (2000 W total) are used to support the multiplicity of TV and cell window viewing requirements. 8 figures

  16. The Swedish Concept for Disposal of Spent Nuclear Fuel: Differences Between Vertical and Horizontal Waste Canister Emplacement

    International Nuclear Information System (INIS)

    Bennett, D.G.; Hicks, T.W.

    2005-10-01

    The Swedish Nuclear Power Inspectorate (SKI) is preparing for the review of licence applications related to the disposal of spent nuclear fuel. The Swedish Nuclear Fuel and Waste Management Company (SKB) refers to its proposals for the disposal of spent nuclear fuel as the KBS-3 concept. In the KBS-3 concept, SKB plans that, after 30 to 40 years of interim storage, spent fuel will be disposed of at a depth of about 500 m in crystalline bedrock, surrounded by a system of engineered barriers. The principle barrier to radionuclide release is a cylindrical copper canister. Within the copper canister, the spent fuel is supported by a cast iron insert. Outside the copper canister is a layer of bentonite clay, known as the buffer, which is designed to provide mechanical protection for the canisters and to limit the access of groundwater and corrosive substances to their surfaces. The bentonite buffer is also designed to sorb radionuclides released from the canisters, and to filter any colloids that may form within the waste. SKB is expected to base its forthcoming licence applications on a repository design in which the waste canisters are emplaced in vertical boreholes (KBS-3V). However, SKB has also indicated that it might be possible and, in some respects, beneficial to dispose of the waste canisters in horizontal tunnels (KBS-3H). There are many similarities between the KBS-3V and KBS-3H designs. There are, however, uncertainties associated with both of the designs and, when compared, both possess relative advantages and disadvantages. SKB has identified many of the key factors that will determine the evolution of a KBS-3H repository and has plans for research and development work in many of the areas where the differences between the KBS-3V and KBS-3H designs mean that they could be significant in terms of repository performance. With respect to the KBS-3H design, key technical issues are associated with: 1. The accuracy of deposition drift construction. 2. Water

  17. High level radioactive waste management facility design criteria

    International Nuclear Information System (INIS)

    Sheikh, N.A.; Salaymeh, S.R.

    1993-01-01

    This paper discusses the engineering systems for the structural design of the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). At the DWPF, high level radioactive liquids will be mixed with glass particles and heated in a melter. This molten glass will then be poured into stainless steel canisters where it will harden. This process will transform the high level waste into a more stable, manageable substance. This paper discuss the structural design requirements for this unique one of a kind facility. A special emphasis will be concentrated on the design criteria pertaining to earthquake, wind and tornado, and flooding

  18. SLUDGE WASHING AND DEMONSTRATION OF THE DWPF FLOWSHEET IN THE SRNL SHIELDED CELLS FOR SLUDGE BATCH 5 QUALIFICATION

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J; Cj Bannochie, C; Damon Click, D; Dan Lambert, D; Michael Stone, M; Bradley Pickenheim, B; Amanda Billings, A; Ned Bibler, N

    2008-11-10

    Sludge Batch 5 (SB5) is predominantly a combination of H-modified (HM) sludge from Tank 11 that underwent aluminum dissolution in late 2007 to reduce the total mass of sludge solids and aluminum being fed to the Defense Waste Processing Facility (DWPF) and Purex sludge transferred from Tank 7. Following aluminum dissolution, the addition of Tank 7 sludge and excess Pu to Tank 51, Liquid Waste Operations (LWO) provided the Savannah River National Laboratory (SRNL) a 3-L sample of Tank 51 sludge for SB5 qualification. SB5 qualification included washing the sample per LWO plans/projections (including the addition of a Pu/Be stream from H Canyon), DWPF Chemical Process Cell (CPC) simulations, waste glass fabrication (vitrification), and waste glass chemical durability evaluation. This report documents: (1) The washing (addition of water to dilute the sludge supernatant) and concentration (decanting of supernatant) of the Tank 51 qualification sample to adjust sodium content and weight percent insoluble solids to Tank Farm projections. (2) The performance of a DWPF CPC simulation using the washed Tank 51 sample. This includes a Sludge Receipt and Adjustment Tank (SRAT) cycle, where acid is added to the sludge to destroy nitrite and remove mercury, and a Slurry Mix Evaporator (SME) cycle, where glass frit is added to the sludge in preparation for vitrification. The SME cycle also included replication of five canister decontamination additions and concentrations. Processing parameters for the CPC processing were based on work with a non radioactive simulant. (3) Vitrification of a portion of the SME product and Product Consistency Test (PCT) evaluation of the resulting glass. (4) Rheology measurements of the initial slurry samples and samples after each phase of CPC processing. This work is controlled by a Task Technical and Quality Assurance Plan (TTQAP) , and analyses are guided by an Analytical Study Plan. This work is Technical Baseline Research and Development (R

  19. Preliminary assessment of the thermal effects of an annular air space surrounding an emplaced nuclear waste canister

    International Nuclear Information System (INIS)

    Davis, B.W.

    1979-01-01

    Modeling results have previously shown that the presence of a large air space (e.g., a repository room) within a nuclear waste repository is expected to cause a waste canister's temperature to remain cooler than it would otherwise be. Results presented herein show that an annular air space surrounding the waste canisters can have similar cooling effects under certain prescribable conditions; for a 16 ft x 1 ft diameter canister containing 650 PWR rods which initially generate a total of 4.61 kw, analysis will show that annular air spaces greater than 11 in will permit the canister surface to attain peak temperatures lower than that which would result from a zero-gap/perfect thermal contact. It was determined that the peak radial temperature gradient in the salt varies in proportion to the inverse of the drill hole radius. Thermal radiation is shown to be the dominant mode of heat transfer across an annular air space during the first two years after emplacement. Finally, a methodology is presented which will allow investigators to easily model radiation and convection heat transfer through air spaces by treating the space as a conduction element that possesses non-linear temperature dependent conductivity

  20. Rates and mechanisms of radioactive release and retention inside a waste disposal canister - in Can Processes

    Energy Technology Data Exchange (ETDEWEB)

    Oversby, V.M. (ed.) [and others

    2003-10-01

    the system that will not be present under long term disposal conditions. A simulation of long-term conditions can be done using uranium dioxide that contains a short-lived isotope of uranium, but this will not include the effects of fission product and higher actinide elements on the behaviour of the spent fuel. We designed a project that had as its objective to improve the scientific understanding of the processes that control release of radioactive species from spent fuel inside a disposal canister and the chemical changes in those species that might limit release of radioactivity from the canister. If the mechanisms that control dissolution of the fuel matrix, including self-irradiation effects, can be clarified, a more realistic assessment of the long-term behaviour of spent fuel under disposal conditions can be made. By removing uncertainties concerning waste form performance, a better assessment of the individual and collective role of the engineered barriers can be made. To achieve the overall objective of the project, the following scientific and technical objectives were set. 1. Measure the actual rate of matrix dissolution of uranium dioxide under oxidising and reducing conditions. 2. Measure the effect of alpha radiolysis on the dissolution rate of uranium dioxide under oxidising and reducing conditions. 3. Measure the dissolution rate of the matrix material of spent fuel and thereby determine the additional effects of beta and gamma radiation on uranium dioxide dissolution rate under oxidising and reducing conditions. 4. Measure the ability of actively corroding iron to reduce oxidised U(VI) to U(IV) when U is present as the complex ion uranyl carbonate. 5. Measure the rate of reduction of Np(V) species in the presence of actively corroding iron. 6. Calculate the expected equilibrium and steady state concentrations of U under the conditions of the experiments used for meeting objectives 1 through 3 and compare the calculated results with those measured in

  1. Corrosion of high-level radioactive waste iron-canisters in contact with bentonite

    Energy Technology Data Exchange (ETDEWEB)

    Kaufhold, Stephan, E-mail: s.kaufhold@bgr.de [BGR, Bundesanstalt für Geowissenschaften und Rohstoffe, Stilleweg 2, D-30655 Hannover (Germany); Hassel, Achim Walter [Max-Planck-Institut für Eisenforschung GmbH, Max-Planck-Straße 1, D-40237 Düsseldorf (Germany); Institute for Chemical Technology of Inorganic Materials, Johannes Kepler University Linz, Altenberger Straße 69, 4040 Linz (Austria); Sanders, Daniel [Max-Planck-Institut für Eisenforschung GmbH, Max-Planck-Straße 1, D-40237 Düsseldorf (Germany); Dohrmann, Reiner [BGR, Bundesanstalt für Geowissenschaften und Rohstoffe, Stilleweg 2, D-30655 Hannover (Germany); LBEG, Landesamt für Bergbau, Energie und Geologie, Stilleweg 2, D-30655 Hannover (Germany)

    2015-03-21

    Graphical abstract: Corrosion at the bentonite iron interface proceeds unaerobically with formation of an 1:1 Fe silicate mineral. A series of exposure tests with different types of bentonites showed that Na–bentonites are slightly less corrosive than Ca–bentonites and highly charges smectites are less corrosive compared to low charged ones. The formation of a patina was observed in some cases and has to be investigated further. - Highlights: • At the iron bentonite interface a 1:1 Fe layer silicate forms upon corrosion. • A series of iron–bentonite corrosion products showed slightly less corrosion for Na-rich and high-charged bentonites. • In some tests the formation of a patina was observed consisting of Fe–silicate, which has to be investigated further. - Abstract: Several countries favor the encapsulation of high-level radioactive waste (HLRW) in iron or steel canisters surrounded by highly compacted bentonite. In the present study the corrosion of iron in contact with different bentonites was investigated. The corrosion product was a 1:1 Fe layer silicate already described in literature (sometimes referred to as berthierine). Seven exposition test series (60 °C, 5 months) showed slightly less corrosion for the Na–bentonites compared to the Ca–bentonites. Two independent exposition tests with iron pellets and 38 different bentonites clearly proved the role of the layer charge density of the swelling clay minerals (smectites). Bentonites with high charged smectites are less corrosive than bentonites dominated by low charged ones. The type of counterion is additionally important because it determines the density of the gel and hence the solid/liquid ratio at the contact to the canister. The present study proves that the integrity of the multibarrier-system is seriously affected by the choice of the bentonite buffer encasing the metal canisters in most of the concepts. In some tests the formation of a patina was observed consisting of Fe

  2. Efficient handling of high-level radioactive cell waste in a vitrification facility analytical laboratory

    International Nuclear Information System (INIS)

    Roberts, D.W.; Collins, K.J.

    1998-01-01

    The Savannah River Site''s (SRS) Defense Waste Processing Facility (DWPF) near Aiken, South Carolina, is the world''s largest and the United State''s first high level waste vitrification facility. For the past 1.5 years, DWPF has been vitrifying high level radioactive liquid waste left over from the Cold War. The vitrification process involves the stabilization of high level radioactive liquid waste into borosilicate glass. The glass is contained in stainless steel canisters. DWPF has filled more than 200 canisters 3.05 meters (10 feet) long and 0.61 meters (2 foot) diameter. Since operations began at DWPF in March of 1996, high level radioactive solid waste continues to be generated due to operating the facility''s analytical laboratory. The waste is referred to as cell waste and is routinely removed from the analytical laboratories. Through facility design, engineering controls, and administrative controls, DWPF has established efficient methods of handling the high level waste generated in its laboratory facility. These methods have resulted in the prevention of undue radiation exposure, wasted man-hours, expenses due to waste disposal, and the spread of contamination. This level of efficiency was not reached overnight, but it involved the collaboration of Radiological Control Operations and Laboratory personnel working together to devise methods that best benefited the facility. This paper discusses the methods that have been incorporated at DWPF for the handling of cell waste. The objective of this paper is to provide insight to good radiological and safety practices that were incorporated to handle high level radioactive waste in a laboratory setting

  3. Canister materials proposed for final disposal of high level nuclear waste - a review with respect to corrosion resistance

    Energy Technology Data Exchange (ETDEWEB)

    Mattsson, E; Odoj, R; Merz, E [eds.

    1981-06-01

    Spent fuel from nuclear reactors has to be disposed of either after reprocessing or without such treatment. Due to toxic radiation the nuclear waste has to be isolated from the biosphere for 300-1000 years, or in extreme cases for more than 100,000 years. The nuclear waste will be enclosed in corrosion resistant canisters. These will be deposited in repositories in geological formations, such as granite, basalt, clay, bedded or domed salt, or the sediments beneath the deep ocean floor. There the canisters will be exposed to groundwater, brine or seawater at an elevated temperature. Species formed by radiolysis may affect the corrosivity of the agent. The corrosion resistance of candidate canister materials is evaluated by corrosion tests and by thermodynamic and mass transport calculations. Examination of ancient metal objects after long exposure in nature may give additional information. On the basis of the work carried out so far, the principal candidate canister materials are titanium materials, copper and high purity alumina.

  4. Sandia studies of high-level waste canisters and overpacks applicable for a salt repository

    International Nuclear Information System (INIS)

    Molecke, M.A.; Schaefer, D.W.; Glass, R.S.; Ruppen, J.A.

    1982-01-01

    An experimental program to develop candidate materials for use as high-level waste (HLW) overpacks or canisters in a salt repository has been in progress at Sandia National Laboratories since 1976. The main objective of this program is to provide a waste package barrier having a long lifetime in the chemical and physical environment of a repository. This paper summarizes the recent corrosion and metallurgical study results for the prime overpack material, TiCode-12, in the areas of uniform corrosion (extremely low rate and extent); local attack, e.g., pits and crevices (none were found); stress corrosion cracking susceptibility (no significant changes in macroscopic tensile properties were detected); hydrogen sorption-embrittlement effects; effects of gamma irradiation in solution; and sensitization effects (testing is still in process in the last three areas). Previous candidate screening analyses on other alloys and recent work on alternate overpack alloys are reviewed. All phases of these interrelated laboratory, hot-cell, and field experimental studies are described. 16 references, 8 figures, 4 tables

  5. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    Rajainmaeki, H.; Nieminen, M.; Laakso, L.

    1991-08-01

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for spent fuel repository are discussed. The capsule design is based on the TVO's new advanced cold process concept where a steel canister is surrounded by the oxygen free copper canister. This study shows that already at present there exist several possible manufacturing routes, which result in consistently high quality canisters. Hot rolling, bending and EB-welding the seam is the best way to assure the small grain size which is preferable for the best inspectability of the final EB-welded seam of the lid. The same route turns out also to be the most economical

  6. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    Rajainmaeki, H.; Nieminen, M.; Laakso, L.

    1991-06-01

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for spent fuel repository are discussed. The capsule design is based on the TVO's new advanced cold process concept where a steel canister is surrounded by the oxygen free copper canister. This study shows that already at present there exist several possible manufacturing routes, which results in consistently high quality canisters. Hot rolling, bending and EB-welding the seam is the best way to assure the small grain size which is preferable for the best inspectability of the final EB-welded seam of the lid. The same route turns out also to be the most economical. (au)

  7. Overview - Defense Waste Processing Facility Operating Experience

    International Nuclear Information System (INIS)

    Norton, M.R.

    2002-01-01

    The Savannah River Site's Defense Waste Processing Facility (DWPF) near Aiken, SC is the world's largest radioactive waste vitrification facility. Radioactive operations began in March 1996 and over 1,000 canisters have been produced. This paper presents an overview of the DWPF process and a summary of recent facility operations and process improvements. These process improvements include efforts to extend the life of the DWPF melter, projects to increase facility throughput, initiatives to reduce the quantity of wastewater generated, improved remote decontamination capabilities, and improvements to remote canyon equipment to extend equipment life span. This paper also includes a review of a melt rate improvement program conducted by Savannah River Technology Center personnel. This program involved identifying the factors that impacted melt rate, conducting small scale testing of proposed process changes and developing a cost effective implementation plan

  8. DWPF simulant CPC studies for SB8

    Energy Technology Data Exchange (ETDEWEB)

    Koopman, D. C.; Zamecnik, J. R.

    2013-06-25

    The Savannah River National Laboratory (SRNL) accepted a technical task request (TTR) from Waste Solidification Engineering to perform simulant tests to support the qualification of Sludge Batch 8 (SB8) and to develop the flowsheet for SB8 in the Defense Waste Processing Facility (DWPF). These efforts pertained to the DWPF Chemical Process Cell (CPC). Separate studies were conducted for frit development and glass properties (including REDOX). The SRNL CPC effort had two primary phases divided by the decision to drop Tank 12 from the SB8 constituents. This report focuses on the second phase with SB8 compositions that do not contain the Tank 12 piece. A separate report will document the initial phase of SB8 testing that included Tank 12. The second phase of SB8 studies consisted of two sets of CPC studies. The first study involved CPC testing of an SB8 simulant for Tank 51 to support the CPC demonstration of the washed Tank 51 qualification sample in the SRNL Shielded Cells facility. SB8-Tank 51 was a high iron-low aluminum waste with fairly high mercury and moderate noble metal concentrations. Tank 51 was ultimately washed to about 1.5 M sodium which is the highest wash endpoint since SB3-Tank 51. This study included three simulations of the DWPF Sludge Receipt and Adjustment Tank (SRAT) cycle and Slurry Mix Evaporator (SME) cycle with the sludge-only flowsheet at nominal DWPF processing conditions and three different acid stoichiometries. These runs produced a set of recommendations that were used to guide the successful SRNL qualification SRAT/SME demonstration with actual Tank 51 washed waste. The second study involved five SRAT/SME runs with SB8-Tank 40 simulant. Four of the runs were designed to define the acid requirements for sludge-only processing in DWPF with respect to nitrite destruction and hydrogen generation. The fifth run was an intermediate acid stoichiometry demonstration of the coupled flowsheet for SB8. These runs produced a set of processing

  9. DWPF simulant CPC studies for SB8

    International Nuclear Information System (INIS)

    Koopman, D. C.; Zamecnik, J. R.

    2013-01-01

    The Savannah River National Laboratory (SRNL) accepted a technical task request (TTR) from Waste Solidification Engineering to perform simulant tests to support the qualification of Sludge Batch 8 (SB8) and to develop the flowsheet for SB8 in the Defense Waste Processing Facility (DWPF). These efforts pertained to the DWPF Chemical Process Cell (CPC). Separate studies were conducted for frit development and glass properties (including REDOX). The SRNL CPC effort had two primary phases divided by the decision to drop Tank 12 from the SB8 constituents. This report focuses on the second phase with SB8 compositions that do not contain the Tank 12 piece. A separate report will document the initial phase of SB8 testing that included Tank 12. The second phase of SB8 studies consisted of two sets of CPC studies. The first study involved CPC testing of an SB8 simulant for Tank 51 to support the CPC demonstration of the washed Tank 51 qualification sample in the SRNL Shielded Cells facility. SB8-Tank 51 was a high iron-low aluminum waste with fairly high mercury and moderate noble metal concentrations. Tank 51 was ultimately washed to about 1.5 M sodium which is the highest wash endpoint since SB3-Tank 51. This study included three simulations of the DWPF Sludge Receipt and Adjustment Tank (SRAT) cycle and Slurry Mix Evaporator (SME) cycle with the sludge-only flowsheet at nominal DWPF processing conditions and three different acid stoichiometries. These runs produced a set of recommendations that were used to guide the successful SRNL qualification SRAT/SME demonstration with actual Tank 51 washed waste. The second study involved five SRAT/SME runs with SB8-Tank 40 simulant. Four of the runs were designed to define the acid requirements for sludge-only processing in DWPF with respect to nitrite destruction and hydrogen generation. The fifth run was an intermediate acid stoichiometry demonstration of the coupled flowsheet for SB8. These runs produced a set of processing

  10. Measurement of the volatility and glass transition temperatures of glasses produced during the DWPF startup test program

    International Nuclear Information System (INIS)

    Marra, J.C.; Harbour, J.R.

    1995-01-01

    The Defense Waste Processing Facility (DWPF) will immobilize high-level radioactive waste currently stored in underground tanks at the Savannah River Site by incorporating the waste into a glass matrix. The molten waste glass will be poured into stainless steel canisters which will be welded shut to produce the final waste form. One specification requires that any volatiles produced as a result of accidentally heating the waste glass to the glass transition temperature be identified. Glass samples from five melter campaigns, run as part of the DWPF Startup Test Program, were analyzed to determine glass transition temperatures and to examine the volatilization (by weight loss). Glass transition temperatures (T g ) for the glasses, determined by differential scanning calorimetry (DSC), ranged between 445 C and 474 C. Thermogravimetric analysis (TGA) scans showed that no overall weight loss occurred in any of the glass samples when heated to 500 C. Therefore, no volatility will occur in the final glass product when heated up to 500 C

  11. Verification Of The Defense Waste Processing Facility's (DWPF) Process Digestion Methods For The Sludge Batch 8 Qualification Sample

    International Nuclear Information System (INIS)

    Click, D. R.; Edwards, T. B.; Wiedenman, B. J.; Brown, L. W.

    2013-01-01

    This report contains the results and comparison of data generated from inductively coupled plasma atomic emission spectroscopy (ICP-AES) analysis of Aqua Regia (AR), Sodium Peroxide/Sodium Hydroxide Fusion Dissolution (PF) and Cold Chem (CC) method digestions and Cold Vapor Atomic Absorption analysis of Hg digestions from the DWPF Hg digestion method of Sludge Batch 8 (SB8) Sludge Receipt and Adjustment Tank (SRAT) Receipt and SB8 SRAT Product samples. The SB8 SRAT Receipt and SB8 SRAT Product samples were prepared in the SRNL Shielded Cells, and the SRAT Receipt material is representative of the sludge that constitutes the SB8 Batch or qualification composition. This is the sludge in Tank 51 that is to be transferred into Tank 40, which will contain the heel of Sludge Batch 7b (SB7b), to form the SB8 Blend composition

  12. Brine: a computer program to compute brine migration adjacent to a nuclear waste canister in a salt repository

    International Nuclear Information System (INIS)

    Duckworth, G.D.; Fuller, M.E.

    1980-01-01

    This report presents a mathematical model used to predict brine migration toward a nuclear waste canister in a bedded salt repository. The mathematical model is implemented in a computer program called BRINE. The program is written in FORTRAN and executes in the batch mode on a CDC 7600. A description of the program input requirements and output available is included. Samples of input and output are given

  13. Systems costs for disposal of Savannah River high-level waste sludge and salt

    International Nuclear Information System (INIS)

    McDonell, W.R.; Goodlett, C.B.

    1984-01-01

    A systems cost model has been developed to support disposal of defense high-level waste sludge and salt generated at the Savannah River Plant. Waste processing activities covered by the model include decontamination of the salt by a precipitation process in the waste storage tanks, incorporation of the sludge and radionuclides removed from the salt into glass in the Defense Waste Processing Facility (DWPF), and, after interim storage, final disposal of the DWPF glass waste canisters in a federal geologic repository. Total costs for processing of waste generated to the year 2000 are estimated to be about $2.9 billion (1984 dollars); incremental unit costs for DWPF and repository disposal activities range from $120,000 to $170,000 per canister depending on DWPF processing schedules. In a representative evaluation of process alternatives, the model is used to demonstrate cost effectiveness of adjustments in the frit content of the waste glass to reduce impacts of wastes generated by the salt decontamination operations. 13 references, 8 tables

  14. The Defense Waste Processing Facility: an innovative process for high-level waste immobilization

    International Nuclear Information System (INIS)

    Cowan, S.P.

    1985-01-01

    The Defense Waste Processing Facility (DWPF), under construction at the Department of Energy's Savannah River Plant (SRP), will process defense high-level radioactive waste so that it can be disposed of safely. The DWPF will immobilize the high activity fraction of the waste in borosilicate glass cast in stainless steel canisters which can be handled, stored, transported and disposed of in a geologic repository. The low-activity fraction of the waste, which represents about 90% of the high-level waste HLW volume, will be decontaminated and disposed of on the SRP site. After decontamination the canister will be welded shut by an upset resistance welding technique. In this process a slightly oversized plug is pressed into the canister opening. At the same time a large current is passed through the canister and plug. The higher resistance of the canister/plug interface causes the heat which welds the plug in place. This process provides a high quality, reliable weld by a process easily operated remotely

  15. High-level waste processing at the Savannah River Site: An update

    International Nuclear Information System (INIS)

    Marra, J.E.; Bennett, W.M.; Elder, H.H.; Lee, E.D.; Marra, S.L.; Rutland, P.L.

    1997-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, SC mg began immobilizing high-level radioactive waste in borosilicate glass in 1996. Currently, the radioactive glass is being produced as a ''sludge-only'' composition by combining washed high-level waste sludge with glass frit. The glass is poured in stainless steel canisters which will eventually be disposed of in a permanent, geological repository. To date, DWPF has produced about 100 canisters of vitrified waste. Future processing operations will, be based on a ''coupled'' feed of washed high-level waste sludge, precipitated cesium, and glass frit. This paper provides an update of the processing activities completed to date, operational/flowsheet problems encountered, and programs underway to increase production rates

  16. Retrieval of canistered experimental waste at the waste isolation pilot plant

    International Nuclear Information System (INIS)

    Stinebaugh, R.E.

    1979-07-01

    To assess the suitability of bedded salt for nuclear waste disposal, an extensive experimental program will be implemented at the Waste Isolation Pilot Plant. In order to evaluate experimental results, it will be necessary to recover certain of these experiments for postmortem examination and analysis. This document describes the equipment and procedures used to effect recovery of one category of WIPP experiments

  17. Corrosion research of alloys for use in nuclear waste disposal canisters

    International Nuclear Information System (INIS)

    Sorensen, N.R.; Diegle, R.B.

    1984-01-01

    All tests involving the corrosion behavior of TiCode-12 (Ti-0.3Mo-0.8Ni) indicate that this alloy is a suitable choice for the containment of nuclear wastes. Although cracking was observed under conditions of hydrogen embrittlement, and pitting occurred in aggressive media, the conditions which produce these two failure modes represented severe overtests and are not expected in a repository. Both Mo and Ni increase the degree of corrosion resistance of Ti alloys. Mo changes the nature of the oxide, while Ni forms intermetallic precipitates which, through galvanic coupling, polarize the Ti into the passive region. All alternate alloys tested (including cast iron) exhibit acceptable corrosion behavior relative to their role as backup canister materials. In the case of cast iron, the material provides for a corrosion allowance design in contrast to the corrosion resistant design offered by Ti alloys. The effects of gamma irradiation still need to be adequately characterized. Research on such effects is underway at Sandia and the University of Minnesota. 11 references, 5 figures, 1 table

  18. Production methods and costs of oxygen free copper canisters for nuclear waste disposal

    International Nuclear Information System (INIS)

    Aalto, H.; Rajainmaeki, H.; Laakso, L.

    1996-10-01

    The fabrication technology and costs of various manufacturing alternatives to make large copper canisters for disposal of spent nuclear fuel from reactors of Teollisuuden Voima Oy (TVO) and Imatran Voima Oy (IVO) are discussed. The canister design is based on the Posiva's concept where solid insert structure is surrounded by the copper mantle. During recent years Outokumpu Copper Products and Posiva have continued their work on development of the copper canisters. Outokumpu Copper Products has also increased capability to manufacture these canisters. In the study the most potential manufacturing methods and their costs are discussed. The cost estimates are based on the assumption that Outokumpu will supply complete copper mantles. At the moment there are at least two commercially available production methods for copper cylinder manufacturing. These routes are based on either hot extrusion of the copper tube or hot rolling, bending and EB-welding of the tube. Trial fabrications has been carried out with both methods for the full size canisters. These trials of the canisters has shown that both the forming from rolled plate and the extrusion are possible methods for fabricating copper canisters on a full scale. (orig.) (26 refs.)

  19. Glass sampling program during DWPF Integrated Cold Runs

    International Nuclear Information System (INIS)

    Plodinec, M.J.

    1990-01-01

    The described glass sampling program is designed to achieve two objectives: To demonstrate Defense Waste Processing Facility (DWPF) ability to control and verify the radionuclide release properties of the glass product; To confirm DWPF's readiness to obtain glass samples during production, and SRL's readiness to analyze and test those samples remotely. The DWPF strategy for control of the radionuclide release properties of the glass product, and verification of its acceptability are described in this report. The basic approach of the test program is then defined

  20. The role of the canister in a system for the final disposal of spent fuel or high-level waste

    International Nuclear Information System (INIS)

    Papp, T.

    1986-01-01

    A final repository for radioactive waste must isolate the toxic substances or distribute their release over time or space to avoid causing harmful concentrations of radionuclides in the biosphere. The Swedish research has focused on a repository 500 m down in crystalline rock where the geochemical environment can give canisters a service life of the order of a million years. These evaluations are discussed and the safety effect of the canister is compared with that of other barriers available in a repository system. Our conclusions are that a combined protection effect of natural and man-made barriers can be achieved that substantially exceeds what could reasonably be required by society. An actual repository design can then be based on an optimization of the cost to reach a level of accepted safety with due regard for the safety margins and redundancy necessary for achieving public confidence. (author)

  1. DWPF Recycle Evaporator Simulant Tests

    International Nuclear Information System (INIS)

    Stone, M

    2005-01-01

    Testing was performed to determine the feasibility and processing characteristics of an evaporation process to reduce the volume of the recycle stream from the Defense Waste Processing Facility (DWPF). The concentrated recycle would be returned to DWPF while the overhead condensate would be transferred to the Effluent Treatment Plant. Various blends of evaporator feed were tested using simulants developed from characterization of actual recycle streams from DWPF and input from DWPF-Engineering. The simulated feed was evaporated in laboratory scale apparatus to target a 30X volume reduction. Condensate and concentrate samples from each run were analyzed and the process characteristics (foaming, scaling, etc) were visually monitored during each run. The following conclusions were made from the testing: Concentration of the ''typical'' recycle stream in DWPF by 30X was feasible. The addition of DWTT recycle streams to the typical recycle stream raises the solids content of the evaporator feed considerably and lowers the amount of concentration that can be achieved. Foaming was noted during all evaporation tests and must be addressed prior to operation of the full-scale evaporator. Tests were conducted that identified Dow Corning 2210 as an antifoam candidate that warrants further evaluation. The condensate has the potential to exceed the ETP WAC for mercury, silicon, and TOC. Controlling the amount of equipment decontamination recycle in the evaporator blend would help meet the TOC limits. The evaporator condensate will be saturated with mercury and elemental mercury will collect in the evaporator condensate collection vessel. No scaling on heating surfaces was noted during the tests, but splatter onto the walls of the evaporation vessels led to a buildup of solids. These solids were difficult to remove with 2M nitric acid. Precipitation of solids was not noted during the testing. Some of the aluminum present in the recycle streams was converted from gibbsite to

  2. CANISTER TRANSFER SYSTEM DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    B. Gorpani

    2000-01-01

    The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane hoist,; DC--loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the; DC--is fully loaded, the Disposal Container Transport System moves the; DC--to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister

  3. Canister Transfer System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

    The Canister Transfer System receives transportation casks containing large and small disposable canisters, unloads the canisters from the casks, stores the canisters as required, loads them into disposal containers (DCs), and prepares the empty casks for re-shipment. Cask unloading begins with cask inspection, sampling, and lid bolt removal operations. The cask lids are removed and the canisters are unloaded. Small canisters are loaded directly into a DC, or are stored until enough canisters are available to fill a DC. Large canisters are loaded directly into a DC. Transportation casks and related components are decontaminated as required, and empty casks are prepared for re-shipment. One independent, remotely operated canister transfer line is provided in the Waste Handling Building System. The canister transfer line consists of a Cask Transport System, Cask Preparation System, Canister Handling System, Disposal Container Transport System, an off-normal canister handling cell with a transfer tunnel connecting the two cells, and Control and Tracking System. The Canister Transfer System operating sequence begins with moving transportation casks to the cask preparation area with the Cask Transport System. The Cask Preparation System prepares the cask for unloading and consists of cask preparation manipulator, cask inspection and sampling equipment, and decontamination equipment. The Canister Handling System unloads the canister(s) and places them into a DC. Handling equipment consists of a bridge crane/hoist, DC loading manipulator, lifting fixtures, and small canister staging racks. Once the cask has been unloaded, the Cask Preparation System decontaminates the cask exterior and returns it to the Carrier/Cask Handling System via the Cask Transport System. After the DC is fully loaded, the Disposal Container Transport System moves the DC to the Disposal Container Handling System for welding. To handle off-normal canisters, a separate off-normal canister handling

  4. Task technical plan: DWPF air permit/dispersion modeling

    International Nuclear Information System (INIS)

    Lambert, D.P.

    1993-01-01

    This Task Technical Plan summarizes work required to project the benzene emissions from the Late Wash Facility (LWF) as well as update the benzene, mercury, and NO x emissions from the remainder of the Defense Waste Processing Facility (DWPF). These calculations will reflect (1) the addition of the LWF and (2) the replacement of formic acid with nitric acid in the melter preparation process. The completed calculations will be used to assist DWPF in applying for the LWF Air Quality Permit

  5. Formic Acid Free Flowsheet Development To Eliminate Catalytic Hydrogen Generation In The Defense Waste Processing

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, Dan P.; Stone, Michael E.; Newell, J. David; Fellinger, Terri L.; Bricker, Jonathan M.

    2012-09-14

    The Defense Waste Processing Facility (DWPF) processes legacy nuclear waste generated at the Savannah River Site (SRS) during production of plutonium and tritium demanded by the Cold War. The nuclear waste is first treated via a complex sequence of controlled chemical reactions and then vitrified into a borosilicate glass form and poured into stainless steel canisters. Converting the nuclear waste into borosilicate glass canisters is a safe, effective way to reduce the volume of the waste and stabilize the radionuclides. Testing was initiated to determine whether the elimination of formic acid from the DWPF's chemical processing flowsheet would eliminate catalytic hydrogen generation. Historically, hydrogen is generated in chemical processing of alkaline High Level Waste sludge in DWPF. In current processing, sludge is combined with nitric and formic acid to neutralize the waste, reduce mercury and manganese, destroy nitrite, and modify (thin) the slurry rheology. The noble metal catalyzed formic acid decomposition produces hydrogen and carbon dioxide. Elimination of formic acid by replacement with glycolic acid has the potential to eliminate the production of catalytic hydrogen. Flowsheet testing was performed to develop the nitric-glycolic acid flowsheet as an alternative to the nitric-formic flowsheet currently being processed at the DWPF. This new flowsheet has shown that mercury can be reduced and removed by steam stripping in DWPF with no catalytic hydrogen generation. All processing objectives were also met, including greatly reducing the Slurry Mix Evaporator (SME) product yield stress as compared to the baseline nitric/formic flowsheet. Ten DWPF tests were performed with nonradioactive simulants designed to cover a broad compositional range. No hydrogen was generated in testing without formic acid.

  6. Development and demonstration of prototype transportation equipment for emplacing HL vitrified waste canisters into small diameter bored horizontal disposal cells

    International Nuclear Information System (INIS)

    Seidler, Wolf K.; Bosgiraud, Jean-Michel; Londe, Louis

    2008-01-01

    Over a period of 4 and years the National Radioactive Waste Management Agency (Andra), working with a variety of Contractors mostly specializing in nuclear orientated mechanical applications, successfully designed, fabricated and demonstrated 2 very different prototype high level waste transport systems. The first system, based on air cushion technology, was developed primarily for very heavy loads (17 to 45 tonnes). The results of this work are described in a separate presentation (Paper 21) at this Conference. The second system, developed by Andra within the framework of the ESDRED Project, generally referred to as the 'Pushing Robot System' for vitrified waste canisters, is the subject of this paper. The 'Pushing Robot System' is a part of the French national disposal concept that is described in Andra's 'Dossier 2005'. The latter is a public document that can be viewed on Andra's web site (www.andra.fr). The 'Pushing Robot System' system is designed for the deep geological disposal (in clay formations) of 'C' type vitrified waste canisters. In its entirety the system provides for the transport, emplacement and, if necessary, the retrieval of those canisters. Nothing in the design of the Andra emplacement equipment would preclude its utilization in horizontal openings in other types of geological settings. Over a period of some 8 years Andra has developed the 'Pushing Robot System' in 3 phases. Initially there was only the 'Conceptual Design' (Phase 1) which was incorporated in the Dossier 2005. This was followed by Phase 2 i.e. the design and fabrication of a simplified full scale prototype system henceforth referred to a P1, which includes a Pushing Robot, a Dummy Canister and a Test Bench. P1 details were also incorporated in the Dossier 2005. Finally, during Phase 3, a second more comprehensive full scale prototype system P2 has been designed and is being assembled and tested this month. This system includes a Transport Shuttle, a Transfer Shielding Cask, a

  7. Conceptual waste package interim product specifications and data requirements for disposal of borosilicate glass defense high-level waste forms in salt geologic repositories

    International Nuclear Information System (INIS)

    1983-06-01

    The conceptual waste package interim product specifications and data requirements presented are applicable specifically to the normal borosilicate glass product of the Defense Waste Processing Facility (DWPF). They provide preliminary numerical values for the defense high-level waste form parameters and properties identified in the waste form performance specification for geologic isolation in salt repositories. Subject areas treated include containment and isolation, operational period safety, criticality control, waste form/production canister identification, and waste package performance testing requirements. This document was generated for use in the development of conceptual waste package designs in salt. It will be revised as additional data, analyses, and regulatory requirements become available

  8. Liquidus Temperature Data for DWPF Glass

    International Nuclear Information System (INIS)

    Piepel, G.F.; Vienna, J.D.; Crum, J.V.; Mika, M.; Hrma, P.

    1999-01-01

    This report provides new liquidus temperature (T L ) versus composition data that can be used to reduce uncertainty in T L calculation for DWPF glass. According to the test plan and test matrix design PNNL has measured T L for 53 glasses within and just outside of the current DWPF processing composition window. The T L database generated under this task will directly support developing and enhancing the current T L process-control model. Preliminary calculations have shown a high probability of increasing HLW loading in glass produced at the SRS and Hanford. This increase in waste loading will decrease the life-cycle tank cleanup costs by decreasing process time and the volume of waste glass produced

  9. Tritium Packages and 17th RH Canister Categories of Transuranic Waste Stored Below Ground within Area G

    Energy Technology Data Exchange (ETDEWEB)

    Hargis, Kenneth Marshall [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-01

    A large wildfire called the Las Conchas Fire burned large areas near Los Alamos National Laboratory (LANL) in 2011 and heightened public concern and news media attention over transuranic (TRU) waste stored at LANL’s Technical Area 54 (TA-54) Area G waste management facility. The removal of TRU waste from Area G had been placed at a lower priority in budget decisions for environmental cleanup at LANL because TRU waste removal is not included in the March 2005 Compliance Order on Consent (Reference 1) that is the primary regulatory driver for environmental cleanup at LANL. The Consent Order is a settlement agreement between LANL and the New Mexico Environment Department (NMED) that contains specific requirements and schedules for cleaning up historical contamination at the LANL site. After the Las Conchas Fire, discussions were held by the U.S. Department of Energy (DOE) with the NMED on accelerating TRU waste removal from LANL and disposing it at the Waste Isolation Pilot Plant (WIPP). This report summarizes available information on the origin, configuration, and composition of the waste containers within the Tritium Packages and 17th RH Canister categories; their physical and radiological characteristics; the results of the radioassays; and potential issues in retrieval and processing of the waste containers.

  10. The defense waste processing facility: the final processing step for defense high-level waste disposal

    International Nuclear Information System (INIS)

    Cowan, S.P.; Sprecher, W.M.; Walton, R.D.

    1983-01-01

    The policy of the U.S. Department of Energy is to pursue an aggressive and credible waste management program that advocates final disposal of government generated (defense) high-level nuclear wastes in a manner consistent with environmental, health, and safety responsibilities and requirements. The Defense Waste Processing Facility (DWPF) is an essential component of the Department's program. It is the first project undertaken in the United States to immobilize government generated high-level nuclear wastes for geologic disposal. The DWPF will be built at the Department's Savannah River Plant near Aiken, South Carolina. When construction is complete in 1989, the DWPF will begin processing the high-level waste at the Savannah River Plant into a borosilicate glass form, a highly insoluble and non-dispersable product, in easily handled canisters. The immobilized waste will be stored on site followed by transportation to and disposal in a Federal repository. The focus of this paper is on the DWPF. The paper discusses issues which justify the project, summarizes its technical attributes, analyzes relevant environmental and insitutional factors, describes the management approach followed in transforming technical and other concepts into concrete and steel, and concludes with observations about the future role of the facility

  11. HLW Canister and Can-In-Canister Drop Calculation

    International Nuclear Information System (INIS)

    H. Marr

    1999-01-01

    The purpose of this calculation is to evaluate the structural response of the standard high-level waste (HLW) canister and the HLW canister containing the cans of immobilized plutonium (''can-in-canister'' throughout this document) to the drop event during the handling operation. The objective of the calculation is to provide the structure parameter information to support the canister design and the waste handling facility design. Finite element solution is performed using the commercially available ANSYS Version (V) 5.4 finite element code. Two-dimensional (2-D) axisymmetric and three-dimensional (3-D) finite element representations for the standard HLW canister and the can-in-canister are developed and analyzed using the dynamic solver

  12. A review of materials and corrosion issues regarding canisters for disposal of spent fuel and high-level waste in Opalinus clay

    International Nuclear Information System (INIS)

    Landolt, D.; Davenport, A.; Payer, J.; Shoesmith, D.

    2009-01-01

    The project 'Entsorgungsnachweis' presented by NAGRA to the Swiss Federal Government in December 2002 assessed the feasibility of disposal of spent fuel (SF), vitrified high level waste (HLW) from reprocessing and long-lived intermediate level waste in an Opalinus Clay repository site in Northern Switzerland. NAGRA proposed the use of carbon steel canisters for disposal of SF/HLW and it also put forward an alternative concept of copper canisters with cast iron insert. In its reply the Federal Government acknowledged that NAGRA had successfully demonstrated the technical feasibility of disposal of SF/HLW. However, some of its experts raised a number of questions related to the choice of steel as canister material. Among others, it was questioned whether hydrogen formed by corrosion of steel in contact with saturated bentonite might adversely affect the barrier function of the Opalinus clay. It was also recommended that alternative canister materials and/or design concepts should be evaluated. To deal with these concerns NAGRA convened an international group of experts, the Canister Materials Review Board (CMRB), who were to review the existing information on canister materials that could be suitable for the proposed repository environment. Based on present knowledge of materials science, the CMRB was to recommend to NAGRA the most suitable material(s) for meeting the performance requirements for SF/HLW canisters. Specifically, the CMRB was to consider corrosion, including hydrogen generation, and stress-assisted failure processes that could affect the integrity and projected life time of SF/HLW canisters or impede the functioning of geological barriers while keeping in mind the overall feasibility of manufacturing, sealing and inspecting the canisters. The CMRB was further asked to identify the needs and provide advice for further studies by NAGRA on the long term performance and safety of SF/HLW canisters in the Swiss repository concept. For the assessment of the

  13. Neutron interrogator assay system for the Idaho Chemical Processing Plant waste canisters and spent fuel: preliminary description and operating procedures manual

    International Nuclear Information System (INIS)

    Menlove, H.O.; Eccleston, G.; Close, D.A.; Speir, L.G.

    1978-05-01

    A neutron interrogation assay system is being designed for the measurement of waste canisters and spent fuel packages at the new Idaho Chemical Processing Plant to be operated by Allied Chemical Corp. The assay samples consist of both waste canisters from the fluorinel dissolution process and spent fuel assemblies. The assay system is a 252 Cf ''Shuffler'' that employs a cyclic sequence of fast-neutron interrogation with a 252 Cf source followed by delayed-neutron counting to determine the 235 U content

  14. Safeguards considerations related to the use of multi-purpose canisters in the Civilian Radioactive Waste Management system

    International Nuclear Information System (INIS)

    Floyd, W.C.

    1995-01-01

    The US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is responsible for disposing of the nation's high-level radioactive waste. Currently, DOE is considering the use of Multi-Purpose Canisters (MPCs) to containerize commercial spent nuclear fuel (SNF) to be handled by the system. To achieve its safeguards and security objectives, OCRWM plans to institute a US Regulatory Commission (NRC)-approved safeguards program. Since the Mined Geologic Disposal System (MGDS) facility and a possible Monitored Retrievable Storage (MRS) facility may be subject to selection for International Atomic Energy Agency (IAEA) inspections, the safeguards program for MPCs may not preclude compliance with the requirements of the IAEA's Annex D, Special Criteria for Difficult-to-Access Fuel Items. MPC safeguards are based on three principles: Verification, Material Control and Accounting, and Physical Protection

  15. DWPF Simulant CPC Studies For SB8

    Energy Technology Data Exchange (ETDEWEB)

    Newell, J. D.

    2013-09-25

    Prior to processing a Sludge Batch (SB) in the Defense Waste Processing Facility (DWPF), flowsheet studies using simulants are performed. Typically, the flowsheet studies are conducted based on projected composition(s). The results from the flowsheet testing are used to 1) guide decisions during sludge batch preparation, 2) serve as a preliminary evaluation of potential processing issues, and 3) provide a basis to support the Shielded Cells qualification runs performed at the Savannah River National Laboratory (SRNL). SB8 was initially projected to be a combination of the Tank 40 heel (Sludge Batch 7b), Tank 13, Tank 12, and the Tank 51 heel. In order to accelerate preparation of SB8, the decision was made to delay the oxalate-rich material from Tank 12 to a future sludge batch. SB8 simulant studies without Tank 12 were reported in a separate report.1 The data presented in this report will be useful when processing future sludge batches containing Tank 12. The wash endpoint target for SB8 was set at a significantly higher sodium concentration to allow acceptable glass compositions at the targeted waste loading. Four non-coupled tests were conducted using simulant representing Tank 40 at 110-146% of the Koopman Minimum Acid requirement. Hydrogen was generated during high acid stoichiometry (146% acid) SRAT testing up to 31% of the DWPF hydrogen limit. SME hydrogen generation reached 48% of of the DWPF limit for the high acid run. Two non-coupled tests were conducted using simulant representing Tank 51 at 110-146% of the Koopman Minimum Acid requirement. Hydrogen was generated during high acid stoichiometry SRAT testing up to 16% of the DWPF limit. SME hydrogen generation reached 49% of the DWPF limit for hydrogen in the SME for the high acid run. Simulant processing was successful using previously established antifoam addition strategy. Foaming during formic acid addition was not observed in any of the runs. Nitrite was destroyed in all runs and no N2O was detected

  16. The suitability of Titanium as a corrosion resistant canister for nuclear waste

    International Nuclear Information System (INIS)

    Henriksson, S.; Pettersson, K.

    1977-08-01

    A literature study and inventory of experience has been carried out, aimed at assessing the possibilities of unalloyed and Pd-alloyed titanium withstanding corrosion for 1000 - 10000 years in contact with Baltic Sea water at 100 percentC and pH 4 - 10. The following assesment can be made: -1. Pitting, crevice corrosion, stress corrosion cracking and corrosion fatigue constitute no problem if the canister is made of unalloyed titanium corresponding to ASTM Grade 1. 2. Linear extrapolation of reported corrosion rates for oxidation and general corrosion gives a life of between 1000 and 10000 years for a 5 mm thick canister. 3. Hydrogen embrittlement resulting from hydrogen pick-up from the deposition environment should not occur. Delayed failure caused by a redistribution of the hydrogen initially present in the titanium can be avoided if its concentration is maximized to 20 ppm. (author)

  17. Comparative risk assessments for the production and interim storage of glass and ceramic waste forms: defense waste processing facility

    International Nuclear Information System (INIS)

    Huang, J.C.; Wright, W.V.

    1982-04-01

    The Defense Waste Processing Facility (DWPF) for immobilizing nuclear high level waste (HLW) is scheduled to be built at the Savannah River Plant (SRP). High level waste is produced when SRP reactor components are subjected to chemical separation operations. Two candidates for immobilizing this HLW are borosilicate glass and crystalline ceramic, either being contained in weld-sealed stainless steel canisters. A number of technical analyses are being conducted to support a selection between these two waste forms. The present document compares the risks associated with the manufacture and interim storage of these two forms in the DWPF. Process information used in the risk analysis was taken primarily from a DWPF processibility analysis. The DWPF environmental analysis provided much of the necessary environmental information. To perform the comparative risk assessments, consequences of the postulated accidents are calculated in terms of: (1) the maximum dose to an off-site individual; and (2) the dose to off-site population within 80 kilometers of the DWPF, both taken in terms of the 50-year inhalation dose commitment. The consequences are then multiplied by the estimated accident probabilities to obtain the risks. The analyses indicate that the maximum exposure risk to an individual resulting from the accidents postulated for both the production and interim storage of either waste form represents only an insignificant fraction of the natural background radiation of about 90 mrem per year per person in the local area. They also show that there is no disaster potential to the off-site population. Therefore, the risks from abnormal events in the production and the interim storage of the DWPF waste forms should not be considered as a dominant factor in the selection of the final waste form

  18. Feasibility of using a high-level waste canister as an engineered barrier in disposal

    International Nuclear Information System (INIS)

    Slate, S.C.; Pitman, S.G.; Nesbitt, J.F.; Partain, W.L.

    1982-08-01

    The objective of this report is to evaluate the feasibility of designing a process canister that could also serve as a barrier canister. To do this a general set of performance criteria is assumed and several metal alloys having a high probability of demonstrating high corrosion resistance under repository conditions are evaluated in a qualitative design assessment. This assessment encompasses canister manufacture, the glass-filling process, interim storage, transportation, and to a limited extent, disposal in a repository. A series of scoping tests were carried out on two titanium alloys and Inconel 625 to determine if the high temperature inherent in the glass-fill processing would seriously affect either the strength or corrosion resistance of these metals. This is a process-related concern unique to the barrier canister concept. The material properties were affected by the heat treatments which simulated both the joule-heated glass melter process (titanium alloys and Inconel 625) and the in-can melter (ICM) process (Inconel 625). However, changes in the material properties were generally within 20% of the original specimens. Accelerated corrosion testing of the heat treated coupons in a highly oxygenated brine showed basic corrosion resistance of titanium grade 12 and Inconel 625 to compare favorably with that of the untreated coupons. The titanium grade 2 coupons experienced severe corrosion pitting. These corrosion tests were of a scoping nature and suitable primarily for the detection of gross sensitivity to the heat treatment inherent in the glass-fill process. They are only suggstive of repository performance since the tests do not adequately model the wide range of repository conditions that could conceivably occur

  19. Radioactive demonstration of DWPF product control strategy

    International Nuclear Information System (INIS)

    Andrews, M.K.; Bibler, N.E.

    1992-01-01

    The effectiveness of the product and process control strategies that will be utilized by the Defense Waste Processing Facility (DWPF) was demonstrated during a campaign in the Shielded Cells Facility (SCF) of the Savannah River Technology Center (SRTC). The remotely operated process included the preparation of the melter feed, vitrification in a slurry-fed 1/100th scale melter and analysis of the glass product both for its composition and durability. The campaign processed approximately 10 kg (on a dry basis) of radioactive sludge from Tank 51. This sludge is representative of the first batch of sludge that will be sent to the DWPF for immobilization into borosilicate glass. Additions to the sludge were made based on calculations using the Product Composition Control System (PCCS). Analysis of the glass produced during the campaign showed that a durable glass was produced with a composition similar to that predicted using the PCCS

  20. Bounding estimate of DWPF mercury emissions

    International Nuclear Information System (INIS)

    Jacobs, R.A.

    1992-01-01

    Purges required for H2 flammability control and verification of elevated Formic Acid Vent Condenser (FAVC) exit temperatures due to NO x reactions have lead to significant changes in Chemical Process Cell (CPC) operating conditions. Accordingly, mercury emissions estimates have been updated based upon the new operating requirements, IDMS (Integrated DWPF Melter System) experience, and development of an NO x /FAVC model which predicts FAVC exit temperatures. Using very conservative assumptions and maximum purge rates, the maximum calculated Hg emissions is approximately 130 lbs/yr. A range of 100 to 120 lbs/yr is conservatively predicted for other operating conditions. Defense Waste Processing Facility (DWPF) permitted Hg emissions are 175 lbs/yr (0.02 lbs/hr annual average)

  1. Preparing, Loading and Shipping Irradiated Metals in Canisters Classified as Remote-Handled (RH) Low-Level Waste (LLW) From Oak Ridge National Laboratory (ORNL) to the Nevada Test Site (NTS)

    International Nuclear Information System (INIS)

    McClelland, B.C.; Moore, T.D.

    2006-01-01

    Irradiated metals, classified as remote-handled low-level waste generated at the Oak Ridge National Laboratory (ORNL) in Oak Ridge, Tennessee, were containerised in various sized canisters for long-term storage. The legacy waste canisters were placed in below-grade wells located at the 7827 Facility until a pathway for final disposal at the Nevada Test Site (NTS) could be identified and approved. Once the pathway was approved, WESKEM, LLC was selected by Bechtel Jacobs Company, LLC to prepare, load, and ship these canisters from ORNL to the NTS. This paper details some of the technical challenges encountered during the retrieval process and solutions implemented to ensure the waste was safely and efficiently over-packed and shipped for final disposal. The technical challenges detailed in this paper include: 1) how to best perform canister/lanyard pre-lift inspections since some canisters had not been moved in ∼10 years, so deterioration was a concern; 2) replacing or removing damaged canister lanyards; 3) correcting a mis-cut waste canister lanyard resulting in a shielded overpack lid not seating properly; 4) retrieving a stuck canister; and 5) developing a path forward after an overstrained lanyard failed causing a well shield plug to fall and come in contact with a waste canister. Several of these methods can serve as positive lessons learned for other projects encountering similar situations. (authors)

  2. Remotely replaceable jumpers and embedded wiring for the DWPF

    International Nuclear Information System (INIS)

    Heckendorn, F.M. II.

    1984-01-01

    The Defense Waste Processing Facility (DWPF) for radioactive waste vitrification at the Savannah River Plant (SRP) is now under construction. Development of specialized electrical/instrument inter-connectors, or jumpers, is now complete. Remote replacement of the associated through-wall wiring using a standard canyon crane has also been demonstrated. 8 figures

  3. GLYCOLIC-NITRIC ACID FLOWSHEET DEMONSTRATION OF THE DWPF CHEMICAL PROCESS CELL WITH SLUDGE AND SUPERNATE SIMULANTS

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, D.; Stone, M.; Newell, J.; Best, D.; Zamecnik, J.

    2012-08-28

    Savannah River Remediation (SRR) is evaluating changes to its current Defense Waste Processing Facility (DWPF) flowsheet to improve processing cycle times. This will enable the facility to support higher canister production while maximizing waste loading. Higher throughput is needed in the Chemical Process Cell (CPC) since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the DWPF gas chromatographs (GC) and the potential for production of flammable quantities of hydrogen, reducing or eliminating the amount of formic acid used in the CPC is being developed. Earlier work at Savannah River National Laboratory has shown that replacing formic acid with an 80:20 molar blend of glycolic and formic acids has the potential to remove mercury in the SRAT without any significant catalytic hydrogen generation. This report summarizes the research completed to determine the feasibility of processing without formic acid. In earlier development of the glycolic-formic acid flowsheet, one run (GF8) was completed without formic acid. It is of particular interest that mercury was successfully removed in GF8, no formic acid at 125% stoichiometry. Glycolic acid did not show the ability to reduce mercury to elemental mercury in initial screening studies, which is why previous testing focused on using the formic/glycolic blend. The objective of the testing detailed in this document is to determine the viability of the nitric-glycolic acid flowsheet in processing sludge over a wide compositional range as requested by DWPF. This work was performed under the guidance of Task Technical and Quality Assurance Plan (TT&QAP). The details regarding the simulant preparation and analysis have been documented previously.

  4. Analysis of heat and mass transport processes near an emplaced nuclear waste canister

    International Nuclear Information System (INIS)

    Keller, C.

    1990-01-01

    A review has been performed of the models and experimental plans for evaluation of the spent fuel canister environment in a nuclear repository, e.g., the planned Yucca Mountain facilities. Special emphasis was placed on the relevance of the models and experiments to the 100 to 10,000 year prediction. The question was addressed whether one could justify testing in materials other than Yucca Mountain rock and obtain results in a relatively short time which would be relevant to the long time in Yucca Mountain. The paper discusses steam evolution in calculations and experiments, fracture models, possible measurements of relative permeability, and long time scale effects. 5 figs. (MB)

  5. Evaluation of a high-level waste radiological maintenance facility

    International Nuclear Information System (INIS)

    Collins, K.J.

    1998-01-01

    The Savannah River Site''s (SRS) Defense Waste Processing Facility (DWPF) near Aiken, SC is the nation''s first and world''s largest high level waste vitrification facility. DWPF began, operations in March 1996 to process radioactive waste, consisting of a matrixed predominantly 137 Cs precipitate and a predominately 90 Sr and alpha emitting sludge, into boro-silicate glass for long term storage. Presently, DWPF is processing only sludge waste and is preparing to process a combination of sludge and precipitate waste. During precipitate operations, canister dose rates are expected to exceed 10 Sv hr -1 (1000 rem hr -1 ). In sludge-only operations, canister contact gamma dose rates are approximately 15 mSv hr -1 (1500 mrem hr -1 ). Transferable contamination levels have been greater than 10 mSv hr -1 (100 cm 2 ) -1 for beta-gamma emitters and into the millions of Bq (100 cm 2 ) -1 for the alpha emitting radionuclides. This paper presents an evaluation of the radiological maintenance areas and their ability to support radiological work

  6. Investigation into the suitability of titanium as a corrosion resistant canister for nuclear waste

    International Nuclear Information System (INIS)

    Henriksson, S.; Pettersson, J.

    A literature study and inventory of experience has been carried out, aimed at assessing the possibilities of unalloyed and Pd-alloyed titanium withstanding corrosion for 1,000 to 10,000 years in contact with Baltic Sea water at 100 0 C and pH 4 to 10. Pitting, crevice corrosion, stress corrosion cracking and corrosion fatigue constitute no problem if the canister is made of unalloyed titanium corresponding to ASTM Grade 1. Titanium alloyed with palladium therefore need not be used. Linear extrapolation of reported corrosion rates for oxidation and general corrosion gives a life of between 1,000 and 10,000 years for a 5 mm thick canister. This estimate must be considered to be conservative since oxidation in fact follows a logarithmic law. Hydrogen embrittlement resulting from hydrogen pick-up from the deposition environment should not occur. Delayed failure caused by a redistribution of the hydrogen initially present in the titanium can be avoided if its concentration is maximized to 20 ppM. Pd-alloyed titanium is more sensitive than unalloyed titanium to hydrogen pick-up, especially in galvanic contact with less noble metals

  7. Comparison of SRP high-level waste disposal costs for borosilicate glass and crystalline ceramic waste forms

    International Nuclear Information System (INIS)

    McDonell, W.R.

    1982-04-01

    An evaluation of costs for the immobilization and repository disposal of SRP high-level wastes indicates that the borosilicate glass waste form is less costly than the crystalline ceramic waste form. The wastes were assumed immobilized as glass with 28% waste loading in 10,300 reference 24-in.-diameter canisters or as crystalline ceramic with 65% waste loading in either 3400 24-in.-diameter canisters or 5900 18-in.-diameter canisters. After an interim period of onsite storage, the canisters would be transported to the federal repository for burial. Total costs in undiscounted 1981 dollars of the waste disposal operations, excluding salt processing for which costs are not yet well defined, were about $2500 million for the borosilicate glass form in reference 24-in.-diameter canisters, compared to about $2900 million for the crystalline ceramic form in 24-in.-diameter canisters and about $3100 million for the crystalline ceramic form in 18-in.-diameter canisters. No large differences in salt processing costs for the borosilicate glass and crystalline ceramic forms are expected. Discounting to present values, because of a projected 2-year delay in startup of the DWPF for the crystalline ceramic form, preserved the overall cost advantage of the borosilicate glass form. The waste immobilization operations for the glass form were much less costly than for the crystalline ceramic form. The waste disposal operations, in contrast, were less costly for the crystalline ceramic form, due to fewer canisters requiring disposal; however, this advantage was not sufficient to offset the higher development and processing costs of the crystalline ceramic form. Changes in proposed Nuclear Regulatory Commission regulations to permit lower cost repository packages for defense high-level wastes would decrease the waste disposal costs of the more numerous borosilicate glass forms relative to the crystalline ceramic forms

  8. SME Acceptability Determination For DWPF Process Control (U)

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-06-12

    The statistical system described in this document is called the Product Composition Control System (PCCS). K. G. Brown and R. L. Postles were the originators and developers of this system as well as the authors of the first three versions of this technical basis document for PCCS. PCCS has guided acceptability decisions for the processing at the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) since the start of radioactive operations in 1996. The author of this revision to the document gratefully acknowledges the firm technical foundation that Brown and Postles established to support the ongoing successful operation at the DWPF. Their integration of the glass propertycomposition models, developed under the direction of C. M. Jantzen, into a coherent and robust control system, has served the DWPF well over the last 20+ years, even as new challenges, such as the introduction into the DWPF flowsheet of auxiliary streams from the Actinide Removal Process (ARP) and other processes, were met. The purpose of this revision is to provide a technical basis for modifications to PCCS required to support the introduction of waste streams from the Salt Waste Processing Facility (SWPF) into the DWPF flowsheet. An expanded glass composition region is anticipated by the introduction of waste streams from SWPF, and property-composition studies of that glass region have been conducted. Jantzen, once again, directed the development of glass property-composition models applicable for this expanded composition region. The author gratefully acknowledges the technical contributions of C.M. Jantzen leading to the development of these glass property-composition models. The integration of these models into the PCCS constraints necessary to administer future acceptability decisions for the processing at DWPF is provided by this sixth revision of this document.

  9. Nuclear hazardous waste cost control management

    International Nuclear Information System (INIS)

    Selg, R.A.

    1991-01-01

    The effects of the waste content of glass waste forms on Savannah River high-level waste disposal costs are currently under study to adjust the glass frit content to optimize the glass waste loadings and therefore significantly reduce the overall waste disposal cost. Changes in waste content affect onsite Defense Waste Changes in waste contents affect onsite Defense Waste Processing Facility (DWPF) costs as well as offsite shipping and repository emplacement charges. A nominal 1% increase over the 28 wt% waste loading of DWPF glass would reduce disposal costs by about $50 million for Savannah River wastes generated to the year 2000. Optimization of the glass waste forms to be produced in the SWPF is being supported by economic evaluations of the impact of the forms on waste disposal costs. Glass compositions are specified for acceptable melt processing and durability characteristics, with economic effects tracked by the number of waste canisters produced. This paper presents an evaluation of the effects of variations in waste content of the glass waste forms on the overall cost of the disposal, including offsite shipment and repository emplacement, of the Savannah River high-level wastes

  10. TRADITIONAL CANISTER-BASED OPEN WASTE MANAGEMENT SYSTEM VERSUS CLOSED SYSTEM: HAZARDOUS EXPOSURE PREVENTION AND OPERATING THEATRE STAFF SATISFACTION.

    Science.gov (United States)

    Horn, M; Patel, N; MacLellan, D M; Millard, N

    2016-06-01

    Exposure to blood and body fluids is a major concern to health care professionals working in operating rooms (ORs). Thus, it is essential that hospitals use fluid waste management systems that minimise risk to staff, while maximising efficiency. The current study compared the utility of a 'closed' system with a traditional canister-based 'open' system in the OR in a private hospital setting. A total of 30 arthroscopy, urology, and orthopaedic cases were observed. The closed system was used in five, four, and six cases, respectively and the open system was used in nine, two, and four cases, respectively. The average number of opportunities for staff to be exposed to hazardous fluids were fewer for the closed system when compared to the open during arthroscopy and urology procedures. The open system required nearly 3.5 times as much staff time for set-up, maintenance during procedures, and post-procedure disposal of waste. Theatre staff expressed greater satisfaction with the closed system than with the open. In conclusion, compared with the open system, the closed system offers a less hazardous and more efficient method of disposing of fluid waste generated in the OR.

  11. GLYCOLIC-NITRIC ACID FLOWSHEET DEMONSTRATION OF THE DWPF CHEMICAL PROCESSING CELL WITH MATRIX SIMULANTS AND SUPERNATE

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, D.; Stone, M.; Newell, J.; Best, D.

    2012-05-07

    Savannah River Remediation (SRR) is evaluating changes to its current DWPF flowsheet to improve processing cycle times. This will enable the facility to support higher canister production while maximizing waste loading. Higher throughput is needed in the CPC since the installation of the bubblers into the melter has increased melt rate. Due to the significant maintenance required for the DWPF gas chromatographs (GC) and the potential for production of flammable quantities of hydrogen, reducing or eliminating the amount of formic acid used in the CPC is being developed. Earlier work at Savannah River National Laboratory has shown that replacing formic acid with an 80:20 molar blend of glycolic and formic acids has the potential to remove mercury in the SRAT without any significant catalytic hydrogen generation. This report summarizes the research completed to determine the feasibility of processing without formic acid. In earlier development of the glycolic-formic acid flowsheet, one run (GF8) was completed without formic acid. It is of particular interest that mercury was successfully removed in GF8, no formic acid at 125% stoichiometry. Glycolic acid did not show the ability to reduce mercury to elemental mercury in initial screening studies, which is why previous testing focused on using the formic/glycolic blend. The objective of the testing detailed in this document is to determine the viability of the nitric-glycolic acid flowsheet in processing sludge over a wide compositional range as requested by DWPF. This work was performed under the guidance of Task Technical and Quality Assurance Plan (TT and QAP). The details regarding the simulant preparation and analysis have been documented previously.

  12. Yield Stress Reduction of DWPF Melter Feed Slurries

    International Nuclear Information System (INIS)

    Stone, M.E.; Smith, M.E.

    2007-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site vitrifies High Level Waste for repository internment. The process consists of three major steps: waste pretreatment, vitrification, and canister decontamination/sealing. The HLW consists of insoluble metal hydroxides and soluble sodium salts. The pretreatment process acidifies the sludge with nitric and formic acids, adds the glass formers as glass frit, then concentrates the resulting slurry to approximately 50 weight percent (wt%) total solids. This slurry is fed to the joule-heated melter where the remaining water is evaporated followed by calcination of the solids and conversion to glass. The Savannah River National Laboratory (SRNL) is currently assisting DWPF efforts to increase throughput of the melter. As part of this effort, SRNL has investigated methods to increase the solids content of the melter feed to reduce the heat load required to complete the evaporation of water and allow more of the energy available to calcine and vitrify the waste. The process equipment in the facility is fixed and cannot process materials with high yield stresses, therefore increasing the solids content will require that the yield stress of the melter feed slurries be reduced. Changing the glass former added during pretreatment from an irregularly shaped glass frit to nearly spherical beads was evaluated. The evaluation required a systems approach which included evaluations of the effectiveness of beads in reducing the melter feed yield stress as well as evaluations of the processing impacts of changing the frit morphology. Processing impacts of beads include changing the settling rate of the glass former (which effects mixing and sampling of the melter feed slurry and the frit addition equipment) as well as impacts on the melt behavior due to decreased surface area of the beads versus frit. Beads were produced from the DWPF process frit by fire polishing. The frit was allowed to free fall through a flame

  13. Geological Disposal of Nuclear Waste: Investigating the Thermo-Hygro-Mechanical-Chemical (THMC) Coupled Processes at the Waste Canister- Bentonite Barrier Interface

    Science.gov (United States)

    Davies, C. W.; Davie, D. C.; Charles, D. A.

    2015-12-01

    Geological disposal of nuclear waste is being increasingly considered to deal with the growing volume of waste resulting from the nuclear legacy of numerous nations. Within the UK there is 650,000 cubic meters of waste safely stored and managed in near-surface interim facilities but with no conclusive permanent disposal route. A Geological Disposal Facility with incorporated Engineered Barrier Systems are currently being considered as a permanent waste management solution (Fig.1). This research focuses on the EBS bentonite buffer/waste canister interface, and experimentally replicates key environmental phases that would occur after canister emplacement. This progresses understanding of the temporal evolution of the EBS and the associated impact on its engineering, mineralogical and physicochemical state and considers any consequences for the EBS safety functions of containment and isolation. Correlation of engineering properties to the physicochemical state is the focus of this research. Changes to geotechnical properties such as Atterberg limits, swelling pressure and swelling kinetics are measured after laboratory exposure to THMC variables from interface and batch experiments. Factors affecting the barrier, post closure, include corrosion product interaction, precipitation of silica, near-field chemical environment, groundwater salinity and temperature. Results show that increasing groundwater salinity has a direct impact on the buffer, reducing swelling capacity and plasticity index by up to 80%. Similarly, thermal loading reduces swelling capacity by 23% and plasticity index by 5%. Bentonite/steel interaction studies show corrosion precipitates diffusing into compacted bentonite up to 3mm from the interface over a 4 month exposure (increasing with temperature), with reduction in swelling capacity in the affected zone, probably due to the development of poorly crystalline iron oxides. These results indicate that groundwater conditions, temperature and corrosion

  14. A pilot scale demonstration of the DWPF process control and product verification strategy

    International Nuclear Information System (INIS)

    Hutson, N.D.; Jantzen, C.M.; Beam, D.C.

    1992-01-01

    The Defense Waste Processing Facility (DWPF) has been designed and constructed to immobilize Savannah River Site high level liquid waste within a durable borosilicate glass matrix for permanent storage. The DWPF will be operated to produce a glass product which must meet a number of product property constraints which are dependent upon the final product composition. During actual operations, the DWPF will control the properties of the glass product by the controlled blending of the waste streams with a glass-forming frit to produce the final melter feed slurry. The DWPF will verify control of the glass product through analysis of vitrified samples of slurry material. In order to demonstrate the DWPF process control and product verification strategy, a pilot-scale vitrification research facility was operated in three discrete batches using simulated DWPF waste streams. All of the DWPF process control methodologies were followed and the glass produce from each experiment was leached according to the Product Consistency Test. Results of the campaign are summarized

  15. A pilot scale demonstration of the DWPF process control and product verification strategy

    International Nuclear Information System (INIS)

    Hutson, N.D.; Jantzen, C.M.; Beam, D.C.

    1992-01-01

    The Defense Waste Processing Facility (DWPF) has been designed and constructed to immobilize Savannah River Site high level liquid waste within a durable borosilicate glass matrix for permanent storage. The DWPF will be operated to produce a glass product which must meet a number of product property constraints which are dependent upon the final product composition. During actual operations, the DWPF will control the properties of the glass product by the controlled blending of the waste streams with a glass-forming frit to produce the final melter feed slurry. The DWPF will verify control of the glass product through analysis of vitrified samples of slurry material. In order to demonstrate the DWPF process control and product verification strategy, a pilot-scale vitrification research facility was operated in three discrete batches using simulated DWPF waste streams. All of the DWPF process control methodologies were followed and the glass product from each experiment was leached according to the Product Consistency Test. In this paper results of the campaign are summarized

  16. Radioactive demonstration of DWPF product control strategy

    International Nuclear Information System (INIS)

    Andrews, M.K.; Bibler, N.E.

    1994-01-01

    The Defense Waste Processing Facility at the Savannah River Site (SRS) will vitrify high-level nuclear waste into borosilicate glass. The waste will be mixed with properly formulated glass-making frit and fed to a melter at 1150 degrees C. Process reliability and product quality are ensured by proper control of the melter feed composition. The effectiveness of the product and process control strategies that will be utilized by the Defense Waste Processing Facility (DWPF) was demonstrated during a campaign in the Shielded Cells Facility of the Savannah River Technology Center (SRTC). The remotely operated process included the preparation of the melter feed, vitrification in a slurry-fed 1/100th scale melter an analysis of the glass product both for its composition an durability. The campaign processed approximately 10 kg (on a dry basis) of radioactive sludge from Tank 51. This sludge is representative of the first batch of sludge that will be sent to the DWPF for immobilization into borosilicate glass. Additions to the sludge were made based on calculations using the Product Composition Control System (PCCS). Analysis of the glass produced during the campaign showed that a durable glass was produced with a composition very close to that predicted using the PCCS. 10 refs., 4 tabs

  17. Evaluation of vitrification factors from DWPF's macro-batch 1

    International Nuclear Information System (INIS)

    Edwards, T.B.

    2000-01-01

    The Defense Waste Processing Facility (DWPF) is evaluating new sampling and analytical methods that may be used to support future Slurry Mix Evaporator (SME) batch acceptability decisions. This report uses data acquired during DWPF's processing of macro-batch 1 to determine a set of vitrification factors covering several SME and Melter Feed Tank (MFT) batches. Such values are needed for converting the cation measurements derived from the new methods to a ''glass'' basis. The available data from macro-batch 1 were used to examine the stability of these vitrification factors, to estimate their uncertainty over the course of a macro-batch, and to provide a recommendation on the use of a single factor for an entire macro-batch. The report is in response to Technical Task Request HLW/DWPF/TTR-980015

  18. The prospective usage of the multi-purpose canister and impacts on the waste management and disposal system

    International Nuclear Information System (INIS)

    McLeod, N.B.

    1993-01-01

    The Multi-Purpose Canister (MPC) is designed to be loaded with spent fuel and sealed at reactors and then serve the functions of transport, storage and disposal without reopening. It can be either self-shielded or unshielded, thus requiring compatible overpacks for transport, storage and disposal. The MPC is not a new concept but it may now be viable because of the particular characteristics at Yucca Mountain: larger MPCs are possible because of ramp access to the repository horizon, and the less difficult temperature limits because of in-drift emplacement, rather than borehole emplacement. This paper describes the advantages and disadvantages of adopting the MPC as the principal technology to be employed in the US program. Use of the MPC permits integration of the utility and DOE portions of the system as well as among the elements within the DOE portion. Paradoxically, the principal disadvantage of the MPC is a direct consequence of its merit as an integrating technology. Full integration includes disposability without reopening, and requires that disposability design decisions be made and implemented well in advance of when waste package licensing uncertainties are resolved. There is, therefore, a risk that MPCs loaded prior to waste package licensing will have to be opened. This risk is discussed in terms of probability and consequences and various alternatives for mitigating this risk are discussed

  19. Cesium release from ceramic waste form materials in simulated canister corrosion product containing solutions

    Energy Technology Data Exchange (ETDEWEB)

    Vittorio, Luca; Drabarek, Elizabeth; Chronis, Harriet; Griffith, Christopher S

    2004-07-01

    It has previously been demonstrated that immobilization of Cs{sup +} and/or Sr{sup 2+} sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials can be achieved by heating the materials in air at temperatures in the range 500 - 1300 deg C. Highly crystalline powdered HTB materials formed by heating at 800 deg C show leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. As a very harsh leaching test, and also to model in a basic manner, leaching in the presence of canister corrosion products in oxidising environments, leaching of the bronzoid phases has been undertaken in Fe(NO{sub 3}){sub 3} solutions of increasing concentration. This is done in comparison with Cs -hollandite materials in order to compare the leaching characteristics of these two materials under such conditions. Both the Cs-loaded bronze and hollandite materials leach severely in Fe(NO{sub 3}){sub 3} losing virtually all of the immobilized Cs in a period of four days at 150 deg C. Total release of Cs and conversion of hollandite to titanium and iron titanium oxides begins to be observed at relatively low concentrations and is virtually complete after four days reaction in 0.5 mol/L Fe(NO{sub 3}){sub 3}. In the case of the bronze, all of the Cs is also extracted but the HTB structure is preserved. The reaction presumably involves an ion-exchange mechanism and iron oxide with a spinel structure is also observed at high Fe concentrations. (authors)

  20. Cesium release from ceramic waste form materials in simulated canister corrosion product containing solutions

    International Nuclear Information System (INIS)

    Vittorio, Luca; Drabarek, Elizabeth; Chronis, Harriet; Griffith, Christopher S.

    2004-01-01

    It has previously been demonstrated that immobilization of Cs + and/or Sr 2+ sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials can be achieved by heating the materials in air at temperatures in the range 500 - 1300 deg C. Highly crystalline powdered HTB materials formed by heating at 800 deg C show leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. As a very harsh leaching test, and also to model in a basic manner, leaching in the presence of canister corrosion products in oxidising environments, leaching of the bronzoid phases has been undertaken in Fe(NO 3 ) 3 solutions of increasing concentration. This is done in comparison with Cs -hollandite materials in order to compare the leaching characteristics of these two materials under such conditions. Both the Cs-loaded bronze and hollandite materials leach severely in Fe(NO 3 ) 3 losing virtually all of the immobilized Cs in a period of four days at 150 deg C. Total release of Cs and conversion of hollandite to titanium and iron titanium oxides begins to be observed at relatively low concentrations and is virtually complete after four days reaction in 0.5 mol/L Fe(NO 3 ) 3 . In the case of the bronze, all of the Cs is also extracted but the HTB structure is preserved. The reaction presumably involves an ion-exchange mechanism and iron oxide with a spinel structure is also observed at high Fe concentrations. (authors)

  1. Technical note 2. A review of the creep ductility of copper for nuclear waste canister application

    International Nuclear Information System (INIS)

    Pettersson, Kjell

    2011-03-01

    Background: The Swedish Radiation Safety Authority (SSM) reviews the Swedish Nuclear Fuel Company's (SKB) applications under the Act on Nuclear Activities (SFS 1984:3) for the construction and operation of a repository for spent nuclear fuel and for an encapsulation facility. As part of the review, SSM commissions consultants to carry out work in order to obtain information on specific issues. The results from the consultants' tasks are reported in SSM's Technical Note series. Objectives of the project: This project is part of SSM:s review of SKB:s license application for final disposal of spent nuclear fuel. The assignment concerns review of creep mechanisms for copper material used as a corrosion barrier in canisters for nal disposal of nuclear fuel in Sweden. Summary by the author: SKB has presented insufficient evidence to justify their position that the OFP copper has an adequate creep ductility during long term storage. Their large body of experiments only serves to prove that the creep ductility is sufficient for much shorter time spans than the intended storage times. There is a clear need for a credible theory of creep brittleness of OFP copper which will permit extrapolations to long term storage. The theory presented by SKB does not in its present state permit credible extrapolations. Alternatively SKB needs to find an explanation to the effect of phosphorus on the creep ductility and that it ensures the absence of creep brittleness in OFP copper. It is interesting to note that SKB has presented experimental evidence that intergranular cracks can form in OFP material tested in cracked specimens. Perhaps it is possible to more systematically study formation and growth of intergranular cracks in specimens of OFP copper with cracks

  2. Parametric testing of a DWPF glass

    International Nuclear Information System (INIS)

    Bazan, F.; Rego, J.

    1985-03-01

    A series of tests has been performed to characterize the chemical stability of a DWPF borosilicate glass sample as part of the Waste Package Task of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. This material was prepared at the Savannah River Laboratory for the purpose of testing the 165-frit matrix doped with a simulated nonradioactive waste. All tests were conducted at 90 0 C using deionized water and J-13 water (a tuffaceous formation ground water). In the deionized water tests, both monoliths and crushed glass were tested at various ratios of surface area of the sample to volume of water in order to compare leach rates for different sample geometries or leaching times. Effects on the leach rates as a result of the presence of crushed tuff and stainless steel material were also investigated in the tests with J-13 water. 3 refs., 12 figs., 7 tabs

  3. The corrosion behavior of DWPF glasses

    International Nuclear Information System (INIS)

    Ebert, W.L.; Bates, J.K.

    1995-01-01

    The authors analyzed the corroded surfaces of reference glasses developed for the Defense Waste Processing Facility (DWPF) to characterize their corrosion behavior. The corrosion mechanism of nuclear waste glasses must be known in order to provide source terms describing radionuclide release for performance assessment calculations. Different DWPF reference glasses were corroded under conditions that highlighted various aspects of the corrosion process and led to different extents of corrosion. The glasses corroded by similar mechanisms, and a phenomenological description of their corrosion behavior is presented here. The initial leaching of soluble glass components results in the formation of an amorphous gel layer on the glass surface. The gel layer is a transient phase that transforms into a layer of clay crystallites, which equilibrates with the solution as corrosion continues. The clay layer does not act as a barrier to either water penetration or glass dissolution, which continues beneath it, and may eventually separate from the glass. Solubility limits for glass components may be established by the eventual precipitation of secondary phases; thus, corrosion of the glass becomes controlled by the chemical equilibrium between the solution and the assemblage of secondary phases. In effect, the solution is an intermediate phase through which the glass transforms to an energetically more favorable assemblage of phases. Implications regarding the prediction of long-term glass corrosion behavior are discussed

  4. Native copper in Permian Mudstones from South Devon: A natural analogue of copper canisters for high-level radioactive waste

    International Nuclear Information System (INIS)

    Milodowski, A.E.; Styles, M.T.; Werme, L.; Oversby, V.M.

    2001-01-01

    Native copper (>99.9% Cu) sheets associated with complex uraniferous and vanadiferous concretions in Upper Permian Mudstones from south Devon (United Kingdom) have been studied as a 'natural analogue' for copper canisters designed to be used in the isolation of spent fuel and high-level radioactive wastes (HLW) for deep geological disposal. Detailed analysis demonstrates that the copper formed before the mudstones were compacted. The copper displays complex corrosion and alteration. The earliest alteration was to copper oxides, followed sequentially by the formation of copper arsenides, nickel arsenide and copper sulphide, and finally nickel arsenide accompanied by nickel-copper arsenide, copper arsenide and uranium silicates. Petrographic observations demonstrate that these alteration products also formed prior to compaction. Consideration of the published history for the region indicates that maximum compaction of the rocks will have occurred by at least the Lower Jurassic (i.e. over 176 Ma ago). Since that time the copper sheets have remained isolated by the compacted mudstones and were unaffected by further corrosion until uplift and exposure to present-day surface weathering

  5. Glass formulation requirements for DWPF coupled operations using crystalline silicotitanates

    International Nuclear Information System (INIS)

    Harbour, J.R.; Andrews, M.K.

    1997-01-01

    The design basis DWPF flowsheet couples the vitrification of two waste streams: (1) a washed sludge and (2) a hydrolyzed sodium tetraphenylborate precipitate product, PHA. The PHA contains cesium-137 which had been precipitated from the tank supernate with sodium tetraphenylborate. Smaller amounts of strontium and plutonium adsorbed on sodium titanate are also present with the PHA feed. Currently, DWPF is running a sludge-only flowsheet while working towards solutions to the problems encountered with In Tank Precipitation (ITP). The sludge loading for the sludge-only flowsheet and for the anticipated coupled operations is 28 wt% on an oxide basis. For the coupled operation, it is essential to balance the treatment of the two waste streams such that no supernate remains after immobilization of all the sludge. An alternative to ITP and sodium titanate is the removal of Cs-137, Sr-90, and plutonium from the tank supernate by ion exchange using crystalline silicotitanate (CST). This material has been shown to effectively sorb these elements from the supernate. It is also known that CST sorbs plutonium. The loaded CST could then be immobilized with the sludge during vitrification. It has recently been demonstrated that CST loadings approaching 70 wt% for a CST-only glass can be achieved using a borosilicate glass formulation which can be processed by the DWPF melter. Initial efforts on coupled waste streams with simulated DWPF sludge show promise that a borosilicate glass formulation can incorporate both sludge and CST. This paper presents the bases for research efforts to develop a glass formulation which will incorporate sludge and CST at loadings appropriate for DWPF operation

  6. Remote automatic plasma arc-closure welding of a dry-storage canister for spent nuclear fuel and high-level radioactive waste

    International Nuclear Information System (INIS)

    Sprecace, R.P.; Blankenship, W.P.

    1982-01-01

    A carbon steel storage canister has been designed for the dry encapsulation of spent nuclear fuel assemblies or of logs of vitrified high level radioactive waste. The canister design is in conformance with the requirements of the ASME Code, Section III, Division 1 for a Class 3 vessel. The canisters will be loaded and sealed as part of a completely remote process sequence to be performed in the hot bay of an experimental encapsulation facility at the Nevada Test Site. The final closure to be made is a full penetration butt weld between the canister body, a 12.75-in O.D. x 0.25-in wall pipe, and a mating semiellipsoidal closure lid. Due to a combination of design, application and facility constraints, the closure weld must be made in the 2G position (canister vertical). The plasma arc welding system is described, and the final welding procedure is described and discussed in detail. Several aspects and results of the procedure development activity, which are of both specific and general interest, are highlighted; these include: The critical welding torch features which must be exactly controlled to permit reproducible energy input to, and gas stream interaction with, the weld puddle. A comparison of results using automatic arc voltage control with those obtained using a mechanically fixed initial arc gap. The optimization of a keyhole initiation procedure. A comparison of results using an autogenous keyhole closure procedure with those obtained using a filler metal addition. The sensitivity of the welding process and procedure to variations in joint configuration and dimensions and to variations in base metal chemistry. Finally, the advantages and disadvantages of the plasma arc process for this application are summarized from the current viewpoint, and the applicability of this process to other similar applications is briefly indicated

  7. Corrosion impact of reductant on DWPF and downstream facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Imrich, K. J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Jantzen, C. M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Murphy, T. H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Wilderman, J. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-12-01

    Glycolic acid is being evaluated as an alternate reductant in the preparation of high level waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid is not completely consumed and small quantities of the glycolate anion are carried forward to other high level waste (HLW) facilities. The impact of the glycolate anion on the corrosion of the materials of construction throughout the waste processing system has not been previously evaluated. A literature review had revealed that corrosion data in glycolate-bearing solution applicable to SRS systems were not available. Therefore, testing was recommended to evaluate the materials of construction of vessels, piping and components within DWPF and downstream facilities. The testing, conducted in non-radioactive simulants, consisted of both accelerated tests (electrochemical and hot-wall) with coupons in laboratory vessels and prototypical tests with coupons immersed in scale-up and mock-up test systems. Eight waste or process streams were identified in which the glycolate anion might impact the performance of the materials of construction. These streams were 70% glycolic acid (DWPF feed vessels and piping), SRAT/SME supernate (Chemical Processing Cell (CPC) vessels and piping), DWPF acidic recycle (DWPF condenser and recycle tanks and piping), basic concentrated recycle (HLW tanks, evaporators, and transfer lines), salt processing (ARP, MCU, and Saltstone tanks and piping), boric acid (MCU separators), and dilute waste (HLW evaporator condensate tanks and transfer line and ETF components). For each stream, high temperature limits and worst-case glycolate concentrations were identified for performing the recommended tests. Test solution chemistries were generally based on analytical results of actual waste samples taken from the various process facilities or of prototypical simulants produced in the laboratory. The materials of construction for most vessels

  8. Defense Waste Processing Facility (DWPF): The vitrification of high-level nuclear waste. (Latest citations from the Bibliographic database). Published Search

    International Nuclear Information System (INIS)

    1993-09-01

    The bibliography contains citations concerning a production-scale facility and the world's largest plant for the vitrification of high-level radioactive nuclear wastes (HLW) located in the United States. Initially based on the selection of borosilicate glass as the reference waste form, the citations present the history of the development including R ampersand D projects and the actual construction of the production facility at the DOE Savannah River Plant (SRP). (Contains a minimum of 177 citations and includes a subject term index and title list.)

  9. Foreign programs for the storage of spent nuclear power plant fuels, high-level waste canisters and transuranic wastes

    International Nuclear Information System (INIS)

    Harmon, K.M.; Johnson, A.B. Jr.

    1984-04-01

    The various national programs for developing and applying technology for the interim storage of spent fuel, high-level radioactive waste, and TRU wastes are summarized. Primary emphasis of the report is on dry storage techniques for uranium dioxide fuels, but data are also provided concerning pool storage

  10. 32-Week Holding-Time Study of SUMMA Polished Canisters and Triple Sorbent Traps Used To Sample Organic Constituents in Radioactive Waste Tank Vapor Headspace

    International Nuclear Information System (INIS)

    Evans, John C.; Huckaby, James L.; Mitroshkov, Alexandre V.; Julya, Janet L.; Hayes, James C.; Edwards, Jeffrey A.; Sasaki, Leela M.

    1997-01-01

    Two sampling methods[SUMMA polished canisters and triple sorbent traps (TSTs)] were compared for long-term storage of trace organic vapor samples collected from the headspaces of high-level radioactive waste tanks at the U.S. Department of Energy's Hanford Site in Washington State. Because safety, quality assurance, radiological controls, the long-term stability of the sampling media during storage needed to be addressed. Samples were analyzed with a gas chromatograph/mass spectrometer (GC/MS) using cryogenic reconcentration or thermal desorption sample introduction techniques. SUMMA canister samples were also analyzed for total non-methane organic compounds (TNMOC) by GC/flame ionization detector (FID) using EPA Compendium Method TO-12 . To verify the long-term stability of the sampling media, multiple samples were collected in parallel from a typical passively ventilated radioactive waste tank known to contain moderately high concentrations of both polar and nonpolar organic compounds. Analyses for organic analytes and TNMOC were conducted at increasing intervals over a 32-week period to determine whether any systematic degradation of sample integrity occurred. Analytes collected in the SUMMA polished canisters generally showed good stability over the full 32 weeks with recoveries at the 80% level or better for all compounds studied. The TST data showed some loss (50-80% recovery) for a few high-volatility compounds even in the refrigerated samples; losses for unrefrigerated samples were far more pronounced with recoveries as low as 20% observed in a few cases

  11. Nuclear waste glass product consistency test (PCT), Version 5.0

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Ramsey, W.G.; Waters, B.J.

    1992-06-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced in the Defense Waste Processing Facility (DWPF), poured into stainless steel canisters, and eventually disposed of in a geologic repository. In order to comply with the Waste Acceptance Preliminary Specifications (WAPS), the durability of the glass needs to be measured during production to assure its long term stability and radionuclide release properties. A durability test, designated the Produce Consistency Test (PCT), was developed for DWPF glass in order to meet the WAPS requirements. The response of the PCT procedure was based on extensive testing with glasses of widely different compositions. The PCT was determined to be very reproducible, to yield reliable results rapidly, and to be easily performed in shielded cell facilities with radioactive samples. Version 5.0 of the PCT procedure is attached

  12. Relaxation of the lower frit loading constraint for DWPF process control

    International Nuclear Information System (INIS)

    Brown, K.G.

    2000-01-01

    The lower limit on the frit loading parameter when measurement uncertainty is introduced has impacted DWPF performance during immobilization of Tank 42 Sludge; therefore, any defensible relaxation or omission of this constraint should correspondingly increase DWPF waste loading and efficiency. Waste loading should be increased because the addition of frit is the current remedy for exceeding the lower frit loading constraint. For example, frit was added to DWPF SME Batches 94, 97 and 98 to remedy these batches for low frit loading. Attempts were also made to add frit in addition to the optimum computed to assure the lower frit loading constraint would be satisfied; however, approximately half of the SME Batches produced after Batch 98 have violated the lower frit loading constraint. If the DWPF batches did not have to be remediated and additional frit added because of the lower frit loading limit, then both, the performance of the DWPF process and the waste loading in the glass produced would be increased. Before determining whether or not the lower frit loading limit can be relaxed or omitted, the origin of this and the other constraints related to durability prediction must be examined. The lower limit loading constraint results from the need to make highly durable glass in DWPF. It is required that DWPF demonstrate that the glass produced would have durability that is at least two standard deviations greater than that of the Environmental Assessment (EA) glass. Glass durability cannot be measured in situ, it must be predicted from composition which can be measured. Fortunately, the leaching characteristics of homogeneous waste glasses is strongly related to the total molar free energy of the constituent species. Thus the waste acceptance specification has been translated into a requirement that the total molar free energy associated with the glass composition that would be produced from a DWPF melter feed batch be less than that of the EA glass accounting for

  13. Conceptual designs of radioactive canister transporters

    International Nuclear Information System (INIS)

    1978-02-01

    This report covers conceptual designs of transporters for the vertical, horizontal, and inclined installation of canisters containing spent-fuel elements, high-level waste, cladding waste, and intermediate-level waste (low-level waste is not discussed). Included in the discussion are cask concepts; transporter vehicle designs; concepts for mechanisms for handling and manipulating casks, canisters, and concrete plugs; transporter and repository operating cycles; shielding calculations; operator radiation dosages; radiation-resistant materials; and criteria for future design efforts

  14. ELIMINATION OF THE CHARACTERIZATION OF DWPF POUR STREAM SAMPLE AND THE GLASS FABRICATION AND TESTING OF THE DWPF SLUDGE BATCH QUALIFICATION SAMPLE

    Energy Technology Data Exchange (ETDEWEB)

    Amoroso, J.; Peeler, D.; Edwards, T.

    2012-05-11

    A recommendation to eliminate all characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification sample was made by a Six-Sigma team chartered to eliminate non-value-added activities for the Defense Waste Processing Facility (DWPF) sludge batch qualification program and is documented in the report SS-PIP-2006-00030. That recommendation was supported through a technical data review by the Savannah River National Laboratory (SRNL) and is documented in the memorandums SRNL-PSE-2007-00079 and SRNL-PSE-2007-00080. At the time of writing those memorandums, the DWPF was processing sludge-only waste but, has since transitioned to a coupled operation (sludge and salt). The SRNL was recently tasked to perform a similar data review relevant to coupled operations and re-evaluate the previous recommendations. This report evaluates the validity of eliminating the characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification samples based on sludge-only and coupled operations. The pour stream sample has confirmed the DWPF's ability to produce an acceptable waste form from Slurry Mix Evaporator (SME) blending and product composition/durability predictions for the previous sixteen years but, ultimately the pour stream analysis has added minimal value to the DWPF's waste qualification strategy. Similarly, the information gained from the glass fabrication and PCT of the sludge batch qualification sample was determined to add minimal value to the waste qualification strategy since that sample is routinely not representative of the waste composition ultimately processed at the DWPF due to blending and salt processing considerations. Moreover, the qualification process has repeatedly confirmed minimal differences in glass behavior from actual radioactive waste to glasses fabricated from simulants or batch chemicals. In

  15. Elimination Of The Characterization Of DWPF Pour Stream Sample And The Glass Fabrication And Testing Of The DWPF Sludge Batch Qualification Sample

    International Nuclear Information System (INIS)

    Amoroso, J.; Peeler, D.; Edwards, T.

    2012-01-01

    A recommendation to eliminate all characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification sample was made by a Six-Sigma team chartered to eliminate non-value-added activities for the Defense Waste Processing Facility (DWPF) sludge batch qualification program and is documented in the report SS-PIP-2006-00030. That recommendation was supported through a technical data review by the Savannah River National Laboratory (SRNL) and is documented in the memorandums SRNL-PSE-2007-00079 and SRNL-PSE-2007-00080. At the time of writing those memorandums, the DWPF was processing sludge-only waste but, has since transitioned to a coupled operation (sludge and salt). The SRNL was recently tasked to perform a similar data review relevant to coupled operations and re-evaluate the previous recommendations. This report evaluates the validity of eliminating the characterization of pour stream glass samples and the glass fabrication and Product Consistency Test (PCT) of the sludge batch qualification samples based on sludge-only and coupled operations. The pour stream sample has confirmed the DWPF's ability to produce an acceptable waste form from Slurry Mix Evaporator (SME) blending and product composition/durability predictions for the previous sixteen years but, ultimately the pour stream analysis has added minimal value to the DWPF's waste qualification strategy. Similarly, the information gained from the glass fabrication and PCT of the sludge batch qualification sample was determined to add minimal value to the waste qualification strategy since that sample is routinely not representative of the waste composition ultimately processed at the DWPF due to blending and salt processing considerations. Moreover, the qualification process has repeatedly confirmed minimal differences in glass behavior from actual radioactive waste to glasses fabricated from simulants or batch chemicals. In contrast, the

  16. Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

    International Nuclear Information System (INIS)

    Shine, E. P.; Poirier, M. R.

    2013-01-01

    Representative sampling is important throughout the Defense Waste Processing Facility (DWPF) process, and the demonstrated success of the DWPF process to achieve glass product quality over the past two decades is a direct result of the quality of information obtained from the process. The objective of this report was to present sampling methods that the Savannah River Site (SRS) used to qualify waste being dispositioned at the DWPF. The goal was to emphasize the methodology, not a list of outcomes from those studies. This methodology includes proven methods for taking representative samples, the use of controlled analytical methods, and data interpretation and reporting that considers the uncertainty of all error sources. Numerous sampling studies were conducted during the development of the DWPF process and still continue to be performed in order to evaluate options for process improvement. Study designs were based on use of statistical tools applicable to the determination of uncertainties associated with the data needs. Successful designs are apt to be repeated, so this report chose only to include prototypic case studies that typify the characteristics of frequently used designs. Case studies have been presented for studying in-tank homogeneity, evaluating the suitability of sampler systems, determining factors that affect mixing and sampling, comparing the final waste glass product chemical composition and durability to that of the glass pour stream sample and other samples from process vessels, and assessing the uniformity of the chemical composition in the waste glass product. Many of these studies efficiently addressed more than one of these areas of concern associated with demonstrating sample representativeness and provide examples of statistical tools in use for DWPF. The time when many of these designs were implemented was in an age when the sampling ideas of Pierre Gy were not as widespread as they are today. Nonetheless, the engineers and

  17. Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

    Energy Technology Data Exchange (ETDEWEB)

    Shine, E. P.; Poirier, M. R.

    2013-10-29

    Representative sampling is important throughout the Defense Waste Processing Facility (DWPF) process, and the demonstrated success of the DWPF process to achieve glass product quality over the past two decades is a direct result of the quality of information obtained from the process. The objective of this report was to present sampling methods that the Savannah River Site (SRS) used to qualify waste being dispositioned at the DWPF. The goal was to emphasize the methodology, not a list of outcomes from those studies. This methodology includes proven methods for taking representative samples, the use of controlled analytical methods, and data interpretation and reporting that considers the uncertainty of all error sources. Numerous sampling studies were conducted during the development of the DWPF process and still continue to be performed in order to evaluate options for process improvement. Study designs were based on use of statistical tools applicable to the determination of uncertainties associated with the data needs. Successful designs are apt to be repeated, so this report chose only to include prototypic case studies that typify the characteristics of frequently used designs. Case studies have been presented for studying in-tank homogeneity, evaluating the suitability of sampler systems, determining factors that affect mixing and sampling, comparing the final waste glass product chemical composition and durability to that of the glass pour stream sample and other samples from process vessels, and assessing the uniformity of the chemical composition in the waste glass product. Many of these studies efficiently addressed more than one of these areas of concern associated with demonstrating sample representativeness and provide examples of statistical tools in use for DWPF. The time when many of these designs were implemented was in an age when the sampling ideas of Pierre Gy were not as widespread as they are today. Nonetheless, the engineers and

  18. Galvanic corrosion of copper-cast iron couples in relation to the Swedish radioactive waste canister concept

    International Nuclear Information System (INIS)

    Smart, N.R.; Fennell, P.A.H.; Rance, A.P.; Werme, L.O.

    2004-01-01

    To ensure the safe encapsulation of spent nuclear fuel rods for geological disposal, SKB are considering using the Copper-Iron Canister, which consists of an outer copper canister and an inner cast iron container. The canister will be placed into boreholes in the bedrock of a geologic repository and surrounded by bentonite clay. In the unlikely event of the outer copper canister being breached, water could enter the annulus between the inner and outer canister and at points of contact between the two metals there would be a possibility of galvanic interactions. To study this effect, copper-cast iron galvanic couples were set up in a number of different environments representing possible conditions in the SKB repository. The tests investigated two artificial pore-waters and a bentonite slurry, under aerated and deaerated conditions, at 30 deg. C and 50 deg. C. The currents passing between the coupled electrodes and the potential of the couples were monitored for several months. In addition, some bimetallic crevice specimens based on the multi-crevice assembly (MCA) design were used to simulate the situation where the copper canister will be in direct contact with the cast iron inner vessel. The effect of growing an oxide film on the surface of the cast iron prior to coupling it with copper was also investigated. The electrochemical results are presented graphically in the form of electrode potentials and galvanic corrosion currents as a function of time. The galvanic currents in aerated conditions were much higher than in deaerated conditions. For example, at 30 deg. C, galvanic corrosion rates as low as 0.02 μm/year were observed for iron in groundwater after de-aeration, but of the order of 100 μm/year for the cast iron at 50 deg. C in the presence of oxygen. The galvanic currents were generally higher at 50 deg. C than at 30 deg. C. None of the MCA specimens exhibited any signs of crevice corrosion under deaerated conditions. It will be shown that in deaerated

  19. Alternative design concept for the second Glass Waste Storage Building

    International Nuclear Information System (INIS)

    Rainisch, R.

    1992-10-01

    This document presents an alternative design concept for storing canisters filled with vitrified waste produced at the Defense Waste Processing Facility (DWPF). The existing Glass Waste Storage Building (GWSB1) has the capacity to store 2,262 canisters and is projected to be completely filled by the year 2000. Current plans for glass waste storage are based on constructing a second Glass Waste Storage Building (GWSB2) once the existing Glass Waste Storage Building (GWSB1) is filled to capacity. The GWSB2 project (Project S-2045) is to provide additional storage capacity for 2,262 canisters. This project was initiated with the issue of a basic data report on March 6, 1989. In response to the basic data report Bechtel National, Inc. (BNI) prepared a draft conceptual design report (CDR) for the GWSB2 project in April 1991. In May 1991 WSRC Systems Engineering issued a revised Functional Design Criteria (FDC), the Rev. I document has not yet been approved by DOE. This document proposes an alternative design for the conceptual design (CDR) completed in April 1991. In June 1992 Project Management Department authorized Systems Engineering to further develop the proposed alternative design. The proposed facility will have a storage capacity for 2,268 canisters and will meet DWPF interim storage requirements for a five-year period. This document contains: a description of the proposed facility; a cost estimate of the proposed design; a cost comparison between the proposed facility and the design outlined in the FDC/CDR; and an overall assessment of the alternative design as compared with the reference FDC/CDR design

  20. SLUDGE WASHING AND DEMONSTRATION OF THE DWPF FLOWSHEET IN THE SRNL SHIELDED CELLS FOR SLUDGE BATCH 7A QUALIFICATION

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J.; Billings, A.; Click, D.

    2011-07-08

    Waste Solidification Engineering (WSE) has requested that characterization and a radioactive demonstration of the next batch of sludge slurry (Sludge Batch 7a*) be completed in the Shielded Cells Facility of the Savannah River National Laboratory (SRNL) via a Technical Task Request (TTR). This characterization and demonstration, or sludge batch qualification process, is required prior to transfer of the sludge from Tank 51 to the Defense Waste Processing Facility (DWPF) feed tank (Tank 40). The current WSE practice is to prepare sludge batches in Tank 51 by transferring sludge from other tanks. Discharges of nuclear materials from H Canyon are often added to Tank 51 during sludge batch preparation. The sludge is washed and transferred to Tank 40, the current DWPF feed tank. Prior to transfer of Tank 51 to Tank 40, SRNL simulates the Tank Farm and DWPF processes with a Tank 51 sample (referred to as the qualification sample). Sludge Batch 7a (SB7a) is composed of portions of Tanks 4, 7, and 12; the Sludge Batch 6 heel in Tank 51; and a plutonium stream from H Canyon. SRNL received the Tank 51 qualification sample (sample ID HTF-51-10-125) following sludge additions to Tank 51. This report documents: (1) The washing (addition of water to dilute the sludge supernate) and concentration (decanting of supernate) of the SB7a - Tank 51 qualification sample to adjust sodium content and weight percent insoluble solids to Tank Farm projections. (2) The performance of a DWPF Chemical Process Cell (CPC) simulation using the washed Tank 51 sample. The simulation included a Sludge Receipt and Adjustment Tank (SRAT) cycle, where acid was added to the sludge to destroy nitrite and reduce mercury, and a Slurry Mix Evaporator (SME) cycle, where glass frit was added to the sludge in preparation for vitrification. The SME cycle also included replication of five canister decontamination additions and concentrations. Processing parameters were based on work with a non

  1. Glass temperatures in free-standing canisters

    International Nuclear Information System (INIS)

    Hardy, B.J.; Hensel, S.J.

    1993-01-01

    The waste-forms produced by the Defense Waste Processing Facility (DWPF) are subject to the requirements of the Waste Acceptance Product Specifications (WAPS). The WAPS sets the maximum post cooldown temperature of the waste-form glass at 400 degrees C. This criterion must be satisfied for the ambient conditions and heat generation rates expected for the waste-forms. As part of the work described in task plan, WSRC-RP-93-1177, Rev. 0, a computer model was used to calculate the maximum glass temperatures in free standing wasteforms for a variety of ambient temperatures and heat generation rates

  2. Effects of radiation on the chemical environment surrounding waste canisters in proposed repository sites and possible effects on the corrosion process

    International Nuclear Information System (INIS)

    Glass, R.S.

    1981-12-01

    This report explores the interaction of ionizing radiation with various environments. In particular, worst case (aqueous) environments for the proposed nuclear waste repository sites are considered. Emphasis is on the fundamental chemical and physical processes involved. The identities of possible radiolysis products (both transient and stable) have been sought through a literature search. The effect of radiation on corrosion processes is discussed. The radiation-induced chemical environment in the worst case repository sites is not well defined. Attention should therefore be given to fundamental studies exploring the interaction of such environments with components of the nuclear waste package, including the canister materials and backfills. Identification and quantification of radiolysis products would be helpful in this regard

  3. Analysis of the DWPF glass pouring system using neural networks

    International Nuclear Information System (INIS)

    Calloway, T.B. Jr.; Jantzen, C.M.

    1997-01-01

    Neural networks were used to determine the sensitivity of 39 selected Melter/Melter Off Gas and Melter Feed System process parameters as related to the Defense Waste Processing Facility (DWPF) Melter Pour Spout Pressure during the overall analysis and resolution of the DWPF glass production and pouring issues. Two different commercial neural network software packages were used for this analysis. Models were developed and used to determine the critical parameters which accurately describe the DWPF Pour Spout Pressure. The model created using a low-end software package has a root mean square error of ± 0.35 inwc ( 2 = 0.77) with respect to the plant data used to validate and test the model. The model created using a high-end software package has a R 2 = 0.97 with respect to the plant data used to validate and test the model. The models developed for this application identified the key process parameters which contribute to the control of the DWPF Melter Pour Spout pressure during glass pouring operations. The relative contribution and ranking of the selected parameters was determined using the modeling software. Neural network computing software was determined to be a cost-effective software tool for process engineers performing troubleshooting and system performance monitoring activities. In remote high-level waste processing environments, neural network software is especially useful as a replacement for sensors which have failed and are costly to replace. The software can be used to accurately model critical remotely installed plant instrumentation. When the instrumentation fails, the software can be used to provide a soft sensor to replace the actual sensor, thereby decreasing the overall operating cost. Additionally, neural network software tools require very little training and are especially useful in mining or selecting critical variables from the vast amounts of data collected from process computers

  4. International technology exchange in support of the Defense Waste Processing Facility wasteform production

    International Nuclear Information System (INIS)

    Kitchen, B.G.

    1989-01-01

    The nearly completed Defense Waste Processing Facility (DWPF) is a Department of Energy (DOE) facility at the Savannah River Site that is designed to immobilize defense high level radioactive waste (HLW) by vitrification in borosilicate glass and containment in stainless steel canisters suitable for storage in the future DOE HLW repository. The DWPF is expected to start cold operation later this year (1990), and will be the first full scale vitrification facility operating in the United States, and the largest in the world. The DOE has been coordinating technology transfer and exchange on issues relating to HLW treatment and disposal through bi-lateral agreements with several nations. For the nearly fifteen years of the vitrification program at Savannah River Laboratory, over two hundred exchanges have been conducted with a dozen international agencies involving about five-hundred foreign national specialists. These international exchanges have been beneficial to the DOE's waste management efforts through confirmation of the choice of the waste form, enhanced understanding of melter operating phenomena, support for paths forward in political/regulatory arenas, confirmation of costs for waste form compliance programs, and establishing the need for enhancements of melter facility designs. This paper will compare designs and schedules of the international vitrification programs, and will discuss technical areas where the exchanges have provided data that have confirmed and aided US research and development efforts, impacted the design of the DWPF and guided the planning for regulatory interaction and product acceptance

  5. Fabrication of remote steam atomized scrubbers for DWPF off-gas system

    International Nuclear Information System (INIS)

    Nielsen, M.G.; Lafferty, J.D.

    1988-01-01

    The defense waste processing facility (DWPF) is being constructed for the purpose of processing high-level waste from sludge to a vitrified borosilicate glass. In the operation of continuous slurry-fed melters, off-gas aerosols are created by entrainment of feed slurries and the vaporization of volatile species from the molten glass mixture. It is necessary to decontaminate these aerosols in order to minimize discharge of airborne radionuclide particulates. A steam atomized scrubber (SAS) has been developed for DWPF which utilizes a patented hydro- sonic system gas scrubbing method. The Hydro-Sonic System utilizes a steam aspirating-type venturi scrubber that requires very precise fabrication tolerances in order to obtain acceptable decontamination factors. In addition to the process-related tolerances, precision mounting and nozzle tolerances are required for remote service at DWPF

  6. SPEEDUP simulation of liquid waste batch processing. Revision 1

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. Freeze and restart of the DWPF Scale Glass Melter

    International Nuclear Information System (INIS)

    Choi, A.S.

    1989-01-01

    After over two years of successful demonstration of many design and operating concepts of the DWPF Melter system, the last Scale Glass Melter campaign was initiated on 6/9/88 and consisted of two parts; (1) simulation of noble metal buildup and (2) freeze and subsequent restart of the melter under various scenarios. The objectives were to simulate a prolonged power loss to major heating elements and to examine the characteristics of transient melter operations during a startup with a limited supply of lid heat. Experimental results indicate that in case of a total power loss to the lower electrodes such as due to noble metal deposition, spinel crystals will begin to form in the SRL 165 composite waste glass pool in 24 hours. The total lid heater power required to initiate joule heating was the same as that during slurry-feeding. Results of a radiative heat transfer analysis in the plenum indicate that under the identical operating conditions, the startup capabilities of the SGM and the DWPF Melter are quite similar, despite a greater lid heater to melt surface area ratio in the DWPF Melter

  8. PLUTONIUM SOLUBILITY IN HIGH-LEVEL WASTE ALKALI BOROSILICATE GLASS

    Energy Technology Data Exchange (ETDEWEB)

    Marra, J.; Crawford, C.; Fox, K.; Bibler, N.

    2011-01-04

    The solubility of plutonium in a Sludge Batch 6 (SB6) reference glass and the effect of incorporation of Pu in the glass on specific glass properties were evaluated. A Pu loading of 1 wt % in glass was studied. Prior to actual plutonium glass testing, surrogate testing (using Hf as a surrogate for Pu) was conducted to evaluate the homogeneity of significant quantities of Hf (Pu) in the glass, determine the most appropriate methods to evaluate homogeneity for Pu glass testing, and to evaluate the impact of Hf loading in the glass on select glass properties. Surrogate testing was conducted using Hf to represent between 0 and 1 wt % Pu in glass on an equivalent molar basis. A Pu loading of 1 wt % in glass translated to {approx}18 kg Pu per Defense Waste Processing Facility (DWPF) canister, or about 10X the current allowed limit per the Waste Acceptance Product Specifications (2500 g/m{sup 3} of glass or about 1700 g/canister) and about 30X the current allowable concentration based on the fissile material concentration limit referenced in the Yucca Mountain Project License Application (897 g/m{sup 3}3 of glass or about 600 g Pu/canister). Based on historical process throughput data, this level was considered to represent a reasonable upper bound for Pu loading based on the ability to provide Pu containing feed to the DWPF. The task elements included evaluating the distribution of Pu in the glass (e.g. homogeneity), evaluating crystallization within the glass, evaluating select glass properties (with surrogates), and evaluating durability using the Product Consistency Test -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions

  9. Plutonium Solubility In High-Level Waste Alkali Borosilicate Glass

    International Nuclear Information System (INIS)

    Marra, J.; Crawford, C.; Fox, K.; Bibler, N.

    2011-01-01

    The solubility of plutonium in a Sludge Batch 6 (SB6) reference glass and the effect of incorporation of Pu in the glass on specific glass properties were evaluated. A Pu loading of 1 wt % in glass was studied. Prior to actual plutonium glass testing, surrogate testing (using Hf as a surrogate for Pu) was conducted to evaluate the homogeneity of significant quantities of Hf (Pu) in the glass, determine the most appropriate methods to evaluate homogeneity for Pu glass testing, and to evaluate the impact of Hf loading in the glass on select glass properties. Surrogate testing was conducted using Hf to represent between 0 and 1 wt % Pu in glass on an equivalent molar basis. A Pu loading of 1 wt % in glass translated to ∼18 kg Pu per Defense Waste Processing Facility (DWPF) canister, or about 10X the current allowed limit per the Waste Acceptance Product Specifications (2500 g/m 3 of glass or about 1700 g/canister) and about 30X the current allowable concentration based on the fissile material concentration limit referenced in the Yucca Mountain Project License Application (897 g/m 3 3 of glass or about 600 g Pu/canister). Based on historical process throughput data, this level was considered to represent a reasonable upper bound for Pu loading based on the ability to provide Pu containing feed to the DWPF. The task elements included evaluating the distribution of Pu in the glass (e.g. homogeneity), evaluating crystallization within the glass, evaluating select glass properties (with surrogates), and evaluating durability using the Product Consistency Test -- Method A (PCT-A). The behavior of Pu in the melter was evaluated using paper studies and corresponding analyses of DWPF melter pour samples.The results of the testing indicated that at 1 wt % Pu in the glass, the Pu was homogeneously distributed and did not result in any formation of plutonium-containing crystalline phases as long as the glass was prepared under 'well-mixed' conditions. The incorporation of 1 wt

  10. Mass transfer between waste canister and water seeping in rock fractures. Revisiting the Q-equivalent model

    International Nuclear Information System (INIS)

    Neretnieks, Ivars; Liu Longcheng; Moreno, Luis

    2010-03-01

    Models are presented for solute transport between seeping water in fractured rock and a copper canister embedded in a clay buffer. The migration through an undamaged buffer is by molecular diffusion only as the clay has so low hydraulic conductivity that water flow can be neglected. In the fractures and in any damaged zone seeping water carries the solutes to or from the vicinity of the buffer in the deposition hole. During the time the water passes the deposition hole molecular diffusion aids in the mass transfer of solutes between the water/buffer interface and the water at some distance from the interface. The residence time of the water and the contact area between the water and the buffer determine the rate of mass transfer between water and buffer. Simple analytical solutions are presented for the mass transfer in the seeping water. For complex migration geometries simplifying assumptions are made that allow analytical solutions to be obtained. The influence of variable apertures on the mass transfer is discussed and is shown to be moderate. The impact of damage to the rock around the deposition hole by spalling and by the presence of a cemented and fractured buffer is also explored. These phenomena lead to an increase of mass transfer between water and buffer. The overall rate of mass transfer between the bulk of the water and the canister is proportional to the overall concentration difference and inversely proportional to the sum of the mass transfer resistances. For visualization purposes the concept of equivalent flowrate is introduced. This entity can be thought as of the flowrate of water that will be depleted of its solute during the water passage past the deposition hole. The equivalent flowrate is also used to assess the release rate of radionuclides from a damaged canister. Examples are presented to illustrate how various factors influence the rate of mass transfer

  11. Durability of glasses from the Hg-doped Integrated DWPF Melter System (IDMS) campaign

    International Nuclear Information System (INIS)

    Jantzen, C.M.

    1992-01-01

    The Integrated DWPF Melter System (IDMS) for the vitrification of high-level radioactive wastes is designed and constructed to be a 1/9th scale prototype of the full scale Defense Waste Processing Facility (DWPF) melter. The IDMS facility is the first engineering scale melter system capable of processing mercury, and flowsheet levels of halides and noble metals. In order to determine the effects of mercury on the feed preparation process, the off-gas chemistry, glass melting behavior, and glass durability, a three-run mercury (Hg) campaign was conducted. The glasses produced during the Hg campaign were composed of Batch 1 sludge, simulated precipitate hydrolysis aqueous product (PHA) from the Precipitate Hydrolysis Experimental Facility (PHEF), and Frit 202. The glasses were produced using the DWPF process/product models for glass durability, viscosity, and liquidus. The durability model indicated that the glasses would all be more durable than the glass qualified in the DWPF Environmental Assessment (EA). The glass quality was verified by performing the Product Consistency Test (PCT) which was designed for glass durability testing in the DWPF

  12. Material compatibility evaluation for DWPF nitric-glycolic acid-literature review

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Skidmore, E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2013-06-01

    Glycolic acid is being evaluated as an alternative for formic and nitric acid in the DWPF flowsheet. Demonstration testing and modeling for this new flowsheet has shown that glycolic acid and glycolate has a potential to remain in certain streams generated during the production of the nuclear waste glass. A literature review was conducted to assess the impact of glycolic acid on the corrosion of the materials of construction for the DWPF facility as well as facilities downstream which may have residual glycolic acid and glycolates present. The literature data was limited to solutions containing principally glycolic acid.

  13. Analysis of high-level radioactive slurries as a method to reduce DWPF turnaround times

    International Nuclear Information System (INIS)

    Coleman, C.J.; Bibler, N.E.; Ferrara, D.M.; Hay, M.S.

    1996-01-01

    Analysis of Defense Waste Processing Facility (DWPF) samples as slurries rather than as dried or vitrified samples is an effective way to reduce sample turnaround times. Slurries can be dissolved with a mixture of concentrated acids to yield solutions for elemental analysis by inductively coupled plasma-atomic emission spectroscopy (ICP-AES). Slurry analyses can be performed in eight hours, whereas analyses of vitrified samples require up to 40 hours to complete. Analyses of melter feed samples consisting of the DWPF borosilicate frit and either simulated or actual DWPF radioactive sludge were typically within a range of 3--5% of the predicted value based on the relative amounts of sludge and frit added to the slurry. The results indicate that the slurry analysis approach yields analytical accuracy and precision competitive with those obtained from analyses of vitrified samples. Slurry analyses offer a viable alternative to analyses of solid samples as a simple way to reduce analytical turnaround times

  14. Defense Waste Processing Facility staged operations: environmental information document

    International Nuclear Information System (INIS)

    1981-11-01

    Environmental information is presented relating to a staged version of the proposed Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The information is intended to provide the basis for an Environmental Impact Statement. In either the integral or the staged design, the DWPF will convert the high-level waste currently stored in tanks into: a leach-resistant form containing about 99.9% of all the radioactivity, and a residual, slightly contaminated salt, which is disposed of as saltcrete. In the first stage of the staged version, the insoluble sludge portion of the waste and the long lived radionuclides contained therein will be vitrified. The waste glass will be sealed in canisters and stored onsite until shipped to a Federal repository. In the second stage, the supernate portion of the waste will be decontaminated by ion exchange. The recovered radionuclides will be transferred to the Stage 1 facility, and mixed with the sludge feed before vitrification. The residual, slightly contaminated salt solution will be mixed with Portland cement to form a concrete product (saltcrete) which will be buried onsite in an engineered landfill. This document describes the conceptual facilities and processes for producing glass waste and decontaminated salt. The environmental effects of facility construction, normal operations, and accidents are then presented. Descriptions of site and environs, alternative sites and waste disposal options, and environmental consultations and permits are given in the base Environmental Information Document

  15. Solidification of Savannah River Plant high level waste

    International Nuclear Information System (INIS)

    Maher, R.; Shafranek, L.F.; Kelley, J.A.; Zeyfang, R.W.

    1981-11-01

    Authorization for construction of the Defense Waste Processing Facility (DWPF) is expected in FY 83. The optimum time for stage 2 authorization is about three years later. Detailed design and construction will require approximately five years for stage 1, with stage 2 construction completed about two to three years later. Production of canisters of waste glass would begin in 1988, and the existing backlog of high level waste sludge stored at SRP would be worked off by about the year 2000. Stage 2 operation could begin in 1990. The technology and engineering are ready for construction and eventual operation of the DWPF for immobilizing high level radioactive waste at Savannah River Plant (SRP). Proceeding with this project will provide the public, and the leadership of this country, with a crucial demonstration that a major quantity of existing high level nuclear wastes can be safely and permanently immobilized. Early demonstration will both expedite and facilitate rational decision making on this aspect of the nuclear program. Delay in providing these facilities will result in significant DOE expenditures at SRP for new tanks just for continued temporary storage of wastes, and would probably result in dissipation of the intellectual and planning momentum that has built up in developing the project

  16. Waste disposal package

    Science.gov (United States)

    Smith, M.J.

    1985-06-19

    This is a claim for a waste disposal package including an inner or primary canister for containing hazardous and/or radioactive wastes. The primary canister is encapsulated by an outer or secondary barrier formed of a porous ceramic material to control ingress of water to the canister and the release rate of wastes upon breach on the canister. 4 figs.

  17. Literature review: Assessment of DWPF melter and melter off-gas system lifetime

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-30

    A glass melter for use in processing radioactive waste is a challenging environment for the materials of construction (MOC) resulting from a combination of high temperatures, chemical attack, and erosion/corrosion; therefore, highly engineered materials must be selected for this application. The focus of this report is to review the testing and evaluations used in the selection of the Defense Waste Processing Facility (DWPF), glass contact MOC specifically the Monofrax® K-3 refractory and Inconel® 690 alloy. The degradation or corrosion mechanisms of these materials during pilot scale testing and in-service operation were analyzed over a range of oxidizing and reducing flowsheets; however, DWPF has primarily processed a reducing flowsheet (i.e., Fe2+/ΣFe of 0.09 to 0.33) since the start of radioactive operations. This report also discusses the materials selection for the DWPF off-gas system and the corrosion evaluation of these materials during pilot scale testing and non-radioactive operations of DWPF Melter #1. Inspection of the off-gas components has not been performed during radioactive operations with the exception of maintenance because of plugging.

  18. Literature review: Assessment of DWPF melter and melter off-gas system lifetime

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-30

    A glass melter for use in processing radioactive waste is a challenging environment for the materials of construction (MOC) resulting from a combination of high temperatures, chemical attack, and erosion/corrosion; therefore, highly engineered materials must be selected for this application. The focus of this report is to review the testing and evaluations used in the selection of the Defense Waste Processing Facility (DWPF), glass contact MOC specifically the Monofrax® K-3 refractory and Inconel® 690 alloy. The degradation or corrosion mechanisms of these materials during pilot scale testing and in-service operation were analyzed over a range of oxidizing and reducing flowsheets; however, DWPF has primarily processed a reducing flowsheet (i.e., Fe2+/ΣFe of 0.09 to 0.33) since the start of radioactive operations. This report also discusses the materials selection for the DWPF off-gas system and the corrosion evaluation of these materials during pilot scale testing and non-radioactive operations of DWPF Melter #1. Inspection of the off-gas components has not been performed during radioactive operations with the exception of maintenance because of plugging.

  19. The Behavior and Effects of the Noble Metals in the DWPF Melter System

    International Nuclear Information System (INIS)

    Smith, M.E.; Bickford, D.F.

    1997-01-01

    Governments worldwide have committed to stabilization of high-level nuclear waste (HLW) by vitrification to a durable glass form for permanent disposal. All of these nuclear wastes contain the fission-product noble metals: ruthenium, rhodium, and palladium. SRS wastes also contain natural silver from iodine scrubbers. Closely associated with the noble metals are the fission products selenium and tellurium which are chemical analogs of sulfur and which combine with noble metals to influence their behavior and properties. Experience has shown that these melt insoluble metals and their compounds tend to settle to the floor of Joule-heated ceramic melters. In fact, almost all of the major research and production facilities have experienced some operational problem which can be associated with the presence of dense accumulations of these relatively conductive metals and/or their compounds. In most cases, these deposits have led to a loss of production capability, in some cases, to the point that melter operation could not continue. HLW nuclear waste vitrification facilities in the United States are the Department of Energy's Defense Waste Processing Facility (DWPF) at the Savannah River Site, the planned Hanford Waste Vitrification Plant (HWVP) at the Hanford Site and the operating West Valley Demonstration Project (WVDP) at West Valley, NY. The Integrated DWPF Melter System (IDMS) is a vitrification test facility at the Savannah River Technology Center (SRTC). It was designed and constructed to provide an engineering-scale representation of the DWPF melter and its associated feed preparation and off-gas treatment systems. An extensive noble metals testing program was begun in 1990. The objectives of this task were to explore the effects of the noble metals on the DWPF melter feed preparation and waste vitrification processes. This report focuses on the vitrification portion of the test program

  20. Program plan: DWPF/HLWDP stirred Melter Program Plan

    International Nuclear Information System (INIS)

    Smith, M.E.

    1994-01-01

    Slurry Fed Melters (SFM) have been developed in the United States, Europe, and Japan for the conversion of high-level radioactive waste (HLW) to borosilicate glass for permanent disposal. The newest design, the stirred melter, combines the high production rates and high glass quality features of the Joule-heated melters with the low-cost, compact, simple maintenance features of the pot melters. However, further engineering design and demonstrations are needed to operate the stirred melter on a large scale. This document outlines the program which develops a full scale stirred melter for the DWPF (240 pph), and provides a basis which will allow further scale-up of the technology for use in the Hanford High Level Waste Disposal Program (HLWDP) for up to four times the reference capacity

  1. DOE requests waiver on double containment for HLW canisters

    International Nuclear Information System (INIS)

    Lobsenz, G.

    1994-01-01

    The Energy Department has asked the Nuclear Regulatory Commission to waive double containment requirements for vitrified high-level radioactive waste canisters, saying the additional protection is not necessary and too costly. NRC said it had received a petition from DOE contending that the vitrified waste canisters were durable enough without double containment to prevent any potential plutonium release during handling and shipping. DOE said testing had shown that the vitrified waste canisters were similar - even superior - in durability to spent reactor fuel shipments, which NRC specifically exempted from the double containment requirement

  2. Grain boundary corrosion of copper canister material

    International Nuclear Information System (INIS)

    Fennell, P.A.H.; Graham, A.J.; Smart, N.R.; Sofield, C.J.

    2001-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister. The potential for grain boundary corrosion was investigated by exposing copper specimens, which had undergone different heat treatments and hence had different grain sizes, to aerated artificial bentonite-equilibrated groundwater with two concentrations of chloride, for increasing periods of time. The degree of grain boundary corrosion was determined by atomic force microscopy (AFM) and optical microscopy. AFM showed no increase in grain boundary 'ditching' for low chloride groundwater. In high chloride groundwater the surface was covered uniformly with a fine-grained oxide. No increases in oxide thickness were observed. No significant grain boundary attack was observed using optical microscopy either. The work suggests that in aerated artificial groundwaters containing chloride ions, grain boundary corrosion of copper is unlikely to adversely affect SKB's copper canisters

  3. Rheological Characterization of Unusual DWPF Slurry Samples

    International Nuclear Information System (INIS)

    Koopman, D. C.

    2005-01-01

    A study was undertaken to identify and clarify examples of unusual rheological behavior in Defense Waste Processing Facility (DWPF) simulant slurry samples. Identification was accomplished by reviewing sludge, Sludge Receipt and Adjustment Tank (SRAT) product, and Slurry Mix Evaporator (SME) product simulant rheological results from the prior year. Clarification of unusual rheological behavior was achieved by developing and implementing new measurement techniques. Development of these new methods is covered in a separate report, WSRC-TR-2004-00334. This report includes a review of recent literature on unusual rheological behavior, followed by a summary of the rheological measurement results obtained on a set of unusual simulant samples. Shifts in rheological behavior of slurries as the wt. % total solids changed have been observed in numerous systems. The main finding of the experimental work was that the various unusual DWPF simulant slurry samples exhibit some degree of time dependent behavior. When a given shear rate is applied to a sample, the apparent viscosity of the slurry changes with time rather than remaining constant. These unusual simulant samples are more rheologically complex than Newtonian liquids or more simple slurries, neither of which shows significant time dependence. The study concludes that the unusual rheological behavior that has been observed is being caused by time dependent rheological properties in the slurries being measured. Most of the changes are due to the effect of time under shear, but SB3 SME products were also changing properties while stored in sample bottles. The most likely source of this shear-related time dependence for sludge is in the simulant preparation. More than a single source of time dependence was inferred for the simulant SME product slurries based on the range of phenomena observed. Rheological property changes were observed on the time-scale of a single measurement (minutes) as well as on a time scale of hours

  4. Production and remediation of low sludge simulated Purex waste glasses, 2: Effects of sludge oxide additions on glass durability

    International Nuclear Information System (INIS)

    Ramsey, W.G.

    1993-01-01

    Glass produced during the Purex 4 campaigns of the Integrated DWPF Melter System (IDMS) and the 774 Research Melter contained a lower fraction of sludge components than targeted by the Product Composition Control System (PCCS). Purex 4 glass was more durable than the benchmark (EA) glass, but was less durable than most other simulated SRS high-level waste glasses. Further, the measured durability of Purex 4 glass was not as well correlated with the durability predicted from the DWPF process control algorithm, probably because the algorithm was developed to predict the durability of SRS high-level waste glasses with higher sludge content than Purex 4. A melter run, designated Purex 4 Remediation, was performed using the 774 Research Melter to determine if the initial PCCS target composition determined for Purex 4 would produce acceptable glass whose durability could be accurately modeled by the DWPF glass durability algorithm. Reagent grade oxides and carbonates were added to Purex 4 melter feed stock to simulate a higher sludge loading. Each canister of glass produced was sampled and the glass durability was determined by the Product Consistency Test method. This document details the durability data and subsequent analysis

  5. Neptunium sorption and co-precipitation of strontium in simulated DWPF salt solution

    International Nuclear Information System (INIS)

    McIntyre, P.F.; Orebaugh, E.G.; King, C.M.

    1988-01-01

    Batch experiments performed using crushed slag saltstone (∼40 mesh) removed >80% of 237 Np from simulated Defense Waste Processing Facility (DWPF) salt solution. The concentration of 237 Np (110 pCi/ml) used was 1000x greater than levels in actual DWPF solutions. Neptunium-239 was used as a tracer and was formed by neutron activation of uranyl nitrate. Results showed that small amounts of crushed saltstone (as little as 0.05 grams), removed >80% of neptunium from 15 ml of simulated DWPF solution after several hours equilibration. The neptunium is sorbed on insoluble carbonates formed in and on the saltstone matrix. Further testing showed that addition of 0.01 and 0.10 ml of 1 molar Ca +2 (ie. Ca (NO 3 ) 2 , CaCl 2 ) into 15 ml of simulated DWPF solution yielded a white carbonate precipitate which also removed >80% of the neptunium after 1 hour equilibration. Further experiments were performed to determine the effectiveness of this procedure to co-precipitate strontium

  6. Drop Calculations of HLW Canister and Pu Can-in-Canister

    International Nuclear Information System (INIS)

    Sreten Mastilovic

    2001-01-01

    The objective of this calculation is to determine the structural response of the standard high-level waste (HLW) canister and the canister containing the cans of immobilized plutonium (Pu) (''can-in-canister'' [CIC] throughout this document) subjected to drop DBEs (design basis events) during the handling operation. The evaluated DBE in the former case is 7-m (23-ft) vertical (flat-bottom) drop. In the latter case, two 2-ft (0.61-m) corner (oblique) drops are evaluated in addition to the 7-m vertical drop. These Pu CIC calculations are performed at three different temperatures: room temperature (RT) (20 C), T = 200 F = 93.3 C , and T = 400 F = 204 C ; in addition to these the calculation characterized by the highest maximum stress intensity is performed at T = 750 F = 399 C as well. The scope of the HLW canister calculation is limited to reporting the calculation results in terms of: stress intensity and effective plastic strain in the canister, directional residual strains at the canister outer surface, and change of canister dimensions. The scope of Pu CIC calculation is limited to reporting the calculation results in terms of stress intensity, and effective plastic strain in the canister. The information provided by the sketches from Reference 26 (Attachments 5.3,5.5,5.8, and 5.9) is that of the potential CIC design considered in this calculation, and all obtained results are valid for this design only. This calculation is associated with the Plutonium Immobilization Project and is performed by the Waste Package Design Section in accordance with Reference 24. It should be noted that the 9-m vertical drop DBE, included in Reference 24, is not included in the objective of this calculation since it did not become a waste acceptance requirement. AP-3.124, ''Calculations'', is used to perform the calculation and develop the document

  7. Mechanical integrity of canisters

    International Nuclear Information System (INIS)

    Nilsson, Fred

    1992-12-01

    This document constitutes the final report from 'SKBs reference group for mechanical integrity of canisters for spent nuclear fuel'. A complete list of all reports initiated by the reference group can be found in the summary report in this document. The main task of the reference group has been to advice SKB regarding the choice (ranking of alternatives) of canister type for different types of storage. The choice should be based on requirements of impermeability for a given time period and identification of possible limiting mechanisms. The main conclusions from the work were: From mechanical point of view, low phosphorous oxygen free copper (Cu-OFP) is a preferred canisters material. It exhibits satisfactory ductility both during tensile and creep testing. The residual stresses in the canisters are of such a magnitude that the estimated time to creep rupture with the data obtained for the Cu-OFP material is essentially infinite. Based on the present knowledge of stress corrosion cracking of copper there appears to be a small risk for such to occur in the projected environment. This risk need some further study. Rock shear movements of the size of 10 cm should pose no direct threat to the integrity of the canisters. Considering mechanical integrity, the composite copper/steel canister is an advantageous alternative. The recommendations for further research included continued studies of the creep properties of copper and of stress corrosion cracking. However, the studies should focus more directly on the design and fabrication aspect of the canister

  8. The concrete canister program

    International Nuclear Information System (INIS)

    Ohta, M.M.

    1978-02-01

    In the spring of 1974, WNRE began development and demonstration of a dry storage concept, called the concrete canister, as a possible alternative to storage of irradiated CANDU fuel in water pools. The canister is a thick-walled concrete monolith containing baskets of fuel in the dry state. The decay heat from the fuel is dissipated to the environment by natural heat transfer. Four canisters were designed and constructed. Two canisters containing electric heaters have been subjected to heat loads of 2.5 times the design, ramp heat-load cycling, and simulated weathering tests. The other two canisters were loaded with irradiated fuel, one containing fuel bundles of uniform decay heat and the other containing bundles of non-uniform decay heat in a non-symmetrical radial and axial array. The collected data were used to verify the analytical tools for prediction of effectiveness of heat transfer and radiation shielding and to verify the design of the basket and canisters. The demonstration canisters have shown that this concept is a viable alternative to water pools for the storage of irradiated CANDU fuel. (author)

  9. Impact of Spherical Frit Beads on Simulated DWPF Slurries

    International Nuclear Information System (INIS)

    SMITH, MICHAEL

    2005-01-01

    It has been shown that the rheological properties of simulated Defense Waste Processing Facility (DWPF) melter feed with the glass former frit as mostly (90 weight percent) solid spherical particles (referred to as beads) were improved as the feed was less viscous as compared to DWPF melter feed that contained the normal irregular shaped frit particles. Because the physical design of the DWPF Slurry Mix Evaporator (SME), Melter Feed Tank (MFT), and melter feed loop are fixed, the impact of changing the rheology might be very beneficial. Most importantly, higher weight percent total solids feed might be processed by reducing the rheological properties (specifically yield stress) of the feed. Additionally, if there are processing problems, such as air entrainment or pumping, these problems might be alleviated by reducing the rheological properties, while maintaining targeted throughputs. Rheology modifiers are chemical, physical, or a combination of the two and can either thin or thicken the rheology of the targeted slurry. The beads are classified as a physical rheological modifier in this case. Even though the improved rheological properties of the feed in the above mentioned DWPF tanks could be quite beneficial, it is the possibility of increased melt rate that is the main driver for the use of beaded glass formers. By improving the rheological properties of the feed, the weight percent solids of the feed could be increased. This higher weight percent solids (less water) feed could be processed faster by the melter as less energy would be required to evaporate the water, and more would be available for the actual melting of the waste and the frit. In addition, the use of beads to thin the feed could possibly allow for the use of a lower targeted acid stoichiometry in the feed preparation process (if in fact acid stoichiometry is being driven by feed rheology as opposed to feed chemistry). Previous work by the Savannah River National Laboratory (SRNL) with the lab

  10. EB-welding of the copper canister for the nuclear waste disposal. Final report of the development programme 1994-1997

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, H. [Outokumpu Oy Poricopper, Pori (Finland)

    1998-10-01

    During 1994-1997 Posiva Oy and Outokumpu Poricopper Oy had a joint project Development of EB-welding method for massive copper canister manufacturing. The project was part of the national technology program `Weld 2000` and it was supported financially by Technology Development Centre (TEKES). The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper canisters and placed deep into the bedrock. The thick copper layer of the canister provides a long time corrosion resistance and prevents deposited nuclear fuel from contact with water. The quality requirements of the copper components are high because of the designed long lifetime of the canister. The EB-welding technology has proved to be applicable method for the production of the copper canisters and the EB-welding technique is needed at least when the lids of the copper canister will be closed. There are a number of parameters in EB-welding which affect weldability. However, the effect of the welding parameters and their optimization has not been extensively studied in welding of thick copper sections using conventional high vacuum EB-welding. One aim of this development work was to extensively study effect of welding parameters on weld quality. The final objective was to minimise welding defects in the main weld and optimize slope out procedure in thick copper EB-welding. Welding of 50 mm thick copper sections was optimized using vertical and horizontal EB-welding techniques. As a result two full scale copper lids were welded to a short cylinder successfully. The resulting weld quality with optimised welding parameters was reasonable good. The optimised welding parameters for horizontal and vertical beam can be applied to the longitudinal body welds of the canister. The optimal slope out procedure for the lid closure needs some additional development work. In addition of extensive EB-welding program ultrasonic inspection and creep strength of the weld were studied. According

  11. EB-welding of the copper canister for the nuclear waste disposal. Final report of the development programme 1994-1997

    International Nuclear Information System (INIS)

    Aalto, H.

    1998-10-01

    During 1994-1997 Posiva Oy and Outokumpu Poricopper Oy had a joint project Development of EB-welding method for massive copper canister manufacturing. The project was part of the national technology program 'Weld 2000' and it was supported financially by Technology Development Centre (TEKES). The spent fuel from Finnish nuclear reactors is planned to be encapsulated in thick-walled copper canisters and placed deep into the bedrock. The thick copper layer of the canister provides a long time corrosion resistance and prevents deposited nuclear fuel from contact with water. The quality requirements of the copper components are high because of the designed long lifetime of the canister. The EB-welding technology has proved to be applicable method for the production of the copper canisters and the EB-welding technique is needed at least when the lids of the copper canister will be closed. There are a number of parameters in EB-welding which affect weldability. However, the effect of the welding parameters and their optimization has not been extensively studied in welding of thick copper sections using conventional high vacuum EB-welding. One aim of this development work was to extensively study effect of welding parameters on weld quality. The final objective was to minimise welding defects in the main weld and optimize slope out procedure in thick copper EB-welding. Welding of 50 mm thick copper sections was optimized using vertical and horizontal EB-welding techniques. As a result two full scale copper lids were welded to a short cylinder successfully. The resulting weld quality with optimised welding parameters was reasonable good. The optimised welding parameters for horizontal and vertical beam can be applied to the longitudinal body welds of the canister. The optimal slope out procedure for the lid closure needs some additional development work. In addition of extensive EB-welding program ultrasonic inspection and creep strength of the weld were studied. According

  12. Groundwork for Universal Canister System Development

    Energy Technology Data Exchange (ETDEWEB)

    Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gross, Mike [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Prouty, Jeralyn L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rigali, Mark J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Craig, Brian [Argonne National Lab. (ANL), Argonne, IL (United States); Han, Zenghu [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, John Hok [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, Yung [Argonne National Lab. (ANL), Argonne, IL (United States); Pope, Ron [Argonne National Lab. (ANL), Argonne, IL (United States); Connolly, Kevin [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Feldman, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jarrell, Josh [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Radulescu, Georgeta [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Scaglione, John [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wells, Alan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The mission of the United States Department of Energy's Office of Environmental Management is to complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and go vernment - sponsored nuclear energy re search. S ome of the waste s that that must be managed have be en identified as good candidates for disposal in a deep borehole in crystalline rock (SNL 2014 a). In particular, wastes that can be disposed of in a small package are good candidates for this disposal concept. A canister - based system that can be used for handling these wastes during the disposition process (i.e., storage, transfers, transportation, and disposal) could facilitate the eventual disposal of these wastes. This report provides information for a program plan for developing specifications regarding a canister - based system that facilitates small waste form packaging and disposal and that is integrated with the overall efforts of the DOE's Office of Nuclear Energy Used Fuel Dis position Camp aign's Deep Borehole Field Test . Groundwork for Universal Ca nister System Development September 2015 ii W astes to be considered as candidates for the universal canister system include capsules containing cesium and strontium currently stored in pools at the Hanford Site, cesium to be processed using elutable or nonelutable resins at the Hanford Site, and calcine waste from Idaho National Laboratory. The initial emphasis will be on disposal of the cesium and strontium capsules in a deep borehole that has been drilled into crystalline rock. Specifications for a universal canister system are derived from operational, performance, and regulatory requirements for storage, transfers, transportation, and disposal of radioactive waste. Agreements between the Department of Energy and the States of Washington and Idaho, as well as the Deep Borehole Field Test plan provide schedule requirements for development of the universal canister system

  13. Physicochemical and Geotechnical Alterations to MX-80 Bentonite at the Waste Canister Interface in an Engineered Barrier System

    Directory of Open Access Journals (Sweden)

    Christopher W. Davies

    2017-08-01

    Full Text Available The study investigated the basic geomechanical and mineralogical evolution of the bentonite barrier under various experimental boundary conditions which replicated the near-field Thermo-Hydro-Chemico (THC conditions in a repository. The relationships between the physicochemical alterations and changes in the geotechnical properties have seldom been studied, especially on a consistent dataset. This paper attempts to link the physicochemical properties of Na-bentonite (MX-80 to the macro-scale engineering functionality of the bentonite post THC exposure. Experiments investigated the impact of THC variables on the engineering and physicochemical functionality of the bentonite with respect to its application within a High-Level Waste (HLW engineered barrier system. Intrinsic alterations to the MX-80 bentonite under relatively short-term exposure to hydrothermal and chemical conditions were measured. Additionally, two long-term tests were conducted under ambient conditions to consider the impact of exposure duration. The intrinsic measurements were then related to the overall performance of the bentonite as a candidate barrier material for application in a UK geological disposal facility. Findings indicate that exposure to thermo-saline-corrosion conditions (i.e., corrosion products derived from structural grade 275 carbon steel inhibits the free swell capacity and plasticity of the bentonite. However, the measured values remained above the design limits set out for the Swedish multi-barrier concept, from which the UK concept may take a lead. Corrosion alone does not appear to significantly affect the geotechnical measurements compared with the influence of thermal loading and high saline pore water after relatively short-term exposure. Thermal and corrosion exposure displayed no impact on the intrinsic swelling of the smectite component, indicating that no significant structural alteration had occurred. However, when exploring more complex saline

  14. Literature Review: Assessment of DWPF Melter and Melter Off-gas System Lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Reigel, M. [Savannah River Site (SRS), Aiken, SC (United States)

    2015-07-30

    Testing to date for the MOC for the Hanford Waste Treatment and Immobilization Plant (WTP) melters is being reviewed with the lessons learned from DWPF in mind and with consideration to the changes in the flowsheet/feed compositions that have occurred since the original testing was performed. This information will be presented in a separate technical report that identifies any potential gaps for WTP processing.

  15. Development Of Remote Hanford Connector Gasket Replacement Tooling For DWPF

    International Nuclear Information System (INIS)

    Krementz, D.; Coughlin, Jeffrey

    2009-01-01

    The Defense Waste Processing Facility (DWPF) requested the Savannah River National Laboratory (SRNL) to develop tooling and equipment to remotely replace gaskets in mechanical Hanford connectors to reduce personnel radiation exposure as compared to the current hands-on method. It is also expected that radiation levels will continually increase with future waste streams. The equipment is operated in the Remote Equipment Decontamination Cell (REDC), which is equipped with compressed air, two master-slave manipulators (MSM's) and an electro-mechanical manipulator (EMM) arm for operation of the remote tools. The REDC does not provide access to electrical power, so the equipment must be manually or pneumatically operated. The MSM's have a load limit at full extension of ten pounds, which limited the weight of the installation tool. In order to remotely replace Hanford connector gaskets several operations must be performed remotely, these include: removal of the spent gasket and retaining ring (retaining ring is also called snap ring), loading the new snap ring and gasket into the installation tool and installation of the new gasket into the Hanford connector. SRNL developed and tested tools that successfully perform all of the necessary tasks. Removal of snap rings from horizontal and vertical connectors is performed by separate air actuated retaining ring removal tools and is manipulated in the cell by the MSM. In order install a new gasket, the snap ring loader is used to load a new snap ring into a groove in the gasket installation tool. A new gasket is placed on the installation tool and retained by custom springs. An MSM lifts the installation tool and presses the mounted gasket against the connector block. Once the installation tool is in position, the gasket and snap ring are installed onto the connector by pneumatic actuation. All of the tools are located on a custom work table with a pneumatic valve station that directs compressed air to the desired tool and

  16. DWPF upgrade, immobilization Programmatic Environmental Impact Statement input. Revision 1

    International Nuclear Information System (INIS)

    Sullivan, I.K.; Bignell, D.

    1994-01-01

    This Programmatic Environmental Impact Statement (PEIS) addresses the immobilization of plutonium by vitrification. Existing engineering documents, analyses, EIS, and technical publications were used and incorporated wherever possible to provide a timely response to this support effort. Although the vitrification technology is proven for the immobilization of high-level radioactive waste, more study and technical detail will be necessary to provide a comprehensive EIS that fully addresses all aspects of introduction of plutonium to the vitrification process. This document describes the concept(s) of plutonium processing as it relates to the upgrade of the DWPF and is therefore conceptual in nature. These concepts are based on technical data and experience at the Savannah River Site and will be detailed and finalized to support execution of this immobilization option

  17. ISOLOK VALVE ACCEPTANCE TESTING FOR DWPF SME SAMPLING PROCESS

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, T.; Hera, K.; Coleman, C.; Jones, M.; Wiedenman, B.

    2011-12-05

    Evaluation of the Defense Waste Processing Facility (DWPF) Chemical Process Cell (CPC) cycle time identified several opportunities to improve the CPC processing time. Of the opportunities, a focus area related to optimizing the equipment and efficiency of the sample turnaround time for DWPF Analytical Laboratory was identified. The Mechanical Systems & Custom Equipment Development (MS&CED) Section of the Savannah River National Laboratory (SRNL) evaluated the possibility of using an Isolok{reg_sign} sampling valve as an alternative to the Hydragard{reg_sign} valve for taking process samples. Previous viability testing was conducted with favorable results using the Isolok sampler and reported in SRNL-STI-2010-00749 (1). This task has the potential to improve operability, reduce maintenance time and decrease CPC cycle time. This report summarizes the results from acceptance testing which was requested in Task Technical Request (TTR) HLW-DWPF-TTR-2010-0036 (2) and which was conducted as outlined in Task Technical and Quality Assurance Plan (TTQAP) SRNL-RP-2011-00145 (3). The Isolok to be tested is the same model which was tested, qualified, and installed in the Sludge Receipt Adjustment Tank (SRAT) sample system. RW-0333P QA requirements apply to this task. This task was to qualify the Isolok sampler for use in the DWPF Slurry Mix Evaporator (SME) sampling process. The Hydragard, which is the current baseline sampling method, was used for comparison to the Isolok sampling data. The Isolok sampler is an air powered grab sampler used to 'pull' a sample volume from a process line. The operation of the sampler is shown in Figure 1. The image on the left shows the Isolok's spool extended into the process line and the image on the right shows the sampler retracted and then dispensing the liquid into the sampling container. To determine tank homogeneity, a Coliwasa sampler was used to grab samples at a high and low location within the mixing tank. Data from

  18. Estimates of radionuclide release from glass waste forms in a tuff repository and the effects on regulatory compliance

    International Nuclear Information System (INIS)

    Aines, R.D.

    1986-04-01

    This paper discusses preliminary estimates of the release of radionuclides from waste packages containing glass-based waste forms under the expected conditions at Yucca Mountain. These estimates can be used to evaluate the contribution of waste package performance toward meeting repository regulatory restrictions on radionuclide release. Glass waste will be held in double stainless steel canisters. After failure of the container sometime after the 300 to 1000 year containment period, the open headspace in these cans will provide the only area where standing water can accumulate and react with the glass. A maximum release rate of 0.177 g/m 2 x year or 1.3 grams per year was obtained. Normalized loss of 1.3 grams per year corresponds to 0.08 parts in 100,000 per year of the 1660 kg reference weight of DWPF glass

  19. Multi-purpose canister project overview

    International Nuclear Information System (INIS)

    Williams, J.

    1995-01-01

    In this presentation, the author lists the approved and proposed dry storage technologies. He discusses the compatibility of dry storage systems with waste management systems. Historical aspects, recent history, key features of the program approach, benefits, specifications, acquisition and potential utility use of the multi-purpose canister (MPC) are covered. The MPCs provide standardization in the waste management system and a cost savings to utilities and government. MPC will be developed to the same level as existing dry storage systems

  20. Design ampersand construction innovations of the defense waste processing facility

    International Nuclear Information System (INIS)

    McKibben, J.M.; Pair, C.R.; Bethmann, H.K.

    1990-01-01

    Construction of the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) is essentially complete. The facility is designed to convert high-level radioactive waste, now contained in large steel tanks as aqueous salts and sludge, into solid borosilicate glass in stainless steel canisters. All processing of the radioactive material and operations in a radioactive environment will be done remotely. The stringent requirements dictated by remote operation and new approaches to the glassification process led to the development of a number of first-of-a-kind pieces of equipment, new construction fabrication and erection techniques, and new applications of old techniques. The design features and construction methods used in the vitrification building and its equipment were to accomplish the objective of providing a state-of-the-art vitrification facility. 3 refs., 10 figs

  1. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    C.E. Sanders

    2005-04-07

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for

  2. CANISTER HANDLING FACILITY CRITICALITY SAFETY CALCULATIONS

    International Nuclear Information System (INIS)

    C.E. Sanders

    2005-01-01

    This design calculation revises and updates the previous criticality evaluation for the canister handling, transfer and staging operations to be performed in the Canister Handling Facility (CHF) documented in BSC [Bechtel SAIC Company] 2004 [DIRS 167614]. The purpose of the calculation is to demonstrate that the handling operations of canisters performed in the CHF meet the nuclear criticality safety design criteria specified in the ''Project Design Criteria (PDC) Document'' (BSC 2004 [DIRS 171599], Section 4.9.2.2), the nuclear facility safety requirement in ''Project Requirements Document'' (Canori and Leitner 2003 [DIRS 166275], p. 4-206), the functional/operational nuclear safety requirement in the ''Project Functional and Operational Requirements'' document (Curry 2004 [DIRS 170557], p. 75), and the functional nuclear criticality safety requirements described in the ''Canister Handling Facility Description Document'' (BSC 2004 [DIRS 168992], Sections 3.1.1.3.4.13 and 3.2.3). Specific scope of work contained in this activity consists of updating the Category 1 and 2 event sequence evaluations as identified in the ''Categorization of Event Sequences for License Application'' (BSC 2004 [DIRS 167268], Section 7). The CHF is limited in throughput capacity to handling sealed U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and high-level radioactive waste (HLW) canisters, defense high-level radioactive waste (DHLW), naval canisters, multicanister overpacks (MCOs), vertical dual-purpose canisters (DPCs), and multipurpose canisters (MPCs) (if and when they become available) (BSC 2004 [DIRS 168992], p. 1-1). It should be noted that the design and safety analyses of the naval canisters are the responsibility of the U.S. Department of the Navy (Naval Nuclear Propulsion Program) and will not be included in this document. In addition, this calculation is valid for the current design of the CHF and may not reflect the ongoing design evolution of the facility

  3. Feasibility of disposal of high-level radioactive waste into the seabed. Volume 8: Review of processes near a buried waste canister

    International Nuclear Information System (INIS)

    Lanza, F.

    1988-01-01

    One of the options suggested for disposal of high-level radioactive waste resulting from the generation of nuclear power is burial beneath the deep ocean floor in geologically stable sediment formations which have no economic value. The 8-volume series provides an assessment of the technical feasibility and radiological safety of this disposal concept based on the results obtained by ten years of co-operation and infomation exchange among the Member countries participating in the NEA Seabed Working Group. This report investigates the phenomena arriving in the proximity of the waste package immersed in the sea sediments

  4. Manufacture of disposal canisters

    International Nuclear Information System (INIS)

    Nolvi, L.

    2009-12-01

    The report summarizes the development work carried out in the manufacturing of disposal canister components, and present status, in readiness for manufacturing, of the components for use in assembly of spent nuclear fuel disposal canister. The disposal canister consist of two major components: the nodular graphite cast iron insert and overpack of oxygen-free copper. The manufacturing process for copper components begins with a cylindrical cast copper billet. Three different manufacturing processes i.e. pierce and draw, extrusion and forging are being developed, which produce a seamless copper tube or a tube with an integrated bottom. The pierce and draw process, Posiva's reference method, makes an integrated bottom possible and only the lid requires welding. Inserts for BWR-element are cast with 12 square channels and inserts for VVER 440-element with 12 round channels. Inserts for EPR-elements have four square channels. Casting of BWR insert type has been studied so far. Experience of casting inserts for PWR, which is similar to the EPR-type, has been got in co-operation with SKB. The report describes the processes being developed for manufacture of disposal canister components and some results of the manufacturing experiments are presented. Quality assurance and quality control in manufacture of canister component is described. (orig.)

  5. SOURCE TERMS FOR HLW GLASS CANISTERS

    International Nuclear Information System (INIS)

    J.S. Tang

    2000-01-01

    This calculation is prepared by the Monitored Geologic Repository (MGR) Waste Package Design Section. The objective of this calculation is to determine the source terms that include radionuclide inventory, decay heat, and radiation sources due to gamma rays and neutrons for the high-level radioactive waste (HLW) from the, West Valley Demonstration Project (WVDP), Savannah River Site (SRS), Hanford Site (HS), and Idaho National Engineering and Environmental Laboratory (INEEL). This calculation also determines the source terms of the canister containing the SRS HLW glass and immobilized plutonium. The scope of this calculation is limited to source terms for a time period out to one million years. The results of this calculation may be used to carry out performance assessment of the potential repository and to evaluate radiation environments surrounding the waste packages (WPs). This calculation was performed in accordance with the Development Plan ''Source Terms for HLW Glass Canisters'' (Ref. 7.24)

  6. Summary Of Cold Crucible Vitrification Tests Results With Savannah River Site High Level Waste Surrogates

    Energy Technology Data Exchange (ETDEWEB)

    Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

    2014-01-13

    The cold crucible inductive melting (CCIM) technology successfully applied for vitrification of low- and intermediate-level waste (LILW) at SIA Radon, Russia, was tested to be implemented for vitrification of high-level waste (HLW) stored at Savannah River Site, USA. Mixtures of Sludge Batch 2 (SB2) and 4 (SB4) waste surrogates and borosilicate frits as slurries were vitrified in bench- (236 mm inner diameter) and full-scale (418 mm inner diameter) cold crucibles. Various process conditions were tested and major process variables were determined. Melts were poured into 10L canisters and cooled to room temperature in air or in heat-insulated boxes by a regime similar to Canister Centerline Cooling (CCC) used at DWPF. The products with waste loading from ~40 to ~65 wt.% were investigated in details. The products contained 40 to 55 wt.% waste oxides were predominantly amorphous; at higher waste loadings (WL) spinel structure phases and nepheline were present. Normalized release values for Li, B, Na, and Si determined by PCT procedure remain lower than those from EA glass at waste loadings of up to 60 wt.%.

  7. Waste removal sequencing using ProdMod

    International Nuclear Information System (INIS)

    Paul, P.K.; Gregory, M.V.; Davis, N.R.; Brooke, J.N.

    1996-01-01

    The Savannah River Site (SRS) is starting to solidify its accumulated high-level radioactive waste into borosilicate glass in stainless steel canisters for eventual permanent storage. The in-tank precipitation process (ITP) and extended sludge processing (ESP) are two key operations in the waste processing complex. The supernate and dissolved salt from the waste storage tanks are transferred to the ITP process tank where the solution is decontaminated in batch processes. Soluble radioactive cesium is precipitated with sodium tetraphenylborate and strontium, uranium, and plutonium are adsorbed on monosodium titanate. The precipitate and adsorbent solids, which now contain the radionuclides, are concentrated using crossflow filters. The concentrated solids are sent to the high-level waste vitrification process. The decontaminated salt solution is sent to the low-level waste solidification process to form cement grout. In parallel with the precipitate operations, insoluble sludges that settled originally to the bottom of the waste tanks are reslurried and sent to ESP to undergo washing to reduce soluble salt content and aluminum dissolution, if required. In the vitrification process in the Defense Waste Processing Facility (DWPF), the concentrated precipitate from the ITP is mixed with the washed sludge from ESP and glass frit in proportion to form a stable borosilicate glass. A novel and fast-running Production Planning Model (ProdMod) has been developed to simulate the waste processing operation. This paper describes the application of ProdMod in sequencing the ITP batches and scheduling the ESP batches

  8. High-Level Waste Systems Plan. Revision 7 (U)

    International Nuclear Information System (INIS)

    Brooke, J.N.; Gregory, M.V.; Paul, P.; Taylor, G.; Wise, F.E.; Davis, N.R.; Wells, M.N.

    1996-10-01

    This revision of the High-Level Waste (HLW) System Plan aligns SRS HLW program planning with the DOE Savannah River (DOE-SR) Ten Year Plan (QC-96-0005, Draft 8/6), which was issued in July 1996. The objective of the Ten Year Plan is to complete cleanup at most nuclear sites within the next ten years. The two key principles of the Ten Year Plan are to accelerate the reduction of the most urgent risks to human health and the environment and to reduce mortgage costs. Accordingly, this System Plan describes the HLW program that will remove HLW from all 24 old-style tanks, and close 20 of those tanks, by 2006 with vitrification of all HLW by 2018. To achieve these goals, the DWPF canister production rate is projected to climb to 300 canisters per year starting in FY06, and remain at that rate through the end of the program in FY18, (Compare that to past System Plans, in which DWPF production peaked at 200 canisters per year, and the program did not complete until 2026.) An additional $247M (FY98 dollars) must be made available as requested over the ten year planning period, including a one-time $10M to enhance Late Wash attainment. If appropriate resources are made available, facility attainment issues are resolved and regulatory support is sufficient, then completion of the HLW program in 2018 would achieve a $3.3 billion cost savings to DOE, versus the cost of completing the program in 2026. Facility status information is current as of October 31, 1996

  9. Hazards analyses of hydrogen evolution and ammonium nitrate accumulation in DWPF -- Revision 1

    International Nuclear Information System (INIS)

    Holtzscheiter, E.W.

    1994-01-01

    This revision consists of two reports, the first of which is an analysis of potential ammonium nitrate explosion hazards in the DWPF (Defense Waste Processing Facility). Sections describe the effect of impurities (organic and inorganic (chlorides, chromates, metals and oxides)); the consequences of a hydrogen deflagration or detonation; the role of confinement; the action of heat on ammonium nitrate; the thermal decomposition of ammonium nitrate; the hazard of spontaneous heating; and the explosive decomposition of ammonium nitrate. The second report, Hazard analysis of hydrogen evolution in DWPF: Process vessels and vent system for the late wash/nitric acid flowsheet, contains a description of a revised model for hydrogen generation based on the late wash/nitric acid process. The second part of the report is a sensitivity analysis of the base case conditions and the hydrogen generation model

  10. Material Compatibility Evaluation for DWPF Nitric-Glycolic Acid - Literature Review

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. I. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Skidmore, T. E. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2015-09-30

    Glycolic acid is being evaluated as an alternative for formic and nitric acid in the DWPF flowsheet. Demonstration testing and modeling for this new flowsheet has shown that glycolic acid and glycolate has a potential to remain in certain streams generated during the production of the nuclear waste glass. A literature review was conducted to assess the impact of glycolic acid on the corrosion of the materials of construction for the DWPF facility as well as facilities downstream which may have residual glycolic acid and glycolates present. The literature data was limited to solutions containing principally glycolic acid. The reported corrosion rates and degradation characteristics have shown the following for the materials of construction.

  11. A natural analogue for copper waste canisters: The copper-uranium mineralised concretions in the Permian mudrocks of south Devon, United Kingdom

    Energy Technology Data Exchange (ETDEWEB)

    Milodowski, A.E.; Styles, M.T.; Hards, V.L. [Natural Environment Research Council (United Kingdom). British Geological Survey

    2000-08-01

    This report presents the results of a small-scale pilot study of the mineralogy and alteration characteristics of unusual sheet-like native copper occurring together with uraniferous and vanadiferous concretions in mudstones and siltstones of the Permian Littleham Mudstone Formation, at Littleham Cove, south Devon, England. The host mudstones and siltstones are smectitic and have been compacted through deep Mesozoic burial. The occurrence of native copper within these rocks represents a natural analogue for the long-term behaviour of copper canisters, sealed in a compacted clay (bentonite) backfill, that will be used for the deep geological disposal of high-level radioactive waste by the SKB. The study was undertaken by the British Geological Survey (BGS) on behalf of SKB between November 1999 and June 2000. The study was based primarily on archived reference material collected by the BGS during regional geological and mineralogical surveys of the area in the 1970's and 1980's. However, a brief visit was made to Littleham Cove in January 2000 to try to examine the native copper in situ and to collect additional material. Unfortunately, recent landslips and mudflows obscured much of the outcrop, and only one new sample of native copper could be collected. The native copper occurs as thin plates, up to 160 mm in diameter, which occur parallel to bedding in the Permian Littleham Mudstone Formation at Littleham Cove (near Budleigh Salterton) in south Devon. Each plate is made up of composite stacks of individual thin copper sheets each 1-2 mm thick. The copper is very pure (>99.4% Cu) but is accompanied by minor amounts of native silver (also pure - >99%) which occurs as small inclusions within the native copper. Detailed mineralogical and petrological studies of the native copper sheets, using optical petrography, backscattered scanning electron microscopy, X-ray diffraction analysis and electron probe microanalytical techniques, reveal a complex history of

  12. A natural analogue for copper waste canisters: The copper-uranium mineralised concretions in the Permian mudrocks of south Devon, United Kingdom

    International Nuclear Information System (INIS)

    Milodowski, A.E.; Styles, M.T.; Hards, V.L.

    2000-08-01

    This report presents the results of a small-scale pilot study of the mineralogy and alteration characteristics of unusual sheet-like native copper occurring together with uraniferous and vanadiferous concretions in mudstones and siltstones of the Permian Littleham Mudstone Formation, at Littleham Cove, south Devon, England. The host mudstones and siltstones are smectitic and have been compacted through deep Mesozoic burial. The occurrence of native copper within these rocks represents a natural analogue for the long-term behaviour of copper canisters, sealed in a compacted clay (bentonite) backfill, that will be used for the deep geological disposal of high-level radioactive waste by the SKB. The study was undertaken by the British Geological Survey (BGS) on behalf of SKB between November 1999 and June 2000. The study was based primarily on archived reference material collected by the BGS during regional geological and mineralogical surveys of the area in the 1970's and 1980's. However, a brief visit was made to Littleham Cove in January 2000 to try to examine the native copper in situ and to collect additional material. Unfortunately, recent landslips and mudflows obscured much of the outcrop, and only one new sample of native copper could be collected. The native copper occurs as thin plates, up to 160 mm in diameter, which occur parallel to bedding in the Permian Littleham Mudstone Formation at Littleham Cove (near Budleigh Salterton) in south Devon. Each plate is made up of composite stacks of individual thin copper sheets each 1-2 mm thick. The copper is very pure (>99.4% Cu) but is accompanied by minor amounts of native silver (also pure - >99%) which occurs as small inclusions within the native copper. Detailed mineralogical and petrological studies of the native copper sheets, using optical petrography, backscattered scanning electron microscopy, X-ray diffraction analysis and electron probe microanalytical techniques, reveal a complex history of

  13. Nuclear waste glass product consistency test (PCT): Version 7.0. Revision 3

    International Nuclear Information System (INIS)

    Jantzen, C.M.; Bibler, N.E.; Beam, D.C.; Ramsey, W.G.

    1994-06-01

    Liquid high-level nuclear waste will be immobilized at the Savannah River Site (SRS) by vitrification in borosilicate glass. The glass will be produced in the Defense Waste Processing Facility (DWPF), poured into stainless steel canisters, and eventually disposed of in a geologic repository. In order to comply with the Waste Acceptance Product Specifications (WAPS), the durability of the glass needs to be measured during production to assure its long term stability and radionuclide release properties. A durability test, designated the Product Consistency Test (PCT), was developed for DWPF glass in order to meet the WAPS requirements. The response of the PCT procedure was based on extensive testing with glasses of widely different compositions. The PCT was determined to be very reproducible, to yield reliable results rapidly, and to be easily performed in shielded cell facilities with radioactive samples. Version 7.0 of the PCT procedure is attached. This draft version has been submitted to ASTM for full committee (C26, Nuclear Fuel Cycle) ballot after being balloted successfully through subcommittee C26.13 on Repository Waste Package Materials Testing

  14. Mechanical design of the storage tubes in the HWVP canister storage building

    International Nuclear Information System (INIS)

    Divona, C.J.; Fages, R.; Janicek, G.P.; Mullally, J.A.

    1993-01-01

    Canisters of high-level waste from the Hanford Waste Vitrification Plant (HWVP) will be stored in an adjacent facility, the Canister Storage Building (CSB). The canisters are stored vertically in an array of tubes within the shielded vault area of the CSB. This paper describes the mechanical design of the storage tubes, the shield floor plugs that confine the waste within the tubes and the impact absorber system used to assure that the canisters are not breached in the event of an accidental drop. Installation and testing of the components is also discussed

  15. SLUDGE WASHING AND DEMONSTRATION OF THE DWPF FLOWSHEET IN THE SRNL SHIELDED CELLS FOR SLUDGE BATCH 6 QUALIFICATION

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J.; Pickenheim, B.; Bannochie, C.; Billings, A.; Bibler, N.; Click, D.

    2010-10-01

    Prior to initiating a new sludge batch in the Defense Waste Processing Facility (DWPF), Savannah River National Laboratory (SRNL) is required to simulate this processing, including Chemical Process Cell (CPC) simulation, waste glass fabrication, and chemical durability testing. This report documents this simulation for the next sludge batch, Sludge Batch 6 (SB6). SB6 consists of Tank 12 material that has been transferred to Tank 51 and subjected to Low Temperature Aluminum Dissolution (LTAD), Tank 4 sludge, and H-Canyon Pu solutions. Following LTAD and the Tank 4 addition, Liquid Waste Operations (LWO) provided SRNL a 3 L sample of Tank 51 sludge for SB6 qualification. Pu solution from H Canyon was also received. SB6 qualification included washing the sample per LWO plans/projections (including the addition of Pu from H Canyon), DWPF CPC simulations, waste glass fabrication (vitrification), and waste glass characterization and chemical durability evaluation. The following are significant observations from this demonstration. Sludge settling improved slightly as the sludge was washed. SRNL recommended (and the Tank Farm implemented) one less wash based on evaluations of Tank 40 heel projections and projections of the glass composition following transfer of Tank 51 to Tank 40. Thorium was detected in significant quantities (>0.1 wt % of total solids) in the sludge. In past sludge batches, thorium has been determined by Inductively Coupled Plasma-Mass Spectroscopy (ICP-MS), seen in small quantities, and reported with the radionuclides. As a result of the high thorium, SRNL-AD has added thorium to their suite of Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES) elements. The acid stoichiometry for the DWPF Sludge Receipt and Adjustment Tank (SRAT) processing of 115%, or 1.3 mol acid per liter of SRAT receipt slurry, was adequate to accomplish some of the goals of SRAT processing: nitrite was destroyed to below 1,000 mg/kg and mercury was removed to

  16. Sludge Washing And Demonstration Of The DWPF Flowsheet In The SRNL Shielded Cells For Sludge Batch 8 Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Pareizs, J. M.; Crawford, C. L.

    2013-04-26

    The current Waste Solidification Engineering (WSE) practice is to prepare sludge batches in Tank 51 by transferring sludge from other tanks to Tank 51. Tank 51 sludge is washed and transferred to Tank 40, the current Defense Waste Processing Facility (DWPF) feed tank. Prior to transfer of Tank 51 to Tank 40, the Savannah River National Laboratory (SRNL) typically simulates the Tank Farm and DWPF processes using a Tank 51 sample (referred to as the qualification sample). WSE requested the SRNL to perform characterization on a Sludge Batch 8 (SB8) sample and demonstrate the DWPF flowsheet in the SRNL shielded cells for SB8 as the final qualification process required prior to SB8 transfer from Tank 51 to Tank 40. A 3-L sample from Tank 51 (the SB8 qualification sample; Tank Farm sample HTF-51-12-80) was received by SRNL on September 20, 2012. The as-received sample was characterized prior to being washed. The washed material was further characterized and used as the material for the DWPF process simulation including a Sludge Receipt and Adjustment Tank (SRAT) cycle, a Slurry Mix Evaporator (SME) cycle, and glass fabrication and chemical durability measurements.

  17. Design basis for the copper/steel canister

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    1996-02-01

    The development of the copper/iron canister which has been proposed by SKB for the containment of high level nuclear waste has been studied from the point of view of choice of materials, manufacturing technology and quality assurance. This report describes the observations on progress which have been made between March 1995 and Feb 1996 and the result of further literature studies. A first trial canister has been produced using a fabricated steel liner and an extruded copper tubular, a second one using a fabricated tubular is at an advanced stage. A change from a fabricated steel inner canister to a proposed cast canister has been justified by a criticality argument but the technology for producing a cast canister is at present untried. The microstructure achieved in the extruded copper tubular for the first canister is unacceptable. Similar problems exist with plate used for the fabricated tubular, but some more favourable structures have been achieved already by this route. Seam welding of the first tubular failed through a suspected material problem. The second fabricated tubular welded without difficulty. Welding of lids and bottoms to the copper canister is problematical.There is as yet no satisfactory non destructive test procedures for the parent metal or the welds in the copper canister material, partly due to the coarse grain size which arise in the proposed material processed by the proposed routes. Further studies are also required on crevice corrosion, galvanic attack and stress corrosion cracking in the copper 50 ppm phosphorus alloy. 28 refs

  18. DWPF liquid sample station: Status of equipment development

    International Nuclear Information System (INIS)

    Caplan, J.R.

    1987-01-01

    This report summarizes operating experience and equipment status of the DWPF liquid sample cell. Operation hours to date, results of equipment inspections and problems encountered and their solutions are discussed. An equipment and instrumentation status updating DPST-85-592, DWPF LIQUID SAMPLE CELL MOCK-UP, is presented. Remaining development items are also outlined

  19. Dissolution rates of DWPF glasses from long-term PCT

    International Nuclear Information System (INIS)

    Ebert, W.L.; Tam, S.W.

    1996-01-01

    We have characterized the corrosion behavior of several Defense Waste Processing Facility (DWPF) reference waste glasses by conducting static dissolution tests with crushed glasses. Glass dissolution rates were calculated from measured B concentrations in tests conducted for up to five years. The dissolution rates of all glasses increased significantly after certain alteration phases precipitated. Calculation of the dissolution rates was complicated by the decrease in the available surface area as the glass dissolves. We took the loss of surface area into account by modeling the particles to be spheres, then extracting from the short-term test results the dissolution rate corresponding to a linear decrease in the radius of spherical particles. The measured extent of dissolution in tests conducted for longer times was less than predicted with this linear dissolution model. This indicates that advanced stages of corrosion are affected by another process besides dissolution, which we believe to be associated with a decrease in the precipitation rate of the alteration phases. These results show that the dissolution rate measured soon after the formation of certain alteration phases provides an upper limit for the long-term dissolution rate, and can be used to determine a bounding value for the source term for radionuclide release from waste glasses. The long-term dissolution rates measured in tests at 20,000 per m at 90 degrees C in tuff groundwater at pH values near 12 for the Environmental Assessment glass and glasses made with SRL 131 and SRL 202 frits, respectively

  20. A Film Canister Colorimeter.

    Science.gov (United States)

    Gordon, James; James, Alan; Harman, Stephanie; Weiss, Kristen

    2002-01-01

    A low-cost, low-tech colorimeter was constructed from a film canister. The student-constructed colorimeter was used to show the Beer-Lambert relationship between absorbance and concentration and to calculate the value of the molar absorptivity for permanganate at the wavelength emission maximum for an LED. Makes comparisons between this instrument…

  1. Control of high-level radioactive waste-glass melters

    International Nuclear Information System (INIS)

    Bickford, D.F.; Coleman, C.J.

    1990-01-01

    The Defense Waste Processing Facility (DWPF) will immobilize Savannah River Site High Level Waste as a durable borosilicate glass for permanent disposal in a repository. The DWPF will be controlled based on glass composition. The following discussion is a preliminary analysis of the capability of the laboratory methods that can be used to control the glass composition, and the relationships between glass durability and glass properties important to glass melting. The glass durability and processing properties will be controlled by controlling the chemical composition of the glass. The glass composition will be controlled by control of the melter feed transferred from the Slurry Mix Evaporator (SME) to the Melter Feed Tank (MFT). During cold runs, tests will be conducted to demonstrate the chemical equivalence of glass sampled from the pour stream and glass removed from cooled canisters. In similar tests, the compositions of glass produced from slurries sampled from the SME and MFT will be compared to final product glass to determine the statistical relationships between melter feed and glass product. The total error is the combination of those associated with homogeneity in the SME or MFT, sampling, preparation of samples for analysis, instrument calibration, analysis, and the composition/property model. This study investigated the sensitivity of estimation of property data to the combination of variations from sampling through analysis. In this or a similar manner, the need for routine glass product sampling will be minimized, and glass product characteristics will be assured before the melter feed is committed to the melter

  2. Assessment of a spent fuel disposal canister. Assessment studies for a copper canister with cast steel inner component

    International Nuclear Information System (INIS)

    Bond, A.E.; Hoch, A.R.; Jones, G.D.; Tomczyk, A.J.; Wiggin, R.M.; Worraker, W.J.

    1997-05-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden, is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in vertical storage holes drilled in a series of caverns excavated from the granite bedrock at a depth of about 500 m. Each canister will be surrounded by compacted bentonite clay. In this report, a simple model of the behaviour of the canister subsequent to a first breach in its copper overpack is developed. This model is used to predict: -the ingress of water to the canister (as a function of the size and the shape of the initial defect, the buffer conductivity, the corrosion rate and the pressure inside the canister); -the build-up of corrosion products in the canister (as a function of the available water in the canister, the corrosion rate and the properties of the corrosion products); -the effect of corrosion on the structural integrity of the canister. A number of different scenarios for the location of the breach in the copper overpack are considered

  3. DWPF Sample Vial Insert Study-Statistical Analysis of DWPF Mock-Up Test Data

    Energy Technology Data Exchange (ETDEWEB)

    Harris, S.P. [Westinghouse Savannah River Company, AIKEN, SC (United States)

    1997-09-18

    This report is prepared as part of Technical/QA Task Plan WSRC-RP-97-351 which was issued in response to Technical Task Request HLW/DWPF/TTR-970132 submitted by DWPF. Presented in this report is a statistical analysis of DWPF Mock-up test data for evaluation of two new analytical methods which use insert samples from the existing HydragardTM sampler. The first is a new hydrofluoric acid based method called the Cold Chemical Method (Cold Chem) and the second is a modified fusion method.Either new DWPF analytical method could result in a two to three fold improvement in sample analysis time.Both new methods use the existing HydragardTM sampler to collect a smaller insert sample from the process sampling system. The insert testing methodology applies to the DWPF Slurry Mix Evaporator (SME) and the Melter Feed Tank (MFT) samples.The insert sample is named after the initial trials which placed the container inside the sample (peanut) vials. Samples in small 3 ml containers (Inserts) are analyzed by either the cold chemical method or a modified fusion method. The current analytical method uses a HydragardTM sample station to obtain nearly full 15 ml peanut vials. The samples are prepared by a multi-step process for Inductively Coupled Plasma (ICP) analysis by drying, vitrification, grinding and finally dissolution by either mixed acid or fusion. In contrast, the insert sample is placed directly in the dissolution vessel, thus eliminating the drying, vitrification and grinding operations for the Cold chem method. Although the modified fusion still requires drying and calcine conversion, the process is rapid due to the decreased sample size and that no vitrification step is required.A slurry feed simulant material was acquired from the TNX pilot facility from the test run designated as PX-7.The Mock-up test data were gathered on the basis of a statistical design presented in SRT-SCS-97004 (Rev. 0). Simulant PX-7 samples were taken in the DWPF Analytical Cell Mock

  4. Evaluation and compilation of DOE waste package test data: Biannual report, August 1986-January 1987

    International Nuclear Information System (INIS)

    Interrante, C.; Escalante, E.; Fraker, A.; Harrison, S.; Shull, R.; Linzer, M.; Ricker, R.; Ruspi, J.

    1987-10-01

    This report summarizes results of the National Bureau of Standards (NBS) evaluations of Department of Energy (DOE) activities on waste packages designed for containment of radioactive high-level nuclear waste (HLW). The waste package is a proposed engineered barrier that is part of a permanent repository for HLW. Metal alloys are the principal barriers within the engineered system. Technical discussions are given for the corrosion of metals proposed for the canister, particularly carbon and stainless steels, and copper. In the section on tuff, the current level of understanding of several canister materials is questioned. Within the Basalt Waste Isolation Project (BWIP) section, discussions are given on problems concerning groundwater, materials for use in the metallic overpack, and diffusion through the packing. For the proposed salt site, questions are raised on the work on both ASTM A216 Steel and Ti-Code 12. NBS work related to the vitrification of HLW borosilicate glass at the West Valley Demonstration Project (WVDP) and the Defense Waste Processing Facility (DWPF) is covered. NBS reviews of selected DOE technical reports and a summary of current waste-package activities of the Materials Characterization Center (MCC) is presented. Using a database management system, a computerized database for storage and retrieval of reviews and evaluations of HLW data has been developed and is described. 17 refs., 2 figs., 2 tabs

  5. Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Amoroso, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Mcclane, D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-09-26

    The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy’s Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanning calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF

  6. Nucleation and crystal growth behavior of nepheline in simulated high-level waste glasses

    International Nuclear Information System (INIS)

    Fox, K.; Amoroso, J.; Mcclane, D.

    2017-01-01

    The Savannah River National Laboratory (SRNL) has been tasked with supporting glass formulation development and process control strategies in key technical areas, relevant to the Department of Energy's Office of River Protection (DOE-ORP) and related to high-level waste (HLW) vitrification at the Waste Treatment and Immobilization Plant (WTP). Of specific interest is the development of predictive models for crystallization of nepheline (NaAlSiO4) in HLW glasses formulated at high alumina concentrations. This report summarizes recent progress by researchers at SRNL towards developing a predicative tool for quantifying nepheline crystallization in HLW glass canisters using laboratory experiments. In this work, differential scanning calorimetry (DSC) was used to obtain the temperature regions over which nucleation and growth of nepheline occur in three simulated HLW glasses - two glasses representative of WTP projections and one glass representative of the Defense Waste Processing Facility (DWPF) product. The DWPF glass, which has been studied previously, was chosen as a reference composition and for comparison purposes. Complementary quantitative X-ray diffraction (XRD) and optical microscopy confirmed the validity of the methodology to determine nucleation and growth behavior as a function of temperature. The nepheline crystallization growth region was determined to generally extend from ~ 500 to >850 °C, with the maximum growth rates occurring between 600 and 700 °C. For select WTP glass compositions (high Al2O3 and B2O3), the nucleation range extended from ~ 450 to 600 °C, with the maximum nucleation rates occurring at ~ 530 °C. For the DWPF glass composition, the nucleation range extended from ~ 450 to 750 °C with the maximum nucleation rate occurring at ~ 640 °C. The nepheline growth at the peak temperature, as determined by XRD, was between 35 - 75 wt.% /hour. A maximum nepheline growth rate of ~ 0.1 mm/hour at 700 °C was measured for the DWPF

  7. Corrosion resistance of copper canister weld material

    International Nuclear Information System (INIS)

    Gubner, Rolf; Andersson, Urban

    2007-03-01

    The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

  8. Corrosion resistance of copper canister weld material

    Energy Technology Data Exchange (ETDEWEB)

    Gubner, Rolf; Andersson, Urban [Corrosion and Metals Research Institute, Sto ckholm (Sweden)

    2007-03-15

    The proposed design for a final repository for spent fuel and other long-lived residues is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast iron insert fitted inside a copper canister. SKB has since several years developed manufacturing processes for the canister components using a network of manufacturers. For the encapsulation process SKB has built the Canister Laboratory to demonstrate and develop the encapsulation technique in full scale. The critical part of the encapsulation of spent fuel is the sealing of the canister which is done by welding the copper lid to the cylindrical part of the canister. Two welding techniques have been developed in parallel, Electron Beam Welding (EBW) and Friction Stir Welding (FSW). During the past two decades, SKB has developed the technology EBW at The Welding Institute (TWI) in Cambridge, UK. The development work at the Canister Laboratory began in 1999. In electron beam welding, a gun is used to generate the electron beam which is aimed at the joint. The beam heats up the material to the melting point allowing a fusion weld to be formed. The gun was developed by TWI and has a unique design for use at reduced pressure. The system has gone through a number of improvements under the last couple of years including implementation of a beam oscillation system. However, during fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow they will tend to concentrate impurities at the new grain boundaries that might pose adverse effects on the corrosion resistance of welds. As a new method for joining, SKB has been developing friction stir welding (FSW) for sealing copper canisters for spent nuclear fuel in cooperation with TWI since 1997. FSW was invented in 1991 at TWI and is a thermo

  9. Bounding estimate of DWPF mercury emissions

    International Nuclear Information System (INIS)

    Jacobs, R.A.

    1993-01-01

    Two factors which have substantial impact on predicted Mercury emissions are the air flows in the Chemical Process Cell (CPC) and the exit temperature of the Formic Acid Vent Condenser (FAVC). The discovery in the IDMS (Integrated DWPF Melter System) of H 2 generation by noble metal catalyzed formic acid decomposition and the resultant required dilution air flow has increased the expected instantaneous CPC air flow by as much as a factor of four. In addition, IDMS has experienced higher than design (10 degrees C) FAVC exit temperatures during certain portions of the operating cycle. These temperatures were subsequently attributed to the exothermic reaction of NO to NO 2 . Moreover, evaluation of the DWPF FAVC indicated it was undersized and unless modified or replaced, routine exit temperatures would be in excess of design. Purges required for H 2 flammability control and verification of elevated FAVC exit temperatures due to NO x reactions have lead to significant changes in CPC operating conditions. Accordingly, mercury emissions estimates have been updated based upon the new operating requirements, IDMS experience, and development of an NO x /FAVC model which predicts FAVC exit temperatures. Using very conservative assumptions and maximum purge rates, the maximum calculated Hg emissions is approximately 130 lbs/yr. A range of 100 to 120 lbs/yr is conservatively predicted for other operating conditions. The peak emission rate calculated is 0.027 lbs/hr. The estimated DWPF Hg emissions for the construction permit are 175 lbs/yr (0.02 lbs/hr annual average)

  10. Feasibility study for a DOE research and production fuel multipurpose canister

    International Nuclear Information System (INIS)

    Lopez, D.A.; Abbott, D.G.

    1994-02-01

    This is a report of the feasibility of multipurpose canisters for transporting, storing, and sing of Department of Energy research and production spent nuclear fuel. Six representative Department of Energy fuel assemblies were selected, and preconceptual canister designs were developed to accommodate these assemblies. The study considered physical interface, structural adequacy, criticality safety, shielding capability, thermal performance of the canisters, and fuel storage site infrastructure. The external envelope of the canisters was designed to fit within the overpack casks for commercial canisters being developed for the Department of Energy Office of Civilian Radioactive Waste Management. The budgetary cost of canisters to handle all fuel considered is estimated at $170.8M. One large conceptual boiling water reactor canister design, developed for the Office of Civilian Radioactive Waste Management, and two new canister designs can accommodate at least 85% of the volume of the Department of Energy fuel considered. Canister use minimizes public radiation exposure and is cost effective compared with bare fuel handling. Results suggest the need for additional study of issues affecting canister use and for conceptual design development of the three canisters

  11. Large-scale continuous process to vitrify nuclear defense waste: operating experience with nonradioactive waste

    International Nuclear Information System (INIS)

    Cosper, M.B.; Randall, C.T.; Traverso, G.M.

    1982-01-01

    The developmental program underway at SRL has demonstrated the vitrification process proposed for the sludge processing facility of the DWPF on a large scale. DWPF design criteria for production rate, equipment lifetime, and operability have all been met. The expected authorization and construction of the DWPF will result in the safe and permanent immobilization of a major quantity of existing high level waste. 11 figures, 4 tables

  12. Crystallization in high level waste (HLW) glass melters: Savannah River Site operational experience

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Kevin M. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Peeler, David K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kruger, Albert A. [USDOE Office of River Protection, Richland, WA (United States)

    2015-06-12

    This paper provides a review of the scaled melter testing that was completed for design input to the Defense Waste Processing Facility (DWPF) melter. Testing with prototype melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by refractory corrosion versus spinels that precipitated from the HLW glass melt pool. A review of the crystallization observed with the prototype melters and the full-scale DWPF melters (DWPF Melter 1 and DWPF Melter 2) is included. Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for a waste treatment and immobilization plant.

  13. Integration of the Uncertainties of Anion and TOC Measurements into the Flammability Control Strategy for Sludge Batch 8 at the DWPF

    International Nuclear Information System (INIS)

    Edwards, T. B.

    2013-01-01

    The Savannah River National Laboratory (SRNL) has been working with the Savannah River Remediation (SRR) Defense Waste Processing Facility (DWPF) in the development and implementation of a flammability control strategy for DWPF's melter operation during the processing of Sludge Batch 8 (SB8). SRNL's support has been in response to technical task requests that have been made by SRR's Waste Solidification Engineering (WSE) organization. The flammability control strategy relies on measurements that are performed on Slurry Mix Evaporator (SME) samples by the DWPF Laboratory. Measurements of nitrate, oxalate, formate, and total organic carbon (TOC) standards generated by the DWPF Laboratory are presented in this report, and an evaluation of the uncertainties of these measurements is provided. The impact of the uncertainties of these measurements on DWPF's strategy for controlling melter flammability also is evaluated. The strategy includes monitoring each SME batch for its nitrate content and its TOC content relative to the nitrate content and relative to the antifoam additions made during the preparation of the SME batch. A linearized approach for monitoring the relationship between TOC and nitrate is developed, equations are provided that integrate the measurement uncertainties into the flammability control strategy, and sample calculations for these equations are shown to illustrate the impact of the uncertainties on the flammability control strategy

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

    International Nuclear Information System (INIS)

    Ha, B.C.

    1994-01-01

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

  15. Impact of Glycolate Anion on Aqueous Corrosion in DWPF and Downstream Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-12

    Glycolic acid is being evaluated as an alternate reductant in the preparation of high level waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid may not be completely consumed with small quantities of the glycolate anion being carried forward to other high level waste (HLW) facilities. The SRS liquid waste contractor requested an assessment of the impact of the glycolate anion on the corrosion of the materials of construction (MoC) throughout the waste processing system since this impact had not been previously evaluated. A literature review revealed that corrosion data were not available for the MoCs in glycolic-bearing solutions applicable to SRS systems. Data on the material compatibility with only glycolic acid or its derivative products were identified; however, data were limited for solutions containing glycolic acid or the glycolate anion. For the proprietary coating systems applied to the DWPF concrete, glycolic acid was deemed compatible since the coatings were resistant to more aggressive chemistries than glycolic acid. Additionally similar coating resins showed acceptable resistance to glycolic acid.

  16. Impact of Glycolate Anion on Aqueous Corrosion in DWPF and Downstream Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Mickalonis, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-11-20

    Glycolic acid is being evaluated as an alternate reductant in the preparation of high level waste for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS). During processing, the glycolic acid may not be completely consumed with small quantities of the glycolate anion being carried forward to other high level waste (HLW) facilities. The SRS liquid waste contractor requested an assessment of the impact of the glycolate anion on the corrosion of the materials of construction (MoC) throughout the waste processing system since this impact had not been previously evaluated. A literature review revealed that corrosion data were not available for the MoCs in glycolic-bearing solutions applicable to SRS systems. Data on the material compatibility with only glycolic acid or its derivative products were identified; however, data were limited for solutions containing glycolic acid or the glycolate anion. For the proprietary coating systems applied to the DWPF concrete, glycolic acid was deemed compatible since the coatings were resistant to more aggressive chemistries than glycolic acid. Additionally, similar coating resins showed acceptable resistance to glycolic acid.

  17. Status report, canister fabrication

    Energy Technology Data Exchange (ETDEWEB)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika [Swedish Nuclear Fuel and Waste Management Co., Stockholm (Sweden); Emilsson, Goeran [CSM Materialteknik AB, Linkoeping (Sweden)

    2004-06-01

    The report gives an account of the development of material and fabrication technology for copper canisters with cast inserts during the period from 2000 until the start of 2004. The engineering design of the canister and the choice of materials in the constituent components described in previous status reports have not been significantly changed. In the reference canister, the thickness of the copper shell is 50 mm. Fabrication of individual components with a thinner copper thickness is done for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. As a part of the development of cast inserts, computer simulations of the casting processes and techniques used at the foundries have been performed for the purpose of optimizing the material properties. These properties have been evaluated by extensive tensile testing and metallographic inspection of test material taken from discs cut at different points along the length of the inserts. The testing results exhibit a relatively large spread. Low elongation values in certain tensile test specimens are due to the presence of poorly formed graphite, porosities, slag or other casting defects. It is concluded in the report that it will not be possible to avoid some presence of observed defects in castings of this size. In the deep repository, the inserts will be exposed to compressive loading and the observed defects are not critical for strength. An analysis of the strength of the inserts and formulation of relevant material requirements must be based on a statistical approach with probabilistic calculations. This work has been initiated and will be concluded during 2004. An initial verifying compression test of a canister in an isostatic press has indicated considerable overstrength in the structure. Seamless copper tubes are fabricated by means of three methods: extrusion, pierce and draw processing, and forging. It can be concluded that extrusion tests have revealed a

  18. Status report, canister fabrication

    International Nuclear Information System (INIS)

    Andersson, Claes-Goeran; Eriksson, Peter; Westman, Marika; Emilsson, Goeran

    2004-06-01

    The report gives an account of the development of material and fabrication technology for copper canisters with cast inserts during the period from 2000 until the start of 2004. The engineering design of the canister and the choice of materials in the constituent components described in previous status reports have not been significantly changed. In the reference canister, the thickness of the copper shell is 50 mm. Fabrication of individual components with a thinner copper thickness is done for the purpose of gaining experience and evaluating fabrication and inspection methods for such canisters. As a part of the development of cast inserts, computer simulations of the casting processes and techniques used at the foundries have been performed for the purpose of optimizing the material properties. These properties have been evaluated by extensive tensile testing and metallographic inspection of test material taken from discs cut at different points along the length of the inserts. The testing results exhibit a relatively large spread. Low elongation values in certain tensile test specimens are due to the presence of poorly formed graphite, porosities, slag or other casting defects. It is concluded in the report that it will not be possible to avoid some presence of observed defects in castings of this size. In the deep repository, the inserts will be exposed to compressive loading and the observed defects are not critical for strength. An analysis of the strength of the inserts and formulation of relevant material requirements must be based on a statistical approach with probabilistic calculations. This work has been initiated and will be concluded during 2004. An initial verifying compression test of a canister in an isostatic press has indicated considerable overstrength in the structure. Seamless copper tubes are fabricated by means of three methods: extrusion, pierce and draw processing, and forging. It can be concluded that extrusion tests have revealed a

  19. Phase II of a Six sigma Initiative to Study DWPF SME Analytical Turnaround Times: SRNL's Evaluation of Carbonate-Based Dissolution Methods

    International Nuclear Information System (INIS)

    Edwards, Thomas

    2005-01-01

    The Analytical Development Section (ADS) and the Statistical Consulting Section (SCS) of the Savannah River National Laboratory (SRNL) are participating in a Six Sigma initiative to improve the Defense Waste Processing Facility (DWPF) Laboratory. The Six Sigma initiative has focused on reducing the analytical turnaround time of samples from the Slurry Mix Evaporator (SME) by developing streamlined sampling and analytical methods [1]. The objective of Phase I was to evaluate the sub-sampling of a larger sample bottle and the performance of a cesium carbonate (Cs 2 CO 3 ) digestion method. Successful implementation of the Cs 2 CO 3 fusion method in the DWPF would have important time savings and convenience benefits because this single digestion would replace the dual digestion scheme now used. A single digestion scheme would result in more efficient operations in both the DWPF shielded cells and the inductively coupled plasma--atomic emission spectroscopy (ICP-AES) laboratory. By taking a small aliquot of SME slurry from a large sample bottle and dissolving the vitrified SME sample with carbonate fusion methods, an analytical turnaround time reduction from 27 hours to 9 hours could be realized in the DWPF. This analytical scheme has the potential for not only dramatically reducing turnaround times, but also streamlining operations to minimize wear and tear on critical shielded cell components that are prone to fail, including the Hydragard(trademark) sampling valves and manipulators. Favorable results from the Phase I tests [2] led to the recommendation for a Phase II effort as outlined in the DWPF Technical Task Request (TTR) [3]. There were three major tasks outlined in the TTR, and SRNL issued a Task Technical and QA Plan [4] with a corresponding set of three major task activities: (1) Compare weight percent (wt%) total solids measurements of large volume samples versus peanut vial samples. (2) Evaluate Cs 2 CO 3 and K 2 CO 3 fusion methods using DWPF simulated

  20. Waste Tank Vapor Project: Vapor characterization of Tank 241-C-103: Report for SUMMA trademark canister samples received 11/29/93 (sample jobs 4 and 5)

    International Nuclear Information System (INIS)

    Clauss, T.R.; Lucke, R.B.; McVeety, B.; Allwine, K.J.; Fruchter, J.S.

    1994-09-01

    The purpose of Sample Jobs 4 and 5 was to determine whether the organic nitrites observed on the outside of tank 241-C-103 originated in the tank or from degradation products of the high-efficiency particulate air (HEPA) filter. The plan was to take samples from either side of the HE-PA filter. The relative level of organic nitrites would help determine whether they were produced in the filter or the tank. Pacific Northwest Laboratory was responsible for analyzing the SUMMA trademark canisters collected in support of this study. The laboratory was to analyze the SUMMA trademark Canister samples according to letters of instruction and report all semivolatile and volatile organic constituents detected in the tank headspace. Pacific Northwest Laboratory was also to submit a letter report to the Program Manager of all qualitative and quantitative analytical data, and estimate concentrations of any aliphatic nitrites identified. This was one of the first sampling activities for this program, and a number of errors were made both in the field and in the laboratory. Because of these errors, the samples and results were of questionable value. Therefore, Westinghouse program management asked that the analysis of the samples for this report not be completed. This report describes the few results that were generated before we were asked to stop work on this activity. In addition to analyzing SUMMA trademark canisters, PNL operates a site portable weather station near tank 241-C-103. Pacific Northwest Laboratory was required to collect atmospheric data starting 11/15/93, but the weather station was already collecting data during the time of both these two sample jobs (11/12/93 and 11/16/93). Therefore, a summary of the atmospheric data is also presented in this report

  1. K West Basin canister survey

    International Nuclear Information System (INIS)

    Pitner, A.L.

    1998-01-01

    A survey was conducted of the K West Basin to determine the distribution of canister types that contain the irradiated N Reactor fuel. An underwater camera was used to conduct the survey during June 1998, and the results were recorded on videotape. A full row-by-row survey of the entire basin was performed, with the distinction between aluminum and stainless steel Mark 1 canisters made by the presence or absence of steel rings on the canister trunions (aluminum canisters have the steel rings). The results of the survey are presented in tables and figures. Grid maps of the three bays show the canister lid ID number and the canister type in each location that contained fuel. The following abbreviations are used in the grid maps for canister type designation: IA = Mark 1 aluminum, IS = Mark 1 stainless steel, and 2 = Mark 2 stainless steel. An overall summary of the canister distribution survey is presented in Table 1. The total number of canisters found to contain fuel was 3842, with 20% being Mark 1 Al, 25% being Mark 1 SS, and 55% being Mark 2 SS. The aluminum canisters were predominantly located in the East and West bays of the basin

  2. Effect of HNO3-cerium(IV) decontamination on stainless steel canister materials

    International Nuclear Information System (INIS)

    Westerman, R.E.; Mackey, D.B.

    1991-01-01

    Stainless steel canisters will be filled with vitrified radioactive waste at the West Valley Demonstration Project (WVDP), West Valley, NY. After they are filled, the sealed canisters will be decontaminated by immersion in a HNO 3 -Ce(IV) solution, which will remove the oxide film and a small amount of metal from the surface of the canisters. Studies were undertaken in support of waste form qualification activities to determine the effect of this decontamination treatment on the legibility of the weld-bead canister identification label, and to determine whether this decontamination treatment could induce stress-corrosion cracking (SCC) in the AISI 304L stainless steel (SS) canister material. Neither the label legibility nor the canister integrity with regard to SCC were found to be prejudiced by the simulated decontamination treatment

  3. Radioactive waste disposal package

    Science.gov (United States)

    Lampe, Robert F.

    1986-11-04

    A radioactive waste disposal package comprising a canister for containing vitrified radioactive waste material and a sealed outer shell encapsulating the canister. A solid block of filler material is supported in said shell and convertible into a liquid state for flow into the space between the canister and outer shell and subsequently hardened to form a solid, impervious layer occupying such space.

  4. The behavior and effects of the noble metals in the DWPF melter system

    International Nuclear Information System (INIS)

    Hutson, N.D.; Smith, M.E.

    1992-01-01

    Fission-product noble metals have caused severe operating problems in numerous worldwide waste vitrification facilities. These dense, highly conductive noble metals have tended to accumulate on the floor of joule-heated glass melters causing electrical distortions which have, in some occurrences, rendered the melter inoperable. A pilot scale vitrification research facility at the U.S. Department of Energy's Savannah River Laboratory has been operated for more than a year with simulated feed streams containing noble metals. In this paper the behavior of these noble metals in the melter system and final glass product and their effects on the scaled DWPF-type melter are discussed

  5. DWPF Sample Vial Insert Study-Statistical Analysis of DWPF Mock-Up Test Data

    International Nuclear Information System (INIS)

    Harris, S.P.

    1997-01-01

    This report is prepared as part of Technical/QA Task Plan WSRC-RP-97-351 which was issued in response to Technical Task Request HLW/DWPF/TTR-970132 submitted by DWPF. Presented in this report is a statistical analysis of DWPF Mock-up test data for evaluation of two new analytical methods which use insert samples from the existing HydragardTM sampler. The first is a new hydrofluoric acid based method called the Cold Chemical Method (Cold Chem) and the second is a modified fusion method.Both new methods use the existing HydragardTM sampler to collect a smaller insert sample from the process sampling system. The insert testing methodology applies to the DWPF Slurry Mix Evaporator (SME) and the Melter Feed Tank (MFT) samples. Samples in small 3 ml containers (Inserts) are analyzed by either the cold chemical method or a modified fusion method. The current analytical method uses a HydragardTM sample station to obtain nearly full 15 ml peanut vials. The samples are prepared by a multi-step process for Inductively Coupled Plasma (ICP) analysis by drying, vitrification, grinding and finally dissolution by either mixed acid or fusion. In contrast, the insert sample is placed directly in the dissolution vessel, thus eliminating the drying, vitrification and grinding operations for the Cold chem method. Although the modified fusion still requires drying and calcine conversion, the process is rapid due to the decreased sample size and that no vitrification step is required.A slurry feed simulant material was acquired from the TNX pilot facility from the test run designated as PX-7.The Mock-up test data were gathered on the basis of a statistical design presented in SRT-SCS-97004 (Rev. 0). Simulant PX-7 samples were taken in the DWPF Analytical Cell Mock-up Facility using 3 ml inserts and 15 ml peanut vials. A number of the insert samples were analyzed by Cold Chem and compared with full peanut vial samples analyzed by the current methods. The remaining inserts were analyzed by

  6. Disposition of excess plutonium using ''off-spec'' MOX pellets as a sintered ceramic waste form

    International Nuclear Information System (INIS)

    Armantrout, G.A.; Jardine, L.J.

    1996-02-01

    The authors describe a potential strategy for the disposition of excess weapons plutonium in a way that minimizes (1) technological risks, (2) implementation costs and completion schedules, and (3) requirements for constructing and operating new or duplicative Pu disposition facilities. This is accomplished by an optimized combination of (1) using existing nuclear power reactors to ''burn'' relatively pure excess Pu inventories as mixed oxide (MOX) fuel and (2) using the same MOX fuel fabrication facilities to fabricate contaminated or impure excess Pu inventories into an ''off-spec'' MOX solid ceramic waste form for geologic disposition. Diversion protection for the SCWF to meet the ''spent fuel standard'' introduced by the National Academy of Sciences can be achieved in at least three ways. (1) One can utilize the radiation field from defense high-level nuclear waste by first packaging the SCWF pellets in 2- to 4-L cans that are subsequently encapsulated in radioactive glass in the Defense Waste Processing Facility (DWPF) glass canisters (a ''can-in-canister'' approach). (2) One can add 137 Cs (recovered from defense wastes at Hanford and currently stored as CsCl in capsules) to an encapsulating matrix such as cement for the SCWF pellets in a small hot-cell facility and thus fabricate large monolithic forms. (3) The SCWF can be fabricated into reactor fuel-like pellets and placed in tubes similar to fuel assemblies, which can then be mixed in sealed repository containers with irradiated spent nuclear fuel for geologic disposition

  7. Statistical analysis of the DWPF prototypic sampler

    International Nuclear Information System (INIS)

    Postles, R.L.; Reeve, C.P.; Jenkins, W.J.; Bickford, D.F.

    1991-01-01

    The DWPF process will be controlled using assay measurements on samples of feed slurry. These slurries are radioactive, and thus will be sampled remotely. A Hydraguard trademark pump-driven sampler system will be used as the remote sampling device. A prototype Hydraguard trademark sampler has been studied in a full-scale mock-up of a DWPF process vessel. Two issues were of dominant interest: (1) what accuracy and precision can be provided by such a pump-driven sampler in the face of the slurry rheology; and, if the Hydraguard trademark sample accurately represents the slurry in its local area, (2) is the slurry homogeneous enough throughout for it to represent the entire vessel? To determine Hydraguard trademark Accuracy, a Grab Sampler of simpler mechanism was used as reference. This (Low) Grab Sampler was located as near to the intake port of the Hydraguard trademark as could be arranged. To determine Homogeneity, a second (High) Grab Sampler was located above the first. The data necessary to these determinations comes from the measurement system, so its important variables also affect the results. Thus, the design of the test involved not just Sampling variables, but also some of the Measurement variables as well. However, the main concern was the Sampler and not the Measurement System, so the test design included only such measurement variables as could not be circumvented (Vials, Dissolution Method, and Aliquoting). The test was executed by, or under the direct oversight of, expert technologists. It thus did not explore the many important particulars of ''routine'' plant operations (such as Remote Sample Preparation or Laboratory Shift Operation)

  8. Material compatibility evaluataion for DWPF nitric-glycolic acid - literature review

    International Nuclear Information System (INIS)

    Mickalonis, J.I; Skidmore, T.E.

    2013-01-01

    Glycolic acid is being evaluated as an alternative for formic and nitric acid in the DWPF flowsheet. Demonstration testing and modeling for this new flowsheet has shown that glycolic acid and glycolate has a potential to remain in certain streams generated during the production of the nuclear waste glass. A literature review was conducted to assess the impact of glycolic acid on the corrosion of the materials of construction for the DWPF facility as well as facilities downstream which may have residual glycolic acid and glycolates present. The literature data was limited to solutions containing principally glycolic acid. The reported corrosion rates and degradation characteristics have shown the following for the materials of construction: For C276 alloy, the primary material of construction for the CPC vessels, corrosion rates of either 2 or 20 mpy were reported up to a temperature of 93 deg C; For the austenitic stainless steels, 304L and 316L, variable rates were reported over a range of temperatures, varying from 2 mpy up to 200 mpy (at 100 deg C); For 690, G30, Allcorr, Ultimet and Stellite alloys no data were available; and, For relevant polymers where data are available, the data suggests that exposure to glycolic acid is not detrimental. The literature data had limited application to the DWPF process since only the storage and feed vessels, pumps and piping used to handle the glycolic acid are directly covered by the available data. These components are either 304L or 316L alloys for which the literature data is inconsistent (See Bullet 2 above). Corrosion rates in pure glycolic acid solutions also are not representative of the DWPF process streams. This stream is complex and contains aggressive species, i.e. chlorides, sulfates, mercury, as well as antifoaming agents which cumulatively have an unknown effect on the corrosion rates of the materials of construction. Therefore, testing is recommended to investigate any synergistic effects of the aggressive

  9. Corrosion resistance of canisters for final disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Mattsson, E.

    1979-01-01

    A group of Swedish scientists has evaluated from the corrosion point of view three alternative canister types for final disposal of waste from nuclear reactors in boreholes in rock 500 m below ground. Titanium canisters with a wall-thickness of 6 mm and 100 mm thick lead lining have been estimated to have a life of at least thousands of years, and probably tens of thousands of years. Copper canisters with 200-mm-thick walls would last for hundreds of thousands of years. The third type, α-alumina sintered under isostatic pressure, is a very promising canister material

  10. Research on corrosion aspects of the advanced cold process canister

    International Nuclear Information System (INIS)

    Blackwood, D.J.; Hoch, A.R.; Naish, C.C.; Rance, A.

    1994-01-01

    The Advanced Cold Process Canister (ACPC) is a waste canister being developed jointly by SKB and TVO for the disposal of spent nuclear fuel. It comprises an outer copper canister, with a carbon steel canister inside. A concern regarding the use of the ACPC is that, in the unlikely event that the outer copper canister is penetrated, the anaerobic corrosion of the carbon steel container may result in the formation of hydrogen gas bubbles. These bubbles could disrupt the backfill, and thus increase water flow through the near field and the flux of radionuclides to the host geology. A number of factors that influence the rate at which hydrogen evolves as a result of the anaerobic corrosion of carbon steel in artificial granitic groundwaters have been investigated. A previously observed, time-dependent decline in the hydrogen evolution rate has been confirmed as being due to the production of magnetite film. Once the magnetite film is about 0.7-1.0 μm thick, the rate of hydrogen evolution reaches a steady state value. The pH and the ionic strength of the groundwater were both found to influence the long-term hydrogen evolution rate. The results of the experimental programme were used to update a model of the corrosion behaviour and hydrogen production from the Advanced Cold Process Canister. 36 figs, 5 tabs, 13 refs

  11. Physical properties of encapsulate spent fuel in canisters

    International Nuclear Information System (INIS)

    1999-01-01

    Spent fuel and high-level wastes will be permanently stored in a deep geological repository (AGP). Prior to this, they will be encapsulated in canisters. The present report is dedicated to the study of such canisters under the different physical demands that they may undergo, be those in operating or accident conditions. The physical demands of interest include mechanical demands, both static and dynamic, and thermal demands. Consideration is given to the complete file of the canister, from the time when it is empty and without lid to the final conditions expected in the repository. Thermal analyses of canisters containing spent fuel are often carried out in two dimensions, some times with hypotheses of axial symmetry and some times using a plane transverse section through the centre of the canister. The results obtained in both types of analyses are compared here to those of complete three-dimensional analyses. The latter generate more reliable information about the temperatures that may be experienced by the canister and its contents; they also allow calibrating the errors embodied in the two-dimensional calculations. (Author)

  12. Process arrangement options for Defense waste immobilization

    International Nuclear Information System (INIS)

    1980-02-01

    Current plans are to immobilize the SRP high-level liquid wastes in a high integrity form. Borosilicate glass was selected in 1977 as the reference waste form and a mjaor effort is currently underway to develop the required technology. A large new facility, referred to as the Defense Waste Processing Facility (DWPF) is being designed to carry out this mission, with project authorization targeted for 1982 and plant startup in 1989. However, a number of other process arrangements or manufacturing strategies, including staging the major elements of the project or using existing SRP facilities for some functions, have been suggested in lieu of building the reference DWPF. This study assesses these various options and compares them on a technical and cost basis with the DWPF. Eleven different manufacturing options for SRP defense waste solidification were examined in detail. These cases are: (1) vitrification of acid waste at current generation rate; (2) vitrification of current rate acid waste and caustic sludge; (3 and 4) vitrification of the sludge portion of neutralized waste; (5) decontamination of salt cake and storage of concentrated cesium and strontium for later immobilization; (6) processing waste in a facility with lower capacity than the DWPF; (7) processing waste in a combination of existing and new facilities; (8) waste immobilization in H Canyon; (9) vitrification of both sludge and salt; (10) DWPF with onsite storage; (11) deferred authorization of DWPF

  13. Grain boundary corrosion of copper canister weld material

    International Nuclear Information System (INIS)

    Gubner, Rolf; Andersson, Urban; Linder, Mats; Nazarov, Andrej; Taxen, Claes

    2006-01-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow, they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister, based on the recommendations of the report SKB-TR--01-09 (INIS ref. 32025363). Grain boundary corrosion of copper is not expected to be a problem for the copper canisters in a repository. However, as one step in the experimental verification it is necessary to study grain boundary corrosion of copper in an environment where it may occur. A literature study aimed to find one or several solutions that are aggressive with respect to grain boundary corrosion of copper. Copper specimens cut from welds of real copper canisters where exposed to aerated ammonium hydroxide solution for a period of 14 days at 80 degrees C and 10 bar pressure. The samples were investigated prior to exposure using the scanning Kelvin probe technique to characterize anodic and cathodic areas on the samples. The degree of corrosion was determined by optical microscopy. No grain boundary corrosion could be observed in the autoclave experiments, however, a higher rate of corrosion was observed for the weld material compared to the base material. The work suggests that grain boundary corrosion of copper weld material is most unlikely to adversely affect SKB's copper canisters under the conditions in the repository

  14. Grain boundary corrosion of copper canister weld material

    Energy Technology Data Exchange (ETDEWEB)

    Gubner, Rolf; Andersson, Urban; Linder, Mats; Nazarov, Andrej; Taxen, Claes [Corrosion and Metals Research Inst. (KIMAB), Stockholm (Sweden)

    2006-01-15

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow, they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister, based on the recommendations of the report SKB-TR--01-09 (INIS ref. 32025363). Grain boundary corrosion of copper is not expected to be a problem for the copper canisters in a repository. However, as one step in the experimental verification it is necessary to study grain boundary corrosion of copper in an environment where it may occur. A literature study aimed to find one or several solutions that are aggressive with respect to grain boundary corrosion of copper. Copper specimens cut from welds of real copper canisters where exposed to aerated ammonium hydroxide solution for a period of 14 days at 80 degrees C and 10 bar pressure. The samples were investigated prior to exposure using the scanning Kelvin probe technique to characterize anodic and cathodic areas on the samples. The degree of corrosion was determined by optical microscopy. No grain boundary corrosion could be observed in the autoclave experiments, however, a higher rate of corrosion was observed for the weld material compared to the base material. The work suggests that grain boundary corrosion of copper weld material is most unlikely to adversely affect SKB's copper canisters under the conditions in the repository.

  15. Final Report - Glass Formulation Development and Testing for DWPF High AI2O3 HLW Sludges, VSL-10R1670-1, Rev. 0, dated 12/20/10

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Pegg, I. L.; Kot, W. K.; Gan, H.; Matlack, K. S.

    2013-11-13

    -hydroxide methods. The above tests were proposed based on previous tests for WTP in which there were few differences in the melter processing characteristics, such as processing rate and melter emissions, between precipitated and direct hydroxide simulants, even though there were differences in rheological properties. To the extent this similarity is found also for simulants for SRS HLW, the direct hydroxide methods may offer the potential for faster, simpler, and cheaper simulant production. There was no plan to match the yield stress and particle size of the direct hydroxide simulant to that of the precipitated simulant because that would have increased the preparation cost and complexity and defeated the purpose of the tests. These objectives were addressed by first developing a series of glass frits and then conducting a crucible scale study to determine the waste loading achievable for the waste composition and to select the preferred frit. Waste loadings were increased until the limits of a glass property were exceeded experimentally. Glass properties for evaluation included: viscosity, electrical conductivity, crystallinity (including liquidus temperature and nepheline formation after canister centerline cooling (CCC) heat-treatment), gross glass phase separation, and the 7- day Product Consistency Test (PCT, ASTM-1285) response. Glass property limits were based upon the constraints used for DWPF process control.

  16. Production and remediation of low-sludge, simulated Purex waste glasses, 1: Effects of sludge oxide additions on melter operation

    International Nuclear Information System (INIS)

    Ramsey, W.G.

    1993-01-01

    Glass produced during the Purex 4 campaigns of the Integrated Defense Waste Processing Facility (DWPF) Melter System (IDMS) and the 774 Research Melter contained a lower fraction of sludge components than targeted by the Product Composition Control System (PCCS). Purex 4 glass was more durable than the benchmark (EA) glass, but less durable than most simulated SRS high-level waste glasses. Also, Purex 4 glass was considerably less durable than predicted by the algorithm which will be used to control production of DWPF glass. A melter run was performed using the 774 Research Melter to determine if the initial PCCS target composition determined for Purex 4 would produce acceptable glass whose durability could be accurately modeled by Hydration Thermodynamics. Reagent grade oxides and carbonates were added to Purex 4 melter feed stock to simulate a higher sludge loading. Each canister of glass produced was sampled and the composition, crystallinity, and durability was determined. This document details the melter operation and composition and crystallinity analyses

  17. Status of the Multipurpose Canister (MPC) Project

    International Nuclear Information System (INIS)

    Hopper, J.P.

    1996-01-01

    The multipurpose canister (MPC) project represents a cornerstone of the current U.S. Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) program for handling spent nuclear fuel. The MPC and associated support equipment is being designed to accommodate the requirements for not only storage and transport but also for the specified disposal requirements of the mined geologic repository system. The phase 1 design effort for the MPC system, being performed by the Westinghouse team on behalf of TRW Environmental Safety Systems (TESS), the OCRWM management ampersand operating (M ampersand O) contractor, is on schedule for delivery of completed safety analysis reports (SARs) in April 1996

  18. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Disbrow, J.

    1994-01-01

    This paper discusses detailed data on canisters and nonfuel components (NFC) at US commercial nuclear power reactors. A wide variety of NFC have been reported on the Form RW-859, open-quotes Nuclear Fuel Dataclose quotes survey. They may have been integral with an assembly, noncanistered in baskets, destined for disposal as low-level radioactive waste, or stored in canisters. Similarly, data on the family of canistered spent nuclear fuel (SNF) in storage pools was compiled. Approximately 85 percent of the 40,194 pieces of nonfuel assembly (NFA) hardware reported were integral with an assembly. This represents data submitted by 95 of the 107 reactors in 10 generic assembly classes. In addition, a total of 286 canisters have been reported as being in storage pools as of December 31, 1992. However, an additional 264 open baskets were also reported to contain miscellaneous SNF and nonfuel materials, garbage and debris. All of these 286 canisters meet the dimensional envelope requirements specified for disposal for open-quotes standard fuelclose quotes under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR 961); most of the baskets do not

  19. Defense waste processing facility at Savannah River Plant. Instrument and power jumpers

    International Nuclear Information System (INIS)

    Heckendorm, F.M. II.

    1983-06-01

    The Defense Waste Processing Facility (DWPF) for waste vitrification at the Savannah River Plant is in the final design stage. Development of equipment interconnecting devices or jumpers for use within the remotely operated processing canyon is now complete. These devices provide for the specialized instrument and electrical requirements of the DWPF process for low-voltage, high-frequency, and high-power interconnections

  20. OCRWM Bulletin: Westinghouse begins designing multi-purpose canister

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-09-01

    This publication consists of two parts: OCRWM (Office of Civilian Radioactive Waste Management) Bulletin; and Of Mountains & Science which has articles on the Yucca Mountain project. The OCRWM provides information about OCRWM activities and in this issue has articles on multi-purpose canister design, and transportation cask trailer.

  1. OCRWM Bulletin: Westinghouse begins designing multi-purpose canister

    International Nuclear Information System (INIS)

    1995-01-01

    This publication consists of two parts: OCRWM (Office of Civilian Radioactive Waste Management) Bulletin; and Of Mountains ampersand Science which has articles on the Yucca Mountain project. The OCRWM provides information about OCRWM activities and in this issue has articles on multi-purpose canister design, and transportation cask trailer

  2. Transport of multiassembly sealed canisters

    International Nuclear Information System (INIS)

    Quinn, R.D.; Lehnert, R.A.; Rosa, J.M.

    1992-01-01

    A significant portion of the commercial spent nuclear fuel in dry storage in the US will be stored in multiassembly sealed canisters before the DOE begins accepting fuel from utilities in 1998. This paper reports that it is desirable from economic and ALARA perspectives to transfer these canisters directly from the plant to the MRS. To this end, it is necessary that the multiassembly sealed canisters, which have been licensed for storage under 10CFR72, be qualified for shipment within a suitable shipping cask under the rules of 10CFR71. Preliminary work performed to date indicates that it is feasible to license a current canister design for transportation, and work is proceeding on obtaining NRC approval

  3. Actual Waste Demonstration of the Nitric-Glycolic Flowsheet for Sludge Batch 9 Qualification

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-01

    For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs qualification testing to demonstrate that the sludge batch is processable. Testing performed by the Savannah River National Laboratory has shown glycolic acid to be effective in replacing the function of formic acid in the DWPF chemical process. The nitric-glycolic flowsheet reduces mercury, significantly lowers the catalytic generation of hydrogen and ammonia which could allow purge reduction in the Sludge Receipt and Adjustment Tank (SRAT), stabilizes the pH and chemistry in the SRAT and the Slurry Mix Evaporator (SME), allows for effective rheology adjustment, and is favorable with respect to melter flammability. In order to implement the new flowsheet, SRAT and SME cycles, designated SC-18, were performed using a Sludge Batch (SB) 9 slurry blended from SB8 Tank 40H and Tank 51H samples. The SRAT cycle involved adding nitric and glycolic acids to the sludge, refluxing to steam strip mercury, and dewatering to a targeted solids concentration. Data collected during the SRAT cycle included offgas analyses, process temperatures, heat transfer, and pH measurements. The SME cycle demonstrated the addition of glass frit and the replication of six canister decontamination additions. The demonstration concluded with dewatering to a targeted solids concentration. Data collected during the SME cycle included offgas analyses, process temperatures, heat transfer, and pH measurements. Slurry and condensate samples were collected for subsequent analysis

  4. Analysis Of DWPF Sludge Batch 7A (Macrobatch 8) Pour Stream Samples

    International Nuclear Information System (INIS)

    Johnson, F.

    2012-01-01

    The Defense Waste Processing Facility (DWPF) began processing Sludge Batch 7a (SB7a), also referred to as Macrobatch 8 (MB8), in June 2011. SB7a is a blend of the heel of Tank 40 from Sludge Batch 6 (SB6) and the SB7a material that was transferred to Tank 40 from Tank 51. SB7a was processed using Frit 418. During processing of each sludge batch, the DWPF is required to take at least one glass sample to meet the objectives of the Glass Product Control Program (GPCP), which is governed by the DWPF Waste Compliance Plan, and to complete the necessary Production Records so that the final glass product may be disposed of at a Federal Repository. Three pour stream glass samples and two Melter Feed Tank (MFT) slurry samples were collected while processing SB7a. These additional samples were taken during SB7a to understand the impact of antifoam and the melter bubblers on glass redox chemistry. The samples were transferred to the Savannah River National Laboratory (SRNL) where they were analyzed. The following conclusions were drawn from the analytical results provided in this report: (1) The sum of oxides for the official SB7a pour stream glass is within the Product Composition Control System (PCCS) limits (95-105 wt%). (2) The average calculated Waste Dilution Factor (WDF) for SB7a is 2.3. In general, the measured radionuclide content of the official SB7a pour stream glass is in good agreement with the calculated values from the Tank 40 dried sludge results from the SB7a Waste Acceptance Program Specification (WAPS) sample. (3) As in previous pour stream samples, ruthenium and rhodium inclusions were detected by Scanning Electron Microscopy-Electron Dispersive Spectroscopy (SEM-EDS) in the official SB7a pour stream sample. (4) The Product Consistency Test (PCT) results indicate that the official SB7a pour stream glass meets the waste acceptance criteria for durability with a normalized boron release of 0.64 g/L, which is an order of magnitude less than the Environmental

  5. Corrosion resistance of metal materials for HLW canister

    International Nuclear Information System (INIS)

    Furuya, Takashi; Muraoka, Susumu; Tashiro, Shingo

    1982-02-01

    In order to verify the materials as an important artificial barrier for canister of vitrified high-level waste from spent fuel reprocessing, data and reports were researched on corrosion resistance of the materials under conditions from glass form production to final disposal. Then, in this report, investigated subjects, improvement methods and future subjects are reviewed. It has become clear that there would be no problem on the inside and outside corrosion of the canister during glass production, but long term corrosion and radiation effect tests and the vitrification methods would be subjects in future on interim storage and final disposal conditions. (author)

  6. Mechanical Integrity of Canisters Using a Fracture Mechanics Approach

    Energy Technology Data Exchange (ETDEWEB)

    Koyama, Tomofumi; Guoxiang Zhang; Lanru Jing [Royal Inst. of Technology, Stockholm (Sweden). Dept. of Land and Water Resources Engineering

    2006-07-15

    This report presents the methods and results of a research project about numerical modeling of mechanical integrity of cast-iron canisters for the final disposal of spent nuclear fuel in Sweden, using combined boundary element (BEM) and finite element (FEM) methods. The objectives of the project are: 1) to investigate the possibility of initiation and growth of fractures in the cast-iron canisters under the mechanical loading conditions defined in the premises of canister design by Swedish Nuclear Fuel and Waste Management Co. (SKB); 2) to investigate the maximum bearing capacity of the cast iron canisters under uniformly distributed and gradually increasing boundary pressure until plastic failure. Achievement of the two objectives may provide some quantitative evidence for the mechanical integrity and overall safety of the cast-iron canisters that are needed for the final safety assessment of the geological repository of the radioactive waste repository in Sweden. The geometrical dimension, distribution and magnitudes of loads and Material properties of the canisters and possible fractures were provided by the latest investigations of SKB. The results of the BEM simulations, using the commercial code BEASY, indicate that under the currently defined loading conditions the possibility of initiation of new fractures or growth of existing fractures (defects) are very small, due to the reasons that: 1) the canisters are under mainly compressive stresses; 2) the induced tensile stress regions are too small in both dimension and magnitude to create new fractures or to induce growth of existing fractures, besides the fact that the toughness of the fractures in the cast iron canisters are much higher that the stress intensity factors in the fracture tips. The results of the FEM simulation show a approximately 75 MPa maximum pressure beyond which plastic collapse of the cast-iron canisters may occur, using an elastoplastic Material model. This figure is smaller compared

  7. COMSOL Multiphysics Model for HLW Canister Filling

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-04-11

    The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. Wastes containing high concentrations of Al2O3 and Na2O can contribute to nepheline (generally NaAlSiO4) crystallization, which can sharply reduce the chemical durability of high level waste (HLW) glass. Nepheline crystallization can occur during slow cooling of the glass within the stainless steel canister. The purpose of this work was to develop a model that can be used to predict temperatures of the glass in a WTP HLW canister during filling and cooling. The intent of the model is to support scoping work in the laboratory. It is not intended to provide precise predictions of temperature profiles, but rather to provide a simplified representation of glass cooling profiles within a full scale, WTP HLW canister under various glass pouring rates. These data will be used to support laboratory studies for an improved understanding of the mechanisms of nepheline crystallization. The model was created using COMSOL Multiphysics, a commercially available software. The model results were compared to available experimental data, TRR-PLT-080, and were found to yield sufficient results for the scoping nature of the study. The simulated temperatures were within 60 ºC for the centerline, 0.0762m (3 inch) from centerline, and 0.2286m (9 inch) from centerline thermocouples once the thermocouples were covered with glass. The temperature difference between the experimental and simulated values reduced to 40 ºC, 4 hours after the thermocouple was covered, and down to 20 ºC, 6 hours after the thermocouple was covered

  8. Proposal of a SiC disposal canister for very deep borehole disposal

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Heui-Joo; Lee, Minsoo; Lee, Jong-Youl; Kim, Kyungsu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    In this paper authors proposed a silicon carbide, SiC, disposal canister for the DBD concept in Korea. A. Kerber et al. first proposed the SiC canister for a geological disposal of HLW, CANDU or HTR spent nuclear fuels. SiC has some drawbacks in welding or manufacturing a large canister. Thus, we designed a double layered disposal canister consisting of a stainless steel outer layer and a SiC inner layer. KAERI has been interested in developing a very deep borehole disposal (DBD) of HLW generated from pyroprocessing of PWR spent nuclear fuel and supported the relevant R and D with very limited its own budget. KAERI team reviewed the DBD concept proposed by Sandia National Laboratories (SNL) and developed its own concept. The SNL concept was based on the steel disposal canister. The authors developed a new technology called cold spray coating method to manufacture a copper-cast iron disposal canister for a geological disposal of high level waste in Korea. With this method, 8 mm thin copper canister with 400 mm in diameter and 1200 mm in height was made. In general, they do not give any credit on the lifetime of a disposal canister in DBD concept unlike the geological disposal. In such case, the expensive copper canister should be replaced with another one. We designed a disposal canister using SiC for DBD. According to an experience in manufacturing a small size canister, the fabrication of a large-size one is a challenge. Also, welding of SiC canister is not easy. Several pathways are being paved to overcome it.

  9. Development of cold sprayed Cu coating for canister

    International Nuclear Information System (INIS)

    Kim, Hyung Jun; Kang, Yoon Ha

    2010-01-01

    Cold sprayed Cu deposition was studied for the application of outer part of canister for high level nuclear waste. Five commercially available pure Cu powders were analyzed and sprayed by high pressure cold spray system. Electrochemical corrosion test using potentiostat in 3.5% NaCl solution was conducted as well as microstructural analysis including hardness and oxygen content measurements. Overall evaluation of corrosion performance of cold sprayed Cu deposition is inferior to forged and extruded Cu plates, but some of Cu depositions are comparable to Cu plates. The simulated corrosion test in 200m underground cave is still in progress. The effect of cold spray process parameters was also studied and the results show that the type of nozzle is the most important other than powder feed rate, spray distance, and scan speed. 1/10 scale miniature of canister was manufactured confirming that the production of full scale canister is possible

  10. Test plan for the Sample Transfer Canister system

    International Nuclear Information System (INIS)

    Flanagan, B.D.

    1998-01-01

    The Sample Transfer Canister will be used by the Waste Receiving and Processing Facility (WRAP) for the transport of small quantity liquid samples that meet the definition of a limited quantity radioactive material, and may also be corrosive and/or flammable. These samples will be packaged and shipped in accordance with the US Department of Transportation (DOT) regulation 49 CFR 173.4, ''Exceptions for small quantities.'' The Sample Transfer Canister is of a ''French Can'' design, intended to be mated with a glove box for loading/unloading. Transport will typically take place north of the Wye Barricade between WRAP and the 222-S Laboratory. The Sample Transfer Canister will be shipped in an insulated ice chest, but the ice chest will not be a part of the small quantity package during prototype testing

  11. IMPACT OF IRRADIATION AND THERMAL AGING ON DWPF SIMULATED SLUDGE PROPERTIES

    International Nuclear Information System (INIS)

    Eibling, R; Michael Stone, M

    2006-01-01

    The research and development programs in support of the Defense Waste Processing Facility (DWPF) and other high-level waste vitrification processes require the use of both nonradioactive waste simulants and actual waste samples. While actual waste samples are the ideal materials to study, acquiring large quantities of actual waste is difficult and expensive. Tests utilizing actual high-level waste require the use of expensive shielded cells facilities to provide sufficient shielding for the researchers. Nonradioactive waste simulants have been used for laboratory testing, pilot-scale testing and full-scale integrated facility testing. These waste simulants were designed to reproduce the chemical and, if possible, the physical properties of the actual high-level waste. This technical report documents a study on the impact of irradiating a Sludge Batch 3 (SB3) simulant and of additional tests on aging a SB3 simulant by additional thermal processing. Prior simulant development studies examined methods of producing sludge and supernate simulants and processes that could be used to alter the physical properties of the simulant to more accurately mimic the properties of actual waste. Development of a precipitated sludge simulant for the River Protection Project (RPP) demonstrated that the application of heat for a period of time could significantly alter the rheology of the sludge simulant. The RPP precipitated simulant used distillation to concentrate the sludge solids and produced a reduction in sludge yield stress of up to 80% compared to the initial sludge properties. Observations at that time suggested that a substantial fraction of the iron hydroxide had converted to the oxide during the distillation. DWPF sludge simulant studies showed a much smaller reduction in yield stress (∼10%), demonstrated the impact of shear on particle size, and showed that smaller particle sizes yielded higher yield stress products. The current study documented in this report focuses

  12. Calibration and Measurement of the Viscosity of DWPF Start-Up Glass

    International Nuclear Information System (INIS)

    Schumacher, R.F.

    2001-01-01

    The Harrop, High-Temperature Viscometer has been in operation at the Savannah River Technology Center (SRTC) for several years and has proven itself to be reasonably accurate and repeatable. This is particularly notable when taking into consideration the small amount of glass required to make the viscosity determination. The volume of glass required is only 2.60 cc or about 6 to 7 grams of glass depending on the glass density. This may be compared to the more traditional viscosity determinations, which generally require between 100 to 1000 grams of glass. Before starting the present investigation, the unit was re-aligned and the furnace thermal gradients measured. The viscometer was again calibrated with available NIST Standard Reference Material glasses (717a and 710a) and a spindle constant equation was determined. Standard DWPF Waste Compliance Glasses (Purex, HM, and Batch 1) were used to provide additional verification for the determinations at low temperature. The Harrop, High-Temperature Viscometer was then used to determine the viscosity of three random samples of ground and blended DWPF, Black, Start -Up Frit, which were obtained from Pacific Northwest National Laboratory (PNNL). The glasses were in powder form and required melting prior to the viscosity determination. The results from this evaluation will be compared to ''Round Robin'' measurements from other DOE laboratories and a number of commercial laboratories

  13. DWPF Melter Off-Gas Flammability Assessment for Sludge Batch 9

    Energy Technology Data Exchange (ETDEWEB)

    Choi, A. S. [Savannah River Site (SRS), Aiken, SC (United States)

    2016-07-11

    The slurry feed to the Defense Waste Processing Facility (DWPF) melter contains several organic carbon species that decompose in the cold cap and produce flammable gases that could accumulate in the off-gas system and create potential flammability hazard. To mitigate such a hazard, DWPF has implemented a strategy to impose the Technical Safety Requirement (TSR) limits on all key operating variables affecting off-gas flammability and operate the melter within those limits using both hardwired/software interlocks and administrative controls. The operating variables that are currently being controlled include; (1) total organic carbon (TOC), (2) air purges for combustion and dilution, (3) melter vapor space temperature, and (4) feed rate. The safety basis limits for these operating variables are determined using two computer models, 4-stage cold cap and Melter Off-Gas (MOG) dynamics models, under the baseline upset scenario - a surge in off-gas flow due to the inherent cold cap instabilities in the slurry-fed melter.

  14. Preliminary Transportation, Aging and Disposal Canister System Performance Specification

    International Nuclear Information System (INIS)

    C.A Kouts

    2006-01-01

    This document provides specifications for selected system components of the Transportation, Aging and Disposal (TAD) canister-based system. A list of system specified components and ancillary components are included in Section 1.2. The TAD canister, in conjunction with specialized overpacks will accomplish a number of functions in the management and disposal of spent nuclear fuel. Some of these functions will be accomplished at purchaser sites where commercial spent nuclear fuel (CSNF) is stored, and some will be performed within the Office of Civilian Radioactive Waste Management (OCRWM) transportation and disposal system. This document contains only those requirements unique to applications within Department of Energy's (DOE's) system. DOE recognizes that TAD canisters may have to perform similar functions at purchaser sites. Requirements to meet reactor functions, such as on-site dry storage, handling, and loading for transportation, are expected to be similar to commercially available canister-based systems. This document is intended to be referenced in the license application for the Monitored Geologic Repository (MGR). As such, the requirements cited herein are needed for TAD system use in OCRWM's disposal system. This document contains specifications for the TAD canister, transportation overpack and aging overpack. The remaining components and equipment that are unique to the OCRWM system or for similar purchaser applications will be supplied by others

  15. Sludge Washing and Demonstration of the DWPF Nitric/Formic Flowsheet in the SRNL Shielded Cells for Sludge Batch 9 Qualification

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-11-01

    Savannah River National Laboratory (SRNL) was requested by Savannah River Remediation (SRR) to qualify the next batch of sludge – Sludge Batch 9 (SB9). Current practice is to prepare sludge batches in Tank 51 by transferring sludge to Tank 51 from other tanks. The sludge is washed and transferred to Tank 40, the current Defense Waste Process Facility (DWPF) feed tank. Prior to sludge transfer from Tank 51 to Tank 40, the Tank 51 sludge must be qualified. SRNL qualifies the sludge in multiple steps. First, a Tank 51 sample is received, then characterized, washed, and again characterized. SRNL then demonstrates the DWPF Chemical Process Cell (CPC) flowsheet with the sludge. The final step of qualification involves chemical durability measurements of glass fabricated in the DWPF CPC demonstrations. In past sludge batches, SRNL had completed the DWPF demonstration with Tank 51 sludge. For SB9, SRNL has been requested to process a blend of Tank 51 and Tank 40 at a targeted ratio of 44% Tank 51 and 56% Tank 40 on an insoluble solids basis.

  16. DWPF Flowsheet Studies with Simulants to Determine Modular Caustic Side Solvent Extraction Unit Solvent Partitioning and Verify Actinide Removal Process Incorporation Strategy

    International Nuclear Information System (INIS)

    Herman, C

    2006-01-01

    The Actinide Removal Process (ARP) facility and the Modular Caustic Side Solvent Extraction Unit (MCU) are scheduled to begin processing salt waste in fiscal year 2007. A portion of the streams generated in the salt processing facilities will be transferred to the Defense Waste Processing Facility (DWPF) to be incorporated in the glass matrix. Before the streams are introduced, a combination of impact analyses and research and development studies must be performed to quantify the impacts on DWPF processing. The Process Science and Engineering (PS and E) section of the Savannah River National Laboratory (SRNL) was requested via Technical Task Request (TTR) HLW/DWPF/TTR-2004-0031 to evaluate the impacts on DWPF processing. Simulant Chemical Process Cell (CPC) flowsheet studies have been performed using previous composition and projected volume estimates for the ARP sludge/monosodium titanate (MST) stream. Due to changes in the flammability control strategy for DWPF for salt processing, the incorporation strategy for ARP has changed and additional ARP flowsheet tests were necessary to validate the new processing strategy. The last round of ARP testing included the incorporation of the MCU stream and identified potential processing issues with the MCU solvent. The identified issues included the potential carry-over and accumulation of the MCU solvent components in the CPC condensers and in the recycle stream to the Tank Farm. Therefore, DWPF requested SRNL to perform additional MCU flowsheet studies to better quantify the organic distribution in the CPC vessels. The previous MCU testing used a Sludge Batch 4 (SB4) simulant since it was anticipated that both of these facilities would begin salt processing during SB4 processing. The same sludge simulant recipe was used in this round of ARP and MCU testing to minimize the number of changes between the two phases of testing so a better comparison could be made. ARP and MCU stream simulants were made for this phase of

  17. DWPF PCCS version 2.0 test case

    International Nuclear Information System (INIS)

    Brown, K.G.; Pickett, M.A.

    1992-01-01

    To verify the operation of the Product Composition Control System (PCCS), a test case specific to DWPF operation was developed. The values and parameters necessary to demonstrate proper DWPF product composition control have been determined and are presented in this paper. If this control information (i.e., for transfers and analyses) is entered into the PCCS as illustrated in this paper, and the results obtained correspond to the independently-generated results, it can safely be said that the PCCS is operating correctly and can thus be used to control the DWPF. The independent results for this test case will be generated and enumerated in a future report. This test case was constructed along the lines of the normal DWPF operation. Many essential parameters are internal to the PCCS (e.g., property constraint and variance information) and can only be manipulated by personnel knowledgeable of the Symbolics reg-sign hardware and software. The validity of these parameters will rely on induction from observed PCCS results. Key process control values are entered into the PCCS as they would during normal operation. Examples of the screens used to input specific process control information are provided. These inputs should be entered into the PCCS database, and the results generated should be checked against the independent, computed results to confirm the validity of the PCCS

  18. Review of information on the radiation chemistry of materials around waste canisters in salt and assessment of the need for additional experimental information

    Energy Technology Data Exchange (ETDEWEB)

    Jenks, G.H.; Baes, C.F. Jr.

    1980-03-01

    The brines, vapors, and salts precipitated from the brines will be exposed to gamma rays and to elevated temperatures in the regions close to a waste package in the salt. Accordingly, they will be subject to changes in composition brought about by reactions induced by the radiations and heat. This report reviews the status of information on the radiation chemistry of brines, gases, and solids which might be present around a waste package in salt and to assess the need for additional laboratory investigations on the radiation chemistry of these materials. The basic aspects of the radiation chemistry of water and aqueous solutions, including concentrated salt solutions, were reviewed briefly and found to be substantially unchanged from those presented in Jenks's 1972 review of radiolysis and hydrolysis in salt-mine brines. Some additional information pertaining to the radiolytic yields and reactions in brine solutions has become available since the previous review, and this information will be useful in the eventual, complete elucidation of the radiation chemistry of the salt-mine brines. 53 references.

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

    International Nuclear Information System (INIS)

    Cornman, W.R.

    1981-12-01

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

  20. Determination of Reportable Radionuclides for DWPF Sludge Batch 2 (Macro Batch 3)

    International Nuclear Information System (INIS)

    Bibler, N.E.

    2002-01-01

    The Waste Acceptance Product Specifications (WAPS) 1.2 require that ''The Producer shall report the inventory of radionuclides (in Curies) that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115''. As part of the strategy to meet WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type, all radionuclides (with half-lives greater than 10 years) that have concentrations greater than 0.01 percent of the total inventory from time of production through the 1100 year period from 2015 through 3115. The initial listing of radionuclides to be included is based on the design-basis glass as identified in the Waste Form Compliance Plan (WCP) and Waste Form Qualification Report (WQR). However, it is required that this list be expanded if other radionuclides with half-lives greater than 10 years are identified that meet the greater than 0.01 percent criterion for Curie content

  1. Mechanical Integrity of Copper Canister Lid and Cylinder. Sensitivity study

    International Nuclear Information System (INIS)

    Karlsson, Marianne

    2002-08-01

    This report is part of a study of the mechanical integrity of canisters used for disposal of nuclear fuel waste. The overall objective is to determine and ensure the static and long-term strength of the copper canister lid and cylinder casing. The canisters used for disposal nuclear fuel waste of type BWR consists of an inner part (insert) of ductile cast iron and an outer part of copper. The copper canister is to provide a sealed barrier between the contents of the canister and the surroundings. The study in this report complements the finite element analyses performed in an earlier study. The analyses aim to evaluate the sensitivity of the canister to tolerances regarding the gap between the copper cylinder and the cast iron insert. Since great uncertainties regarding the material's long term creep properties prevail, analyses are also performed to evaluate the effect of different creep data on the resulting strain and stress state. The report analyses the mechanical response of the lid and flange of the copper canister when subjected to loads caused by pressure from swelling bentonite and from groundwater at a depth of 500 meter. The loads acting on the canister are somewhat uncertain and the cases investigated in this report are possible cases. Load cases analysed are: Pressure 15 MPa uniformly distributed on lid and 5 MPa uniformly distributed on cylinder; Pressure 5 MPa uniformly distributed on lid and 15 MPa uniformly distributed on cylinder; Pressure 20 MPa uniformly distributed on lid and cylinder; and Side pressures 10 MPa and 20 MPa uniformly distributed on part of the cylinder. Creep analyses are performed for two of the load cases. For all considered designs high principal stresses appear on the outside of the copper cylinder in the region from the weld down to the level of the lid lower edge. Altering the gap between lid and cylinder and/or between cylinder and insert only marginally affects the resulting stress state. Fitting the lid in the cylinder

  2. Stakeholder involvement in the evaluation of a multipurpose canister system

    International Nuclear Information System (INIS)

    Williams, J.R.; Kane, D.; Smith, T.B. Jr.

    1994-01-01

    The U.S. Department of Energy (DOE), Office of Civilian Radioactive Waste Management (OCRWM), began evaluating a multipurpose canister (MPC) concept in October of 1992. This followed recommendations by the Nuclear Waste Technical Review Board (NWTRB) and the U.S. Nuclear Regulatory Commission (NRC) that DOE develop a nuclear waste management system that achieves system integration, standardization, and reduced fuel-handling operations. Industry organizations such as Edison Electric Institute (EEI) and Electric Power Research Institute (EPRI) had conducted earlier studies that concluded advantages to the nuclear waste management system may be offered by such a concept. The MPC concept involves a metal canister which would contain multiple spent nuclear fuel assemblies. The canister would be sealed at the nuclear power plant and would not be reopened. The MPC would then be placed inside separate casks or overpacks for storage, transportation, and disposal. An important factor in DOE's evaluation of the MPC concept was the involvement of external parties. This paper describes that involvement process for the OCRWM's MPC implementation program. External parties who have an interest or stake in the program are referred to as stakeholders

  3. DWPF nitric-glycolic flowsheet chemical process cell chemistry. Part 1

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-02-01

    The conversions of nitrite to nitrate, the destruction of glycolate, and the conversion of glycolate to formate and oxalate were modeled for the Nitric-Glycolic flowsheet using data from Chemical Process Cell (CPC) simulant runs conducted by SRNL from 2011 to 2015. The goal of this work was to develop empirical correlations for these variables versus measureable variables from the chemical process so that these quantities could be predicted a-priori from the sludge composition and measurable processing variables. The need for these predictions arises from the need to predict the REDuction/OXidation (REDOX) state of the glass from the Defense Waste Processing Facility (DWPF) melter. This report summarizes the initial work on these correlations based on the aforementioned data. Further refinement of the models as additional data is collected is recommended.

  4. Summary of Preliminary Criticality Analysis for Peach Bottom Fuel in the DOE Standardized Spent Nuclear Fuel Canister

    International Nuclear Information System (INIS)

    Henrikson, D.J.

    1999-01-01

    The Department of Energy's (DOE's) National Spent Nuclear Fuel Program is developing a standardized set of canisters for DOE spent nuclear fuel (SNF). These canisters will be used for DOE SNF handling, interim storage, transportation, and disposal in the national repository. Several fuels are being examined in conjunction with the DOE SNF canisters. This report summarizes the preliminary criticality safety analysis that addresses general fissile loading limits for Peach Bottom graphite fuel in the DOE SNF canister. The canister is considered both alone and inside the 5-HLW/DOE Long Spent Fuel Co-disposal Waste Package, and in intact and degraded conditions. Results are appropriate for a single DOE SNF canister. Specific facilities, equipment, canister internal structures, and scenarios for handling, storage, and transportation have not yet been defined and are not evaluated in this analysis. The analysis assumes that the DOE SNF canister is designed so that it maintains reasonable geometric integrity. Parameters important to the results are the canister outer diameter, inner diameter, and wall thickness. These parameters are assumed to have nominal dimensions of 45.7-cm (18.0-in.), 43.815-cm (17.25-in), and 0.953-cm (0.375-in.), respectively. Based on the analysis results, the recommended fissile loading for the DOE SNF canister is 13 Peach Bottom fuel elements if no internal steel is present, and 15 Peach Bottom fuel elements if credit is taken for internal steel

  5. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    This paper covers the design, development and testing of the magazines (cylinders containing cans of plutonium-ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed robotic arm that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102mm) diameter canister throat

  6. Test design requirements: Canister-scale heater test

    International Nuclear Information System (INIS)

    Schauer, M.I.; Craig, P.A.; Stickney, R.G.

    1986-03-01

    This document establishes the Test Design Requirements for the design of a canister scale heater test to be performed in the Exploratory Shaft test facility. The purpose of the test is to obtain thermomechanical rock mass response data for use in validation of the numerical models. The canister scale heater test is a full scale simulation of a high-level nuclear waste container in a prototypic emplacement borehole. Electric heaters are used to simulate the heat loads expected in an actual waste container. This document presents an overview of the test including objectives and justification for the test. A description of the test as it is presently envisioned is included. Discussions on Quality Assurance and Safety are also included in the document. 12 refs., 1 fig

  7. Evaluation of a molybdenum assay canister

    International Nuclear Information System (INIS)

    Yoshizumi, T.T.; Keener, S.J.

    1988-01-01

    The performance characteristics of a commercial molybdenum assay canister were evaluated. The geometrical variation of the technetium-99m (/sup 99m/Tc) activity reading was studied as a function of the elution volume for the standard vials. It was found that the /sup 99m/Tc canister activity reading was ∼ 5% lower than that of the standard method. This is due to attenuation by the canister wall. However, the effect of the geometric variation on the clinical dose preparation was found to be insignificant. The molybdenum-99 ( 99 Mo) contamination level was compared by two methods: (1) the commercial canister and (2) the standard assay kit. The 99 Mo contamination measurements with the canister indicated consistently lower readings than those with the standard 99 Mo assay kit. The authors conclude that the canister may be used in the clinical settings. However, the user must be aware of the problems and the limitations associated with this canister

  8. Paper Study Evaluations Of The Introduction Of Small Column Ion Exchange Waste Streams To The Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Fox, K.; Edwards, T.; Stone, M.; Koopman, D.

    2010-01-01

    The objective of this paper study is to provide guidance on the impact of Monosodium Titanate (MST) and Crystalline Silicotitanate (CST) streams from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) flowsheet and glass waste form. A series of waste processing scenarios was evaluated, including projected compositions of Sludge Batches 8 through 17 (SB8 through SB17), MST additions, CST additions to Tank 40 or to a sludge batch preparation tank (Tank 42 or Tank 51, referred to generically as Tank 51 in this report), streams from the Salt Waste Processing Facility (SWPF), and two canister production rates. A wide array of potential glass frit compositions was used to support this assessment. The sludge and frit combinations were evaluated using the predictive models in the current DWPF Product Composition Control System (PCCS). The results were evaluated based on the number of frit compositions available for a particular sludge composition scenario. A large number of candidate frit compositions (e.g., several dozen to several hundred) is typically a good indicator of a sludge composition for which there is flexibility in forming an acceptable waste glass and meeting canister production rate commitments. The MST and CST streams will significantly increase the concentrations of certain components in glass, such as Nb 2 O 5 , TiO 2 , and ZrO 2 , to levels much higher than have been previously processed at DWPF. Therefore, several important assumptions, described in detail in the report, had to be made in performing the evaluations. The results of the paper studies, which must be applied carefully given the assumptions made concerning the impact of higher Ti, Zr, and Nb concentrations on model validity, provided several observations: (1) There was difficulty in identifying a reasonable number of candidate frits (and in some cases an inability to identify any candidate frits) when a waste loading of 40% is targeted for Sludge

  9. Chemical durability of copper canisters under crystalline bedrock repository conditions

    International Nuclear Information System (INIS)

    Sjoeblom, R.; Hermansson, H.P.; Amcoff, Oe.

    1995-01-01

    In the Swedish waste management programme, the copper canister is expected to provide containment of the radionuclides for a very long time, perhaps million of years. The purpose of the present paper is to analyze prerequisites for assessments of corrosion lifetimes for copper canisters. The analysis is based on compilations of literature from the following areas: chemical literature on copper and copper corrosion, mineralogical literature with emphasis on the stability of copper in near surface environments, and chemical and mineralogical literature with emphasis on the stabilities and thermodynamics of species and phases that may exist in a repository environment. Three main types of situations are identified: (1) under oxidizing and low chloride conditions, passivating oxide type of layers may form on the copper surface; (2) under oxidizing and high chloride conditions, the species formed may all be dissolved; and (3) under reducing conditions, non-passivating sulfide type layers may form on the copper surface. Considerable variability and uncertainty exists regarding the chemical environment for the canister, especially in certain scenarios. Thus, the mechanisms for corrosion can be expected to differ greatly for different situations. The lifetime of a thick-walled copper canister subjected to general corrosion appears to be long for most reasonable chemistries. Localized corrosion may appear for types (1) and (3) above but the mechanisms are widely different in character. The penetration caused by localized corrosion can be expected to be very sensitive to details in the chemistry. 20 refs, 3 figs, 1 tab

  10. Simulation of heat transfer around a canister placed horizontally in a drift

    International Nuclear Information System (INIS)

    Moujaes, S.; Bhargava, A.

    1994-01-01

    The Yucca Mountain Site Characterization Project is investigating the feasibility of locating a high level radioactive nuclear waste repository at Yucca Mountain, Nevada. The bore hole and the in-drift waste emplacement schemes are under evaluation as potential repository drift geometries. This paper presents a two-dimensional finite element thermal analysis of the nuclear waste canister placed horizontally in a drift. Simulation has been carried out for 1000 years and the peak temperatures at the walls of the drift and at the center of the canister have been determined. The effect of the three modes of heat transfer, conduction, natural convection and radiation, is also discussed

  11. Liquid level measurement in high level nuclear waste slurries

    International Nuclear Information System (INIS)

    Weeks, G.E.; Heckendorn, F.M.; Postles, R.L.

    1990-01-01

    Accurate liquid level measurement has been a difficult problem to solve for the Defense Waste Processing Facility (DWPF). The nuclear waste sludge tends to plug or degrade most commercially available liquid-level measurement sensors. A liquid-level measurement system that meets demanding accuracy requirements for the DWPF has been developed. The system uses a pneumatic 1:1 pressure repeater as a sensor and a computerized error correction system. 2 figs

  12. The defense waste processing facility: A status report

    International Nuclear Information System (INIS)

    Cowan, S.P.; Fulmer, D.C.

    1987-01-01

    The Defense Waste Processing Fascility (DWPF) will be the nation's first production scale facility for immobilizing high-level waste for disposal. It will also be the largest facility of its kind in the world. The technology, design, and construction efforts are on schedule for ''hot'' operation in fiscal year 1990. This paper provides a status report on the DWPF technology, design, and construction, and describes some of the challenges that have arisen during design and construction

  13. Environmental information document defense waste processing facility

    International Nuclear Information System (INIS)

    1981-07-01

    This report documents the impact analysis of a proposed Defense Waste Processing Facility (DWPF) for immobilizing high-level waste currently being stored on an interim basis at the Savannah River Plant (SRP). The DWPF will process the waste into a form suitable for shipment to and disposal in a federal repository. The DWPF will convert the high-level waste into: a leach-resistant form containing above 99.9% of all the radioactivity, and a residue of slightly contaminated salt. The document describes the SRP site and environs, including population, land and water uses; surface and subsurface soils and waters; meteorology; and ecology. A conceptual integrated facility for concurrently producing glass waste and saltcrete is described, and the environmental effects of constructing and operating the facility are presented. Alternative sites and waste disposal options are addressed. Also environmental consultations and permits are discussed

  14. MAR Assessments Of The High Level Waste System Plan Revision 16

    International Nuclear Information System (INIS)

    Peeler, D.; Edwards, T.

    2011-01-01

    High-level waste (HLW) throughput (i.e., the amount of waste processed per unit of time) is primarily a function of two critical parameters: waste loading (WL) and melt rate. For the Defense Waste Processing Facility (DWPF), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). Significant increases in waste throughput have been achieved at DWPF since initial radioactive operations began in 1996. Key technical and operational initiatives that supported increased waste throughput included improvements in facility attainment, the Chemical Processing Cell (CPC) flowsheet, process control models and frit formulations. As a result of these key initiatives, DWPF increased WLs from a nominal 28% for Sludge Batch 2 (SB2) to ∼34 to 38% for SB3 through SB6 while maintaining or slightly improving canister fill times. Although considerable improvements in waste throughput have been obtained, future contractual waste loading targets are nominally 40%, while canister production rates are also expected to increase (to a rate of 325 to 400 canisters per year). Although implementation of bubblers have made a positive impact on increasing melt rate for recent sludge batches targeting WLs in the mid30s, higher WLs will ultimately make the feeds to DWPF more challenging to process. Savannah River Remediation (SRR) recently requested the Savannah River National Laboratory (SRNL) to perform a paper study assessment using future sludge projections to evaluate whether the current Process Composition Control System (PCCS) algorithms would provide projected operating windows to allow future contractual WL targets to be met. More specifically, the objective of this study was to evaluate future sludge batch projections (based on Revision 16 of the HLW Systems Plan) with respect to projected operating windows using current PCCS models and associated constraints. Based on the assessments, the waste loading interval over

  15. Preliminary Evaluation Of DWPF Impacts Of Boric Acid Use In Cesium Strip FOR SWPF And MCU

    International Nuclear Information System (INIS)

    Stone, M.

    2010-01-01

    A new solvent system is being evaluated for use in the Modular Caustic-Side Solvent Extraction Unit (MCU) and in the Salt Waste Processing Facility (SWPF). The new system includes the option to replace the current dilute nitric acid strip solution with boric acid. To support this effort, the impact of using 0.01M, 0.1M, 0.25M and 0.5M boric acid in place of 0.001M nitric acid was evaluated for impacts on the DWPF facility. The evaluation only covered the impacts of boric acid in the strip effluent and does not address the other changes in solvents (i.e., the new extractant, called MaxCalix, or the new suppressor, guanidine). Boric acid additions may lead to increased hydrogen generation during the SRAT and SME cycles as well as change the rheological properties of the feed. The boron in the strip effluent will impact glass composition and could require each SME batch to be trimmed with boric acid to account for any changes in the boron from strip effluent additions. Addition of boron with the strip effluent will require changes in the frit composition and could lead to changes in melt behavior. The severity of the impacts from the boric acid additions is dependent on the amount of boric acid added by the strip effluent. The use of 0.1M or higher concentrations of boric acid in the strip effluent was found to significantly impact DWPF operations while the impact of 0.01M boric acid is expected to be relatively minor. Experimental testing is required to resolve the issues identified during the preliminary evaluation. The issues to be addressed by the testing are: (1) Impact on SRAT acid addition and hydrogen generation; (2) Impact on melter feed rheology; (3) Impact on glass composition control; (4) Impact on frit production; and (5) Impact on melter offgas. A new solvent system is being evaluated for use in the Modular Caustic-Side Solvent Extraction Unit (MCU) and in the Salt Waste Processing Facility (SWPF). The new system includes the option to replace the

  16. DWPF Glass Melter Technology Manual: Volume 4

    International Nuclear Information System (INIS)

    Iverson, D.C.

    1993-01-01

    This document details information about the design of a glass melter to be used at the Defense Waste Processing Facility located at the Savannah River Plant. Information contained in this document consists solely of a machine drawing and parts list and purchase orders with specifications of equipment used in the development of the melter

  17. DWPF Glass Melter Technology Manual: Volume 4

    Energy Technology Data Exchange (ETDEWEB)

    Iverson, D.C.

    1993-12-31

    This document details information about the design of a glass melter to be used at the Defense Waste Processing Facility located at the Savannah River Plant. Information contained in this document consists solely of a machine drawing and parts list and purchase orders with specifications of equipment used in the development of the melter.

  18. DWPF Glass Melter Technology Manual: Volume 3

    Energy Technology Data Exchange (ETDEWEB)

    Iverson, D.C.

    1993-12-31

    This document details information about the design of a glass melter to be used at the Defense Waste Processing Facility located at the Savannah River Site. Topics discussed include: Information collected during testing, equipment, materials, design basis, feed tubes, and an evaluation of the performance of various components. Information is conveyed using many diagrams and photographs.

  19. DWPF Glass Melter Technology Manual: Volume 3

    International Nuclear Information System (INIS)

    Iverson, D.C.

    1993-01-01

    This document details information about the design of a glass melter to be used at the Defense Waste Processing Facility located at the Savannah River Site. Topics discussed include: Information collected during testing, equipment, materials, design basis, feed tubes, and an evaluation of the performance of various components. Information is conveyed using many diagrams and photographs

  20. IMPACTS OF SMALL COLUMN ION EXCHANGE STREAMS ON DWPF GLASS FORMULATION: KT01, KT02, KT03, AND KT04-SERIES GLASS COMPOSITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Fox, K.; Edwards, T.

    2010-11-01

    Four series of glass compositions were selected, fabricated, and characterized as part of a study to determine the impacts of the addition of Crystalline Silicotitanate (CST) and Monosodium Titanate (MST) from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) glass waste form and the applicability of the DWPF process control models. The KT01 and KT02-series of glasses were chosen to allow for the identification of the influence of the concentrations of major components of the glass on the retention of TiO{sub 2}. The KT03 series of glasses was chosen to allow for the identification of these influences when higher Nb{sub 2}O{sub 5} and ZrO{sub 2} concentrations are included along with TiO2. The KT04 series of glasses was chosen to investigate the properties and performance of glasses based on the best available projections of actual compositions to be processed at the DWPF (i.e., future sludge batches including the SCIX streams).

  1. FEASIBILITY EVALUATION AND RETROFIT PLAN FOR COLD CRUCIBLE INDUCTION MELTER DEPLOYMENT IN THE DEFENSE WASTE PROCESSING FACILITY AT SAVANNAH RIVER SITE 8118

    International Nuclear Information System (INIS)

    Barnes, A; Dan Iverson, D; Brannen Adkins, B

    2008-01-01

    increase heat transfer to the slurry fed High Level Waste (HLW) sludge, the CCIM may be equipped with bubblers and/or water cooled mechanical agitators. The DWPF could benefit from use of CCIM technology, especially in light of our latest projections of waste volume to be vitrified. Increased waste loading and increased throughput could result in substantial life cycle cost reduction. In order to significantly surpass the waste throughput capability of the currently installed JHM, it may be necessary to install two 950 mm CCIMs in the DWPF Melt Cell. A cursory evaluation of system design requirements and modifications to the facility that may be required to support installation and operation of two 950 mm CCIMs was performed. Based on this evaluation, it appears technically feasible to position two CCIMs in the Melt Cell of the DWPF within the existing footprint of the current melter. Interfaces with support systems and controls including Melter Feed, Power, Melter Cooling Water, Melter Off-gas, and Canister Operations must be designed to support dual CCIM operations. This paper describes the CCIM technology and identifies technical challenges that must be addressed in order to implement CCIMs in the DWPF

  2. Preliminary technical data summary defense waste processing facility stage 2

    International Nuclear Information System (INIS)

    1980-12-01

    This Preliminary Technical Data Summary presents the technical basis for design of Stage 2 of the Staged Defense Waste Processing Facility (DWPF). Process changes incorporated in the staged DWPF relative to the Alternative DWPF described in PTDS No. 3 (DPSTD-77-13-3) are the result of ongoing research and development and are aimed at reducing initial capital investment and developing a process to efficiently immobilize the radionuclides in Savannah River Plant (SRP) high-level liquid waste. The radionuclides in SRP waste are present in sludge that has settled to the bottom of waste storage tanks and in crystallized salt and salt solution (supernate). Stage 1 of the DWPF receives washed, aluminum dissolved sludge from the waste tank farms and immobilizes it in a borosilicate glass matrix. The supernate is retained in the waste tank farms until completion of Stage 2 of the DWPF at which time it is filtered and decontaminated by ion exchange in the Stage 2 facility. The decontaminated supernate is concentrated by evaporation and mixed with cement for burial. The radioactivity removed from the supernate is fixed in borosilicate glass along with the sludge. This document gives flowsheets, material and curie balances, material and curie balance bases, and other technical data for design of Stage 2 of the DWPF. Stage 1 technical data are presented in DPSTD-80-38

  3. Spent fuel canister docking station

    International Nuclear Information System (INIS)

    Suikki, M.

    2006-01-01

    The working report for the spent fuel canister docking station presents a design for the operation and structure of the docking equipment located in the fuel handling cell for the spent fuel in the encapsulation plant. The report contains a description of the basic requirements for the docking station equipment and their implementation, the operation of the equipment, maintenance and a cost estimate. In the designing of the equipment all the problems related with the operation have been solved at the level of principle, nevertheless, detailed designing and the selection of final components have not yet been carried out. In case of defects and failures, solutions have been considered for postulated problems, and furthermore, the entire equipment was gone through by the means of systematic risk analysis (PFMEA). During the docking station designing we came across with needs to influence the structure of the actual disposal canister for spent nuclear fuel, too. Proposed changes for the structure of the steel lid fastening screw were included in the report. The report also contains a description of installation with the fuel handling cell structures. The purpose of the docking station for the fuel handling cell is to position and to seal the disposal canister for spent nuclear fuel into a penetration located on the cell floor and to provide suitable means for executing the loading of the disposal canister and the changing of atmosphere. The designed docking station consists of a docking ring, a covering hatch, a protective cone and an atmosphere-changing cap as well as the vacuum technology pertaining to the changing of atmosphere and the inert gas system. As far as the solutions are concerned, we have arrived at rather simple structures and most of the actuators of the system are situated outside of the actual fuel handling cell. When necessary, the equipment can also be used for the dismantling of a faulty disposal canister, cut from its upper end by machining. The

  4. DETERMINATION OF REPORTABLE RADIONUCLIDES FOR DWPF SLUDGE BATCH 4 MACROBATCH 5

    International Nuclear Information System (INIS)

    Bannochie, C; Ned Bibler, N; David Diprete, D

    2008-01-01

    The Waste Acceptance Product Specifications (WAPS)1 1.2 require that 'The Producer shall report the inventory of radionuclides (in Curies) that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115'. As part of the strategy to meet WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type, all radionuclides (with half-lives greater than 10 years) that have concentrations greater than 0.01 percent of the total inventory from time of production through the 1100 year period from 2015 through 3115. The initial listing of radionuclides to be included is based on the design-basis glass as identified in the Waste Form Compliance Plan (WCP)2 and Waste Form Qualification Report (WQR)3. However, it is required that this list be expanded if other radionuclides with half-lives greater than 10 years are identified that may meet the greater than 0.01% criterion for Curie content. Specification 1.6 of the WAPS, International Atomic Energy Agency (IAEA) Safeguards Reporting for High Level Waste (HLW), requires that the ratio by weights of the following uranium and plutonium isotopes be reported: U-233, U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. Therefore, the complete set of reportable radionuclides must also include this set of U and Pu isotopes. The DWPF is receiving radioactive sludge slurry from HLW Tank 40. The radioactive sludge slurry in Tank 40 is a blend of the previous contents of Tank 40 (Sludge Batch 3) and the sludge that was transferred to Tank 40 from Tank 51. The blend of sludge from Tank 51 and Tank 40 defines Sludge Batch 4 (also referred to as Macrobatch 5 (MB5)). This report develops the list of reportable radionuclides and associated activities and determines the radionuclide activities as a function of time. The DWPF will use this list and the activities as one of

  5. Control of DWPF melter feed composition

    International Nuclear Information System (INIS)

    Brown, K.G.; Edwards, R.E.; Postles, R.L.; Randall, C.T.

    1989-01-01

    The Defense Waste Processing Facility will be used to immobilize Savannah River Site high-level waste into a stable borosilicate glass for disposal in a geologic repository. Proper control of the melter feed composition in this facility is essential to the production of glass which meets product durability constraints dictated by repository regulations and facility processing constraints dictated by melter design. A technique has been developed which utilizes glass property models to determine acceptable processing regions based on the multiple constraints imposed on the glass product and to display these regions graphically. This system along with the batch simulation of the process is being used to form the basis for the statistical process control system for the facility

  6. DWPF Glass Melter Technology Manual: Volume 1

    International Nuclear Information System (INIS)

    Iverson, D.C.

    1993-01-01

    This document details information about the design of a glass melter to be used at the Defense Waste Processing Facility located at the Savannah River Site. Topics include: melter overview, design basis, materials, vessel configuration, insulation, refractory configuration, electrical isolation, electrodes, riser and pour spout heater design, dome heaters, feed tubes, drain valves, differential pressure pouring, and melter test results. Information is conveyed using many diagrams and photographs

  7. Organics Characterization Of DWPF Alternative Reductant Simulants, Glycolic Acid, And Antifoam 747

    Energy Technology Data Exchange (ETDEWEB)

    White, T. L. [Savannah River Site (SRS), Aiken, SC (United States); Wiedenman, B. J. [Savannah River Site (SRS), Aiken, SC (United States); Lambert, D. P. [Savannah River Site (SRS), Aiken, SC (United States); Crump, S. L. [Savannah River Site (SRS), Aiken, SC (United States); Fondeur, F. F. [Savannah River Site (SRS), Aiken, SC (United States); Papathanassiu, A. E. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States); Kot, W. K. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States); Pegg, I. L. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States)

    2013-10-01

    The present study examines the fate of glycolic acid and other organics added in the Chemical Processing Cell (CPC) of the Defense Waste Processing Facility (DWPF) as part of the glycolic alternate flowsheet. Adoption of this flowsheet is expected to provide certain benefits in terms of a reduction in the processing time, a decrease in hydrogen generation, simplification of chemical storage and handling issues, and an improvement in the processing characteristics of the waste stream including an increase in the amount of nitrate allowed in the CPC process. Understanding the fate of organics in this flowsheet is imperative because tank farm waste processed in the CPC is eventually immobilized by vitrification; thus, the type and amount of organics present in the melter feed may affect optimal melt processing and the quality of the final glass product as well as alter flammability calculations on the DWPF melter off gas. To evaluate the fate of the organic compounds added as the part of the glycolic flowsheet, mainly glycolic acid and antifoam 747, samples of simulated waste that was processed using the DWPF CPC protocol for tank farm sludge feed were generated and analyzed for organic compounds using a variety of analytical techniques at the Savannah River National Laboratory (SRNL). These techniques included Ion Chromatography (IC), Gas Chromatography-Mass Spectrometry (GC-MS), Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES), and Nuclear Magnetic Resonance (NMR) Spectroscopy. A set of samples were also sent to the Catholic University of America Vitreous State Laboratory (VSL) for analysis by NMR Spectroscopy at the University of Maryland, College Park. Analytical methods developed and executed at SRNL collectively showed that glycolic acid was the most prevalent organic compound in the supernatants of Slurry Mix Evaporator (SME) products examined. Furthermore, the studies suggested that commercially available glycolic acid contained minor amounts

  8. Organics Characterization Of DWPF Alternative Reductant Simulants, Glycolic Acid, And Antifoam 747

    International Nuclear Information System (INIS)

    White, T. L.; Wiedenman, B. J.; Lambert, D. P.; Crump, S. L.; Fondeur, F. F.; Papathanassiu, A. E.; Kot, W. K.; Pegg, I. L.

    2013-01-01

    The present study examines the fate of glycolic acid and other organics added in the Chemical Processing Cell (CPC) of the Defense Waste Processing Facility (DWPF) as part of the glycolic alternate flowsheet. Adoption of this flowsheet is expected to provide certain benefits in terms of a reduction in the processing time, a decrease in hydrogen generation, simplification of chemical storage and handling issues, and an improvement in the processing characteristics of the waste stream including an increase in the amount of nitrate allowed in the CPC process. Understanding the fate of organics in this flowsheet is imperative because tank farm waste processed in the CPC is eventually immobilized by vitrification; thus, the type and amount of organics present in the melter feed may affect optimal melt processing and the quality of the final glass product as well as alter flammability calculations on the DWPF melter off gas. To evaluate the fate of the organic compounds added as the part of the glycolic flowsheet, mainly glycolic acid and antifoam 747, samples of simulated waste that was processed using the DWPF CPC protocol for tank farm sludge feed were generated and analyzed for organic compounds using a variety of analytical techniques at the Savannah River National Laboratory (SRNL). These techniques included Ion Chromatography (IC), Gas Chromatography-Mass Spectrometry (GC-MS), Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES), and Nuclear Magnetic Resonance (NMR) Spectroscopy. A set of samples were also sent to the Catholic University of America Vitreous State Laboratory (VSL) for analysis by NMR Spectroscopy at the University of Maryland, College Park. Analytical methods developed and executed at SRNL collectively showed that glycolic acid was the most prevalent organic compound in the supernatants of Slurry Mix Evaporator (SME) products examined. Furthermore, the studies suggested that commercially available glycolic acid contained minor amounts

  9. Integration of SWPF into the DWPF Flowsheet: Gap Analysis and Test Matrix Development

    Energy Technology Data Exchange (ETDEWEB)

    Peeler, D. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2014-12-10

    Based on Revision 19 of the High Level Waste (HLW) System Plan, it is anticipated that the Salt Waste Processing Facility (SWPF) will be integrated into the Defense Waste Processing Facility (DWPF) flowsheet in October 2018 (or with Sludge Batch 11 (SB11)). Given that, Savannah River Remediation (SRR) has requested a technical basis be developed that validates the current Product Composition Control System (PCCS) models for use during the processing of the SWPF-based coupled flowsheet or that leads to the refinements of or modifications to the models that are needed so that the models may be used during the processing of the SWPF-based coupled flowsheet. To support this objective, Savannah River National Laboratory (SRNL) has completed three key interim activities prior to validation of the current or development of refined PCCS models over the anticipated glass composition region for SWPF processing. These three key activities include: (1) defining the glass compositional region over which SWPF is anticipated to be processed, (2) comparing the current PCCS model validation ranges to the SWPF glass compositional region from which compositional gaps can be identified, and (3) developing a test matrix to cover the compositional gaps.

  10. Design premises for canister for spent nuclear fuel

    International Nuclear Information System (INIS)

    Werme, L.

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel

  11. Multi-purpose canisters as an alternative for storage, transportation, and disposal of spent nuclear fuel

    International Nuclear Information System (INIS)

    Hollaway, W.R.; Rozier, R.; Nitti, D.A.; Williams, J.R.

    1993-01-01

    A study was conducted to assess the feasibility of using multi-purpose canisters to handle spent nuclear fuel throughout the Civilian Radioactive Waste Management System. Multi-purpose canisters would be sealed, metallic containers maintaining multiple spent fuel assemblies in a dry, inert environment and overpacked separately and uniquely for the various system elements of storage, transportation, and disposal. Using five implementation scenarios, the multi-purpose canister was evaluated with regard to several measures of effectiveness, including number of handlings, radiation exposure, cost, schedule and licensing considerations, and public perception. Advantages and disadvantages of the multi-purpose canister were identified relative to the current reference system within each scenario, and the scenarios were compared to determine the most effective method of implementation

  12. Multiple-canister flow and transport code in 2-dimensional space. MCFT2D: user's manual

    International Nuclear Information System (INIS)

    Lim, Doo-Hyun

    2006-03-01

    A two-dimensional numerical code, MCFT2D (Multiple-Canister Flow and Transport code in 2-Dimensional space), has been developed for groundwater flow and radionuclide transport analyses in a water-saturated high-level radioactive waste (HLW) repository with multiple canisters. A multiple-canister configuration and a non-uniform flow field of the host rock are incorporated in the MCFT2D code. Effects of heterogeneous flow field of the host rock on migration of nuclides can be investigated using MCFT2D. The MCFT2D enables to take into account the various degrees of the dependency of canister configuration for nuclide migration in a water-saturated HLW repository, while the dependency was assumed to be either independent or perfectly dependent in previous studies. This report presents features of the MCFT2D code, numerical simulation using MCFT2D code, and graphical representation of the numerical results. (author)

  13. Commercial Submersible Mixing Pump For SRS Tank Waste Removal - 15223

    International Nuclear Information System (INIS)

    Hubbard, Mike; Herbert, James E.; Scheele, Patrick W.

    2015-01-01

    The Savannah River Site Tank Farms have 45 active underground waste tanks used to store and process nuclear waste materials. There are 4 different tank types, ranging in capacity from 2839 m 3 to 4921 m 3 (750,000 to 1,300,000 gallons). Eighteen of the tanks are older style and do not meet all current federal standards for secondary containment. The older style tanks are the initial focus of waste removal efforts for tank closure and are referred to as closure tanks. Of the original 51 underground waste tanks, six of the original 24 older style tanks have completed waste removal and are filled with grout. The insoluble waste fraction that resides within most waste tanks at SRS requires vigorous agitation to suspend the solids within the waste liquid in order to transfer this material for eventual processing into glass filled canisters at the Defense Waste Processing Facility (DWPF). SRS suspends the solid waste by use of recirculating mixing pumps. Older style tanks generally have limited riser openings which will not support larger mixing pumps, since the riser access is typically 58.4 cm (23 inches) in diameter. Agitation for these tanks has been provided by four long shafted standard slurry pumps (SLP) powered by an above tank 112KW (150 HP) electric motor. The pump shaft is lubricated and cooled in a pressurized water column that is sealed from the surrounding waste in the tank. Closure of four waste tanks has been accomplished utilizing long shafted pump technology combined with heel removal using multiple technologies. Newer style waste tanks at SRS have larger riser openings, allowing the processing of waste solids to be accomplished with four large diameter SLPs equipped with 224KW (300 HP) motors. These tanks are used to process the waste from closure tanks for DWPF. In addition to the SLPs, a 224KW (300 HP) submersible mixer pump (SMP) has also been developed and deployed within older style tanks. The SMPs are product cooled and product lubricated canned

  14. Commercial Submersible Mixing Pump For SRS Tank Waste Removal - 15223

    Energy Technology Data Exchange (ETDEWEB)

    Hubbard, Mike [Savannah River Remediation, LLC., Aiken, SC (United States); Herbert, James E. [Savannah River Remediation, LLC., Aiken, SC (United States); Scheele, Patrick W. [Savannah River Remediation, LLC., Aiken, SC (United States)

    2015-01-12

    The Savannah River Site Tank Farms have 45 active underground waste tanks used to store and process nuclear waste materials. There are 4 different tank types, ranging in capacity from 2839 m3 to 4921 m3 (750,000 to 1,300,000 gallons). Eighteen of the tanks are older style and do not meet all current federal standards for secondary containment. The older style tanks are the initial focus of waste removal efforts for tank closure and are referred to as closure tanks. Of the original 51 underground waste tanks, six of the original 24 older style tanks have completed waste removal and are filled with grout. The insoluble waste fraction that resides within most waste tanks at SRS requires vigorous agitation to suspend the solids within the waste liquid in order to transfer this material for eventual processing into glass filled canisters at the Defense Waste Processing Facility (DWPF). SRS suspends the solid waste by use of recirculating mixing pumps. Older style tanks generally have limited riser openings which will not support larger mixing pumps, since the riser access is typically 58.4 cm (23 inches) in diameter. Agitation for these tanks has been provided by four long shafted standard slurry pumps (SLP) powered by an above tank 112KW (150 HP) electric motor. The pump shaft is lubricated and cooled in a pressurized water column that is sealed from the surrounding waste in the tank. Closure of four waste tanks has been accomplished utilizing long shafted pump technology combined with heel removal using multiple technologies. Newer style waste tanks at SRS have larger riser openings, allowing the processing of waste solids to be accomplished with four large diameter SLPs equipped with 224KW (300 HP) motors. These tanks are used to process the waste from closure tanks for DWPF. In addition to the SLPs, a 224KW (300 HP) submersible mixer pump (SMP) has also been developed and deployed within older style tanks. The SMPs are product cooled and

  15. Multi-purpose canister system evaluation: A systems engineering approach

    International Nuclear Information System (INIS)

    1994-09-01

    This report summarizes Department of Energy (DOE) efforts to investigate various container systems for handling, transporting, storing, and disposing of spent nuclear fuel (SNF) assemblies in the Civilian Radioactive Waste Management System (CRWMS). The primary goal of DOE's investigations was to select a container technology that could handle the vast majority of commercial SNF at a reasonable cost, while ensuring the safety of the public and protecting the environment. Several alternative cask and canister concepts were evaluated for SNF assembly packaging to determine the most suitable concept. Of these alternatives, the multi-purpose canister (MPC) system was determined to be the most suitable. Based on the results of these evaluations, the decision was made to proceed with design and certification of the MPC system. A decision to fabricate and deploy MPCs will be made after further studies and preparation of an environmental impact statement

  16. Theoretical predictions for glass flow into an evacuated canister

    International Nuclear Information System (INIS)

    Routt, K.R.; Crow, K.R.

    1983-01-01

    Radioactive waste currently stored at the Savannah River Plant in liquid form is to be immobilized by incorporating it into a borosilicate glass. The glass melter for this process will consist of a refractory lined, steel vessel operated at a glass temperature of 1150 0 C. At the end of a two-year projected melter lifetime, the glass inside the melter is to be drained prior to disposition of the melter vessel. One proposed technique for accomplishing this drainage is by sucking the glass into an evacuated canister. The theoretical bases for design of an evacuated canister for draining a glass melter have been developed and tested. The theoretical equations governing transient and steady-state flow were substantiated with both a silicone glass simulant and molten glass

  17. Canisters for spent-fuel disposal: Design measures against localized corrosion

    International Nuclear Information System (INIS)

    Werme, L.O.; Oversby, V.M.

    2000-01-01

    Common to all high-level-waste disposal concepts is the encapsulation of the waste into metal canisters. The purpose of this waste canister is to isolate the radioactive waste from contact with its surroundings for a desired time period. The design service life ranges from hundreds to thousands of years depending on the disposal concept. After the isolation has been breached, other barriers in the disposal system will delay and attenuate the radioactive releases to acceptable levels. In a deep geologic repository, the waste package will be exposed to chemical attack and, depending on the type of repository, to mechanical stresses. Each of these factors will by itself or in combination inevitably lead to loss of confinement some time in the future. In the design of the Swedish waste canister, the corrosion resistance is provided by an outer shell of pure copper while an insert supplies the mechanical strength cast nodular iron. The close fit between the insert and the copper results in very small tensile stresses in the copper over very limited areas once the repository has been saturated. Measurements of stress corrosion crack growth show that annealed copper cannot maintain sufficiently high stress intensity factors for cracks to grow. For annealed copper, the stress intensity factor was limited to 25 MPa·m 1/2 because of extensive plastic deformation. For cold-worked copper, no crack growth could be observed for stress intensity factors 1/2 . Through the choices of canister material, canister, and repository design, and considering the expected chemical conditions, the risks for localized corrosion can be lowered to an acceptable level, if not eliminated altogether, and the releases from prematurely failed canisters can be kept well within acceptable dose levels

  18. Errors of DWPF frit analysis: Final report

    International Nuclear Information System (INIS)

    Schumacher, R.F.

    1993-01-01

    Glass frit will be a major raw material for the operation of the Defense Waste Processing Facility. The frit will be controlled by certificate of conformance and a confirmatory analysis from a commercial analytical laboratory. The following effort provides additional quantitative information on the variability of frit chemical analyses at two commercial laboratories. Identical samples of IDMS Frit 202 were chemically analyzed at two commercial laboratories and at three different times over a period of four months. The SRL-ADS analyses, after correction with the reference standard and normalization, provided confirmatory information, but did not detect the low silica level in one of the frit samples. A methodology utilizing elliptical limits for confirming the certificate of conformance or confirmatory analysis was introduced and recommended for use when the analysis values are close but not within the specification limits. It was also suggested that the lithia specification limits might be reduced as long as CELS is used to confirm the analysis

  19. Maximum total organic carbon limit for DWPF melter feed

    International Nuclear Information System (INIS)

    Choi, A.S.

    1995-01-01

    DWPF recently decided to control the potential flammability of melter off-gas by limiting the total carbon content in the melter feed and maintaining adequate conditions for combustion in the melter plenum. With this new strategy, all the LFL analyzers and associated interlocks and alarms were removed from both the primary and backup melter off-gas systems. Subsequently, D. Iverson of DWPF- T ampersand E requested that SRTC determine the maximum allowable total organic carbon (TOC) content in the melter feed which can be implemented as part of the Process Requirements for melter feed preparation (PR-S04). The maximum TOC limit thus determined in this study was about 24,000 ppm on an aqueous slurry basis. At the TOC levels below this, the peak concentration of combustible components in the quenched off-gas will not exceed 60 percent of the LFL during off-gas surges of magnitudes up to three times nominal, provided that the melter plenum temperature and the air purge rate to the BUFC are monitored and controlled above 650 degrees C and 220 lb/hr, respectively. Appropriate interlocks should discontinue the feeding when one or both of these conditions are not met. Both the magnitude and duration of an off-gas surge have a major impact on the maximum TOC limit, since they directly affect the melter plenum temperature and combustion. Although the data obtained during recent DWPF melter startup tests showed that the peak magnitude of a surge can be greater than three times nominal, the observed duration was considerably shorter, on the order of several seconds. The long surge duration assumed in this study has a greater impact on the plenum temperature than the peak magnitude, thus making the maximum TOC estimate conservative. Two models were used to make the necessary calculations to determine the TOC limit

  20. DWPF integrated cold runs revised technical bases for precipitate hydrolysis

    International Nuclear Information System (INIS)

    Landon, L.F.

    1992-01-01

    The report defines new precipitate hydrolysis process operating parameters for DWPF Chemical runs assuming the precipitate feed simulants to be processed reflect the decision to implement a final wash of the tetraphenylborate slurry before transfer to DWPF (i.e. the Late Wash Facility). Control of the nitrite content of the tetraphenylborate slurry to 0.01M or less has eliminated the need for hydroxylamine nitrate (HAN) during hydrolysis. Consequently, the oxidant nitrous oxide will not be generated. However, nitric oxide (NO) is expected to be generated (reaction of formic acid with nitrite) and some fraction of the NO can be expected to be oxidized to nitrogen dioxide. The rate of NO generation with low nitrite feed has not been quantified at this time nor is the extent to which the NO is oxidized to NO 2 known. A mass spectrometer is being installed in the Precipitate Hydrolysis Experimental Facility (PHEF) which will enable the NO generation rate to be defined as well as the extent to which the NO is oxidized to NO 2 . There is some undocumented data available for C 6 H 6 /NO and C 6 H 6 /NO 2 with N 2 as the diluent but no similar data for CO 2 . Development of test data in the required time frame is not possible. However, MOC's will be estimated for benzene/NO/NO 2 /CO 2 gas mixtures (the MOC is expected to be approximately 60% less than for the HAN process). Once these data are obtained, and NO/NO 2 concentration profiles are obtained from PHEF hydrolysis process demonstrations, a flammability control strategy for the DWPF Salt Processing Cell will be developed. Implementation of the HAN process purge strategy upon startup of the SPC with the late wash process would be conservative

  1. DWPF Melter No.2 Prototype Bus Bar Test Report

    International Nuclear Information System (INIS)

    Gordon, J.

    2003-01-01

    Characterization and performance testing of a prototype DWPF Melter No.2 Dome Heater Bus Bar are described. The prototype bus bar was designed to address the design features of the existing system which may have contributed to water leaks on Melter No.1. Performance testing of the prototype revealed significant improvement over the existing design in reduction of both bus bar and heater connection maximum temperature, while characterization revealed a few minor design and manufacturing flaws in the bar. The prototype is recommended as an improvement over the existing design. Recommendations are also made in the area of quality control to ensure that critical design requirements are met

  2. Inspection of disposal canisters components

    International Nuclear Information System (INIS)

    Pitkaenen, J.

    2013-12-01

    This report presents the inspection techniques of disposal canister components. Manufacturing methods and a description of the defects related to different manufacturing methods are described briefly. The defect types form a basis for the design of non-destructive testing because the defect types, which occur in the inspected components, affect to choice of inspection methods. The canister components are to nodular cast iron insert, steel lid, lid screw, metal gasket, copper tube with integrated or separate bottom, and copper lid. The inspection of copper material is challenging due to the anisotropic properties of the material and local changes in the grain size of the copper material. The cast iron insert has some acoustical material property variation (attenuation, velocity changes, scattering properties), which make the ultrasonic inspection demanding from calibration point of view. Mainly three different methods are used for inspection. Ultrasonic testing technique is used for inspection of volume, eddy current technique, for copper components only, and visual testing technique are used for inspection of the surface and near surface area

  3. BEHAVIOR OF MERCURY DURING DWPF CHEMICAL PROCESS CELL PROCESSING

    Energy Technology Data Exchange (ETDEWEB)

    Zamecnik, J.; Koopman, D.

    2012-04-09

    The Defense Waste Processing Facility has experienced significant issues with the stripping and recovery of mercury in the Chemical Processing Cell (CPC). The stripping rate has been inconsistent, often resulting in extended processing times to remove mercury to the required endpoint concentration. The recovery of mercury in the Mercury Water Wash Tank has never been high, and has decreased significantly since the Mercury Water Wash Tank was replaced after the seventh batch of Sludge Batch 5. Since this time, essentially no recovery of mercury has been seen. Pertinent literature was reviewed, previous lab-scale data on mercury stripping and recovery was examined, and new lab-scale CPC Sludge Receipt and Adjustment Tank (SRAT) runs were conducted. For previous lab-scale data, many of the runs with sufficient mercury recovery data were examined to determine what factors affect the stripping and recovery of mercury and to improve closure of the mercury material balance. Ten new lab-scale SRAT runs (HG runs) were performed to examine the effects of acid stoichiometry, sludge solids concentration, antifoam concentration, form of mercury added to simulant, presence of a SRAT heel, operation of the SRAT condenser at higher than prototypic temperature, varying noble metals from none to very high concentrations, and higher agitation rate. Data from simulant runs from SB6, SB7a, glycolic/formic, and the HG tests showed that a significant amount of Hg metal was found on the vessel bottom at the end of tests. Material balance closure improved from 12-71% to 48-93% when this segregated Hg was considered. The amount of Hg segregated as elemental Hg on the vessel bottom was 4-77% of the amount added. The highest recovery of mercury in the offgas system generally correlated with the highest retention of Hg in the slurry. Low retention in the slurry (high segregation on the vessel bottom) resulted in low recovery in the offgas system. High agitation rates appear to result in lower

  4. Hydrogen generation during treatment of simulated high-level radioactive waste with formic acid

    International Nuclear Information System (INIS)

    Ritter, J.A.; Zamecnik, J.R.; Hsu, C.W.

    1992-01-01

    The Integrated Defense Waste Processing Facility (DWPF) Melter System (IDMS), operated by the Savannah River Laboratory, is a one-fifth scale pilot facility used in support of the start-up and operation of the Department of Energy's DWPF. Five IDMS runs determined the effect of the presence of noble metals in HLW sludge on the H 2 generation rate during the preparation of melter feed with formic acid. Overall, the results clearly showed that H 2 generation in the DWPF SRAT could, at times, exceed the lower flammable limit of H 2 in air (4 vol%), depending on such factors as offgas generation and air inleakage of the DWPF vessels. Therefore, the installation of a forced air purge system and H 2 monitors were recommended to the DWPF to control the generation of H 2 during melter feed preparation by fuel dilution

  5. Effect of Canister Movement on Water Turbidity

    International Nuclear Information System (INIS)

    TRIMBLE, D.J.

    2000-01-01

    Requirements for evaluating the adherence characteristics of sludge on the fuel stored in the K East Basin and the effect of canister movement on basin water turbidity are documented in Briggs (1996). The results of the sludge adherence testing have been documented (Bergmann 1996). This report documents the results of the canister movement tests. The purpose of the canister movement tests was to characterize water turbidity under controlled canister movements (Briggs 1996). The tests were designed to evaluate methods for minimizing the plumes and controlling water turbidity during fuel movements leading to multi-canister overpack (MCO) loading. It was expected that the test data would provide qualitative visual information for use in the design of the fuel retrieval and water treatment systems. Video recordings of the tests were to be the only information collected

  6. Defense Waste Processing Facility Process Simulation Package Life Cycle

    International Nuclear Information System (INIS)

    Reuter, K.

    1991-01-01

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

  7. Lot No. 1 of Frit 202 for DWPF cold runs

    International Nuclear Information System (INIS)

    Schumacher, R.F.

    1993-01-01

    This report was prepared at the end of 1992 and summarizes the evaluation of the first lot sample of DWPF Frit 202 from Cataphote Inc. Publication of this report was delayed until the results from the carbon analyses could be included. To avoid confusion the frit specifications presented in this report were those available at the end of 1992. The specifications were slightly modified early in 1993. The frit was received and evaluated for moisture, particle size distribution, organic-inorganic carbon and chemical composition. Moisture content and particle size distribution were determined on a representative sample at SRTC. These properties were within the DWPF specifications for Frit 202. A representative sample was submitted to Corning Engineering Laboratory Services for chemical analyses. The sample was split and two dissolutions prepared. Each dissolution was analyzed on two separate days. The results indicate that there is a high probability (>95%) that the silica content of this frit is below the specification limit of 77.0 ± 1.0 wt %. The average of the four analyzed values was 75.1 wt % with a standard deviation of 0.28 wt %. All other oxides were within the elliptical two sigma limits. Control standard frit samples were submitted and analyzed at the same time and the results were very similar to previous analyses of these materials

  8. Final Report: Characterization of Canister Mockup Weld Residual Stresses

    International Nuclear Information System (INIS)

    Enos, David; Bryan, Charles R.

    2016-01-01

    Stress corrosion cracking (SCC) of interim storage containers has been indicated as a high priority data gap by the Department of Energy (DOE) (Hanson et al., 2012), the Electric Power Research Institute (EPRI, 2011), the Nuclear Waste Technical Review Board (NWTRB, 2010a), and the Nuclear Regulatory Commission (NRC, 2012a, 2012b). Uncertainties exist in terms of the environmental conditions that prevail on the surface of the storage containers, the stress state within the container walls associated both with weldments as well as within the base metal itself, and the electrochemical properties of the storage containers themselves. The goal of the work described in this document is to determine the stress states that exists at various locations within a typical storage canister by evaluating the properties of a full-diameter cylindrical mockup of an interim storage canister. This mockup has been produced using the same manufacturing procedures as the majority of the fielded spent nuclear fuel interim storage canisters. This document describes the design and procurement of the mockup and the characterization of the stress state associated with various portions of the container. It also describes the cutting of the mockup into sections for further analyses, and a discussion of the potential impact of the results from the stress characterization effort.

  9. Final Report: Characterization of Canister Mockup Weld Residual Stresses

    Energy Technology Data Exchange (ETDEWEB)

    Enos, David [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bryan, Charles R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2016-12-01

    Stress corrosion cracking (SCC) of interim storage containers has been indicated as a high priority data gap by the Department of Energy (DOE) (Hanson et al., 2012), the Electric Power Research Institute (EPRI, 2011), the Nuclear Waste Technical Review Board (NWTRB, 2010a), and the Nuclear Regulatory Commission (NRC, 2012a, 2012b). Uncertainties exist in terms of the environmental conditions that prevail on the surface of the storage containers, the stress state within the container walls associated both with weldments as well as within the base metal itself, and the electrochemical properties of the storage containers themselves. The goal of the work described in this document is to determine the stress states that exists at various locations within a typical storage canister by evaluating the properties of a full-diameter cylindrical mockup of an interim storage canister. This mockup has been produced using the same manufacturing procedures as the majority of the fielded spent nuclear fuel interim storage canisters. This document describes the design and procurement of the mockup and the characterization of the stress state associated with various portions of the container. It also describes the cutting of the mockup into sections for further analyses, and a discussion of the potential impact of the results from the stress characterization effort.

  10. Physical properties of encapsulate spent fuel in canisters; Comportamiento fisico de las capsulas de almacenamiento

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-07-01

    Spent fuel and high-level wastes will be permanently stored in a deep geological repository (AGP). Prior to this, they will be encapsulated in canisters. The present report is dedicated to the study of such canisters under the different physical demands that they may undergo, be those in operating or accident conditions. The physical demands of interest include mechanical demands, both static and dynamic, and thermal demands. Consideration is given to the complete file of the canister, from the time when it is empty and without lid to the final conditions expected in the repository. Thermal analyses of canisters containing spent fuel are often carried out in two dimensions, some times with hypotheses of axial symmetry and some times using a plane transverse section through the centre of the canister. The results obtained in both types of analyses are compared here to those of complete three-dimensional analyses. The latter generate more reliable information about the temperatures that may be experienced by the canister and its contents; they also allow calibrating the errors embodied in the two-dimensional calculations. (Author)

  11. Topical safety analysis report for the transportation of the NUHOMS reg-sign dry shielded canister

    International Nuclear Information System (INIS)

    1993-08-01

    This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS reg-sign) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS reg-sign DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS reg-sign Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport

  12. Corrosion test plan to guide canister material selection and design for a tuff repository

    International Nuclear Information System (INIS)

    McCright, R.D.; van Konynenburg, R.A.; Ballou, L.B.

    1983-11-01

    Corrosion rates and the mode of corrosion attack form a most important basis for selection of canister materials and design of a nuclear waste package. Type 304L stainless steel was selected as the reference material for canister fabrication because of its generally excellent corrosion resistance in water, steam and air. However, 304L may be susceptible to localized and stress-assisted forms of corrosion under certain conditions. Alternative alloys are also investigated; these alloys were chosen because of their improved resistance to these forms of corrosion. The fabrication and welding processes, as well as the glass pouring operation for defense and commercial high-level wastes, may influence the susceptibility of the canister to localized and stress forms of corrosion. 12 references, 2 figures, 4 tables

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-09

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

  14. Remote instrument/electrical wall nozzle replaement in the Defense Waste Processing Facility

    International Nuclear Information System (INIS)

    Heckendorn, F.M. II.

    1983-09-01

    The Defense Waste Processing Facility (DWPF) for waste vitrification at the Savannah River Plant is in the final design stage. Development of remotely replaceable instrument and electrical through-wall wiring is now complete. These assemblies connect the power and control signals from the high radiation environment to the personnel access areas. The ability to replace them will extend the life and lower the cost of the DWPF. 3 references, 22 figures, 2 tables

  15. CANISTER HANDLING FACILITY DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    J.F. Beesley

    2005-04-21

    The purpose of this facility description document (FDD) is to establish requirements and associated bases that drive the design of the Canister Handling Facility (CHF), which will allow the design effort to proceed to license application. This FDD will be revised at strategic points as the design matures. This FDD identifies the requirements and describes the facility design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This FDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This FDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the facility. Knowledge of these requirements is essential in performing the design process. The FDD follows the design with regard to the description of the facility. The description provided in this FDD reflects the current results of the design process.

  16. CANISTER HANDLING FACILITY DESCRIPTION DOCUMENT

    International Nuclear Information System (INIS)

    Beesley. J.F.

    2005-01-01

    The purpose of this facility description document (FDD) is to establish requirements and associated bases that drive the design of the Canister Handling Facility (CHF), which will allow the design effort to proceed to license application. This FDD will be revised at strategic points as the design matures. This FDD identifies the requirements and describes the facility design, as it currently exists, with emphasis on attributes of the design provided to meet the requirements. This FDD is an engineering tool for design control; accordingly, the primary audience and users are design engineers. This FDD is part of an iterative design process. It leads the design process with regard to the flowdown of upper tier requirements onto the facility. Knowledge of these requirements is essential in performing the design process. The FDD follows the design with regard to the description of the facility. The description provided in this FDD reflects the current results of the design process

  17. Settlement of Canisters with smectite clay envelopes in deposition holes

    International Nuclear Information System (INIS)

    Pusch, R.

    1986-12-01

    Settlement of canisters containing radioactive waste and being surrounded by dense smectite clay is caused by the stresses and heat induced in the clay. Consolidation by water expulsion of the clay underlying a model canister with 5 cm diameter and 30 cm length would theoretically account for a maximum finite settlement of about 70 my m in a few weeks, while shear-induced creep would yield a settlement of only a few microns in the same time period. These predictions were checked by running a laboratory test in which a dead load of 80 kg was applied to a small cylindrical copper canister embedded in Na bentonite. The settlement, which increased in proportion to log time, turned out to be about 6 my m in the first 2.5 months. After the first loading period at room temperature, heating to 50 degrees C and, after a 4 months long 'room temperature' period, to 70 degrees C took place. This cycling gave strong, instant settlement and upheaval because of the different thermal expansion of the interacting components of the system. After the development of constant temperature conditions in the entire system and completion of the consolidation or expansion that followed from the thermo-mechanical interactions, the settlement proceeded at a rather high rate at 70 degrees C, still following a log time creep law, but with somewhat stronger retardation. At room temperature, i.e. in the post-heating periods, the settlement seemed to cease, on the other hand. The conclusion from the study is that the canister movements under isothermal conditions were in accordance with the log t-type creep settlement that was predicted in theoretical grounds. Pre-heating and low stresses may account for extraordinary retardation of the settlement. (author)

  18. Analysis of probability of defects in the disposal canisters

    International Nuclear Information System (INIS)

    Holmberg, J.-E.; Kuusela, P.

    2011-06-01

    This report presents a probability model for the reliability of the spent nuclear waste final disposal canister. Reliability means here that the welding of the canister lid has no critical defects from the long-term safety point of view. From the reliability point of view, both the reliability of the welding process (that no critical defects will be born) and the non-destructive testing (NDT) process (all critical defects will be detected) are equally important. In the probability model, critical defects in a weld were simplified into a few types. Also the possibility of human errors in the NDT process was taken into account in a simple manner. At this moment there is very little representative data to determine the reliability of welding and also the data on NDT is not well suited for the needs of this study. Therefore calculations presented here are based on expert judgements and on several assumptions that have not been verified yet. The Bayesian probability model shows the importance of the uncertainty in the estimation of the reliability parameters. The effect of uncertainty is that the probability distribution of the number of defective canisters becomes flat for larger numbers of canisters compared to the binomial probability distribution in case of known parameter values. In order to reduce the uncertainty, more information is needed from both the reliability of the welding and NDT processes. It would also be important to analyse the role of human factors in these processes since their role is not reflected in typical test data which is used to estimate 'normal process variation'.The reported model should be seen as a tool to quantify the roles of different methods and procedures in the weld inspection process. (orig.)

  19. Application of the TEMPEST computer code to canister-filling heat transfer problems

    International Nuclear Information System (INIS)

    Farnsworth, R.K.; Faletti, D.W.; Budden, M.J.

    1988-03-01

    Pacific Northwest Laboratory (PNL) researchers used the TEMPEST computer code to simulate thermal cooldown behavior of nuclear waste glass after it was poured into steel canisters for long-term storage. The objective of this work was to determine the accuracy and applicability of the TEMPEST code when used to compute canister thermal histories. First, experimental data were obtained to provide the basis for comparing TEMPEST-generated predictions. Five canisters were instrumented with appropriately located radial and axial thermocouples. The canister were filled using the pilot-scale ceramic melter (PSCM) at PNL. Each canister was filled in either a continous or a batch filling mode. One of the canisters was also filled within a turntable simulant (a group of cylindrical shells with heat transfer resistances similar to those in an actual melter turntable). This was necessary to provide a basis for assessing the ability of the TEMPEST code to also model the transient cooling of canisters in a melter turntable. The continous-fill model, Version M, was found to predict temperatures with more accuracy. The turntable simulant experiment demonstrated that TEMPEST can adequately model the asymmetric temperature field caused by the turntable geometry. Further, TEMPEST can acceptably predict the canister cooling history within a turntable, despite code limitations in computing simultaneous radiation and convection heat transfer between shells, along with uncertainty in stainless-steel surface emissivities. Based on the successful performance of TEMPEST Version M, development was initiated to incorporate 1) full viscous glass convection, 2) a dynamically adaptive grid that automatically follows the glass/air interface throughout the transient, and 3) a full enclosure radiation model to allow radiation heat transfer to non-nearest neighbor cells. 5 refs., 47 figs., 17 tabs

  20. DWPF GLASS BEADS AND GLASS FRIT TRANSPORT DEMONSTRATION

    Energy Technology Data Exchange (ETDEWEB)

    Adamson, D; Bradley Pickenheim, B

    2008-11-24

    DWPF is considering replacing irregularly shaped glass frit with spherical glass beads in the Slurry Mix Evaporator (SME) process to decrease the yield stress of the melter feed (a non-Newtonian Bingham Plastic). Pilot-scale testing was conducted on spherical glass beads and glass frit to determine how well the glass beads would transfer when compared to the glass frit. Process Engineering Development designed and constructed the test apparatus to aid in the understanding and impacts that spherical glass beads may have on the existing DWPF Frit Transfer System. Testing was conducted to determine if the lines would plug with the glass beads and the glass frit slurry and what is required to unplug the lines. The flow loop consisted of vertical and horizontal runs of clear PVC piping, similar in geometry to the existing system. Two different batches of glass slurry were tested: a batch of 50 wt% spherical glass beads and a batch of 50 wt% glass frit in process water. No chemicals such as formic acid was used in slurry, only water and glass formers. The glass beads used for this testing were commercially available borosilicate glass of mesh size -100+200. The glass frit was Frit 418 obtained from DWPF and is nominally -45+200 mesh. The spherical glass beads did not have a negative impact on the frit transfer system. The transferring of the spherical glass beads was much easier than the glass frit. It was difficult to create a plug with glass bead slurry in the pilot transfer system. When a small plug occurred from setting overnight with the spherical glass beads, the plug was easy to displace using only the pump. In the case of creating a man made plug in a vertical line, by filling the line with spherical glass beads and allowing the slurry to settle for days, the plug was easy to remove by using flush water. The glass frit proved to be much more difficult to transfer when compared to the spherical glass beads. The glass frit impacted the transfer system to the point

  1. DWPF GLASS BEADS AND GLASS FRIT TRANSPORT DEMONSTRATION

    International Nuclear Information System (INIS)

    Adamson, D.; Pickenheim, Bradley

    2008-01-01

    DWPF is considering replacing irregularly shaped glass frit with spherical glass beads in the Slurry Mix Evaporator (SME) process to decrease the yield stress of the melter feed (a non-Newtonian Bingham Plastic). Pilot-scale testing was conducted on spherical glass beads and glass frit to determine how well the glass beads would transfer when compared to the glass frit. Process Engineering Development designed and constructed the test apparatus to aid in the understanding and impacts that spherical glass beads may have on the existing DWPF Frit Transfer System. Testing was conducted to determine if the lines would plug with the glass beads and the glass frit slurry and what is required to unplug the lines. The flow loop consisted of vertical and horizontal runs of clear PVC piping, similar in geometry to the existing system. Two different batches of glass slurry were tested: a batch of 50 wt% spherical glass beads and a batch of 50 wt% glass frit in process water. No chemicals such as formic acid was used in slurry, only water and glass formers. The glass beads used for this testing were commercially available borosilicate glass of mesh size -100+200. The glass frit was Frit 418 obtained from DWPF and is nominally -45+200 mesh. The spherical glass beads did not have a negative impact on the frit transfer system. The transferring of the spherical glass beads was much easier than the glass frit. It was difficult to create a plug with glass bead slurry in the pilot transfer system. When a small plug occurred from setting overnight with the spherical glass beads, the plug was easy to displace using only the pump. In the case of creating a man made plug in a vertical line, by filling the line with spherical glass beads and allowing the slurry to settle for days, the plug was easy to remove by using flush water. The glass frit proved to be much more difficult to transfer when compared to the spherical glass beads. The glass frit impacted the transfer system to the point

  2. Deep Borehole Disposal Concept: Development of Universal Canister Concept of Operations

    Energy Technology Data Exchange (ETDEWEB)

    Rigali, Mark J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Applied Systems Analysis and Research; Price, Laura L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States). Applied Systems Analysis and Research

    2016-08-01

    This report documents key elements of the conceptual design for deep borehole disposal of radioactive waste to support the development of a universal canister concept of operations. A universal canister is a canister that is designed to be able to store, transport, and dispose of radioactive waste without the canister having to be reopened to treat or repackage the waste. This report focuses on the conceptual design for disposal of radioactive waste contained in a universal canister in a deep borehole. The general deep borehole disposal concept consists of drilling a borehole into crystalline basement rock to a depth of about 5 km, emplacing WPs in the lower 2 km of the borehole, and sealing and plugging the upper 3 km. Research and development programs for deep borehole disposal have been ongoing for several years in the United States and the United Kingdom; these studies have shown that deep borehole disposal of radioactive waste could be safe, cost effective, and technically feasible. The design concepts described in this report are workable solutions based on expert judgment, and are intended to guide follow-on design activities. Both preclosure and postclosure safety were considered in the development of the reference design concept. The requirements and assumptions that form the basis for the deep borehole disposal concept include WP performance requirements, radiological protection requirements, surface handling and transport requirements, and emplacement requirements. The key features of the reference disposal concept include borehole drilling and construction concepts, WP designs, and waste handling and emplacement concepts. These features are supported by engineering analyses.

  3. SPENT NUCLEAR FUEL NUMBER DENSITIES FOR MULTI-PURPOSE CANISTER CRITICALITY CALCULATIONS

    International Nuclear Information System (INIS)

    D. A. Thomas

    1996-01-01

    The purpose of this analysis is to calculate the number densities for spent nuclear fuel (SNF) to be used in criticality evaluations of the Multi-Purpose Canister (MPC) waste packages. The objective of this analysis is to provide material number density information which will be referenced by future MPC criticality design analyses, such as for those supporting the Conceptual Design Report

  4. Final Report - Engineering Study for DWPF Bubblers, VSL-10R1770-1, Rev. 0, dated 12/22/10

    Energy Technology Data Exchange (ETDEWEB)

    Kruger, Albert A.; Joseph, I.; Matlack, K. S.; Kot, W. K.; Diener, G. A.; Pegg, I. L.; Callow, R. A.

    2013-11-13

    The objective of this work was to perform an engineering assessment of the impact of implementation of bubblers to improve mixing of the glass pool, and thereby increase throughput, in the Defense Waste Processing Facility (DWPF) on the melter and off-gas system. Most of the data used for this evaluation were from extensive melter tests performed on non-SRS feeds. This information was supplemented by more recent results on SRS HLW simulants that were tested on a melter system at VSL under contracts from ORP and SRR. Per the work scope, the evaluation focused on the following areas: Glass production rate; Corrosion of melter components; Power requirements; Pouring stability; Off-gas characteristics; Safety and flammability.

  5. Review of Statistical Analyses Resulting from Performance of HLDWD-DWPF-005

    International Nuclear Information System (INIS)

    Beck, R.S.

    1997-01-01

    The Engineering Department at the Defense Waste Processing Facility (DWPF) has reviewed two reports from the Statistical Consulting Section (SCS) involving the statistical analysis of test results for analysis of small sample inserts (references 1 ampersand 2). The test results cover two proposed analytical methods, a room temperature hydrofluoric acid preparation (Cold Chem) and a sodium peroxide/sodium hydroxide fusion modified for insert samples (Modified Fusion). The reports support implementation of the proposed small sample containers and analytical methods at DWPF. Hydragard sampler valve performance was typical of previous results (reference 3). Using an element from each major feed stream. lithium from the frit and iron from the sludge, the sampler was determined to deliver a uniform mixture in either sample container.The lithium to iron ratios were equivalent for the standard 15 ml vial and the 3 ml insert.The proposed method provide equivalent analyses as compared to the current methods. The biases associated with the proposed methods on a vitrified basis are less than 5% for major elements. The sum of oxides for the proposed method compares favorably with the sum of oxides for the conventional methods. However, the average sum of oxides for the Cold Chem method was 94.3% which is below the minimum required recovery of 95%. Both proposed methods, cold Chem and Modified Fusion, will be required at first to provide an accurate analysis which will routinely meet the 95% and 105% average sum of oxides limit for Product Composition Control System (PCCS).Issued to be resolved during phased implementation are as follows: (1) Determine calcine/vitrification factor for radioactive feed; (2) Evaluate covariance matrix change against process operating ranges to determine optimum sample size; (3) Evaluate sources for low sum of oxides; and (4) Improve remote operability of production versions of equipment and instruments for installation in 221-S.The specifics of

  6. Canister storage building natural phenomena design loads

    International Nuclear Information System (INIS)

    Tallman, A.M.

    1996-02-01

    This document presents natural phenomena hazard (NPH) loads for use in the design and construction of the Canister Storage Building (CSB), which will be located in the 200 East Area of the Hanford Site

  7. Canister transfer into repository in shaft alternative

    International Nuclear Information System (INIS)

    Raiko, H.; Kukkola, T.; Autio, J.

    2005-09-01

    In this report, a study of lift transportation of a massive canister for spent nuclear fuel is considered. The canister is transferred from ground level to repository, which lies in the depth of 400 to 500 m in the bedrock. The canister is a massive metal vessel, whose weight is 19 to 29 tons, and which is strongly irradiant (gamma and neutrons), and which contains 1.4 to 2.2 tons of very strongly radio-active material, the activity of the fuel should not be spread in the environment even during postulated accidents. The study observes that the lift alternative is possible to be built and through good design practices and good maintenance procedures its safety, reliability and usability can be kept on such high level that canister transport is estimated to be licensable. (orig.)

  8. Investigation and modelling of friction stir welded copper canisters

    Energy Technology Data Exchange (ETDEWEB)

    Kaellgren, Therese

    2010-02-15

    This work has been focused on characterisation of FSW joints, and modelling of the process, both analytically and numerically. The Swedish model for final deposit of nuclear fuel waste is based on copper canisters as a corrosion barrier with an inner pressure holding insert of cast iron. Friction Stir Welding (FSW) is the method to seal the copper canister, a technique invented by The Welding Institute (TWI). The first simulations were based on Rosenthal's analytical medium plate model. The model is simple to use, but has limitations. Finite element models (FEM) were developed, initially with a two-dimensional geometry. Due to the requirements of describing both the heat flow and the tool movement, three-dimensional models were developed. These models take into account heat transfer, material flow, and continuum mechanics. The geometries of the models are based on the simulation experiments carried out at TWI and at Swedish Nuclear Fuel Waste and Management Co (SKB). Temperature distribution, material flow and their effects on the thermal expansion were predicted for a full-scale canister and lid. The steady state solutions have been compared with temperature measurements, showing good agreement. In order to understand the material flow during welding a marker technique is used, which involves inserting dissimilar material into the weld zone before joining. Different materials are tested showing that brass rods are the most suitable material in these welds. After welding, the weld line is sliced, etched and examined by optical microscope. To understand the material flow further, and in the future predict the flow, a FEM is developed. This model and the etched samples are compared showing similar features. Furthermore, by using this model the area that is recrystallised can be predicted. The predicted area and the grain size and hardness profile agree well

  9. Design premises for canister for spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Werme, L

    1998-09-01

    The purpose of this report is to establish the basic premises for designing canisters for the disposal of spent nuclear fuel, the requirements for canister characteristics, and the design criteria, and to present alternative canister designs that satisfy these premises. The point of departure for canister design has been that the canister must be able to be used for both BWR and PWR fuel 43 refs, 4 figs, 6 tabs

  10. Preparation and Heat-Treatment of DWPF Simulants With and Without Co-Precipitated Noble Metals

    International Nuclear Information System (INIS)

    Koopman, David C.:Eibling, Russel E

    2005-01-01

    The Savannah River National Laboratory is in the process of investigating factors suspected of impacting catalytic hydrogen generation in the Chemical Process Cell of the Defense Waste Processing Facility, DWPF. Noble metal catalyzed hydrogen generation in simulation work constrains the allowable acid addition operating window in DWPF. This constraint potentially impacts washing strategies during sludge batch preparation. It can also influence decisions related to the addition of secondary waste streams to a sludge batch. Noble metals have historically been added as trim chemicals to process simulations. The present study investigated the potential conservatism that might be present from adding the catalytic species as trim chemicals to the final sludge simulant versus co-precipitating the noble metals into the insoluble sludge solids matrix. Parallel preparations of two sludge simulants targeting the composition of Sludge Batch 3 were performed in order to evaluate the impact of the form of noble metals. Identical steps were used except that one simulant had dissolved palladium, rhodium, and ruthenium present during the precipitation of the insoluble solids. Noble metals were trimmed into the other stimulant prior to process tests. Portions of both sludge simulants were held at 97 C for about eight hours to qualitatively simulate the effects of long term storage on particle morphology and speciation. The simulants were used as feeds for Sludge Receipt and Adjustment Tank, SRAT, process simulations. The following conclusions were drawn from the simulant preparation work: (1) The first preparation of a waste slurry simulant with co-precipitated noble metals was successful, based on the data obtained. It appears that 99+% of the noble metals were retained in the simulant. (2) Better control of carbonate, hydroxide, and post-wash trim chemical additions is needed before the new method of simulant preparation will be as reproducible as the old method. (3) The two new

  11. Am/Cm canister temperature evaluation in CIM5

    International Nuclear Information System (INIS)

    Baich, M.A.

    2000-01-01

    To facilitate the evaluation of alternate canister designs, 2 canisters were outfitted with thermocouples at elevations of 1/2, 3 1/2, and 6 1/2 inches from the canister bottom. The canisters were fabricated from two inch diameter schedule 10 and two inch diameter schedule 40 stainless steel pipe. Each canister was filled with approximately 2 kilograms of 49 wt percent lanthanide (Ln) loaded 25SrABS glass during 5 inch Cylindrical Induction Melter (CIM5) runs for TTR Tasks 3.03 and 4.03. Melter temperature, total mass of glass poured, and the glass pour rates were almost identical in both runs. The schedule 40 canister has a slightly smaller ID compared to the schedule 10 canister and therefore filled to a level of 9.5 inches compared to 8.0 inches for the schedule 40 canister. The schedule 40 canister had an empty mass of 1906 grams compared to 919 grams for the schedule 10 canister. The schedule 10 canister was found to have a higher maximum surface temperature by about 50--100 C (depending on height) during the glass pour compared to the schedule 40 canister. The additional thermal mass of the schedule 40 canister accounts for this difference. Once filled with glass, each of the canisters cooled at about the same rate, taking about an hour to cool below a maximum surface temperature of 200 C. No significant deformation of the either of the canisters was visually observed

  12. APET methodology for Defense Waste Processing Facility: Mode C operation

    International Nuclear Information System (INIS)

    Taylor, R.P. Jr.; Massey, W.M.

    1995-04-01

    Safe operation of SRS facilities continues to be the highest priority of the Savannah River Site (SRS). One of these facilities, the Defense Waste Processing Facility or DWPF, is currently undergoing cold chemical runs to verify the design and construction preparatory to hot startup in 1995. The DWPFF is a facility designed to convert the waste currently stored in tanks at the 200-Area tank farm into a form that is suitable for long term storage in engineered surface facilities and, ultimately, geologic isolation. As a part of the program to ensure safe operation of the DWPF, a probabilistic Safety Assessment of the DWPF has been completed. The results of this analysis are incorporated into the Safety Analysis Report (SAR) for DWPF. The usual practice in preparation of Safety Analysis Reports is to include only a conservative analysis of certain design basis accidents. A major part of a Probabilistic Safety Assessment is the development and quantification of an Accident Progression Event Tree or APET. The APET provides a probabilistic representation of potential sequences along which an accident may progress. The methodology used to determine the risk of operation of the DWPF borrows heavily from methods applied to the Probabilistic Safety Assessment of SRS reactors and to some commercial reactors. This report describes the Accident Progression Event Tree developed for the Probabilistic Safety Assessment of the DWPF

  13. Pilot scale processing of simulated Savannah River Site high level radioactive waste

    International Nuclear Information System (INIS)

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

    1991-01-01

    The Savannah River Laboratory operates the Integrated DWPF Melter System (IDMS), which is a pilot-scale test facility used in support of the start-up and operation of the US Department of Energy's Defense Waste Processing Facility (DWPF). Specifically, the IDMS is used in the evaluation of the DWPF melter and its associated feed preparation and offgass treatment systems. This article provides a general overview of some of the test work which has been conducted in the IDMS facility. The chemistry associated with the chemical treatment of the sludge (via formic acid adjustment) is discussed. Operating experiences with simulated sludge containing high levels of nitrite, mercury, and noble metals are summarized

  14. Thermal studies of the canister staging pit in a hypothetical Yucca Mountain canister handling facility using computational fluid dynamics

    International Nuclear Information System (INIS)

    Soltani, Mehdi; Barringer, Chris; Bues, Timothy T. de

    2007-01-01

    The proposed Yucca Mountain nuclear waste storage site will contain facilities for preparing the radioactive waste canisters for burial. A previous facility design considered was the Canister Handling Facility Staging Pit. This design is no longer used, but its thermal evaluation is typical of such facilities. Structural concrete can be adversely affected by the heat from radioactive decay. Consequently, facilities must have heating ventilation and air conditioning (HVAC) systems for cooling. Concrete temperatures are a function of conductive, convective and radiative heat transfer. The prediction of concrete temperatures under such complex conditions can only be adequately handled by computational fluid dynamics (CFD). The objective of the CFD analysis was to predict concrete temperatures under normal and off-normal conditions. Normal operation assumed steady state conditions with constant HVAC flow and temperatures. However, off-normal operation was an unsteady scenario which assumed a total HVAC failure for a period of 30 days. This scenario was particularly complex in that the concrete temperatures would gradually rise, and air flows would be buoyancy driven. The CFD analysis concluded that concrete wall temperatures would be at or below the maximum temperature limits in both the normal and off-normal scenarios. While this analysis was specific to a facility design that is no longer used, it demonstrates that such facilities are reasonably expected to have satisfactory thermal performance. (author)

  15. Preliminary design for spent fuel canister handling systems in a canister transfer and installation vehicle

    International Nuclear Information System (INIS)

    Wendelin, T.; Suikki, M.

    2008-12-01

    The report presents a spent fuel canister transfer and installation vehicle. The vehicle is used for carrying the fuel canister into a disposal tunnel and installing it into a deposition hole. The report outlines basic requirements and a design for canister handling equipment used in a canister transfer and installation vehicle, a description regarding the operation and maintenance of the equipment, as well as a cost estimate. Specific vehicles will be manufactured for all canister types in order to minimize the height of the disposal tunnels. This report is only focused on a transfer and installation vehicle for OL1-2 fuel canisters. Detailed designing and selection of final components have not yet been carried out. The report also describes the vehicle's requirements for the structures of a repository system, as well as actions in possible malfunction or fault situations. The spent fuel canister is brought from an encapsulation plant by a canister lift down to the repository level. The fuel canister is driven from the canister lift by an automated guided vehicle onto a canister hoist at a canister loading station. The canister transfer and installation vehicle is waiting for the canister with its radiation shield in an upright position above the canister hoist. The hoist carries the canister upward until the vehicle's own lifting means grab hold of the canister and raise it up into the vehicle's radiation shield. This is followed by turning the radiation shield to a transport position and by closing it in a radiation-proof manner against a rear radiation shield. The vehicle is driven along the central tunnel into the disposal tunnel and parked on top of the deposition hole. The vehicle's radiation shield is turned to the upright position and the canister is lowered with the vehicle's hydraulic winches into a bentonite-lined deposition hole. The radiation shield is turned back to the transport position and the vehicle can be driven out of the disposal tunnel

  16. Numerical Modelling of Mechanical Integrity of the Copper-Cast Iron Canister. A Literature Review

    International Nuclear Information System (INIS)

    Lanru Jing

    2004-04-01

    This review article presents a summary of the research works on the numerical modelling of the mechanical integrity of the composite copper-cast iron canisters for the final disposal of Swedish nuclear wastes, conducted by SKB and SKI since 1992. The objective of the review is to evaluate the outstanding issues existing today about the basic design concepts and premises, fundamental issues on processes, properties and parameters considered for the functions and requirements of canisters under the conditions of a deep geological repository. The focus is placed on the adequacy of numerical modelling approaches adopted in regards to the overall mechanical integrity of the canisters, especially the initial state of canisters regarding defects and the consequences of their evolution under external and internal loading mechanisms adopted in the design premises. The emphasis is the stress-strain behaviour and failure/strength, with creep and plasticity involved. Corrosion, although one of the major concerns in the field of canister safety, was not included

  17. Drop of canistered spent fuel segments into a deep borehole and subsequent aerosol release

    International Nuclear Information System (INIS)

    Bantle, S.; Herbe, H.; Miu, J.

    1991-09-01

    The source term of the released aerosols is estimated. First, the number of failing canisters is calculated for the case of an axial symmetric canister (POLLUX) pile, and then, for the case of a 'zig-zag' pile, as found in reality. The weight-specific energy acting on the fuel - a measure for the degree of fuel fractioning - is determined from the acceleration acting on the pin segments. In the borehole prevails a steady-state flow pattern which is stimulated by the heat of the disposed waste canister, and is also influenced by the ventilation of the drift above the borehole. Based on this stationary flow pattern flow velocities are calculated by means of fluid mechanical methods. Further investigations deal with the unsteady case which occurs during and immediately after the canister drop as well as with the wake behind the canister. The most relevant result is that under the considered boundary conditions no release form the borehole into the repository is to be expected. (orig./HP) [de

  18. Progress in the understanding of the long-term corrosion behaviour of copper canisters

    Science.gov (United States)

    King, Fraser; Lilja, Christina; Vähänen, Marjut

    2013-07-01

    Copper has been proposed as a canister material for the disposal of spent nuclear fuel in a deep geologic repository in a number of countries worldwide. Since it was first proposed for this purpose in 1978, a significant number of studies have been performed to assess the corrosion performance of copper under repository conditions. These studies are critically reviewed and the suitability of copper as a canister material for nuclear waste is re-assessed. Over the past 30-35 years there has been considerable progress in our understanding of the expected corrosion behaviour of copper canisters. Crucial to this progress has been the improvement in the understanding of the nature of the repository environment and how it will evolve over time. With this improved understanding, it has been possible to predict the evolution of the corrosion behaviour from the initial period of warm, aerobic conditions in the repository to the long-term phase of cool, anoxic conditions dominated by the presence of sulphide. An historical review of the treatment of the corrosion behaviour of copper canisters is presented, from the initial corrosion assessment in 1978, through a major review of the corrosion behaviour in 2001, through to the current level of understanding based on the results of on-going studies. Compared with the initial corrosion assessment, there has been considerable progress in the treatment of localised corrosion, stress corrosion cracking, and microbiologically influenced corrosion of the canisters. Progress in the mechanistic modelling of the evolution of the corrosion behaviour of the canister is also reviewed, as is the continuing debate about the thermodynamic stability of copper in pure water. The overall conclusion of this critical review is that copper is a suitable material for the disposal of spent nuclear fuel and offers the prospect of containment of the waste for an extended period of time. The corrosion behaviour is influenced by the presence of the

  19. Remote controlled mover for disposal canister transfer

    Energy Technology Data Exchange (ETDEWEB)

    Suikki, M. [Optimik Oy, Turku (Finland)

    2013-10-15

    This working report is an update for an earlier automatic guided vehicle design (Pietikaeinen 2003). The short horizontal transfers of disposal canisters manufactured in the encapsulation process are conducted with remote controlled movers both in the encapsulation plant and in the underground areas at the canister loading station of the disposal facility. The canister mover is a remote controlled transfer vehicle mobile on wheels. The handling of canisters is conducted with the assistance of transport platforms (pallets). The very small automatic guided vehicle of the earlier design was replaced with a commercial type mover. The most important reasons for this being the increased loadbearing requirement and the simpler, proven technology of the vehicle. The larger size of the vehicle induced changes to the plant layouts and in the principles for dealing with fault conditions. The selected mover is a vehicle, which is normally operated from alongside. In this application, the vehicle steering technology must be remote controlled. In addition, the area utilization must be as efficient as possible. This is why the vehicle was downsized in its outer dimensions and supplemented with certain auxiliary equipment and structures. This enables both remote controlled operation and improves the vehicle in terms of its failure tolerance. Operation of the vehicle was subjected to a risk analysis (PFMEA) and to a separate additional calculation conserning possible canister toppling risks. The total cost estimate, without value added tax for manufacturing the system amounts to 730 000 euros. (orig.)

  20. Drop Testing Representative Multi-Canister Overpacks

    Energy Technology Data Exchange (ETDEWEB)

    Snow, Spencer D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Morton, Dana K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-06-01

    The objective of the work reported herein was to determine the ability of the Multi- Canister Overpack (MCO) canister design to maintain its containment boundary after an accidental drop event. Two test MCO canisters were assembled at Hanford, prepared for testing at the Idaho National Engineering and Environmental Laboratory (INEEL), drop tested at Sandia National Laboratories, and evaluated back at the INEEL. In addition to the actual testing efforts, finite element plastic analysis techniques were used to make both pre-test and post-test predictions of the test MCOs structural deformations. The completed effort has demonstrated that the canister design is capable of maintaining a 50 psig pressure boundary after drop testing. Based on helium leak testing methods, one test MCO was determined to have a leakage rate not greater than 1x10-5 std cc/sec (prior internal helium presence prevented a more rigorous test) and the remaining test MCO had a measured leakage rate less than 1x10-7 std cc/sec (i.e., a leaktight containment) after the drop test. The effort has also demonstrated the capability of finite element methods using plastic analysis techniques to accurately predict the structural deformations of canisters subjected to an accidental drop event.

  1. Remote controlled mover for disposal canister transfer

    International Nuclear Information System (INIS)

    Suikki, M.

    2013-10-01

    This working report is an update for an earlier automatic guided vehicle design (Pietikaeinen 2003). The short horizontal transfers of disposal canisters manufactured in the encapsulation process are conducted with remote controlled movers both in the encapsulation plant and in the underground areas at the canister loading station of the disposal facility. The canister mover is a remote controlled transfer vehicle mobile on wheels. The handling of canisters is conducted with the assistance of transport platforms (pallets). The very small automatic guided vehicle of the earlier design was replaced with a commercial type mover. The most important reasons for this being the increased loadbearing requirement and the simpler, proven technology of the vehicle. The larger size of the vehicle induced changes to the plant layouts and in the principles for dealing with fault conditions. The selected mover is a vehicle, which is normally operated from alongside. In this application, the vehicle steering technology must be remote controlled. In addition, the area utilization must be as efficient as possible. This is why the vehicle was downsized in its outer dimensions and supplemented with certain auxiliary equipment and structures. This enables both remote controlled operation and improves the vehicle in terms of its failure tolerance. Operation of the vehicle was subjected to a risk analysis (PFMEA) and to a separate additional calculation conserning possible canister toppling risks. The total cost estimate, without value added tax for manufacturing the system amounts to 730 000 euros. (orig.)

  2. DWPF SB6 Initial CPC Flowsheet Testing SB6-1 TO SB6-4L Tests Of SB6-A And SB6-B Simulants

    International Nuclear Information System (INIS)

    Lambert, D.; Pickenheim, B.; Best, D.

    2009-01-01

    The Defense Waste Processing Facility (DWPF) will transition from Sludge Batch 5 (SB5) processing to Sludge Batch 6 (SB6) processing in late fiscal year 2010. Tests were conducted using non-radioactive simulants of the expected SB6 composition to determine the impact of varying the acid stoichiometry during the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) processes. The work was conducted to meet the Technical Task Request (TTR) HLW/DWPF/TTR-2008-0043, Rev.0 and followed the guidelines of a Task Technical and Quality Assurance Plan (TT and QAP). The flowsheet studies are performed to evaluate the potential chemical processing issues, hydrogen generation rates, and process slurry rheological properties as a function of acid stoichiometry. These studies were conducted with the estimated SB6 composition at the time of the study. This composition assumed a blend of 101,085 kg of Tank 4 insoluble solids and 179,000 kg of Tank 12 insoluble solids. The current plans are to subject Tank 12 sludge to aluminum dissolution. Liquid Waste Operations assumed that 75% of the aluminum would be dissolved during this process. After dissolution and blending of Tank 4 sludge slurry, plans included washing the contents of Tank 51 to ∼1M Na. After the completion of washing, the plan assumes that 40 inches on Tank 40 slurry would remain for blending with the qualified SB6 material. There are several parameters that are noteworthy concerning SB6 sludge: (1) This is the second batch DWPF will be processing that contains sludge that has had a significant fraction of aluminum removed through aluminum dissolution; (2) The sludge is high in mercury, but the projected concentration is lower than SB5; (3) The sludge is high in noble metals, but the projected concentrations are lower than SB5; and(4) The sludge is high in U and Pu - components that are not added in sludge simulants. Six DWPF process simulations were completed in 4-L laboratory-scale equipment using

  3. Miscibility Evaluation Of The Next Generation Solvent With Polymers Currently Used At DWPF, MCU, And Saltstone

    Energy Technology Data Exchange (ETDEWEB)

    Fondeur, F. F.

    2013-04-17

    The Office of Waste Processing, within the Office of Technology Innovation and Development, funded the development of an enhanced Caustic-Side Solvent Extraction (CSSX) solvent for deployment at the Savannah River Site for removal of cesium from High Level Waste. This effort lead to the development of the Next Generation Solvent (NGS) with Tris (3,7-dimethyl octyl) guanidine (TiDG). The first deployment target for the NGS solvent is within the Modular CSSX Unit (MCU). Deployment of a new chemical within an existing facility requires verification that the new chemical components are compatible with the installed equipment. In the instance of a new organic solvent, the primary focus is on compatibility of the solvent with organic polymers used in the affected facility. This report provides the calculated data from exposing these polymers to the Next Generation Solvent. An assessment of the dimensional stability of polymers known to be used or present in the MCU, Defense Waste Processing Facility (DWPF), and Saltstone facilities that will be exposed to the NGS showed that TiDG could selectively affect the elastomers and some thermoplastics to varying extents, but the typical use of these polymers in a confined geometry will likely prevent the NGS from impacting component performance. The polymers identified as of primary concern include Grafoil® (flexible graphite), Tefzel®, Isolast®, ethylene-propylene-diene monomer (EPDM) rubber, nitrile-butadiene rubber (NBR), styrene-butadiene rubber (SBR), ultra high molecular weight polyethylene (UHMWPE), and fluorocarbon rubber (FKM). Certain polymers like NBR and EPDM were found to interact mildly with NGS but their calculated swelling and the confined geometry will impede interaction with NGS. In addition, it was found that Vellumoid (cellulose fibers-reinforced glycerin and protein) may leach protein and Polyvinyl Chloride (PVC) may leach plasticizer (such as Bis-Ethylhexyl-Phthalates) into the NGS solvent. Either case

  4. Ecological studies related to construction of the Defense Waste Processing Facility on the Savannah River Site

    International Nuclear Information System (INIS)

    Scott, D.E.; Pechmann, J.H.K.; Knox, J.N.; Estes, R.A.; McGregor, J.H.; Bailey, K.

    1988-12-01

    The Savannah River Ecology Laboratory has completed 10 years of ecological studies related to the construction of the Defense Waste Processing Facility (DWPF) on the Savannah River Site. This progress report examines water quality studies on streams peripheral to the DWPF construction site and examines the effectiveness of ''refuge ponds'' in ameliorating the effects of construction on local amphibians. Individual papers on these topics are indexed separately. 93 refs., 15 figs., 15 tabs

  5. Ecological studies related to construction of the Defense Waste Processing Facility on the Savannah River Site

    Energy Technology Data Exchange (ETDEWEB)

    Scott, D.E.; Pechmann, J.H.K.; Knox, J.N.; Estes, R.A.; McGregor, J.H.; Bailey, K. (ed.)

    1988-12-01

    The Savannah River Ecology Laboratory has completed 10 years of ecological studies related to the construction of the Defense Waste Processing Facility (DWPF) on the Savannah River Site. This progress report examines water quality studies on streams peripheral to the DWPF construction site and examines the effectiveness of refuge ponds'' in ameliorating the effects of construction on local amphibians. Individual papers on these topics are indexed separately. 93 refs., 15 figs., 15 tabs. (MHB)

  6. Testing of the Defense Waste Processing Facility Cold Chemical Dissolution Method in Sludge Batch 9 Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, T. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Pareizs, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Coleman, C. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Young, J. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Brown, L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2016-05-10

    For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) tests the applicability of the digestion methods used by the DWPF Laboratory for elemental analysis of Sludge Receipt and Adjustment Tank (SRAT) Receipt samples and SRAT Product process control samples. DWPF SRAT samples are typically dissolved using a method referred to as the DWPF Cold Chemical or Cold Chem Method (CC), (see DWPF Procedure SW4- 15.201). Testing indicates that the CC method produced mixed results. The CC method did not result in complete dissolution of either the SRAT Receipt or SRAT Product with some fine, dark solids remaining. However, elemental analyses did not reveal extreme biases for the major elements in the sludge when compared with analyses obtained following dissolution by hot aqua regia (AR) or sodium peroxide fusion (PF) methods. The CC elemental analyses agreed with the AR and PF methods well enough that it should be adequate for routine process control analyses in the DWPF after much more extensive side-by-side tests of the CC method and the PF method are performed on the first 10 SRAT cycles of the Sludge Batch 9 (SB9) campaign. The DWPF Laboratory should continue with their plans for further tests of the CC method during these 10 SRAT cycles.

  7. Erosion Modeling Analysis For Modified DWPF SME Tank

    International Nuclear Information System (INIS)

    LEE, SI

    2004-01-01

    In support of an erosion evaluation for the modified cooling coil guide and its supporting structure in the DWPF SME vessel, a computational model was developed to identify potential sites of high erosion using the same methodology established by previous work. The erosion mechanism identified in the previous work was applied to the evaluation of high erosion locations representative of the actual flow process in the modified coil guide of the SME vessel, abrasive erosion which occurs by high wall shear of viscous liquid. The results show that primary locations of the highest erosion due to the abrasive wall erosion are at the leading edge of the guide, external surface of the insert plate, the tank floor next to the insert plate of the coil guide support, and the upstream lead-in plate. The present modeling results show a good comparison between the original and the modified cases in terms of high erosion sites, as well as the degree of erosion and the calculated shear stress. Wall she ar of the tank floor is reduced by about 30 per cent because of the new coil support plate. Calculations for the impeller speed lower than 103 rpm in the SME showed similar erosion patterns but significantly reduced wall shear stresses and reduced overall erosion. Comparisons of the 103 rpm results with SME measurements indicated that no significant erosion of the tank floor in the SME is to be expected. Thus, it is recommended that the agitator speed of SME does not exceed 103 rpm

  8. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a jointed arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  9. Plutonium Immobilization Project - Can-In-Canister Hardware Development/Selection

    International Nuclear Information System (INIS)

    Hamilton, L.

    2001-01-01

    The Plutonium Immobilization Project (PIP) is a program funded by the U.S. Department of Energy to develop technology to disposition excess weapons grade plutonium. This program introduces the ''Can-in-Canister'' (CIC) technology that immobilizes the plutonium by encapsulating it in ceramic forms (or pucks) and ultimately surrounding it with high-level waste glass to provide a deterrent to recovery. Since there are significant radiation, contamination and security concerns, the project team is developing unique technologies to remotely perform plutonium immobilization tasks. This paper covers the design, development and testing of the magazines (cylinders containing cans of ceramic pucks) and the rack that holds them in place inside the waste glass canister. Several magazine and rack concepts were evaluated to produce a design that gives the optimal balance between resistance to thermal degradation and facilitation of remote handling. This paper also reviews the effort to develop a join ted arm robot that can remotely load seven magazines into defined locations inside a stationary canister working only through the 4 inch (102 mm) diameter canister throat

  10. The Impact Of The Mcu Life Extension Solvent On Dwpf Glass Formulation Efforts

    International Nuclear Information System (INIS)

    Peeler, D.; Edwards, T.

    2011-01-01

    As a part of the Actinide Removal Process (ARP)/Modular Caustic Side Solvent Extraction Unit (MCU) Life Extension Project, a next generation solvent (NG-CSSX), a new strip acid, and modified monosodium titanate (mMST) will be deployed. The strip acid will be changed from dilute nitric acid to dilute boric acid (0.01 M). Because of these changes, experimental testing with the next generation solvent and mMST is required to determine the impact of these changes in 512-S operations as well as Chemical Process Cell (CPC), Defense Waste Processing Facility (DWPF) glass formulation activities, and melter operations at DWPF. To support programmatic objectives, the downstream impacts of the boric acid strip effluent (SE) to the glass formulation activities and melter operations are considered in this study. More specifically, the impacts of boric acid additions to the projected SB7b operating windows, potential impacts to frit production temperatures, and the potential impact of boron volatility are evaluated. Although various boric acid molarities have been reported and discussed, the baseline flowsheet used to support this assessment was 0.01M boric acid. The results of the paper study assessment indicate that Frit 418 and Frit 418-7D are robust to the implementation of the 0.01M boric acid SE into the SB7b flowsheet (sludge-only or ARP-added). More specifically, the projected operating windows for the nominal SB7b projections remain essentially constant (i.e., 25-43 or 25-44% waste loading (WL)) regardless of the flowsheet options (sludge-only, ARP added, and/or the presence of the new SE). These results indicate that even if SE is not transferred to the Sludge Receipt and Adjustment Tank (SRAT), there would be no need to add boric acid (from a trim tank) to compositionally compensate for the absence of the boric acid SE in either a sludge-only or ARP-added SB7b flowsheet. With respect to boron volatility, the Measurement Acceptability Region (MAR) assessments also

  11. Mechanical Integrity of Copper Canister Lid and Cylinder

    International Nuclear Information System (INIS)

    Karlsson, Marianne

    2002-01-01

    This report compiles finite element analyses performed to ensure the structural integrity of canisters used for storing of nuclear fuel waste of type BWR. The report comprises analyses performed on the canister lid and cylinder casing in order to determine static and long-term strength of the structure. The report analyses the mechanical response of the lid and flange of the copper canister when subjected to loads caused by pressure from swelling bentonite and from ground water at a depth of 500 meter. The loads acting on the canister are somewhat uncertain and the cases investigated in this report are possible cases. Load cases analysed are: Pressure 15 MPa uniformly distributed on lid and 5 MPa uniformly distributed on cylinder; Pressure 5 MPa uniformly distributed on lid and 15 MPa uniformly distributed on cylinder; Pressure 20 MPa uniformly distributed on lid and cylinder; Side pressures 10 MPa and 20 MPa uniformly distributed on part of the cylinder. Creep analyses are also performed in order to estimate the stresses that will arise when the canister is placed in the repository. The analyses in this report are recreated from the original analyses but the models differ in geometry. Also, there is no information in the original reports on material data, time-independent as well as creep data, and analysis procedure. The data used in the recreated analyses are based on information from References 2, 3, 6 and 7. The results presented in this report are based on the supplementary analyses. These results differ from the original results. Most likely this is due to differences in model geometry. The original results are appended to the report and are summarised for comparison with results from the supplementary analyses. Otherwise, these results are not further discussed. For all load cases, high tensile stresses are found in the lid fillet between the planar part and the flange. High tensile stresses are also found in the weld surface and on the outer side of the

  12. NOBLE METAL CHEMISTRY AND HYDROGEN GENERATION DURING SIMULATED DWPF MELTER FEED PREPARATION

    Energy Technology Data Exchange (ETDEWEB)

    Koopman, D

    2008-06-25

    Simulations of the Defense Waste Processing Facility (DWPF) Chemical Processing Cell vessels were performed with the primary purpose of producing melter feeds for the beaded frit program plus obtaining samples of simulated slurries containing high concentrations of noble metals for off-site analytical studies for the hydrogen program. Eight pairs of 22-L simulations were performed of the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles. These sixteen simulations did not contain mercury. Six pairs were trimmed with a single noble metal (Ag, Pd, Rh, or Ru). One pair had all four noble metals, and one pair had no noble metals. One supporting 4-L simulation was completed with Ru and Hg. Several other 4-L supporting tests with mercury have not yet been performed. This report covers the calculations performed on SRNL analytical and process data related to the noble metals and hydrogen generation. It was originally envisioned as a supporting document for the off-site analytical studies. Significant new findings were made, and many previous hypotheses and findings were given additional support as summarized below. The timing of hydrogen generation events was reproduced very well within each of the eight pairs of runs, e.g. the onset of hydrogen, peak in hydrogen, etc. occurred at nearly identical times. Peak generation rates and total SRAT masses of CO{sub 2} and oxides of nitrogen were reproduced well. Comparable measures for hydrogen were reproduced with more variability, but still reasonably well. The extent of the reproducibility of the results validates the conclusions that were drawn from the data.

  13. Uncertainty analysis of multiple canister repository model by large-scale calculation

    International Nuclear Information System (INIS)

    Tsujimoto, K.; Okuda, H.; Ahn, J.

    2007-01-01

    A prototype uncertainty analysis has been made by using the multiple-canister radionuclide transport code, VR, for performance assessment for the high-level radioactive waste repository. Fractures in the host rock determine main conduit of groundwater, and thus significantly affect the magnitude of radionuclide release rates from the repository. In this study, the probability distribution function (PDF) for the number of connected canisters in the same fracture cluster that bears water flow has been determined in a Monte-Carlo fashion by running the FFDF code with assumed PDFs for fracture geometry. The uncertainty for the release rate of 237 Np from a hypothetical repository containing 100 canisters has been quantitatively evaluated by using the VR code with PDFs for the number of connected canisters and the near field rock porosity. The calculation results show that the mass transport is greatly affected by (1) the magnitude of the radionuclide source determined by the number of connected canisters by the fracture cluster, and (2) the canister concentration effect in the same fracture network. The results also show the two conflicting tendencies that the more fractures in the repository model space, the greater average value but the smaller uncertainty of the peak fractional release rate is. To perform a vast amount of calculation, we have utilized the Earth Simulator and SR8000. The multi-level hybrid programming method is applied in the optimization to exploit high performance of the Earth Simulator. The Latin Hypercube Sampling has been utilized to reduce the number of samplings in Monte-Carlo calculation. (authors)

  14. Design basis for the copper/steel canister. Stage three. Final report

    International Nuclear Information System (INIS)

    Bowyer, W.H.

    1997-02-01

    The development of the copper/iron canister proposed for the containment of high-level waste in the Swedish disposal programme has been studied from the points of view of choice of materials, manufacturing technology and Q A. This report describes the observations on progress which has been made between March 1995 and February 1996 and the results of further literature studies. A first trial canister has been produced by SKB using a fabricated steel liner and an extruded copper tubular, a second one using a fabricated tubular is at an advanced stage. A change from a fabricated steel inner canister to a proposed cast canister has been justified by a criticality argument but the technology for producing a cast canister is at present untried. It is considered that such a change will require a significant development programme. The microstructure achieved in the extruded copper tubular for the first canister is unacceptable. An improved microstructure may be achieved by extruding at a lower temperature but this remains to be demonstrated. Similar problems exist with plate used for the fabricated tubular but some more favourable structures have been achieved already by this route. Seam welding of the first tubular failed through a suspected material problem. The second fabricated tubular welded without difficulty. However it was necessary to constrain it during welding and it subsequently distorted during machining. There was some evidence of hot tearing close to the weld. The distortion problem may be overcome by a stress relieving anneal but this could cause further grain size problems. 19 refs

  15. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    Werme, L.

    1990-09-01

    A near-field performance evaluation of an Advanced Cold Process Canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. The canister design was originally proposed by TVO. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. Throughout the analysis, present day underground conditions has been assumed to persist during the service life of the canister. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie localized corrosion for the steel or copper canisters can be dismissed as a failure mechanism. The evaluation of the effects of processes outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. This factor will ensure the safety of the concept. (orig.)

  16. Techniques for freeing deposited canisters. Final report

    International Nuclear Information System (INIS)

    Kalbantner, P.; Sjoeblom, R.

    2000-06-01

    Four different techniques for removal of the bentonite buffer around a deposited canister have been identified, studied and evaluated: mechanical, hydrodynamical, thermal, and electrical techniques. Different techniques to determine the position of the canister in the buffer have also been studied: mechanical, electromagnetic, thermal and acoustic techniques. The mechanical techniques studied are full-face boring, milling and core-drilling. It is expected that the bentonite can be machined relatively easily. It is assessed that cooling by means of flushing water over the outer surfaces of the tools is not feasible in view of the tendency of bentonite to form a gel. The mechanical techniques are characterized by the potential of damaging the canister, a high degree of complexity, and high requirements of energy/power input. The generated byproduct is solid and cannot be removed by means of flushing. Removal is assessed to be simplest in conjunction with full-face boring and most difficult when coredrilling is applied. The hydrodynamical techniques comprise high-pressure hydrodynamic techniques, where pressures above and below 100 bar, and low pressure hydrodynamical techniques (< 10 bar) are separated. At pressures above 100 bar, a water jet with a diameter of approximately a millimetre cuts through the material. If desired, sand can be added to the jet. At pressures below 100 bar the jet has a diameter of one or a few centimetres. The liquid contains a few percent of salt, which is essential for the efficiency of the process. The flushing is important not only because it removes the modified bentonite but also because it frees previously unaffected bentonite and thereby makes it accessible to chemical modification. All of the hydrodynamical techniques are applicable for freeing the end surface as well as the mantle surface. The degree of complexity and the requirement on energy/power decrease with a decrease in pressure. A significant potential for damaging the

  17. Topical safety analysis report for the transportation of the NUHOMS{reg_sign} dry shielded canister. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    None

    1993-08-01

    This Topical Safety Analysis Report (SAR) describes the design and the generic transportation licensing basis for utilizing the NUTECH HORIZONTAL MODULAR STORAGE (NUHOMS{reg_sign}) system dry shielded canister (DSC) containing twenty-four pressurized water reactor (PWR) spent fuel assemblies (SFA) in conjunction with a conceptually designed Transportation Cask. This SAR documents the design qualification of the NUHOMS{reg_sign} DSC as an integral part of a 10CFR71 Fissile Material Class III, Type B(M) Transportation Package. The package consists of the canister and a conceptual transportation cask (NUHOMS{reg_sign} Transportation Cask) with impact limiters. Engineering analysis is performed for the canister to confirm that the existing canister design complies with 10CFR71 transportation requirements. Evaluations and/or analyses is performed for criticality safety, shielding, structural, and thermal performance. Detailed engineering analysis for the transportation cask will be submitted in a future SAR requesting 10CFR71 certification of the complete waste package. Transportation operational considerations describe various operational aspects of the canister/transportation cask system. operational sequences are developed for canister transfer from storage to the transportation cask and interfaces with the cask auxiliary equipment for on- and off-site transport.

  18. Design of double containment canister cask storage system

    International Nuclear Information System (INIS)

    Asami, M.; Matsumoto, T.; Oohama, T.; Kuriyama, K.; Kawakami, K.

    2004-01-01

    Spent fuels discharged from Japanese LWR will be stored as recycled-fuel-resources in interim storage facilities. The concrete cask storage system is one of important forms for the spent fuel interim storage. In Japan, the interim storage facility will be located near the coast, therefore it is important to prevent SCC (Stress Corrosion Cracking) caused by sea salt particles and to assure the containment integrity of the canister which contains spent fuels. KEPCO, NFT and OCL have designed the double containment canister cask storage system that can assure the long-term containment integrity and monitor the containment performance without storage capacity decrease. Major features of the combined canister cask system are shown as follows: This system can survey containment integrity of dual canisters by monitoring the pressure of the gap between canisters. The primary canister has dual lids sealed by welding. The secondary canister has single lid tightened by bolts and sealed by metallic gaskets. The primary canister is contained in the transport cask during transportation, and the gap between the primary canister and the transport cask is filled with He gas. Under storage condition in the concrete cask, the primary canister is contained in the secondary canister, and the gap between these canisters is filled with helium gas. Hence this system can prevent the primary canister to contact sea salt particle in the air and from SCC. Decrease of cooling performance because of the double canister is compensated by fins fitted on the secondary canister surface. Then, this system can prevent the decrease of storage capacity determined by the fuel temperature limit. This system can assure that the primary canister will keep intact for long term storage. Therefore, in the case of pressure down of the gap between canisters, it can be considered that the secondary canister containment is damaged, and the primary canister will be transferred to another secondary canister at the

  19. Defense Waste Processing Facility Recycle Stream Evaporation

    International Nuclear Information System (INIS)

    STONE, MICHAEL

    2006-01-01

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) stabilizes high level radioactive waste (HLW) by vitrification of the waste slurries. DWPF currently produces approximately five gallons of dilute recycle for each gallon of waste vitrified. This recycle stream is currently sent to the HLW tank farm at SRS where it is processed through the HLW evaporators with the concentrate eventually sent back to the DWPF for stabilization. Limitations of the HLW evaporators and storage space constraints in the tank farm have the potential to impact the operation of the DWPF and could limit the rate that HLW is stabilized. After an evaluation of various alternatives, installation of a dedicated evaporator for the DWPF recycle stream was selected for further evaluation. The recycle stream consists primarily of process condensates from the pretreatment and vitrification processes. Other recycle streams consist of process samples, sample line flushes, sump flushes, and cleaning solutions from the decontamination and filter dissolution processes. The condensate from the vitrification process contains some species, such as sulfate, that are not appreciably volatile at low temperature and could accumulate in the system if 100% of the evaporator concentrate was returned to DWPF. These species are currently removed as required by solids washing in the tank farm. The cleaning solutions are much higher in solids content than the other streams and are generated 5-6 times per year. The proposed evaporator would be required to concentrate the recycle stream by a factor of 30 to allow the concentrate to be recycled directly to the DWPF process, with a purge stream sent to the tank farm as required to prevent buildup of sulfate and similar species in the process. The overheads are required to meet stringent constraints to allow the condensate to be sent directly to an effluent treatment plant. The proposed evaporator would nearly de-couple the DWPF process from the

  20. Design report of the canister for nuclear fuel disposal

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    1996-12-01

    The report provides a summary of the design of the canister for final disposal of nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 11 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (26 refs.)

  1. Status Of The Development Of In-Tank/At-Tank Separations Technologies For High-Level Waste Processing For The U.S. Department Of Energy

    International Nuclear Information System (INIS)

    Aaron, G.; Wilmarth, B.

    2011-01-01

    salt and sludge processing life cycle, thereby reducing the Defense Waste Processing Facility (DWPF) mission by 7 years. Additionally at the Hanford site, problematic waste streams, such as high boehmite and phosphate wastes, could be treated prior to receipt by WTP and thus dramatically improve the capacity of the facility to process HLW. Treatment of boehmite by continuous sludge leaching (CSL) before receipt by WTP will dramatically reduce the process cycle time for the WTP pretreatment facility, while treatment of phosphate will significantly reduce the number of HLW borosilicate glass canisters produced at the WTP. These and other promising technologies will be discussed.

  2. Performance Assessment and Sensitivity Analyses of Disposal of Plutonium as Can-in-Canister Ceramic

    International Nuclear Information System (INIS)

    Rainer Senger

    2001-01-01

    The purpose of this analysis is to examine whether there is a justification for using high-level waste (HLW) as a surrogate for plutonium disposal in can-in-canister ceramic in the total-system performance assessment (TSPA) model for the Site Recommendation (SR). In the TSPA-SR model, the immobilized plutonium waste form is not explicitly represented, but is implicitly represented as an equal number of canisters of HLW. There are about 50 metric tons of plutonium in the U. S. Department of Energy inventory of surplus fissile material that could be disposed. Approximately 17 tons of this material contain significant quantities of impurities and are considered unsuitable for mixed-oxide (MOX) reactor fuel. This material has been designated for direct disposal by immobilization in a ceramic waste form and encapsulating this waste form in high-level waste (HLW). The remaining plutonium is suitable for incorporation into MOX fuel assemblies for commercial reactors (Shaw 1999, Section 2). In this analysis, two cases of immobilized plutonium disposal are analyzed, the 17-ton case and the 13-ton case (Shaw et al. 2001, Section 2.2). The MOX spent-fuel disposal is not analyzed in this report. In the TSPA-VA (CRWMS M and O 1998a, Appendix B, Section B-4), the calculated dose release from immobilized plutonium waste form (can-in-canister ceramic) did not exceed that from an equivalent amount of HLW glass. This indicates that the HLW could be used as a surrogate for the plutonium can-in-canister ceramic. Representation of can-in-canister ceramic as a surrogate is necessary to reduce the number of waste forms in the TSPA model. This reduction reduces the complexity and running time of the TSPA model and makes the analyses tractable. This document was developed under a Technical Work Plan (CRWMS M and O 2000a), and is compliant with that plan. The application of the Quality Assurance (QA) program to the development of that plan (CRWMS M and O 2000a) and of this Analysis is

  3. Integrity of copper/steel canisters under crystalline bedrock repository conditions

    International Nuclear Information System (INIS)

    Bowyer, W.H.; Sjoblom, R.; Trolle, M.

    1996-01-01

    In the Swedish nuclear waste disposal programme, the need to store the spent nuclear fuel safely for very long times has prompted a strategy which includes a long life canister. Technical as well as economical considerations related to design, choice of materials and manufacturing technology have lead to the selection of a reference design to be used for the continued development work. The canisters are cylindrical with a diameter close to 1 meter and a height of about 5 meters. In order to meet the need for an appropriate combination of mechanical strength, toughness, durability and corrosion resistance, the canisters comprise an inner vessel made of steel or cast iron to cope with mechanical stresses and an outer vessel made of almost pure copper to provide corrosion resistance. The Swedish nuclear industry has recently extended its development work to full-scale tests. Such experience is needed not least for the evaluation of the long-term integrity of the canister. This work has been closely followed by the Swedish Nuclear Power Inspectorate (SKI) who have also carried out independent investigations and analyses. It should be emphasized that the findings relate to a canister which is under development and cannot, in general, be expected to be relevant for the fully developed canister. Significant results of the analyses include the identification of conceivable modes of canister failures. Such failures may be related to defects, segregation, limitations in inspectability, long term creep properties, adverse mechanical load situations, etc. It is assessed that the distribution functions of these failures might have their largest uncertainties at the tails extending to comparatively short times. Specific issues related to canister manufacture, scaling and non destructive testing which have been found to warrant further investigation are: defects in the copper ingot which may transfer to the rolled copper plate; the amount of work applied during the rolling or

  4. Near-field performance of the advanced cold process canister

    International Nuclear Information System (INIS)

    Werme, L.

    1991-12-01

    A near-field performance evaluation of an advanced cold process canister for spent fuel disposal has been performed jointly by TVO, Finland and SKB, Sweden. The canister consists of a steel canister as a load bearing element, with an outer corrosion shield of copper. In the analysis, as well internal (ie corrosion processes from the inside of the canister) as external processes (mechanical and chemical) have been considered both prior to and after canister breach. The major conclusions for the evaluation are: Internal processes cannot cause the canister breach under foreseen conditions, ie local-iced corrosion for the steel or copper canisters can be dismissed as a failure mechanism; The evaluation of the effects of processed outside the canister indicate that there is no rapid mechanism to endanger the integrity of the canister. Consequently the service life of the canister will be several million years. For completeness also evaluation of post-failure behaviour was carried out. Analyses were focussed on low probability phenomena from faults in canisters. Some items were identified where further research is justified in order to increase knowledge of the phenomena and thus strengthen the confidence of safety margins. However, it can be concluded that the risks of these scenarios can be judged to be acceptable. This is due to the fact that firstly, the probability of occurrence of most of these scenarios can be controlled to a large extent through technical measures. Secondly, these analyses indicated that the consequences would not be severe

  5. Design report of the disposal canister for twelve fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Raiko, H. [VTT Energy, Espoo (Finland); Salo, J.P. [Posiva Oy, Helsinki (Finland)

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.) 35 refs.

  6. Design report of the disposal canister for twelve fuel assemblies

    International Nuclear Information System (INIS)

    Raiko, H.; Salo, J.P.

    1999-05-01

    The report provides a summary of the design of the canister for final disposal of spent nuclear fuel. The canister structure consists of a cylindrical massive nodular graphite cast iron insert covered by a 50 mm thick copper overlay. The capacity of the canister is 12 assemblies of BWR or VVER 440 fuel. The canister shall be tight with a high probability for about 100 000 years. The good and long lasting tightness requires: (1) The good initial tightness that is achieved by high quality requirements and extensive quality control, (2) The good corrosion resistance, which is obtained by the overpack of oxygen free copper, and (3) Mechanical strength of the canister, that is ensured by analyses (the following loads are considered: hydrostatic pressure, even and uneven swelling pressure of bentonite, thermal effects, and elevated hydrostatic pressure during glaciation. The allowed stresses and strains are set in such a way that reasonable engineering safety factors are obtained in all assessed design base loading cases). The canister shall limit the radiation dose rate outside the canister to minimise the radiolysis of the water in the vicinity of the canister. The canister insert shall keep the fuel assemblies in a subcritical configuration even if the void in the canister is filled with water due to postulated leakage. The design basis of the canister is set, the performed analyses are summarised and the results are assessed and discussed in the report. (orig.)

  7. High Level Waste (HLW) Processing Experience with Increased Waste Loading

    International Nuclear Information System (INIS)

    JANTZEN, CAROL

    2004-01-01

    The Defense Waste Processing Facility (DWPF) Engineering requested characterization of glass samples that were taken after the second melter had been operational for about 5 months. After the new melter had been installed, the waste loading had been increased to about 38 weight percentage after a new quasicrystalli