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
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
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
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
Waste canister for storage of nuclear wastes
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.
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
Canister arrangement for storing radioactive waste
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.
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.
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
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.
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
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
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
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
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
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
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
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
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
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
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
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
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)
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
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
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
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
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
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.
Radiation transport in high-level waste form
International Nuclear Information System (INIS)
Arakali, V.S.; Barnes, S.M.
1992-01-01
The waste form selected for vitrifying high-level nuclear waste stored in underground tanks at West Valley, NY is borosilicate glass. The maximum radiation level at the surface of a canister filled with the high-level waste form is prescribed by repository design criteria for handling and disposition of the vitrified waste. This paper presents an evaluation of the radiation transport characteristics for the vitreous waste form expected to be produced at West Valley and the resulting neutron and gamma dose rates. The maximum gamma and neutron dose rates are estimated to be less than 7500 R/h and 10 mRem/h respectively at the surface of a West Valley canister filled with borosilicate waste glass
Characteristics of borosilicate waste glass form for high-level radioactive waste
Energy Technology Data Exchange (ETDEWEB)
Kim, Seung Soo; Chun, Kwan Sik; Choi, Jong Won; Kang, Chul Hyung
2001-03-01
Basic data, required for the design and the performance assessment of a repository of HLW, suchas the chemical composition and the characteristics of the borosilicate waste glass have been identified according to the burn-ups of spent PWR fuels. The diemnsion of waste canister is 430mm in diameter and 1135mm in length, and the canister should hold less than 2kwatts of heat from their decay of radionuclides contained in the HLW. Based on the reprocessing of 5 years-cooled spent fuel, one canister could hold about 11.5wt.% and 10.8wt.% of oxidized HLW corresponding to their burn-ups of 45,000MWD/MTU and 55,000MWD/MTU, respectively. These waste forms have been recommanded as the reference waste forms of HLW. The characteristics of these wastes as a function of decay time been evaluated. However, after a specific waste form and a specific site for the disposal would be selected, the characteristics of the waste should be reevaluated under the consideration of solidification period, loaded waste, storage condition and duration, site circumstances for the repository system and its performance assessment.
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.)
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
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
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
Alternatives for high-level waste forms, containers, and container processing systems
International Nuclear Information System (INIS)
Crawford, T.W.
1995-01-01
This study evaluates alternatives for high-level waste forms, containers, container processing systems, and onsite interim storage. Glass waste forms considered are cullet, marbles, gems, and monolithic glass. Small and large containers configured with several combinations of overpack confinement and shield casks are evaluated for these waste forms. Onsite interim storage concepts including canister storage building, bore holes, and storage pad were configured with various glass forms and canister alternatives. All favorable options include the monolithic glass production process as the waste form. Of the favorable options the unshielded 4- and 7-canister overpack options have the greatest technical assurance associated with their design concepts due to their process packaging and storage methods. These canisters are 0.68 m and 0.54 m in diameter respectively and 4.57 m tall. Life-cycle costs are not a discriminating factor in most cases, varying typically less than 15 percent
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)
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
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
Review of high-level waste form properties
International Nuclear Information System (INIS)
Rusin, J.M.
1980-12-01
This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison
International Nuclear Information System (INIS)
Anon.
1980-01-01
The schedules for waste package development for the various host rocks were presented. The waste form subtask activities were reviewed, with the papers focusing on high-level waste, transuranic waste, and spent fuel. The following ten papers were presented: (1) Waste Package Development Approach; (2) Borosilicate Glass as a Matrix for Savannah River Plant Waste; (3) Development of Alternative High-Level Waste Forms; (4) Overview of the Transuranic Waste Management Program; (5) Assessment of the Impacts of Spent Fuel Disassembly - Alternatives on the Nuclear Waste Isolation System; (6) Reactions of Spent Fuel and Reprocessing Waste Forms with Water in the Presence of Basalt; (7) Spent Fuel Stabilizer Screening Studies; (8) Chemical Interactions of Shale Rock, Prototype Waste Forms, and Prototype Canister Metals in a Simulated Wet Repository Environment; (9) Impact of Fission Gas and Volatiles on Spent Fuel During Geologic Disposal; and (10) Spent Fuel Assembly Decay Heat Measurement and Analysis
Development of solid radionuclide waste forms in the United States
International Nuclear Information System (INIS)
Crandall, J.L.
1979-01-01
New ways of reworking the wastes require a new classification in terms of the final waste forms. This paper surveys the candidate forms: encapsulation binders, in-place solidification waste forms, glass and ceramic waste forms, mineral waste forms, matrix waste forms, gaseous waste forms (fixation), and canisters and engineered barriers. Participants in the US-high-level waste form development program are listed. Requirements and selection of waste forms are also discussed. 26 references
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.
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
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
Alternative solidified forms for nuclear wastes
International Nuclear Information System (INIS)
McElroy, J.L.; Ross, W.A.
1976-01-01
Radioactive wastes will occur in various parts of the nuclear fuel cycle. These wastes have been classified in this paper as high-level waste, intermediate and low-level waste, cladding hulls, and residues. Solidification methods for each type of waste are discussed in a multiple barrier context of primary waste form, applicable coatings or films, matrix encapsulation, canister, engineered structures, and geological storage. The four major primary forms which have been most highly developed are glass for HLW, cement for ILW, organics for LLW, and metals for hulls
International Nuclear Information System (INIS)
Ross, W.A.; Lokken, R.O.; May, R.P.; Roberts, F.P.; Timmerman, C.L.; Treat, R.L.; Westsik, J.H. Jr.
1982-09-01
This study provides an assesses seven waste forms and eight processes for immobilizing transuranic (TRU) wastes. The waste forms considered are cast cement, cold-pressed cement, FUETAP (formed under elevated temperature and pressure) cement, borosilicate glass, aluminosilicate glass, basalt glass-ceramic, and cold-pressed and sintered silicate ceramic. The waste-immobilization processes considered are in-can glass melting, joule-heated glass melting, glass marble forming, cement casting, cement cold-pressing, FUETAP cement processing, ceramic cold-pressing and sintering, basalt glass-ceramic processing. Properties considered included gas generation, chemical durability, mechanical strength, thermal stability, and radiation stability. The ceramic products demonstrated the best properties, except for plutonium release during leaching. The glass and ceramic products had similar properties. The cement products generally had poorer properties than the other forms, except for plutonium release during leaching. Calculations of the Pu release indicated that the waste forms met the proposed NRC release rate limit of 1 part in 10 5 per year in most test conditions. The cast-cement process had the lowest processing cost, followed closely by the cold-pressed and FUETAP cement processes. Joule-heated glass melting had the lower cost of the glass processes. In-can melting in a high-quality canister had the highest cost, and cold-pressed and sintered ceramic the second highest. Labor and canister costs for in-can melting were identified. The major contributor to costs of disposing of TRU wastes in a defense waste repository is waste processing costs. Repository costs could become the dominant cost for disposing of TRU wastes in a commercial repository. It is recommended that cast and FUETAP cement and borosilicate glass waste-form systems be considered. 13 figures, 16 tables
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
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
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.
Disposal costs for SRP high-level wastes in borosilicate glass and crystalline ceramic waste forms
International Nuclear Information System (INIS)
Rozsa, R.B.; Campbell, J.H.
1982-01-01
Purpose of this document is to compare and contrast the overall burial costs of the glass and ceramic waste forms, including processing, storage, transportation, packaging, and emplacement in a repository. Amount of waste will require approximately 10,300 standard (24 in. i.d. x 9-5/6 ft length) canisters of waste glass, each containing about 3260 lb of waste at 28% waste loading. The ceramic waste form requires about one-third the above number of standard canisters. Approximately $2.5 billion is required to process and dispose of this waste, and the total cost is independent of waste form (glass or ceramic). The major cost items (about 80% of the total cost) for all cases are capital and operating expenses. The capital and 20-year operating costs for the processing facility are the same order of magnitude, and their sum ranges from about one-half of the total for the reference glass case to two-thirds of the total for the ceramic cases
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
Review of high-level waste form properties. [146 bibliographies
Energy Technology Data Exchange (ETDEWEB)
Rusin, J.M.
1980-12-01
This report is a review of waste form options for the immobilization of high-level-liquid wastes from the nuclear fuel cycle. This review covers the status of international research and development on waste forms as of May 1979. Although the emphasis in this report is on waste form properties, process parameters are discussed where they may affect final waste form properties. A summary table is provided listing properties of various nuclear waste form options. It is concluded that proposed waste forms have properties falling within a relatively narrow range. In regard to crystalline versus glass waste forms, the conclusion is that either glass of crystalline materials can be shown to have some advantage when a single property is considered; however, at this date no single waste form offers optimum properties over the entire range of characteristics investigated. A long-term effort has been applied to the development of glass and calcine waste forms. Several additional waste forms have enough promise to warrant continued research and development to bring their state of development up to that of glass and calcine. Synthetic minerals, the multibarrier approach with coated particles in a metal matrix, and high pressure-high temperature ceramics offer potential advantages and need further study. Although this report discusses waste form properties, the total waste management system should be considered in the final selection of a waste form option. Canister design, canister materials, overpacks, engineered barriers, and repository characteristics, as well as the waste form, affect the overall performance of a waste management system. These parameters were not considered in this comparison.
Processes for production of alternative waste forms
International Nuclear Information System (INIS)
Ross, W.A.; Rusin, J.M.; McElroy, J.L.
1979-01-01
During the past 20 years, numerous waste forms and processes have been proposed for solidification of high-level radioactive wastes (HLW). The number has increased significantly during the past 3 to 4 years. At least five factors must be considered in selecting the waste form and process method: 1) processing flexibility, 2) waste loading, 3) canister size and stability, 4) waste form inertness and stability, and 5) processing complexity. This paper describes various waste form processes and operations, and a simple system is proposed for making comparisons. This system suggests that one goal for processes would be to reduce the number of process steps, thereby providing less complex processing systems
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
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.)
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
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
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
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
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
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
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
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.
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
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
International Nuclear Information System (INIS)
1983-10-01
The conceptual waste package interim product specifications and data requirements presented are applicable to the reference glass composition described in PNL-3838 and carbon steel canister described in ONWI-438. They provide preliminary numerical values for the commercial 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. 13 references, 1 figure
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)
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
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
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
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
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
Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form
Energy Technology Data Exchange (ETDEWEB)
Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.
2011-09-12
The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.
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)
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)
Low-risk alternative waste forms for problematic high-level and long-lived nuclear wastes
International Nuclear Information System (INIS)
Stewart, M.W.A.; Begg, B.D.; Moricca, S.; Day, R.A.
2006-01-01
Full text: The highest cost component the nuclear waste clean up challenge centres on high-level waste (HLW) and consequently the greatest opportunity for cost and schedule savings lies with optimising the approach to HLW cleanup. The waste form is the key component of the immobilisation process. To achieve maximum cost savings and optimum performance the selection of the waste form should be driven by the characteristics of the specific nuclear waste to be immobilised, rather than adopting a single baseline approach. This is particularly true for problematic nuclear wastes that are often not amenable to a single baseline approach. The use of tailored, high-performance, alternative waste forms that include ceramics and glass-ceramics, coupled with mature process technologies offer significant performance improvements and efficiency savings for a nuclear waste cleanup program. It is the waste form that determines how well the waste is locked up (chemical durability), and the number of repository disposal canisters required (waste loading efficiency). The use of alternative waste forms for problematic wastes also lowers the overall risk by providing high performance HLW treatment alternatives. The benefits tailored alternative waste forms bring to the HLW cleanup program will be briefly reviewed with reference to work carried out on the following: The HLW calcines at the Idaho National Laboratory; SYNROC ANSTO has developed a process utilising a glass-ceramic combined with mature hot-isostatic pressing (HIP) technology and has demonstrated this at a waste loading of 80 % and at a 30 kg HIP scale. The use of this technology has recently been estimated to result in a 70 % reduction in waste canisters, compared to the baseline borosilicate glass technology; Actinide-rich waste streams, particularly the work being done by SYNROC ANSTO with Nexia Solutions on the Plutonium-residues wastes at Sellafield in the UK, which if implemented is forecast to result in substantial
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%
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
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
International Nuclear Information System (INIS)
Hand, R.L.
1990-12-01
The US Department of Energy-Office of Civilian Radioactive Waste Management has been designated the national high-level waste repository licensee and the recipient for the canistered waste forms. The Office of Waste Operations executes overall responsibility for producing the canistered waste form. The Hanford Waste Vitrification Plant Project, as part of the waste form producer organization, consists of a vertical relationship. Overall control is provided by the US Department of Energy-Environmental Restoration and Waste Management Headquarters; with the US Department of Energy-Office of Waste Operations; the US Department of Energy- Headquarters/Vitrification Project Branch; the US Department of Energy-Richland Operations Office/Vitrification Project Office; and the Westinghouse Hanford Company, operations and engineering contractor. This document has been prepared in response to direction from the US Department of Energy-Office of Civilian Radioactive Waste Management through the US Department of Energy-Richland Operations Office for a quality assurance program that meets the requirements of the US Department of Energy. This document provides guidance and direction for implementing a quality assurance program that applies to the Hanford Waste Vitrification Plant Project. The Hanford Waste Vitrification Plant Project management commits to implementing the quality assurance program activities; reviewing the program periodically, and revising it as necessary to keep it current and effective. 12 refs., 6 figs., 1 tab
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
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
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
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
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
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
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
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.
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
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
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
Corrosion of high-level radioactive waste iron-canisters in contact with bentonite.
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.
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
Evaluation of conditioned high-level waste forms
International Nuclear Information System (INIS)
Mendel, J.E.; Turcotte, R.P.; Chikalla, T.D.; Hench, L.L.
1983-01-01
The evaluation of conditioned high-level waste forms requires an understanding of radiation and thermal effects, mechanical properties, volatility, and chemical durability. As a result of nuclear waste research and development programs in many countries, a good understanding of these factors is available for borosilicate glass containing high-level waste. The IAEA through its coordinated research program has contributed to this understanding. Methods used in the evaluation of conditioned high-level waste forms are reviewed. In the US, this evaluation has been facilitated by the definition of standard test methods by the Materials Characterization Center (MCC), which was established by the Department of Energy (DOE) in 1979. The DOE has also established a 20-member Materials Review Board to peer-review the activities of the MCC. In addition to comparing waste forms, testing must be done to evaluate the behavior of waste forms in geologic repositories. Such testing is complex; accelerated tests are required to predict expected behavior for thousands of years. The tests must be multicomponent tests to ensure that all potential interactions between waste form, canister/overpack and corrosion products, backfill, intruding ground water and the repository rock, are accounted for. An overview of the status of such multicomponent testing is presented
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
Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams
International Nuclear Information System (INIS)
Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.
2010-01-01
In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.(1) The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste
Glass Ceramic Waste Forms for Combined CS+LN+TM Fission Products Waste Streams
Energy Technology Data Exchange (ETDEWEB)
Crum, Jarrod V.; Turo, Laura A.; Riley, Brian J.; Tang, Ming; Kossoy, Anna; Sickafus, Kurt E.
2010-09-23
In this study, glass ceramics were explored as an alternative waste form for glass, the current baseline, to be used for immobilizing alkaline/alkaline earth + lanthanide (CS+LN) or CS+LN+transition metal (TM) fission-product waste streams generated by a uranium extraction (UREX+) aqueous separations type process. Results from past work on a glass waste form for the combined CS+LN waste streams showed that as waste loading increased, large fractions of crystalline phases precipitated upon slow cooling.[1] The crystalline phases had no noticeable impact on the waste form performance by the 7-day product consistency test (PCT). These results point towards the development of a glass ceramic waste form for treating CS+LN or CS+LN+TM combined waste streams. Three main benefits for exploring glass ceramics are: (1) Glass ceramics offer increased solubility of troublesome components in crystalline phases as compared to glass, leading to increased waste loading; (2) The crystalline network formed in the glass ceramic results in higher heat tolerance than glass; and (3) These glass ceramics are designed to be processed by the same melter technology as the current baseline glass waste form. It will only require adding controlled canister cooling for crystallization into a glass ceramic waste form. Highly annealed waste form (essentially crack free) with up to 50X lower surface area than a typical High-Level Waste (HLW) glass canister. Lower surface area translates directly into increased durability. This was the first full year of exploring glass ceramics for the Option 1 and 2 combined waste stream options. This work has shown that dramatic increases in waste loading are achievable by designing a glass ceramic waste form as an alternative to glass. Table S1 shows the upper limits for heat, waste loading (based on solubility), and the decay time needed before treatment can occur for glass and glass ceramic waste forms. The improvements are significant for both combined waste
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.
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...
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)
Compatibility testing of vitrified waste forms
International Nuclear Information System (INIS)
Rankin, W.N.
1978-01-01
The compatibility of vitrified radioactive waste with candidate canister materials will be evaluated with both cast and in-can melted vitrified waste. Both real and simulated sludges will be used. In addition, the compatibility of these materials with salt from a possible final storage location will be determined. Cast vitrified waste will be tested with ASTM A 333 and ASTM A 516 low-carbon steels and Type 304L stainless steel at 100, 600 and 800 0 C. Cast vitrified waste that has been devitrified by heat treatment will be tested at 100 0 C. Two types of test specimens will be used with either simulated or real sludges: (1) unsealed capsules made of pieces of mill-finished pipe into which vitrified waste is cast, and (2) sealed capsules containing a small container of vitrified waste identical to the ones in the unsealed capsule. In-can melted vitrified waste will be tested with synthetic sludge only and with ASTM A 333 and ASTM A 516 low-carbon steels, Type 304L stainless steel and Inconel 600. Two types of tests will be carried out: (1) melting vitrified waste in miniature metal canisters and (2) exposure of small (carefully measured) metal coupons to molten glass. The air oxidation rates of candidate canister materials will be determined, and specimens will also be exposed to salt from Drill Hole AEC-8 in Carlsbad, New Mexico. Sealed capsules containing an ASTM A 516 low-carbon steel or Type 304L stainless steel specimen partially embedded in a small block of salt will be heated
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.
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
Waste package designs for disposal of high-level waste in salt formations
International Nuclear Information System (INIS)
Basham, S.J. Jr.; Carr, J.A.
1984-01-01
In the United States of America the selected method for disposal of radioactive waste is mined repositories located in suitable geohydrological settings. Currently four types of host rocks are under consideration: tuff, basalt, crystalline rock and salt. Development of waste package designs for incorporation in mined salt repositories is discussed. The three pertinent high-level waste forms are: spent fuel, as disassembled and close-packed fuel pins in a mild steel canister; commercial high-level waste (CHLW), as borosilicate glass in stainless-steel canisters; defence high-level waste (DHLW), as borosilicate glass in stainless-steel canisters. The canisters are production and handling items only. They have no planned long-term isolation function. Each waste form requires a different approach in package design. However, the general geometry and the materials of the three designs are identical. The selected waste package design is an overpack of low carbon steel with a welded closure. This container surrounds the waste forms. Studies to better define brine quantity and composition, radiation effects on the salt and brines, long-term corrosion behaviour of the low carbon steel, and the leaching behaviour of the spent fuel and borosilicate glass waste forms are continuing. (author)
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.
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.
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
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
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
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
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
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
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
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
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
Hanford Waste Vitrification Plant quality assurance program description: Overview and applications
International Nuclear Information System (INIS)
Caplinger, W.H.
1990-12-01
This document describes the Hanford Waste Vitrification Plant Project Quality Assurance Program. This program is being implemented to ensure the acceptability of high-level radioactive canistered waste forms produced by the Hanford Waste Vitrification Plant for disposal in a licensed federal repository. The Hanford Waste Vitrification Plant Quality Assurance Program is comprised of this Quality Assurance Program Description as well as the associated contractors' quality assurance programs. The objective of this Quality Assurance Program Description is to provide the Hanford Waste Vitrification Plant Project participants with guidance and direction for program implementation while satisfying the US Department of Energy Office of Civilian Radioactive Waste Management needs in repository licensing activities with regard to canistered waste forms. To accomplish this objective, this description will be prepared in three parts: Part 1 - Overview and applications document; Part 2 - Development and qualification of the canistered waste form; Part 3 - Production of canistered waste forms. Part 1 describes the background, strategy, application, and content of the Hanford Waste Vitrification Plant Quality Assurance Program. This Quality Assurance Program Description, when complete, is designed to provide a level of confidence in the integrity of the canistered waste forms. 8 refs
Economic comparison of crystalline ceramic and glass waste forms for HLW disposal
International Nuclear Information System (INIS)
McKee, R.W.; Daling, P.M.; Wiles, L.E.
1983-05-01
A titanate-based, crystalline ceramic produced by hot isostatic pressing has been proposed as a potentially more stable and improved waste form for high-level nuclear waste disposal compared to the currently favored borosilicate glass waste form. This paper describes the results of a study to evaluate the relative costs for disposal of high-level waste from a 70,000 metric ton equivalent (MTE) system. The entire waste management system, including waste processing and encapsulation, transportation, and final repository disposal, was included in this analysis. The repository concept is based on the current basalt waste isolation project (BWIP) reference design. A range of design basis alternatives is considered to determine if this would influence the relative economics of the two waste forms. A thermal analysis procedure was utilized to define optimum canister sizes to assure that each waste form was compared under favorable conditions. Repository costs are found to favor the borosilicate glass waste form while transportation costs greatly favor the crystalline ceramic waste form. The determining component in the cost comparison is the waste processing cost, which strongly favors the borosilicate glass process because of its relative simplicity. A net cost advantage on the order of 12% to 15% on a waste management system basis is indicated for the glass waste form
Energy Technology Data Exchange (ETDEWEB)
B. A. Staples; T. P. O' Holleran
1999-05-01
The Department of Energy (DOE) has specific technical and documentation requirements for high-level waste (HLW) that is to be placed in a federal repository. This document describes in general terms the strategy to be used at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that vitrified HLW, if produced at the INEEL, meets these requirements. Waste form, canister, quality assurance, and documentation specifications are discussed. Compliance strategy is given, followed by an overview of how this strategy would be implemented for each specification.
International Nuclear Information System (INIS)
Staples, B. A.; O'Holleran, T. P.
1999-01-01
The Department of Energy (DOE) has specific technical and documentation requirements for high-level waste (HLW) that is to be placed in a federal repository. This document describes in general terms the strategy to be used at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that vitrified HLW, if produced at the INEEL, meets these requirements. Waste form, canister, quality assurance, and documentation specifications are discussed. Compliance strategy is given, followed by an overview of how this strategy would be implemented for each specification
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
International Nuclear Information System (INIS)
Bateman, K. J.; DiSanto, T.; Goff, K. M.; Johnson, S. G.; O'Holleran, T.; Riley, W. P. Jr.
1999-01-01
Argonne National Laboratory is developing a process for the conditioning of spent nuclear fuel to prepare the material for final disposal. Two waste streams will result from the treatment process, a stainless steel based form and a ceramic based form. The ceramic waste form will be enclosed in a stainless steel container. In order to assess the performance of the ceramic waste form in a repository two factors must be examined, the surface area increases caused by waste form cracking and any ceramic/canister interactions that may release toxic material. The results indicate that the surface area increases are less than the High Level Waste glass and any toxic releases are below regulatory limits
Radioactive waste disposal package
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.
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.)
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
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
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)
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)
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
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)
Review of metal-matrix encapsulation of solidified radioactive high-level waste
International Nuclear Information System (INIS)
Jardine, L.J.; Steindler, M.J.
1978-05-01
Literature describing previous and current work on the encapsulation of solidified high-level waste forms in a metal matrix was reviewed. Encapsulation of either stabilized calcine pellets or glass beads in alloys by casting techniques was concluded to be the most developed and direct approach to fabricating solid metal-matrix waste forms. Further characterizations of the physical and chemical properties of metal-matrix waste forms are still needed to assess the net attributes of metal-encapsulation alternatives. Steady-state heat transfer properties of waste canisters in air and water environments were calculated for four reference waste forms: (1) calcine, (2) glass monoliths, (3) metal-encapsulated calcine, and (4) metal-encapsulated glass beads. A set of criteria for the maximum allowable canister centerline and surface temperatures and heat generation rates per canister at the time of shipment to a Federal repository was assumed, and comparisons were made between canisters of these reference waste forms of the shortest time after reactor discharge that canisters could be filled and the subsequent ''interim'' storage times prior to shipment to a Federal repository for various canister diameters and waste ages. A reference conceptual flowsheet based on existing or developing technology for encapsulation of stabilized calcine pellets is discussed. Conclusions and recommendations are presented
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
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
Defense High Level Waste Disposal Container System Description Document
International Nuclear Information System (INIS)
Pettit, N. E.
2001-01-01
The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by
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
both of them) of applying the vibrational forces? 2) What is best mode of operation: first fill the canister with calcined waste and then vibrate it and refill it again, or apply vibrational forces during the filling process. By optimum or best we mean less creation of stress/strain forces during the volume reduction vibration process. Lessons learnt: This preliminary study shows that; 1) The maximum stress concentration always occurs in the canister wall, however its location varies and depends on the loading condition, and vibration process. 2) The proposed vibrational process would not cause any damages to the granulated calcined waste. 3) The first natural frequency of the longitudinal vibration of the canister is around 400 Hz, which is far away from the applied vibrational frequencies and from possibility of resonance phenomena that may cause damage to the canister 4) The relationship between the maximum internal stress and the frequency of the applied load is not parabolic. 5) The mechanical properties of the granulated calcined nuclear waste have small impact on the internal stress of the canister. Finally, the calculated data suggested that applying vibrational forces will keep the entire canister whole without any indication of development defects, and will have significant economical benefits of handling HLW and LLW in calcined forms, from waste manipulation, storage and transportation
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.)
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
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
External Criticality Risk of Immobilized Plutonium Waste Form in a Geologic Repository
International Nuclear Information System (INIS)
McClure, J.
2001-01-01
This purpose of this technical report is to provide a comprehensive summary of the waste package (WP) external criticality-related risk of the Plutonium Disposition ceramic waste form, which is being developed and evaluated by the Office of Fissile Materials Disposition of the United States Department of Energy (DOE). Potential accumulation of the fissile materials, 239 Pu and 235 U, in rock formations having a favorable chemical environment for such actions, requires analysis because autocatalytic configurations, while unlikely to form, never-the-less have consequences which are undesirable and require evaluation. Secondly, the WP design has evolved necessitating a re-evaluation of the internal WP degradation scenarios that contribute to the external source terms. The scope of this study includes a summary of the revised WP degradation calculations, a summary of the accumulation mechanisms in fractures and lithophysae in the tuff beneath the WP footprint, and a summary of the criticality risk calculations from any accumulated fissile material. Accumulations of fissile material external to the WP sufficient to pose a potential criticality risk require a deposition mechanism operating over sufficient time to reach required levels. The transporting solution concentrations themselves are well below critical levels (CRWMS 2001e). The ceramic waste form consists of Pu immobilized in ceramic disks, which would be embedded in High-Level Waste (HLW) glass in the standard HLW glass disposal canister. The ceramic disks would occupy approximately 12% of the HLW canister volume, while most of the remaining 88% of the volume would be occupied by HLW glass
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
Testing of high-level waste forms under repository conditions
International Nuclear Information System (INIS)
Mc Menamin, T.
1989-01-01
The workshop on testing of high-level waste forms under repository conditions was held on 17 to 21 October 1988 in Cadarache, France, and sponsored by the Commission of the European Communities (CEC), the Commissariat a l'energie atomique (CEA) and the Savannah River Laboratory (US DOE). Participants included representatives from Australia, Belgium, Denmark, France, Germany, Italy, Japan, the Netherlands, Sweden, Switzerland, The United Kingdom and the United States. The first part of the conference featured a workshop on in situ testing of simulated nuclear waste forms and proposed package components, with an emphasis on the materials interface interactions tests (MIIT). MIIT is a sevent-part programme that involves field testing of 15 glass and waste form systems supplied by seven countries, along with potential canister and overpack materials as well as geologic samples, in the salt geology at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico, USA. This effort is still in progress and these proceedings document studies and findings obtained thus far. The second part of the meeting emphasized multinational experimental studies and results derived from repository systems simulation tests (RSST), which were performed in granite, clay and salt environments
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
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
International Nuclear Information System (INIS)
1981-01-01
A discussion of the relative strengths and weaknesses of the alternative forms and recommendations for future program directions are presented in the body of this report. In addition to the relative ranking, the Peer Review Panel makes the following observations and recommendations: (1) Differences in overall performance of most of the uncoated waste forms are relatively small when compared under approximately equivalent conditions. (2) The increased scientific basis for this class of waste forms has not yet been sufficient to achieve reliably large improvements in waste form performance over the best borosilicate glasses. (3) The increased leach rates at elevated temperatures and the uncertainty regarding mechanisms of leaching under repository conditions continue to indicate that surface temperatures of waste canisters and especially any waste form-water interfaces should be restricted to less than 100 0 C, until more data is available to indicate otherwise. (4) Improvements are noteworthy, but there is still a need for adopting additional standardized tests, standard reference materials, common units and standardized methods of data presentation in the nuclear waste program. (5) Comparative data on leach rates in waters equilibrated with candidate rocks and potential geologic environments are almost non-existent and are essential to establish relevant long term extrapolation of waste form performance.(6) Understanding radiation damage effects on the microstructure and leaching mechanisms of polycrystalline ceramics is still insufficient to judge long term reliability of this class of waste forms. (7) More extensive data on rates and mechanisms of leaching of all waste forms under radiation and repository conditions are needed. (8) Additional studies of fundamental mechanisms controlling long term reliability of glass and alternative waste forms are strongly encouraged
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
Evaluation of lead-iron-phosphate glass as a high-level waste form
International Nuclear Information System (INIS)
Chick, L.A.; Bunnell, L.R.; Strachan, D.M.; Kissinger, H.E.; Hodges, F.N.
1986-01-01
The lead-iron-phosphate (Pb-Fe-P) nuclear waste glass developed at Oak Ridge National Laboratory (ORNL) was evaluated for its potential as an improvement over the current reference waste form, borosilicate (B-Si) glass. Vitreous Pb-Fe-P glass appears to have substantially better chemical durability than B-Si glass. However, severe crystallization leading to deteriorated chemical durability would result if this glass were poured into large canisters, as is presently done with B-Si glass. Cesium leach rates from this crystallized material are orders of magnitude greater than those from B-Si glass. Therefore, to realize the performance advantages of the Pb-Fe-P material in a nuclear waste form, it would be necessary to process it so that it is cooled rapidly, thus retaining its vitreous structure
Evaluation of lead-iron-phosphate glass as a high-level waste form
International Nuclear Information System (INIS)
Chick, L.A.; Bunnell, L.R.; Strachan, D.M.; Kissinger, H.E.; Hodges, F.N.
1986-01-01
The lead-iron-phosphate nuclear waste glass developed at Oak Ridge National Laboratory (ORNL) was evaluated for its potential as an improvement over the current reference waste form, borosilicate glass. Vitreous lead-iron-phosphate glass appears to have substantially better chemical durability than borosilicate glass. However, severe crystallization leading to deteriorated chemical durability would result if this glass were poured into large canisters as is presently done with borosilicate glass. Cesium leach rates from this crystallized material are orders of magnitude greater than those from borosilicate glass. Therefore, in order to realize the performance advantages of the lead-iron-phosphate material in a nuclear waste form, it would be necessary to process it so that it is rapidly cooled, thus retaining its vitreous structure. 22 refs., 4 figs., 4 tabs
Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste
International Nuclear Information System (INIS)
Ebert, W. L.; Snyder, C. T.; Frank, Steven; Riley, Brian
2016-01-01
loading from about 12% to 10% on a mass basis, but this will not significantly impact the waste loading on a volume basis. It is likely that heat output will limit the amount of waste salt that can be accommodated in a waste canister rather than the salt loading in an ACWF, and that the increase from 8 mass% to about 10 mass% salt loadings in ACWF materials will be sufficient to optimize these waste forms. Although the waste salt composition used in this study contained a moderate amount of NaCl, the test results suggest waste salts with little or no NaCl can be accommodated in ACWF materials by using the new binder glass, albeit at waste loadings lower than 8 mass%. The higher glass contents that will be required for ACWF materials made with salt wastes that do not contain NaCl are expected to result in much lower porosities in those waste forms.
Designing Advanced Ceramic Waste Forms for Electrochemical Processing Salt Waste
Energy Technology Data Exchange (ETDEWEB)
Ebert, W. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Snyder, C. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Frank, Steven [Argonne National Lab. (ANL), Argonne, IL (United States); Riley, Brian [Argonne National Lab. (ANL), Argonne, IL (United States)
2016-03-01
the waste loading from about 12% to 10% on a mass basis, but this will not significantly impact the waste loading on a volume basis. It is likely that heat output will limit the amount of waste salt that can be accommodated in a waste canister rather than the salt loading in an ACWF, and that the increase from 8 mass% to about 10 mass% salt loadings in ACWF materials will be sufficient to optimize these waste forms. Although the waste salt composition used in this study contained a moderate amount of NaCl, the test results suggest waste salts with little or no NaCl can be accommodated in ACWF materials by using the new binder glass, albeit at waste loadings lower than 8 mass%. The higher glass contents that will be required for ACWF materials made with salt wastes that do not contain NaCl are expected to result in much lower porosities in those waste forms.
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
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)
Strategy for Qualification of the Plutonium Immobilized Form
International Nuclear Information System (INIS)
Marra, J.C.; Marra, S.L.; Bibler, N.E.; Strachan, D.M.; Shaw, H.F.
1998-01-01
In order to dispose of a radioactive waste form in a federal high- level waste repository, a waste form qualification strategy and program must be developed. The waste form qualification program must include the acceptance specifications for the product, show how compliance with these specifications will be met, and then demonstrate compliance. An important element of this program is developing a measure to demonstrate product consistency. For the can- in-canister option, waste form qualification is needed not only for the Pu immobilized form but also for the high-level waste glass canister containing the Pu waste form. The latter will require close interaction and coordination with the U.S. Department of Energy - Office of Environmental Management (DOE-EM) who, through their contractor at the Savannah River Site (SRS), is the producer of the high-level waste form at the Defense Waste Processing Facility (DWPF). In this paper a waste form qualification strategy for the plutonium ceramic form is described that utilizes, as much as possible, the qualification strategy successfully used for vitrified high-level waste
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.
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
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
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
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
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
Identification of items and activities important to waste form acceptance by Westinghouse GoCo sites
International Nuclear Information System (INIS)
Plodinec, M.J.; Marra, S.L.; Dempster, J.; Randklev, E.H.
1993-01-01
The Department of Energy has established specifications (Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms, or WAPS) for canistered waste forms produced at Hanford, Savannah River, and West Valley. Compliance with these specifications requires that each waste form producer identify the items and activities which must be controlled to ensure compliance. As part of quality assurance oversight activities, reviewers have tried to compare the methodologies used by the waste form producers to identify items and activities important to waste form acceptance. Due to the lack of a documented comparison of the methods used by each producer, confusion has resulted over whether the methods being used are consistent. This confusion has been exacerbated by different systems of nomenclature used by each producer, and the different stages of development of each project. The waste form producers have met three times in the last two years, most recently on June 28, 1993, to exchange information on each producer's program. These meetings have been sponsored by the Westinghouse GoCo HLW Vitrification Committee. This document is the result of this most recent exchange. It fills the need for a documented comparison of the methodologies used to identify items and activities important to waste form acceptance. In this document, the methodology being used by each waste form producer is summarized, and the degree of consistency among the waste form producers is determined
Evaluation of lead-iron-phosphate glass as a high-level waste form
International Nuclear Information System (INIS)
Chick, L.A.; Bunnell, L.R.; Strachan, D.M.; Kissinger, H.E.; Hodges, F.N.
1986-09-01
The lead-iron-phosphate (Pb-Fe-P) glass developed at Oak Ridge National Laboratory was evaluated for its potential as an improvement over the current reference nuclear waste form, borosilicate (B-Si) glass. The evaluation was conducted as part of the Second Generation HLW Technology Subtask of the Nuclear Waste Treatment Program at Pacific Northwest Laboratory. The purpose of this work was to investigate possible alternatives to B-Si glass as second-generation waste forms. While vitreous Pb-Fe-P glass appears to have substantially better chemical durability than B-Si glass, severe crystallization or devitrification leading to deteriorated chemical durability would result if this glass were poured into large canisters as is the procedure with B-Si glass. Cesium leach rates from this crystallized material are orders of magnitude greater than those from B-Si glass. Therefore, to realize the potential performance advantages of the Pb-Fe-P material in a nuclear waste form, the processing method would have to cool the material rapidly to retain its vitreous structure
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)
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
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
International Nuclear Information System (INIS)
El Mrabet, Said; Castro, Miguel A.; Orta, M. Mar; Pazos, M. Carolina; Alba, Maria D.; Astudillo, Julio; Rueda, Silvia; Hurtado, Santiago; Villa, Mara
2012-01-01
Document available in extended abstract form only. The disposal of high level radioactive wastes (HLW) such as spent fuel or reprocessing waste resulting from the operation and dismantling of nuclear reactors is one of the most problems facing the worlds because of its long half life and radionuclide migration to the biosphere. For long term performance assessment of radioactive waste disposal, knowledge concerning radionuclide retention processes on materials composing the engineered barrier (clay and container waste) is required. Steel waste containers and bentonite have been proposed as candidate materials for overpack and buffer respectively in most of the proposed repositories designs for nuclear waste disposal. This contribution aims to study the competitiveness of the bentonite and the metallic canister in the retention process of some kinds of radioactive waste such as 152 Eu. The europium was chosen because it is a toxic metal and usually taken as a simulator of the trivalent high level radioactive waste. In order to elucidate the mechanisms involved in the retention processes of europium by both bentonite and metallic canister, a cylindrical steel mini-reactor was designed and prepared from the same material as the steel reactor AISI-316L. The bentonite was then introduced and compacted within the mini-reactor forming a set mini-reactor- bentonite. The system mini-reactor -bentonite was then subjected to hydrothermal treatments at 300 deg. C for 4.5 days. The morphology and chemical composition of both steel and bentonite were analyzed by XRD and SEM. SEM and XRD results revealed that both the bentonite and the metallic canister were involved in sorption mechanism of europium by the formation of insoluble phases of europium silicates originated from the mixed solution of bentonite, Eu 3+ and canister. The pH-Redox potential (Eh) indicated that the interlayer cations of bentonite were replaced by Eu 3+ with higher acidity and Eh which means that the active
Defense High Level Waste Disposal Container System Description Document
International Nuclear Information System (INIS)
2000-01-01
The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials
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
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
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
High level waste forms: glass marbles and thermal spray coatings
International Nuclear Information System (INIS)
Treat, R.L.; Oma, K.H.; Slate, S.C.
1982-01-01
A process that converts high-level waste to glass marbles and then coats the marbles has been developed at Pacific Northwest Laboratory (PNL) under sponsorship of the US Department of Energy. The process consists of a joule-heated glass melter, a marble-making device based on a patent issued to Corning Glass Works, and a coating system that includes a plasma spray coater and a marble tumbler. The process was developed under the Alternative Waste Forms Program which strived to improve upon monolithic glass for immobilizing high-level wastes. Coated glass marbles were found to be more leach-resistant, and the marbles, before coating were found to be very homogeneous, highly impact resistant, and conductive to encapsulation in a metal matric for improved heat transfer and containment. Marbles are also ideally suited for quality assurance and recycling. However, the marble process is more complex, and marbles require a larger number of canisters for waste containment and have a higher surface area than do glass monoliths
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
International Nuclear Information System (INIS)
Vogler, S.; Trevorrow, L.E.; Ziegler, A.A.; Steindler, M.J.
1981-08-01
Techniques for isolating and solidifying the nuclear wastes in the storage tanks at the Western New York Nuclear Service Center plant have been examined. One technique involves evaporating the water and forming a molten salt containing the precipitated sludge. The salt is allowed to solidify and is stored in canisters until processing into a final waste form is to be done. Other techniques involve calcining the waste material, then agglomerating the calcine with sodium silicate to reduce its dispersibility. This option can also involve a prior separation and decontamination of the supernatant salt. The sludge and all resins containing fission-product activity are then calcined together. The technique of removing the water and solidifying the salt may be the simplest method for removing the waste from the West Valley Plant
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
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
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
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)
Hydroxylated ceramic waste forms and the absurdity of leach tests
Energy Technology Data Exchange (ETDEWEB)
Roy, R; Odoj, R; Merz, E [eds.
1981-06-01
The repository pressure and temperature conditions during the thermal period projected in US repositories have been drastically lowered in the last year or two to new values of say 175 +- 50/sup 0/K. Using the argument that the evidence from natural models indicates the most stable mineral (= ceramic) hosts for radionuclides, one finds that under these new repository conditions such crystalline assemblages would be micas, clays, zeolites and other hydrated minerals, plus the tetravalent anhydrous oxide families. A waste form consisting of specific hydroxylated candidate phases can be made via a simple in-can technology (demonstrated by Oak Ridge) by reacting liquid wastes with precursor gels or phyllo or tektosilicates at <200/sup 0/C under modest pressure within the final disposal canister. The data on the rate of reaction of typical oxide materials to yield hydroxylated phases under these conditions show that the typical leach test (at 25 to 100/sup 0/C in deionized water) does not provide a simulation of the reactions which will occur. Hence such tests are not only totally meaningless with respect to qualifying a waste form for its role in a repository, they can be downright misleading.
Hydroxylated ceramic waste forms and the absurdity of 'leach tests'
Energy Technology Data Exchange (ETDEWEB)
Roy, R; Odoj, R; Merz, E [eds.
1981-06-01
The repository pressure and temperature conditions during the thermal period projected in U.S. repositories have been drastically lowered in the last year or two to new values of say 175 +- 50 K. Using the argument that the evidence from natural models indicates the most stable mineral (= ceramic) hosts for radionuclides, one finds that under these new repository conditions such crystalline assemblages would be micas, clays, zeolites, and other hydrated minerals, plus the tetravalent anhydrous oxide families. A waste form consisting of specific hydroxylated candidate phase can be made via a simple in-can technology (demonstrated by Oak Ridge) by reacting liquid wastes with precursor gels or phyllo or tektosilicates at <200/sup 0/C under modest pressure within the final disposal canister. The data on the rate of reaction of typical oxide materials to yield hydroxylated phases under these conditions show that the typical leach test (at 25-100/sup 0/C in deionized water) does not provide a simulation of the reactions which will occur. Hence such tests are not only totally meaningless with respect to qualifying a waste form for its role in a repository, they can be downright misleading.
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
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)
Preliminary formulation studies for a ''hydroceramic'' alternative waste form for INEEL HLW
International Nuclear Information System (INIS)
Siemer, D.D.; Gougar, M.L.D.; Grutzeck, M.W.; Scheetz, B.E.
1999-01-01
Herein the authors discuss scoping studies performed to develop an efficient way to prepare the Idaho National Engineering and Environmental Laboratory (INEEL) nominally high-level (∼40 W/m 3 ) calcined radioactive waste (HLW) and liquid metal (sodium) reactor coolants for disposal. The investigated approach implements the chemistry of Hanford's cancrinite-making clay reaction process via Oak Ridge National Laboratory's (ORNL's) formed-under-elevated-temperatures-and-pressures concrete monolith-making technology to make hydroceramics (HCs). The HCs differ from conventional Portland cement/blast furnace slag (PC/BFS) grouts in that the binder minerals formed during the curing process are hydrated alkali-aluminosilicates (feldspathoids-sodalites, cancrinites, and zeolites) rather than hydrated calcium silicates (CSH). This is desirable because (a) US defense-type radioactive wastes generally contain much more sodium and aluminum than calcium; (b) sodalites/cancrinites do a much better job of retaining the anionic components of real radioactive waste (e.g., nitrate) than do calcium silicates; (c) natural feldspathoids form from glasses (and therefore are more stable) in that region of the United States where a repository for this sort of waste could be sited; and (d) if eventually deemed necessary, feldspathoid-type concrete wasteforms could be hot-isostatically-pressed into even more durable materials without removing them from their original canisters
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
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
International program to study subseabed disposal of high-level radioactive wastes
International Nuclear Information System (INIS)
Carlin, E.M.; Hinga, K.R.; Knauss, J.A.
1984-01-01
This report provides an overview of the international program to study seabed disposal of nuclear wastes. Its purpose is to inform legislators, other policy makers, and the general public as to the history of the program, technological requirements necessary for feasibility assessment, legal questions involved, international coordination of research, national policies, and research and development activities. Each of these major aspects of the program is presented in a separate section. The objective of seabed burial, similar to its continental counterparts, is to contain and to isolate the wastes. The subseabed option should not be confuesed with past practices of ocean dumping which have introduced wastes into ocean waters. Seabed disposal refers to the emplacement of solidified high-level radioactive waste (with or without reprocessing) in certain geologically stable sediments of the deep ocean floor. Specially designed surface ships would transport waste canisters from a port facility to the disposal site. Canisters would be buried from a few tens to a few hundreds of meters below the surface of ocean bottom sediments, and hence would not be in contact with the overlying ocean water. The concept is a multi-barrier approach for disposal. Barriers, including waste form, canister, ad deep ocean sediments, will separate wastes from the ocean environment. High-level wastes (HLW) would be stabilized by conversion into a leach-resistant solid form such as glass. This solid would be placed inside a metallic canister or other type of package which represents a second barrier. The deep ocean sediments, a third barrier, are discussed in the Feasibility Assessment section. The waste form and canister would provide a barrier for several hundred years, and the sediments would be relied upon as a barrier for thousands of years. 62 references, 3 figures, 2 tables
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.
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
Hsu, Jen-Hsien; Bai, Jincheng; Kim, Cheol-Woon; Brow, Richard K.; Szabo, Joe; Zervos, Adam
2018-03-01
The effects of cooling rate on the chemical durability of iron phosphate waste forms containing up to 40 wt% of a high MoO3 Collins-CLT waste simulant were determined at 90 °C using the product consistency test (PCT). The waste form, designated 40wt%-5, meets appropriate Department of Energy (DOE) standards when rapidly quenched from the melt (as-cast) and after slow cooling following the CCC (canister centerline cooling)-protocol, although the quenched glass is more durable. The analysis of samples from the vapor hydration test (VHT) and the aqueous corrosion test (differential recession test) reveals that rare earth orthophosphate (monazite) and Zr-pyrophosphate crystals that form on cooling are more durable than the residual glass in the 40wt%-5 waste form. The residual glass in the CCC-treated samples has a greater average phosphate chain length and a lower Fe/P ratio, and those contribute to its faster corrosion kinetics.
International Nuclear Information System (INIS)
Wicks, G.G.; Lodding, A.R.; Macedo, P.B.; Molecke, M.A.
1989-01-01
The first in-situ tests involving burial of simulated high-level waste (HLW) forms conducted in the United States were started on July 22, 1986. This effort, called the Materials Interface Interactions Tests (MIIT), comprises the largest, most cooperative field testing venture in the international waste management community. Included in the study are over 900 waste form samples comprising 15 different systems supplied by seven countries. Also included are almost 300 potential canister or overpack metal samples of 11 different metals along with more than 500 geologic and backfill specimens. There are a total of 1926 relevant interactions that characterize this effort which is being conducted in the bedded salt site at the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico
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
Densified waste form and method for forming
Garino, Terry J.; Nenoff, Tina M.; Sava Gallis, Dorina Florentina
2015-08-25
Materials and methods of making densified waste forms for temperature sensitive waste material, such as nuclear waste, formed with low temperature processing using metallic powder that forms the matrix that encapsulates the temperature sensitive waste material. The densified waste form includes a temperature sensitive waste material in a physically densified matrix, the matrix is a compacted metallic powder. The method for forming the densified waste form includes mixing a metallic powder and a temperature sensitive waste material to form a waste form precursor. The waste form precursor is compacted with sufficient pressure to densify the waste precursor and encapsulate the temperature sensitive waste material in a physically densified matrix.
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
DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER
International Nuclear Information System (INIS)
Radulesscu, G.; Tang, J.S.
2000-01-01
The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M andO [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M andO 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M andQ 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M andO 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this
DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER
Energy Technology Data Exchange (ETDEWEB)
G. Radulesscu; J.S. Tang
2000-06-07
The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable
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
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.
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
Permitting plan for the high-level waste interim storage
International Nuclear Information System (INIS)
Deffenbaugh, M.L.
1997-01-01
This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist
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
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
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
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
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)
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)
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
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
Parametric study of the effects of thermal environment on a waste package for a tuff repository
Energy Technology Data Exchange (ETDEWEB)
Johnstone, J K; Sundberg, W D; Krumhansl, J L [Sandia National Laboratories Albuquerque, NM, (USA)
1982-12-31
The thermal environment has been modeled in a simple reference waste package in a tuff repository for a variety of variables. The waste package was composed of the waste form, canister, overpack and backfill. The emplacement hole was 122cm dia. Waste forms used in the calculations were commercial high level waste (CHLW) and spent fuel (SF). Canister loadings varied from 50 to 100 kW/acre. Primary attention was focused on the backfill behavior in the thermal and chemical environment. Results are related to the maximum temperature calculated for the backfill. These calculations raise serious concerns about the effectiveness of the backfill within the context of the total waste package.
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
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)
Using process instrumentation to obviate destructive examination of canisters of HLW glass
International Nuclear Information System (INIS)
Kuhn, W.L.; Slate, S.C.
1983-01-01
An important concern of a manufacturer of packages of solidified high-level waste (HLW) is quality assurance of the waste form. The vitrification of HLW as a borosilicate glass is considered, and, based on a reference vitrification process, it is proposed that information from process instrumentation may be used to assure quality without the need for additional information obtained by destructive examining (core drilling) canisters of glass. This follows mainly because models of product performance and process behavior must be previously established in order to confidently select the desired glass formulation, and to have confidence that the process is well enough developed to be installed and operated in a nuclear facility
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
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
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
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
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.
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
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
WIPP waste package testing on simulated DHLW: emplacement
International Nuclear Information System (INIS)
Molecke, M.A.
1984-01-01
Several series of simulated (nonradioactive) defense high-level waste (DHLW) package tests have been emplaced in the WIPP, a research and development facility authorized to demonstrate the safe disposal of defense-related wastes. The primary purpose of these 3-to-7 year duration tests is to evaluate the in situ materials performance of waste package barriers (canisters, overpacks, backfills, and nonradioactive DHLW glass waste form) for possible future application to a licensed waste repository in salt. This paper describes all test materials, instrumentation, and emplacement and testing techniques, and discusses progress of the various tests. These tests are intended to provide information on materials behavior (i.e., corrosion, metallurgical and geochemical alterations, waste form durability, surface interactions, etc.), as well as comparison between several waste package designs, fabrications details, and actual costs. These experiments involve 18 full-size simulated DHLW packages (approximately 3.0 m x 0.6 m diameter) emplaced in vertical boreholes in the salt drift floor. Six of the test packages contain internal electrical heaters (470 W/canister), and were emplace under approximately reference DHLW repository conditions. Twelve other simulated DHLW packages were emplaced under accelerated-aging or overtest conditions, including the artificial introduction of brine, and a thermal loading approximately three to four times higher than reference. Eight of these 12 test packages contain 1500 W/canister electrical heaters; the other four are filled with DHLW glass. 9 refs., 1 fig
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
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
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)
Desludging of N Reactor fuel canisters: Analysis, Test, and data requirements
International Nuclear Information System (INIS)
Johnson, A.B. Jr.
1996-01-01
The N Reactor fuel is currently stored in canisters in the K East (KE) and K West (KW) Basins. In KE, the canisters have open tops; in KW, the cans have sealed lids, but are vented to release gases. Corrosion products have formed on exposed uranium metal fuel, on carbon steel basin component surfaces, and on aluminum alloy canister surfaces. Much of the corrosion product is retained on the corroding surfaces; however, large inventories of particulates have been released. Some of the corrosion product particulates form sludge on the basin floors; some particulates are retained within the canisters. The floor sludge inventories are much greater in the KE Basin than in the KW Basin because KE Basin operated longer and its water chemistry was less controlled. Another important factor is the absence of lids on the KE canisters, allowing uranium corrosion products to escape and water-borne species, principally iron oxides, to settle in the canisters. The inventories of corrosion products, including those released as particulates inside the canisters, are only beginning to be characterized for the closed canisters in KW Basin. The dominant species in the KE floor sludge are oxides of aluminum, iron, and uranium. A large fraction of the aluminum and uranium floor sludge particulates may have been released during a major fuel segregation campaign in the 1980s, when fuel was emptied from 4990 canisters. Handling and jarring of the fuel and aluminum canisters seems likely to have released particulates from the heavily corroded surfaces. Four candidate methods are discussed for dealing with canister sludge emerged in the N Reactor fuel path forward: place fuel in multi-canister overpacks (MCOs) without desludging; drill holes in canisters and drain; drill holes in canisters and flush with water; and remove sludge and repackage the fuel
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
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
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
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.
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
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.
Secondary Waste Cast Stone Waste Form Qualification Testing Plan
Energy Technology Data Exchange (ETDEWEB)
Westsik, Joseph H.; Serne, R. Jeffrey
2012-09-26
The Hanford Tank Waste Treatment and Immobilization Plant (WTP) is being constructed to treat the 56 million gallons of radioactive waste stored in 177 underground tanks at the Hanford Site. The WTP includes a pretreatment facility to separate the wastes into high-level waste (HLW) and low-activity waste (LAW) fractions for vitrification and disposal. The LAW will be converted to glass for final disposal at the Integrated Disposal Facility (IDF). Cast Stone – a cementitious waste form, has been selected for solidification of this secondary waste stream after treatment in the ETF. The secondary-waste Cast Stone waste form must be acceptable for disposal in the IDF. This secondary waste Cast Stone waste form qualification testing plan outlines the testing of the waste form and immobilization process to demonstrate that the Cast Stone waste form can comply with the disposal requirements. Specifications for the secondary-waste Cast Stone waste form have not been established. For this testing plan, Cast Stone specifications are derived from specifications for the immobilized LAW glass in the WTP contract, the waste acceptance criteria for the IDF, and the waste acceptance criteria in the IDF Permit issued by the State of Washington. This testing plan outlines the testing needed to demonstrate that the waste form can comply with these waste form specifications and acceptance criteria. The testing program must also demonstrate that the immobilization process can be controlled to consistently provide an acceptable waste form product. This testing plan also outlines the testing needed to provide the technical basis for understanding the long-term performance of the waste form in the disposal environment. These waste form performance data are needed to support performance assessment analyses of the long-term environmental impact of the secondary-waste Cast Stone waste form in the IDF
TSA waste stream and final waste form composition
International Nuclear Information System (INIS)
Grandy, J.D.; Eddy, T.L.; Anderson, G.L.
1993-01-01
A final vitrified waste form composition, based upon the chemical compositions of the input waste streams, is recommended for the transuranic-contaminated waste stored at the Transuranic Storage Area of the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. The quantities of waste are large with a considerable uncertainty in the distribution of various waste materials. It is therefore impractical to mix the input waste streams into an ''average'' transuranic-contaminated waste. As a result, waste stream input to a melter could vary widely in composition, with the potential of affecting the composition and properties of the final waste form. This work examines the extent of the variation in the input waste streams, as well as the final waste form under conditions of adding different amounts of soil. Five prominent Rocky Flats Plant 740 waste streams are considered, as well as nonspecial metals and the ''average'' transuranic-contaminated waste streams. The metals waste stream is the most extreme variation and results indicate that if an average of approximately 60 wt% of the mixture is soil, the final waste form will be predominantly silica, alumina, alkaline earth oxides, and iron oxide. This composition will have consistent properties in the final waste form, including high leach resistance, irrespective of the variation in waste stream. For other waste streams, much less or no soil could be required to yield a leach resistant waste form but with varying properties
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
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.
International Nuclear Information System (INIS)
Hoekmark, Harald; Faelth, Billy
2003-12-01
The report addresses the problem of the minimum spacing required between neighbouring canisters in the deep repository. That spacing is calculated for a number of assumptions regarding the conditions that govern the temperature in the nearfield and at the surfaces of the canisters. The spacing criterion is that the temperature at the canister surfaces must not exceed 100 deg C .The results are given in the form of nomographic charts, such that it is in principle possible to determine the spacing as soon as site data, i.e. the initial undisturbed rock temperature and the host rock heat transport properties, are available. Results of canister spacing calculations are given for the KBS-3V concept as well as for the KBS-3H concept. A combination of numerical and analytical methods is used for the KBS-3H calculations, while the KBS-3V calculations are purely analytical. Both methods are described in detail. Open gaps are assigned equivalent heat conductivities, calculated such that the conduction across the gaps will include also the heat transferred by radiation. The equivalent heat conductivities are based on the emissivities of the different gap surfaces. For the canister copper surface, the emissivity is determined by back-calculation of temperatures measured in the Prototype experiment at Aespoe HRL. The size of the different gaps and the emissivity values are of great importance for the results and will be investigated further in the future
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
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
Post treatment of high-level nuclear fuel wastes
International Nuclear Information System (INIS)
Berreth, J.R.; Cole, H.S.; Hoskins, A.P.; Lewis, L.C.; Samsel, E.G.
1975-01-01
The glass-ceramic product prepared from fluidized-bed calcined synthetic commercial wastes, based on data obtained to date, has many of the properties desired for long-term storage. Although more characterization is necessitated, the product's high-calcine content will decrease the number of storage canisters required and use a minimum of product-forming additives, resulting in significant process cost savings. The product remains in a solid, nonflowing form at temperatures close to the preparation temperature and yet is prepared at relatively low temperatures. The product has void spaces to accommodate radiolytic gas formation, but is hard and dense and has very low leach rates. Process features, such as no direct product contact with furnace or storage canisters, will minimize corrosion of both process equipment and storage canisters
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
Transuranic waste management program waste form development
International Nuclear Information System (INIS)
Bennett, W.S.; Crisler, L.R.
1981-01-01
To ensure that all technology necessary for long term management of transuranic (TRU) wastes is available, the Department of Energy has established the Transuranic Waste Management Program. A principal focus of the program is development of waste forms that can accommodate the very diverse TRU waste inventory and meet geologic isolation criteria. The TRU Program is following two approaches. First, decontamination processes are being developed to allow removal of sufficient surface contamination to permit management of some of the waste as low level waste. The other approach is to develop processes which will allow immobilization by encapsulation of the solids or incorporate head end processes which will make the solids compatible with more typical waste form processes. The assessment of available data indicates that dewatered concretes, synthetic basalts, and borosilicate glass waste forms appear to be viable candidates for immobilization of large fractions of the TRU waste inventory in a geologic repository
Testing and evaluation of solidified high-level waste forms. Joint annual progress report 1983
International Nuclear Information System (INIS)
Malow, G.
1985-01-01
A second joint programme of the European Atomic Community was started in 1981 under the indirect action programme (1980-84), Action No 5 'Testing and evaluation of the properties of various potential materials for immobilizing high activity waste'. The overall objective of the research is to test various European potential solidified high-level radioactive waste forms so as to predict their behaviour after disposal. The most important aspect is to produce data to calculate the activity release from the waste products under the attack of various aqueous solutions. The experiments were partly performed under waste repository relevant conditions and partly under simplified conditions for investigating basic activity release mechanisms. The topics of the programme were: (i) studies of basic leaching mechanisms; (ii) studies of hydrothermal leaching and surface attack of waste glasses; (iii) leach test carried out in contact with granite at low water flow rates; (iv) static leach tests with specimen surrounded by canister and backfill materials; (v) specific isotope leach tests in slowly flowing water; (vi) leach test of actinide spiked samples; (vii) leach tests of highly radioactive samples; (viii) leach tests of alpha radiation stability; (ix) studies of mechanical stability; (x) studies of mineral phases as model compounds and phase relations
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
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
Storage of nuclear waste in long boreholes
International Nuclear Information System (INIS)
Sandstedt, H.; Wichmann, C.; Pusch, R.; Boergesson, L.; Loennerberg, B.
1991-08-01
This report constitutes a feasibility study for the storage of high level radioactive waste in long TBM drilled tunnels. The report will form the basis for a comparison with other concepts in future analysis of the isolation performance in a typical Swedish rock structure. The suggested repository concept consists of three parallel, 4.5 km long, horizontal tunnels at a depth of 500 m constructed using TBM technology. The tunnel diameter will be about 2.4 m for deployment of canisters with a diameter of 1.6 m. The space between the canisters and rock will be totally sealed off by bentonite. The study comprises the design of canisters, canister handling and deposition, near field design, near field sealing and behaviour, and technical design of the repository. The report also includes a tentative time schedule and cost estimate, incorporating the construction phase and deployment of canisters. (au)
Progress in the understanding of the long-term corrosion behaviour of copper canisters
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
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
Liquid secondary waste: Waste form formulation and qualification
Energy Technology Data Exchange (ETDEWEB)
Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)
2017-07-31
The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity
International Nuclear Information System (INIS)
Wald, J.W.; Lokken, R.O.; Shade, J.W.; Rusin, J.M.
1980-12-01
A number of alternative process and waste form options exist for the immobilization of nuclear wastes. Although data exists on the characterization of these alternative waste forms, a straightforward comparison of product properties is difficult, due to the lack of standardized testing procedures. The characterization study described in this report involved the application of the same volatility, mechanical strength and leach tests to ten alternative waste forms, to assess product durability. Bulk property, phase analysis and microstructural examination of the simulated products, whose waste loading varied from 5% to 100% was also conducted. The specific waste forms investigated were as follows: Cold Pressed and Sintered PW-9 Calcine; Hot Pressed PW-9 Calcine; Hot Isostatic Pressed PW-9 Calcine; Cold Pressed and Sintered SPC-5B Supercalcine; Hot Isostatic pressed SPC-5B Supercalcine; Sintered PW-9 and 50% Glass Frit; Glass 76-68; Celsian Glass Ceramic; Type II Portland Cement and 10% PW-9 Calcine; and Type II Portland Cement and 10% SPC-5B Supercalcine. Bulk property data were used to calculate and compare the relative quantities of waste form volume produced at a spent fuel processing rate of 5 metric ton uranium/day. This quantity ranged from 3173 L/day (5280 Kg/day) for 10% SPC-5B supercalcine in cement to 83 L/day (294 Kg/day) for 100% calcine. Mechanical strength, volatility, and leach resistance tests provide data related to waste form durability. Glass, glass-ceramic and supercalcine ranked high in waste form durability where as the 100% PW-9 calcine ranked low. All other materials ranked between these two groupings
Crystallization behavior of nuclear waste forms
International Nuclear Information System (INIS)
Rusin, J.M.; Lokken, R.O.; May, R.P.; Wald, J.W.
1981-09-01
Several waste form options have been or are being developed for the immobilization of high-level wastes. The final selection of a waste form must take into consideration both waste form product as well as process factors. Crystallization behavior has an important role in nuclear waste form technology. For glass or vitreous waste forms, crystallization is generally controlled to a minimum by appropriate glass formulation and heat treatment schedules. With glass ceramic waste forms, crystallization is essential to convert glass products to highly crystalline waste forms with a minimum residual glass content. In the case of ceramic waste forms, additives and controlled sintering schedules are used to contain the radionuclides in specific tailored crystalline phases
International Nuclear Information System (INIS)
Reimus, P.W.; Kuhn, W.L.; Peters, R.D.; Pulsipher, B.A.
1986-07-01
During fiscal year 1982, the US Department of Energy (DOE) assigned responsibility for managing civilian nuclear waste treatment programs in the United States to the Nuclear Waste Treatment Program (NWTP) at the Pacific Northwest Laboratory (PNL). One of the principal objectives of this program is to establish relationships between vitrification process control and glass quality. Users of the liquid-fed ceramic melter (LFCM) process will need such relationships in order to establish acceptance of vitrified high-level nuclear waste at a licensed federal repository without resorting to destructive examination of the canisters. The objective is to be able to supply a regulatory agency with an estimate of the composition, durability, and integrity of the glass in each waste glass canister produced from an LFCM process simply by examining the process data collected during the operation of the LFCM. The work described here will continue through FY-1987 and culminate in a final report on the ability to control and monitor an LFCM process through sampling and process control charting of the LFCM feed system
International Nuclear Information System (INIS)
Neilson, R.M. Jr.; Colombo, P.
1982-01-01
Contemporary solidification agents are being investigated relative to their applications to major fuel cycle and non-fuel cycle low-level waste (LLW) streams. Work is being conducted to determine the range of conditions under which these solidification agents can be applied to specific LLW streams. These studies are directed primarily towards defining operating parameters for both improved solidification of problem wastes and solidification of new LLW streams generated from advanced volume reduction technologies. Work is being conducted to measure relevant waste form properties. These data will be compiled and evaluated to demonstrate compliance with waste form performance and shallow land burial acceptance criteria and transportation requirements
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
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)
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
Liquid secondary waste. Waste form formulation and qualification
Energy Technology Data Exchange (ETDEWEB)
Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); King, W. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)
2016-03-01
The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during Site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the Integrated Disposal Facility IDF). Waste form testing to support this plan is composed of work in the near term to demonstrate the waste form will provide data as input to a performance assessment (PA) for Hanford’s IDF.
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
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
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
Mixed Waste Focus Area - Waste form initiative
International Nuclear Information System (INIS)
Nakaoka, R.; Waters, R.; Pohl, P.; Roach, J.
1998-01-01
The mission of the US Department of Energy's (DOE) Mixed Waste Focus Area (MWFA) is to provide acceptable technologies that enable implementation of mixed waste treatment systems which are developed in partnership with end-users, stakeholders, tribal governments, and regulators. To accomplish this mission, a technical baseline was established in 1996 and revised in 1997. The technical baseline forms the basis for determining which technology development activities will be supported by the MWFA. The primary attribute of the technical baseline is a set of prioritized technical deficiencies or roadblocks related to implementation of mixed waste treatment systems. The Waste Form Initiative (WFI) was established to address an identified technical deficiency related to waste form performance. The primary goal of the WFI was to ensure that the mixed low-level waste (MLLW) treatment technologies being developed, currently used, or planned for use by DOE would produce final waste forms that meet the waste acceptance criteria (WAC) of the existing and/or planned MLLW disposal facilities. The WFI was limited to an evaluation of the disposal requirements for the radioactive component of MLLW. Disposal requirements for the hazardous component are dictated by the Resource Conservation and Recovery Act (RCRA), and were not addressed. This paper summarizes the technical basis, strategy, and results of the activities performed as part of the WFI
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)
Backfill barriers for nuclear waste repositories in salt
Energy Technology Data Exchange (ETDEWEB)
Nowak, E J; Odoj, R; Merz, E [eds.
1981-06-01
Backfill mixtures surrounding the waste form and canister can provide a neutral or slightly acidic, potentially reducing environment, prevent convective aqueous flow, and act as an effective radionuclide migration barrier. Bentonite is likely to remain hydrothermally stable but potentially sensitive to waste package interactions which could alter the pH, the ratio of dissolved wires, or the sorption properties of radionuclide species.
International Nuclear Information System (INIS)
Neilson, R.M. Jr.; Colombo, P.
1982-01-01
In this program, contemporary solidification agents are being investigated relative to their applications to major fuel cycle and non-fuel cycle low-level waste (LLW) streams. Work is being conducted to determine the range of conditions under which these solidification agents can be applied to specific LLW streams. These studies are directed primarily towards defining operating parameters for both improved solidification of problem wastes and solidification of new LLW streams generated from advanced volume reduction technologies. Work is being conducted to measure relevant waste form properties. These data will be compiled and evaluated to demonstrate compliance with waste form performance and shallow land burial acceptance criteria and transportation requirements (both as they exist and as they are modified with time). 6 tables
Waste acceptance product specifications for vitrified high-level waste forms
International Nuclear Information System (INIS)
Applewhite-Ramsey, A.; Sproull, J.F.
1993-01-01
The Nuclear Waste Policy Act of 1982 mandated that all high-level waste (HLW) be sent to a federal geologic repository for permanent disposal. DOE published the Environmental Assessment in 1982 which identified borosilicate glass as the chosen HLW form. 1 In 1985 the Department of Energy instituted a Waste Acceptance Process to assure that DWPF glass waste forms would be acceptable to such a repository. This assurance was important since production of waste forms will precede repository construction and licensing. As part of this Waste Acceptance Process, the DOE Office of Civilian Radioactive Waste Management (RW) formed the Waste Acceptance Committee (WAC). The WAC included representatives from the candidate repository sites, the waste producing sites and DOE. The WAC was responsible for developing the Waste Acceptance Preliminary Specifications (WAPS) which defined the requirements the waste forms must meet to be compatible with the candidate repository geologies
Preliminary Hanford Waste Vitrification Plan Waste Form Qualification Plan
International Nuclear Information System (INIS)
Nelson, J.L.
1987-09-01
This Waste Form Qualification Plan describes the waste form qualification activities that will be followed during the design and operation of the Hanford Waste Vitrification Plant to ensure that the vitrified Hanford defense high-level wastes will meet the acceptance requirements of the candidate geologic repositories for nuclear waste. This plan is based on the defense waste processing facility requirements. The content of this plan is based on the assumption that the Hanford Waste Vitrification Plant high-level waste form will be disposed of in one of the geologic repository projects. Proposed legislation currently under consideration by Congress may change or delay the repository site selection process. The impacts of this change will be assessed as details of the new legislation become available. The Plan describes activities, schedules, and programmatic interfaces. The Waste Form Qualification Plan is updated regularly to incorporate Hanford Waste Vitrification Plant-specific waste acceptance requirements and to serve as a controlled baseline plan from which changes in related programs can be incorporated. 10 refs., 5 figs., 5 tabs
Selection of barrier metals for a waste package in tuff
International Nuclear Information System (INIS)
Russell, E.W.; McCright, R.D.; O'Neal, W.C.
1983-09-01
The Nevada Nuclear Waste Storage Investigation (NNWSI) project under the Civilian Radioactive Waste Management Program is planning a repository at Yucca Mountain at the Nevada Test Site for isolation of high-level nuclear waste. LLNL is developing designs for an engineered barrier system containing several barriers such as the waste form, a canister and/or an overpack, packing, and near field host rock. The selection of metal containment barriers is addressed. 13 references
Hanford Waste Vitrification Plant
International Nuclear Information System (INIS)
Larson, D.E.; Allen, C.R.; Kruger, O.L.; Weber, E.T.
1991-10-01
The Hanford Waste Vitrification Plant (HWVP) is being designed to immobilize pretreated Hanford high-level waste and transuranic waste in borosilicate glass contained in stainless steel canisters. Testing is being conducted in the HWVP Technology Development Project to ensure that adapted technologies are applicable to the candidate Hanford wastes and to generate information for waste form qualification. Empirical modeling is being conducted to define a glass composition range consistent with process and waste form qualification requirements. Laboratory studies are conducted to determine process stream properties, characterize the redox chemistry of the melter feed as a basis for controlling melt foaming and evaluate zeolite sorption materials for process waste treatment. Pilot-scale tests have been performed with simulated melter feed to access filtration for solids removal from process wastes, evaluate vitrification process performance and assess offgas equipment performance. Process equipment construction materials are being selected based on literature review, corrosion testing, and performance in pilot-scale testing. 3 figs., 6 tabs
Materials interactions relating to long-term geologic disposal of nuclear waste glass
International Nuclear Information System (INIS)
Bibler, N.E.; Jantzen, C.M.
1987-01-01
In the geologic disposal of nuclear waste glass, the glass will eventually interact with groundwater in the repository system. Interactions can also occur between the glass and other waste package materials that are present. These include the steel canister that holds the glass, the metal overpack over the canister, backfill materials that may be used, and the repository host rock. This review paper systematizes the additional interactions that materials in the waste package will impose on the borosilicate glass waste form-groundwater interactions. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g. oxic vs anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interactions is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. It is noted when further tests of repository interactions are needed before long-term predictions can be made. 63 references, 1 table
Waste-rock interactions in the immediate repository
International Nuclear Information System (INIS)
McCarthy, G.J.
1977-01-01
The high level wastes (HLW's) to be placed underground in rock formations will contain significant amounts of radioactive decay heat for the first hundred-or-so years of isolation. Several physical-chemical changes analogous to natural geochemical processes can occur during this ''thermal period.'' The waste canister can act as a heat source and cause changes in the mineralogy and properties of the surrounding rocks. Geochemically, this is ''contact metamorphism.'' In the event that the canister is corroded and breached, chemical reactions can occur between the HLW, the surrounding rock and possibly the remains of the canister. In a dry repository which has not been backfilled (and thus pressurized) these interactions could be slow at best and with rates decreasing rapidly as the HLW cools. However, significant interactions can occur in years, months or even days under hydrothermal conditions. These conditions could be created by the combination of HLW heat, overburden pressure and water mobilized from the rocks or derived from groundwater intrusion. At the end of the thermal period these interaction products would constitute the actual HLW form (or ''source term'') subject to the low temperature leaching and migration processes under investigation in other laboratories. It is quite possible that these interaction product waste forms will have superior properties compared to the original HLW. Experimental programs initiated at Penn State during the last year aim at determining the nature of any chemical or mineralogical changes in, or interactions between, HLW solids and host rocks under various repository ambients. The accompanying figures describe the simulated HLW forms and the experimental approach and techniques. Studies with basalts as the repository rock are supported by Rockwell Hanford Operations and with shales by the Office of Waste Isolation
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
White Paper: Multi-purpose canister (MPC) for DOE-owned spent nuclear fuel (SNF)
International Nuclear Information System (INIS)
Knecht, D.A.
1994-04-01
The paper examines the issue, What are the advantages, disadvantages, and other considerations for using the MPC concept as part of the strategy for interim storage and disposal of DOE-owned SNF? The paper is based in part on the results of an evaluation made for the DOE National Spent Fuel Program by the Waste Form Barrier/Canister Team, which is composed of knowledgeable DOE and DOE-contractor personnel. The paper reviews the MPC and DOE SNF status, provides criteria and other considerations applicable to the issue, and presents an evaluation, conclusions, and recommendations. The primary conclusion is that while most of DOE SNF is not currently sufficiently characterized to be sealed into an MPC, the advantages of standardized packages in handling, reduced radiation exposure, and improved human factors should be considered in DOE SNF program planning. While the design of MPCs for DOE SNF are likely premature at this time, the use of canisters should be considered which are consistent with interim storage options and the MPC design envelope
Site-to-canister scale flow and transport in Haestholmen, Kivetty, Olkiluoto and Romuvaara
Energy Technology Data Exchange (ETDEWEB)
Poteri, A.; Laitinen, M. [VTT Energy, Espoo (Finland)
1999-05-01
Radioactive waste is originating from production of electricity in nuclear power plants. Most of the waste has only low or intermediate levels of radioactivity. However, the spent nuclear fuel is highly radioactive and it has to be isolated from the biosphere. The current nuclear waste management plan in Finland is based on direct disposal of the spent nuclear fuel deep underground. The only feasible mechanism for the radionuclides to escape from an underground repository is to be carried by the groundwater flow after the failure of waste containers. The scope of this study is to examine the groundwater flow situation and transport properties in the vicinity of the disposal canister and along the potential release paths from the repository into the biosphere. The results of this study are further applied in the site specific safety analysis of a spent fuel repository. Synthesis is made of the porous medium estimates of the groundwater flow in the regional and site scales and the detailed fracture network analysis of the flow in the canister scale. This synthesis includes estimation of the transport properties from the canister into the biosphere and flow rates around the deposition holes of the waste canisters. The modelling has been carried out for four different sites: Hastholmen, Kivetty, Olkiluoto and Romavaara. According to the simulations groundwater flow rate around the deposition holes is less than about 1 litre/a for about 75 % of the deposition holes. For about 5 % of the deposition holes the flow rates are a few litres per year or higher. The highest flow rates resulted at Hastholmen, in fresh water conditions 10 000 years after present, and at Kivetty. The transport resistances were calculated for the `worst` flow paths that might have impact on the safety of the repository. The total transport resistances from the repository into the biosphere along those flow paths varied between about 40 000 a/m and 5-10{sup 6} a/m. Most of the total transport
Test plan for K Basin Sludge Canister and Floor Sampling Device
International Nuclear Information System (INIS)
Meling, T.A.
1995-01-01
This document provides the test plan and procedure forms for conducting the functional and operational acceptance testing of the K Basin Sludge Canister and Floor Sampling Device(s). These samplers samples sludge off the floor of the 100K Basins and out of 100K fuel storage canisters
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)
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
On the road to WIPP: Or remote packaging of transuranic waste
International Nuclear Information System (INIS)
Ledbetter, J.M.; Field, L.R.
1994-01-01
At the Los Alamos National Laboratory (LANL) Hot Cell facility, highly productive programs in reactor research spanning three decades have generated appreciable quantities of legacy waste. Hot cell capability had become virtually useless due to the storage of this waste. As a result of concentrated efforts by LANL staff, in cooperation with Westinghouse Waste Isolation Pilot Plant (WIPP), a solution was arrived at that allowed the facility to become productive once again. Equipment has been designed and fabricated to remotely handle 55-gal. waste drums, load waste canisters, perform canister weld closure, leak test welds, grapple the waste canister and transport the canister to an interim storage site. It is our contention that the technology and acquired equipment produced from this effort should be used to further benefit other DOE sites
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
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
Radionuclide Retention in Concrete Waste Forms
Energy Technology Data Exchange (ETDEWEB)
Mattigod, Shas V.; Bovaird, Chase C.; Wellman, Dawn M.; Wood, Marcus I.
2010-09-30
Assessing long-term performance of Category 3 waste cement grouts for radionuclide encasement requires knowledge of the radionuclide-cement interactions and mechanisms of retention (i.e., sorption or precipitation); the mechanism of contaminant release; the significance of contaminant release pathways; how waste form performance is affected by the full range of environmental conditions within the disposal facility; the process of waste form aging under conditions that are representative of processes occurring in response to changing environmental conditions within the disposal facility; the effect of waste form aging on chemical, physical, and radiological properties; and the associated impact on contaminant release. This knowledge will enable accurate prediction of radionuclide fate when the waste forms come in contact with groundwater. The information presented in the report provides data that 1) quantify radionuclide retention within concrete waste form materials similar to those used to encapsulate waste in the Low-Level Waste Burial Grounds (LLBG); 2) measure the effect of concrete waste form properties likely to influence radionuclide migration; and 3) quantify the stability of uranium-bearing solid phases of limited solubility in concrete.
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.
Final waste classification and waste form technical position papers
International Nuclear Information System (INIS)
1983-05-01
The waste classification technical position paper describes overall procedures acceptable to NRC staff which may be used by licensees to determine the presence and concentrations of the radionuclides listed in section 61.55, and thereby classifying waste for near-surface disposal. This technical position paper also provides guidance on the types of information which should be included in shipment manifests accompanying waste shipments to near-surface disposal facilities. The technical position paper on waste form provides guidance to waste generators on test methods and results acceptable to NRC staff for implementing the 10 CFR Part 61 waste form requirements. It can be used as an acceptable approach for demonstrating compliance with the 10 CFR Part 61 waste structural stability criteria. This technical position paper includes guidance on processing waste into an acceptable stable form, designing acceptable high-integrity containers, packaging cartridge filters, and minimizing radiation effects on organic ion-exchange resins. The guidance in the waste form technical position paper may be used by licensees as the basis for qualifying process control programs to meet the waste form stability requirements, including tests which can be used to demonstrate resistance to degradation arising from the effects of compression, moisture, microbial activity, radiation, and chemical changes. Generic test data (e.g., topical reports prepared by vendors who market solidification technology) may be used for process control program qualification where such generic data is applicable to the particular types of waste generated by a licensee
Method of ground disposal of radioactive waste
International Nuclear Information System (INIS)
Harashina, Heihachi.
1991-01-01
Rock bases are drilled to form a disposal hole, an overhanging hole and a burying hole each as a shaft. An appropriate number of canisters prepared by vitrification of high level radioactive wastes are charged in the disposal hole with a gap to the inner wall of the hole. Shock absorbers each made of bentonite are filled between each of the canisters and between the canister and the inner wall of the disposal hole, and the canisters are entirely covered with the layer of the shock absorbers. Further, plucking materials having water sealing property such as cement mortar are filled thereover. With such a constitution, in a case if water should intrude into the overhung portion, since the disposal hole is covered with the large flange portion in addition to the water sealing performance of the plucking, the shock absorbers and the canisters undergo no undesirable effects. Further, in a case if water should intrude to the disposal hole, the shock absorber layers are swollen by water absorption, to suppress the intrusion of water. (T.M.)
Research needs in cement-based waste forms
International Nuclear Information System (INIS)
McDaniel, E.W.; Spence, R.D.; Tallent, O.K.
1990-01-01
Cement-based waste forms are one of the most widely used waste disposal options, yet definitive knowledge of the fate of the waste species inside the waste form is lacking. A fundamental understanding of the chemistry and microstructure of the waste forms would lead to a better understanding of the mass transfer of the waste species, more confidence in predicting and extrapolating waste form performance, and design of better waste forms. Better and cheaper leach tests would lead to quicker and more cost effective screening of waste form alternatives. In addition, assessment of durability may be important to predicting waste form performance in the field. It should be noted that the research needs discussed in this report are from the perspective of investigators working in applied waste management areas, while the proposed investigations are fundamental or basic. Details as to experimental methods and tools to be used in achieving the objectives of the proposed are research beyond the scope of this paper and are better filled in by others. In broad terms, the research topics discussed are correlation of cement-based waste form physical properties to performance, waste-form fundamental chemistry and microstructure, and product performance testing
Ceramic and glass radioactive waste forms
Energy Technology Data Exchange (ETDEWEB)
Readey, D.W.; Cooley, C.R. (comps.)
1977-01-01
This report contains 14 individual presentations and 6 group reports on the subject of glass and polycrystalline ceramic radioactive waste forms. It was the general consensus that the information available on glass as a waste form provided a good basis for planning on the use of glass as an initial waste form, that crystalline ceramic forms could also be good waste forms if much more development work were completed, and that prediction of the chemical and physical stability of the waste form far into the future would be much improved if the basic synergistic effects of low temperature, radiation and long times were better understood. Continuing development of the polycrystalline ceramic forms was recommended. It was concluded that the leach rate of radioactive species from the waste form is an important criterion for evaluating its suitability, particularly for the time period before solidified waste is permanently placed in the geologic isolation of a Federal repository. Separate abstracts were prepared for 12 of the individual papers; the remaining two were previously abstracted.
Material interactions relating to long-term geologic disposal of nuclear waste glass
International Nuclear Information System (INIS)
Bibler, N.E.; Jantzen, C.M.
1986-01-01
This review paper systematizes the additional interactions that materials in a geologic repository will impose on the borosilicate glass waste form-groundwater interactions. These materials are the steel canister that holds the glass, the steel overpack over the canister, backfill materials that may be used, and last, the repository host rock. The repository geologies reviewed are tuff, salt, basalt, and granite. The interactions emphasized are those appropriate to conditions expected after repository closure, e.g., oxic vs anoxic conditions. Whenever possible, the effect of radiation from the waste form on the interaction(s) is examined. The interactions are evaluated based on their effect on the release and speciation of various elements including radionuclides from the glass. Repository relevant interactions testing that requires further study before long-term predictions can be made are noted. 62 refs
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)
Radioactive waste material melter apparatus
Newman, D.F.; Ross, W.A.
1990-04-24
An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs.
Radioactive waste material melter apparatus
International Nuclear Information System (INIS)
Newman, D.F.; Ross, W.A.
1990-01-01
An apparatus for preparing metallic radioactive waste material for storage is disclosed. The radioactive waste material is placed in a radiation shielded enclosure. The waste material is then melted with a plasma torch and cast into a plurality of successive horizontal layers in a mold to form a radioactive ingot in the shape of a spent nuclear fuel rod storage canister. The apparatus comprises a radiation shielded enclosure having an opening adapted for receiving a conventional transfer cask within which radioactive waste material is transferred to the apparatus. A plasma torch is mounted within the enclosure. A mold is also received within the enclosure for receiving the melted waste material and cooling it to form an ingot. The enclosure is preferably constructed in at least two parts to enable easy transport of the apparatus from one nuclear site to another. 8 figs
Synroc tailored waste forms for actinide immobilization
Energy Technology Data Exchange (ETDEWEB)
Gregg, Daniel J.; Vance, Eric R. [Australian Nuclear Science and Technology Organisation, Kirrawee (Australia). ANSTOsynroc, Inst. of Materials Engineering
2017-07-01
Since the end of the 1970s, Synroc at the Australian Nuclear Science and Technology Organisation (ANSTO) has evolved from a focus on titanate ceramics directed at PUREX waste to a platform waste treatment technology to fabricate tailored glass-ceramic and ceramic waste forms for different types of actinide, high- and intermediate level wastes. The particular emphasis for Synroc is on wastes which are problematic for glass matrices or existing vitrification process technologies. In particular, nuclear wastes containing actinides, notably plutonium, pose a unique set of requirements for a waste form, which Synroc ceramic and glass-ceramic waste forms can be tailored to meet. Key aspects to waste form design include maximising the waste loading, producing a chemically durable product, maintaining flexibility to accommodate waste variations, a proliferation resistance to prevent theft and diversion, and appropriate process technology to produce waste forms that meet requirements for actinide waste streams. Synroc waste forms incorporate the actinides within mineral phases, producing products which are much more durable in water than baseline borosilicate glasses. Further, Synroc waste forms can incorporate neutron absorbers and {sup 238}U which provide criticality control both during processing and whilst within the repository. Synroc waste forms offer proliferation resistance advantages over baseline borosilicate glasses as it is much more difficult to retrieve the actinide and they can reduce the radiation dose to workers compared to borosilicate glasses. Major research and development into Synroc at ANSTO over the past 40 years has included the development of waste forms for excess weapons plutonium immobilization in collaboration with the US and for impure plutonium residues in collaboration with the UK, as examples. With a waste loading of 40-50 wt.%, Synroc would also be considered a strong candidate as an engineered waste form for used nuclear fuel and highly
Method of encapsulating solid radioactive waste material for storage
International Nuclear Information System (INIS)
Bunnell, L.R.; Bates, J.L.
1976-01-01
High-level radioactive wastes are encapsulated in vitreous carbon for long-term storage by mixing the wastes as finely divided solids with a suitable resin, formed into an appropriate shape and cured. The cured resin is carbonized by heating under a vacuum to form vitreous carbon. The vitreous carbon shapes may be further protected for storage by encasement in a canister containing a low melting temperature matrix material such as aluminum to increase impact resistance and improve heat dissipation. 8 claims
TMI-2 fuel canister interface requirements for INEL. Revision 1
International Nuclear Information System (INIS)
Wilkins, D.E.; Martz, D.E.; Reno, H.W.
1984-06-01
This report focuses on fuel canister interface requirements at INEL which should be incorporated into the canister design criteria. The requirements will ensure compatibility with existing INEL structures and equipment to be used for receipt, unloading, and storage of fuel canisters. INEL can and does receive and store radioactive materials in many different forms, including reactor fuel. INEL requires detailed descriptions of canisters and casks. Therefore, requirements listed represent engineering design features which will simplify the handling and storage operations; consequently, they are not to be viewed as absolute or non-negotiable. However, the core acquisition contract was negotiated with certain storage assumptions which effect costs of storage. Deviations from those assumptions which significantly effect costs would require approval by DOE-Idaho. If some stated requirements are too restrictive, modifications based on sound engineering principles may be negotiated with INEL. 11 figures
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)
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
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
Selection of barrier metals for a waste package in tuff
International Nuclear Information System (INIS)
Russell, E.W.; McCright, R.D.; O'Neal, W.C.
1983-10-01
The Nevada Nuclear Waste Storage Investigations (NNWSI) project under the Civilian Radioactive Waste Management Program is planning a repository at Yucca Mountain at the Nevada Test Site for isolation of high-level nuclear waste. Lawrence Livermore National Laboratory is developing designs for an engineered barrier system containing several barriers such as the waste form, a canister and/or an overpack, packing, and near field host rock. In this paper we address the selection of metal containment barriers. 13 references, 4 tables
Standardized waste form test methods
International Nuclear Information System (INIS)
Slate, S.C.
1984-01-01
The Materials Characterization Center (MCC) is developing standard tests to characterize nuclear waste forms. Development of the first thirteen tests was originally initiated to provide data to compare different high-level waste (HLW) forms and to characterize their basic performance. The current status of the first thirteen MCC tests and some sample test results are presented: the radiation stability tests (MCC-6 and 12) and the tensile-strength test (MCC-11) are approved; the static leach tests (MCC-1, 2, and 3) are being reviewed for full approval; the thermal stability (MCC-7) and microstructure evaluation (MCC-13) methods are being considered for the first time; and the flowing leach test methods (MCC-4 and 5), the gas generation methods (MCC-8 and 9), and the brittle fracture method (MCC-10) are indefinitely delayed. Sample static leach test data on the ARM-1 approved reference material are presented. Established tests and proposed new tests will be used to meet new testing needs. For waste form production, tests on stability and composition measurement are needed to provide data to ensure waste form quality. In transporation, data are needed to evaluate the effects of accidents on canisterized waste forms. The new MCC-15 accident test method and some data are presented. Compliance testing needs required by the recent draft repository waste acceptance specifications are described. These specifications will control waste form contents, processing, and performance
Standardized waste form test methods
International Nuclear Information System (INIS)
Slate, S.C.
1984-11-01
The Materials Characterization Center (MCC) is developing standard tests to characterize nuclear waste forms. Development of the first thirteen tests was originally initiated to provide data to compare different high-level waste (HLW) forms and to characterize their basic performance. The current status of the first thirteen MCC tests and some sample test results is presented: The radiation stability tests (MCC-6 and 12) and the tensile-strength test (MCC-11) are approved; the static leach tests (MCC-1, 2, and 3) are being reviewed for full approval; the thermal stability (MCC-7) and microstructure evaluation (MCC-13) methods are being considered for the first time; and the flowing leach tests methods (MCC-4 and 5), the gas generation methods (MCC-8 and 9), and the brittle fracture method (MCC-10) are indefinitely delayed. Sample static leach test data on the ARM-1 approved reference material are presented. Established tests and proposed new tests will be used to meet new testing needs. For waste form production, tests on stability and composition measurement are needed to provide data to ensure waste form quality. In transportation, data are needed to evaluate the effects of accidents on canisterized waste forms. The new MCC-15 accident test method and some data are presented. Compliance testing needs required by the recent draft repository waste acceptance specifications are described. These specifications will control waste form contents, processing, and performance. 2 references, 2 figures
SNF Interim Storage Canister Corrosion and Surface Environment Investigations
International Nuclear Information System (INIS)
Bryan, Charles R.; Enos, David G.
2015-01-01
This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.
SNF Interim Storage Canister Corrosion and Surface Environment Investigations
Energy Technology Data Exchange (ETDEWEB)
Bryan, Charles R. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States); Enos, David G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)
2015-09-01
This progress report describes work being done at Sandia National Laboratories (SNL) to assess the localized corrosion performance of container/cask materials used in the interim storage of spent nuclear fuel (SNF). Of particular concern is stress corrosion cracking (SCC), by which a through-wall crack could potentially form in a canister outer wall over time intervals that are shorter than possible dry storage times. In order for SCC to occur, three criteria must be met. A corrosive environment must be present on the canister surface, the metal must susceptible to SCC, and sufficient tensile stress to support SCC must be present through the entire thickness of the canister wall. SNL is currently evaluating the potential for each of these criteria to be met.
International Nuclear Information System (INIS)
Schneider, K.J.; Lakey, L.T.; Silviera, D.J.
1989-09-01
This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100 degree C. 52 refs., 9 figs
ANSTO's waste forms for the 31. century
International Nuclear Information System (INIS)
Vance, E.R.; Begg, B. D.; Day, R. A.; Moricca, S.; Perera, D. S.; Stewart, M. W. A.; Carter, M. L.; McGlinn, P. J.; Smith, K. L.; Walls, P. A.; Robina, M. La
2004-01-01
ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and 99 Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)
ANSTO's waste forms for the 31. century
Energy Technology Data Exchange (ETDEWEB)
Vance, E R; Begg, B D; Day, R A; Moricca, S; Perera, D S; Stewart, M W. A.; Carter, M L; McGlinn, P J; Smith, K L; Walls, P A; Robina, M La
2004-07-01
ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and {sup 99}Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)
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
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
Combined Waste Form Cost Trade Study
International Nuclear Information System (INIS)
Gombert, Dirk; Piet, Steve; Trickel, Timothy; Carter, Joe; Vienna, John; Ebert, Bill; Matthern, Gretchen
2008-01-01
A new generation of aqueous nuclear fuel reprocessing, now in development under the auspices of the DOE Office of Nuclear Energy (NE), separates fuel into several fractions, thereby partitioning the wastes into groups of common chemistry. This technology advance enables development of waste management strategies that were not conceivable with simple PUREX reprocessing. Conventional wisdom suggests minimizing high level waste (HLW) volume is desirable, but logical extrapolation of this concept suggests that at some point the cost of reducing volume further will reach a point of diminishing return and may cease to be cost-effective. This report summarizes an evaluation considering three groupings of wastes in terms of cost-benefit for the reprocessing system. Internationally, the typical waste form for HLW from the PUREX process is borosilicate glass containing waste elements as oxides. Unfortunately several fission products (primarily Mo and the noble metals Ru, Rh, Pd) have limited solubility in glass, yielding relatively low waste loading, producing more glass, and greater disposal costs. Advanced separations allow matching the waste form to waste stream chemistry, allowing the disposal system to achieve more optimum waste loading with improved performance. Metals can be segregated from oxides and each can be stabilized in forms to minimize the HLW volume for repository disposal. Thus, a more efficient waste management system making the most effective use of advanced waste forms and disposal design for each waste is enabled by advanced separations and how the waste streams are combined. This trade-study was designed to juxtapose a combined waste form baseline waste treatment scheme with two options and to evaluate the cost-benefit using available data from the conceptual design studies supported by DOE-NE
Scoping calculations for canister-tunnel migration of corrodants, oxidants and radionuclides
International Nuclear Information System (INIS)
Shaw, W.; Worth, D.
1992-03-01
This report presents the mathematical models and results obtained for a set of scooping calculations which estimate the possible extent of the vertical migration of canister corrodants, oxidants (forming a redox front) and radionuclides between a copper canister containing spent nuclear fuel, and an overlying emplacement tunnel. The KBS-3 concept for the disposal of spent nuclear fuel is copper canisters, vertically emplaced in deposition holes bored in the floor of a tunnel, situated deep underground. The deposition holes are filled with a buffer of bentonite and the tunnel is backfilled with a mixture of sand and bentonite. (au)
Stainless steel-zirconium alloy waste forms
International Nuclear Information System (INIS)
McDeavitt, S.M.; Abraham, D.P.; Keiser, D.D. Jr.; Park, J.Y.
1996-01-01
An electrometallurgical treatment process has been developed by Argonne National Laboratory to convert various types of spent nuclear fuels into stable storage forms and waste forms for repository disposal. The first application of this process will be to treat spent fuel alloys from the Experimental Breeder Reactor-II. Three distinct product streams emanate from the electrorefining process: (1) refined uranium; (2) fission products and actinides extracted from the electrolyte salt that are processed into a mineral waste form; and (3) metallic wastes left behind at the completion of the electrorefining step. The third product stream (i.e., the metal waste stream) is the subject of this paper. The metal waste stream contains components of the chopped spent fuel that are unaffected by the electrorefining process because of their electrochemically ''noble'' nature; this includes the cladding hulls, noble metal fission products (NMFP), and, in specific cases, zirconium from metal fuel alloys. The selected method for the consolidation and stabilization of the metal waste stream is melting and casting into a uniform, corrosion-resistant alloy. The waste form casting process will be carried out in a controlled-atmosphere furnace at high temperatures with a molten salt flux. Spent fuels with both stainless steel and Zircaloy cladding are being evaluated for treatment; thus, stainless steel-rich and Zircaloy-rich waste forms are being developed. Although the primary disposition option for the actinides is the mineral waste form, the concept of incorporating the TRU-bearing product into the metal waste form has enough potential to warrant investigation
IGNEOUS INTRUSION IMPACTS ON WASTE PACKAGES AND WASTE FORMS
International Nuclear Information System (INIS)
Bernot, P.
2004-01-01
The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The models are based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. The models described in this report constitute the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of LA (BSC 2004 [DIRS:167796]) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2003 [DIRS: 166296]). The technical work plan was prepared in accordance with AP-2.27Q, Planning for Science Activities. Any deviations from the technical work plan are documented in the following sections as they occur. The TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model assessments: (1) Mechanical and thermal impacts of basalt magma intrusion on the invert, waste packages and waste forms of the intersected emplacement drifts of Zone 1. (2) Temperature and pressure trends of basaltic magma intrusion intersecting Zone 1 and their potential effects on waste packages and waste forms in Zone 2 emplacement drifts. (3) Deleterious volatile gases, exsolving from the intruded basalt magma and their potential effects on waste packages of Zone 2 emplacement drifts. (4) Post-intrusive physical
Defining a metal-based waste form for IFR pyroprocessing wastes
International Nuclear Information System (INIS)
McDeavitt, S.M.; Park, J.Y.; Ackerman, J.P.
1994-01-01
Pyrochemical electrorefining to recover actinides from metal nuclear fuel is a key element of the Integral Fast Reactor (IFR) fuel cycle. The process separates the radioactive fission products from the long-lived actinides in a molten LiCl-KCl salt, and it generates a lower waste volume with significantly less long-term toxicity as compared to spent nuclear fuel. The process waste forms include a mineral-based waste form that will contain fission products removed from an electrolyte salt and a metal-based waste form that will contain metallic fission products and the fuel cladding and process materials. Two concepts for the metal-based waste form are being investigated: (1) encapsulating the metal constituents in a Cu-Al alloy and (2) alloying the metal constituents into a uniform stainless steel-based waste form. Results are given from our recent studies of these two concepts
Modeling transient heat transfer in nuclear waste repositories.
Yang, Shaw-Yang; Yeh, Hund-Der
2009-09-30
The heat of high-level nuclear waste may be generated and released from a canister at final disposal sites. The waste heat may affect the engineering properties of waste canisters, buffers, and backfill material in the emplacement tunnel and the host rock. This study addresses the problem of the heat generated from the waste canister and analyzes the heat distribution between the buffer and the host rock, which is considered as a radial two-layer heat flux problem. A conceptual model is first constructed for the heat conduction in a nuclear waste repository and then mathematical equations are formulated for modeling heat flow distribution at repository sites. The Laplace transforms are employed to develop a solution for the temperature distributions in the buffer and the host rock in the Laplace domain, which is numerically inverted to the time-domain solution using the modified Crump method. The transient temperature distributions for both the single- and multi-borehole cases are simulated in the hypothetical geological repositories of nuclear waste. The results show that the temperature distributions in the thermal field are significantly affected by the decay heat of the waste canister, the thermal properties of the buffer and the host rock, the disposal spacing, and the thickness of the host rock at a nuclear waste repository.
Liquid Secondary Waste Grout Formulation and Waste Form Qualification
Energy Technology Data Exchange (ETDEWEB)
Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Williams, B. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Snyder, Michelle M. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2016-05-23
This report describes the results from liquid secondary waste (LSW) grout formulation and waste form qualification tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate new formulations for preparing a grout waste form with high-sulfate secondary waste simulants and the release of key constituents from these grout monoliths. Specific objectives of the LSW grout formulation and waste form qualification tests described in this report focused on five activities: 1.preparing new formulations for the LSW grout waste form with high-sulfate LSW simulants and solid characterization of the cured LSW grout waste form; 2.conducting the U.S. Environmental Protection Agency (EPA) Method 1313 leach test (EPA 2012) on the grout prepared with the new formulations, which solidify sulfate-rich Hanford Tank Waste Treatment and Immobilization Plant (WTP) off-gas condensate secondary waste simulant, using deionized water (DIW); 3.conducting the EPA Method 1315 leach tests (EPA 2013) on the grout monoliths made with the new dry blend formulations and three LSW simulants (242-A evaporator condensate, Environmental Restoration Disposal Facility (ERDF) leachate, and WTP off-gas condensate) using two leachants, DIW and simulated Hanford Integrated Disposal Facility (IDF) Site vadose zone pore water (VZPW); 4.estimating the 99Tc desorption Kd (distribution coefficient) values for 99Tc transport in oxidizing conditions to support the IDF performance assessment (PA); 5.estimating the solubility of 99Tc(IV)-bearing solid phases for 99Tc transport in reducing conditions to support the IDF PA.
Waste-rock interactions and bedrock reactions
International Nuclear Information System (INIS)
White, W.B.
1977-01-01
The experimental program is designed to discover possible reactions between shale repository rocks and radioactive wastes. The canister can be regarded in three ways: (a) As a source of heat that modifies the mineralogy and therefore the physical properties of the surrounding rock (dry heat). (b) As a source of heat that activates reactions between minerals in the surrounding rock and slowly percolating ground water. (c) As a source of reaction materials of different composition from the surrounding rock and which therefore may react to form completely new ''minerals'' in a contact aureole around the canister. The matrix of interactions contains two composition axes. The waste compositions are defined by the various prototype waste forms usually investigated: glass, calcine, ''spent fuel'' and the ceramic supercalcine. The temperatures and pressures at which these reactions take place must be investigated. Thus each node on the ''wiring diagram'' is itself a matrix of experiments in which the T and to some extent P are varied. Experiments at higher pressure and temperature allow reactions to take place on a laboratory time scale and thus identify what could happen. These reactions are then followed downward in temperature to determine both phase boundaries and kinetic cut-offs below which equilibrium cannot be achieved on a laboratory time scale
Design basis for the copper/steel canister. Stage four. Final report
Energy Technology Data Exchange (ETDEWEB)
Bowyer, W.H. [Meadow End Farm, Farnham (United Kingdom)
1998-06-01
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 points of view of choice of materials, manufacturing technology and quality assurance. Cast steel has been rejected in favour of cast iron as a candidate material for the load bearing liner. Nodular (or ductile) iron is selected and this is capable of providing mechanical properties which are equally suitable as those of the originally selected high strength low alloy steel. The material specified for the overpack is Oxygen free copper with 50 ppm of phosphorus added. Corrosion studies supported by SKB indicate that in the absence of mechanical failure or accelerated localised corrosion the overpack should provide corrosion shielding of the canister for its full design life. Published work claiming that the nodular iron liner would have corrosion characteristics similar to the carbon steel which had been examined in depth is flawed since the microstructures of the iron and carbon steel specimens used were not investigated. It is highly unlikely that nodular irons in the form used for the experiments would have similar structures to nodular iron in the canisters by chance. If the overpack were breached during the aerobic period of the repository life then very rapid penetration of the inner liner could occur. It has been recognised that the roll forming method is not suitable for serial production and alternatives are being sought. The electron beam welding process has been explored with tenacity but has so far failed to produce a satisfactory lid weld. A new welder is being developed for supply to the SKB pilot plant where development will be continued. An alternative welding process, friction stir welding, is being examined as a candidate for attaching lids. Surface breaking defects may be detected using eddy current methods but there is currently no reliable way of detecting small sub surface defects in the overpack
High-level waste-form-product performance evaluation
International Nuclear Information System (INIS)
Bernadzikowski, T.A.; Allender, J.S.; Stone, J.A.; Gordon, D.E.; Gould, T.H. Jr.; Westberry, C.F. III.
1982-01-01
Seven candidate waste forms were evaluated for immobilization and geologic disposal of high-level radioactive wastes. The waste forms were compared on the basis of leach resistance, mechanical stability, and waste loading. All forms performed well at leaching temperatures of 40, 90, and 150 0 C. Ceramic forms ranked highest, followed by glasses, a metal matrix form, and concrete. 11 tables
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
Updated Liquid Secondary Waste Grout Formulation and Preliminary Waste Form Qualification
Energy Technology Data Exchange (ETDEWEB)
Saslow, Sarah A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Russell, Renee L. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Asmussen, Robert M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Sahajpal, Rahul [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2017-07-01
This report describes the results from liquid secondary waste grout (LSWG) formulation and cementitious waste form qualification tests performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). New formulations for preparing a cementitious waste form from a high-sulfate liquid secondary waste stream simulant, developed for Effluent Management Facility (EMF) process condensates merged with low activity waste (LAW) caustic scrubber, and the release of key constituents (e.g. 99Tc and 129I) from these monoliths were evaluated. This work supports a technology development program to address the technology needs for Hanford Site Effluent Treatment Facility (ETF) liquid secondary waste (LSW) solidification and supports future Direct Feed Low-Activity Waste (DFLAW) operations. High-priority activities included simulant development, LSWG formulation, and waste form qualification. The work contained within this report relates to waste form development and testing and does not directly support the 2017 integrated disposal facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY17, and for future waste form development efforts. The provided data should be used by (i) cementitious waste form scientists to further understanding of cementitious dissolution behavior, (ii) IDF PA modelers who use quantified constituent leachability, effective diffusivity, and partitioning coefficients to advance PA modeling efforts, and (iii) the U.S. Department of Energy (DOE) contractors and decision makers as they assess the IDF PA program. The results obtained help fill existing data gaps, support final selection of a LSWG waste form, and improve the technical defensibility of long-term waste form performance estimates.
Special waste-form lysimeters: Arid
International Nuclear Information System (INIS)
Jones, T.L.; Serne, R.J.
1987-08-01
The release of contaminant from solidified low-level waste forms is being studied in a field lysimeter facility at the Hanford Site in southeastern Washington State. Duplicate samples of five different waste forms have been buried in 10 lysimeters since March 1984. Waste-form samples represent three different waste streams and four solidification agents (masonry cement, Portland III cement, Dow polymer /sup (a)/, and bitumen). Most precipitation at the Hanford Site arrives as winter snow; this contributes to a strong seasonal pattern in water storage and drainage observed in the lysimeters. The result is an annual range in the volumetric soil water content from 11% in late winter to 7% in the late summer and early fall, as well as annual changes in pore water velocities from approximately 1 cm/wk in early spring to less than 0.05 cm/wk in early fall. Measurable quantities of tritium and cobalt-60 are being collected in lysimeter drainage water. Approximately 30% of the original tritium inventory has been leached from two lysimeters originally containing tritium. Cobalt-60 is present in all waste forms; it is being collected in the leachate from five lysimeters. The total amount released varies, but in each case it is less than 0.1% of the original cobalt inventory of the waste sample. Nonradioactive constituents contained in the waste form, such as sodium, boron, and sulfate, are also being leached
Nuclear waste disposal: alternatives to solidification in glass proposed
International Nuclear Information System (INIS)
Kerr, R.A.
1979-01-01
More than a quarter-million cubic meters of liquid radioactive wastes are now being held at government installations awaiting final disposal. During the past 20 years, the disposal plan of choice has been to incorporate the 40 to 50 radioactive elements dissolved in liquid wastes into blocks of glass, seal the glass in metal canisters, and insert the canisters into deep, geologically stable salt beds. Over the last few years, some geologists and materials scientists have become concerned that perhaps not enough is known yet about the interaction of waste, container, and salt (or any rock) to have a reasonable assurance that the hazardous wastes will be contained successfully. The biggest advantage of glass at present is the demonstrated practicality of producing large, highly radioactive blocks of it. The frontrunner as a successor to glass is ceramics, which are nonmetallic crystalline materials formed at high temperature, such as chinaware or natural minerals. An apparent advantage of ceramics is that they already have an ordered atomic structure, whose properties can be tailored to a particular waste element and to conditions of a specific disposal site. A ceramic tailored for waste disposal called supercalcine-ceramic has been developed. It was emphasized that the best minerals for waste solidification may be those that have proved most stable under natural conditions over geologic time. Disadvantage to ceramics are radiation damage and transmutation. However, it is now obvious that some ceramics are more stable than glass under certain conditions. Metal-encapsulated ceramic, called cermet, is being developed as a waste form. Cermets are considerably more resistant at 100 0 C than a borosilicate waste glass. Researchers are now testing prospective waste forms under the most extreme conditions that might prevail in a waste disposal site
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)
A U-bearing composite waste form for electrochemical processing wastes
Energy Technology Data Exchange (ETDEWEB)
Chen, X.; Ebert, W. L.; Indacochea, J. E.
2018-04-01
Metallic/ceramic composite waste forms are being developed to immobilize combined metallic and oxide waste streams generated during electrochemical recycling of used nuclear fuel. Composites were made for corrosion testing by reacting HT9 steel to represent fuel cladding, Zr and Mo to simulate metallic fuel waste, and a mixture of ZrO2, Nd2O3, and UO2 to represent oxide wastes. More than half of the added UO2 was reduced to metal and formed Fe-Zr-U intermetallics and most of the remaining UO2 and all of the Nd2O3 reacted to form zirconates. Fe-Cr-Mo intermetallics were also formed. Microstructure characterization of the intermetallic and ceramic phases that were generated and tests conducted to evaluate their corrosion behaviors indicate composite waste forms can accommodate both metallic and oxidized waste streams in durable host phases. (c) 2018 Elsevier B.V. All rights reserved.
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.
Development of high-frequency induction melting system for radioactive solid wastes
International Nuclear Information System (INIS)
Kawaguchi, Ichiro; Yamazaki, Seichiro; Takahashi, Noriaki; Kugai, Katsutoshi; Yokozawa, Minoru
2004-01-01
Kawasaki Heavy Industries, Ltd. developed an active insulation (AI) method radiofrequency melting system as a new melting treatment system of radioactive solid wastes and proved production of waste satisfied the treatment performances and burying by repeating many practical melting tests. The melting vessel uses a low-priced ceramic canister with nonelectrical conductivity, which is able to treat wastes with large amount of inorganic substances. The wastes melted in the canister is taken out the canister itself from radiofrequency melting reactor and solidified after cooling. The cool canister is stored in 2001 metal drum filling up a gap with mortal for laying underground. New radiofrequency melting reactor, 1/3 scale melting test, estimation of scale effects, melting tests for practical use and the total system are explained. (S.Y.)
International Nuclear Information System (INIS)
BOEHNKE, W.M.
2001-01-01
A plan is currently in place to process the high-level radioactive wastes that resulted from uranium and plutonium recovery operations from Spent Nuclear Fuel at the Hanford Site, Richland, Washington. Currently, millions of gallons of high-level radioactive waste in the form of liquids, sludges, and saltcake are stored in many large underground tanks onsite. This waste will be processed and separated into high-level and low-activity fractions. Both fractions will then be vitrified (i.e., blended with molten borosilicate glass) in order to encapsulate the toxic radionuclides. The immobilized low-activity waste (ILAW) glass will be poured into LAW canisters, allowed to cool and harden to solid form, sealed by welding, and then transported to a double-lined trench in the 200 East Area for permanent disposal. This document presents the packaging design criteria (PDC) for an onsite LAW transportation system, which includes the ILAW canister, ILAW package, and transport vehicle and defines normal and accident conditions. This PDC provides the basis for the ILAW onsite transportation system design and fabrication and establishes the transportation safety criteria that the design will be evaluated against in the Package Specific Safety Document (PSSD). It provides the criteria for the ILAW canister, cask and transport vehicles and defines normal and accident conditions. The LAW transportation system is designed to transport stabilized waste from the vitrification facility to the ILAW disposal facility developed by Project W-520. All ILAW transport will take place within the 200 East Area (all within the Hanford Site)
Survey of concrete waste forms
International Nuclear Information System (INIS)
Moore, J.G.
1981-01-01
The incorporation of radioactive waste in cement has been widely studied for many years. It has been routinely used at nuclear research and production sites for some types of nuclear waste for almost three decades and at power reactor plants for nearly two decades. Cement has many favorable characteristics that have contributed to its popularity. It is a readily available material and has not required complex and/or expensive equipment to solidify radioactive waste. The resulting solid products are noncombustible, strong, radiation resistant, and have reasonable chemical and thermal stability. As knowledge increased on the possible dangers from radioactive waste, requirements for waste fixation became more stringent. A brief survey of some of the research efforts used to extend and improve cementitious waste hosts to meet these requirements is given in this paper. Selected data are presented from the rather extensive study of the applicability of concrete as a waste form for Savannah River defense waste and the use of polymer impregnation to reduce the leachability and improve the durability of such waste forms. Hot-pressed concretes that were developed as prospective host solids for high-level wastes are described. Highlights are given from two decades of research on cementitious waste forms at Oak Ridge National Laboratory. The development of the hydrofracture process for the disposal of all locally generated radioactive waste led to a process for the disposal of I-129 and to the current research on the German in-situ solidification process for medium-level waste and the Oak Ridge FUETAP process for all classes of waste including commercial and defense high-level wastes. Finally, some of the more recent ORNL concepts are presented for the use of cement in the disposal of inorganic and biological sludges, waste inorganic salts, trash, and krypton
Numerical analysis of a natural convection cooling system for radioactive canisters storage
Energy Technology Data Exchange (ETDEWEB)
Tsal, R.J.; Anwar, S.; Mercada, M.G. [Fluor Daniel Inc., Irvine, CA (United States)
1995-02-01
This paper describes the use of numerical analysis for studying natural convection cooling systems for long term storage of heat producing radioactive materials, including special nuclear materials and nuclear waste. The paper explains the major design philosophy, and shares the experiences of numerical modeling. The strategy of storing radioactive material is to immobilize nuclear high-level waste by a vitrification process, convertion it into borosilicate glass, and cast the glass into stainless steel canisters. These canisters are seal welded, decontaminated, inspected, and temporarily stored in an underground vault until they can be sent to a geologic repository for permanent storage. These canisters generate heat by nuclear decay of radioactive isotopes. The function of the storage facility ventilation system is to ensure that the glass centerline temperature does not exceed the glass transition temperature during storage and the vault concrete temperatures remain within the specified limits. A natural convection cooling system was proposed to meet these functions. The effectiveness of a natural convection cooling system is dependent on two major factors that affect air movement through the vault for cooling the canisters: (1) thermal buoyancy forces inside the vault which create a stack effect, and (2) external wind forces, that may assist or oppose airflow through the vault. Several numerical computer models were developed to analyze the thermal and hydraulic regimes in the storage vault. The Site Model is used to simulate the airflow around the building and to analyze different air inlet/outlet devices. The Airflow Model simulates the natural convection, thermal regime, and hydraulic resistance in the vault. The Vault Model, internal vault temperature stratification; and, finally, the Hot Area Model is used for modeling concrete temperatures within the vault.
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
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
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.
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.
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
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
Development of iron phosphate ceramic waste form to immobilize radioactive waste solution
Energy Technology Data Exchange (ETDEWEB)
Choi, Jongkwon [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Um, Wooyong, E-mail: wooyong.um@pnnl.gov [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of); Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Choung, Sungwook [Division of Advanced Nuclear Engineering, Pohang University of Science and Technology (POSTECH), San 31, Hyoja-Dong, Pohang (Korea, Republic of)
2014-09-15
The objective of this research was to develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl–KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Compressive strengths of the IPC waste form prepared with different types of waste solutions were 16 MPa and 19 MPa for LiCl–KCl eutectic salt and the off-gas scrubber simulant, respectively, which meet the minimum compressive strength of 3.45 MPa (500 psi) for waste forms to be accepted into the radioactive waste repository. The reduction capacity of converter slag, a main dry ingredient used to prepare the IPC waste form, was 4136 meq/kg by the Ce(IV) method, which is much higher than those of the conventional Fe oxides used for the IPC waste form and the blast furnace slag materials. Average leachability indexes of Tc, Li, and K for the IPC waste form were higher than 6.0, and the IPC waste form demonstrated stable durability even after 63-day leaching. In addition, the Toxicity Characteristic Leach Procedure measurements of converter slag and the IPC waste form with LiCl–KCl eutectic salt met the universal treatment standard of the leachability limit for metals regulated by the Resource Conservation and Recovery Act. This study confirms the possibility of development of the IPC waste form using converter slag, showing its immobilization capability for radionuclides in both LiCl–KCl eutectic salt and off-gas scrubber solutions with significant cost savings.
A generalized definition for waste form durability
International Nuclear Information System (INIS)
Fanning, T. H.; Bauer, T. H.; Morris, E. E.; Wigeland, R. A.
2002-01-01
When evaluating waste form performance, the term ''durability'' often appears in casual discourse, but in the technical literature, the focus is often on waste form ''degradation'' in terms of mass lost per unit area per unit time. Waste form degradation plays a key role in developing models of the long-term performance in a repository environment, but other factors also influence waste form performance. These include waste form geometry; density, porosity, and cracking; the presence of cladding; in-package chemistry feedback; etc. The paper proposes a formal definition of waste form ''durability'' which accounts for these effects. Examples from simple systems as well as from complex models used in the Total System Performance Assessment of Yucca Mountain are provided. The application of ''durability'' in the selection of bounding models is also discussed
International Nuclear Information System (INIS)
Treat, R.L.; Nesbitt, J.F.; Blair, H.T.; Carter, J.G.; Gorton, P.S.; Partain, W.L.; Timmerman, C.L.
1980-04-01
This document contains preconceptual design data on 11 processes for the solidification and isolation of nuclear high-level liquid wastes (HLLW). The processes are: in-can glass melting (ICGM) process, joule-heated glass melting (JHGM) process, glass-ceramic (GC) process, marbles-in-lead (MIL) matrix process, supercalcine pellets-in-metal (SCPIM) matrix process, pyrolytic-carbon coated pellets-in-metal (PCCPIM) matrix process, supercalcine hot-isostatic-pressing (SCHIP) process, SYNROC hot-isostatic-pressing (SYNROC HIP) process, titanate process, concrete process, and cermet process. For the purposes of this study, it was assumed that each of the solidification processes is capable of handling similar amounts of HLLW generated in a production-sized fuel reprocessing plant. It was also assumed that each of the processes would be enclosed in a shielded canyon or cells within a waste facility located at the fuel reprocessing plant. Finally, it was assumed that all of the processes would be subject to the same set of regulations, codes and standards. Each of the solidification processes converts waste into forms that may be acceptable for geological disposal. Each process begins with the receipt of HLLW from the fuel reprocessing plant. In this study, it was assumed that the original composition of the HLLW would be the same for each process. The process ends when the different waste forms are enclosed in canisters or containers that are acceptable for interim storage. Overviews of each of the 11 processes and the bases used for their identification are presented in the first part of this report. Each process, including its equipment and its requirements, is covered in more detail in Appendices A through K. Pertinent information on the current state of the art and the research and development required for the implementation of each process are also noted in the appendices
Hanford Waste Vitrification Plant Project Waste Form Qualification Program Plan
International Nuclear Information System (INIS)
Randklev, E.H.
1993-06-01
The US Department of Energy has created a waste acceptance process to help guide the overall program for the disposal of high-level nuclear waste in a federal repository. This Waste Form Qualification Program Plan describes the hierarchy of strategies used by the Hanford Waste Vitrification Plant Project to satisfy the waste form qualification obligations of that waste acceptance process. A description of the functional relationship of the participants contributing to completing this objective is provided. The major activities, products, providers, and associated scheduling for implementing the strategies also are presented
Canisters and nonfuel components at commercial nuclear reactors
International Nuclear Information System (INIS)
Gibbard, K.; Thorpe, J.; Moore, R.S.
1995-01-01
The Energy Information Administration of the U.S. Department of Energy (DOE) collects data annually from commercial nuclear power reactors via the Nuclear Fuel Data survey, Form RW-859. Over the past three years, the survey has collected data on the quantities and types of nonfuel components and on the quantities and contents of canisters in storage at reactor sites. This paper focuses on the annual changes in the data, specific implications of these changes, and lessons that should be applied to future revisions of the study. The total number of canisters reported by utilities for each year from 1986 to 1993 is listed. Changes in the quantities of nonfuel components report by General Reactors from 1992 to 1993 are also provided. Comparisons of canister and nonfuel components components data from year to year and from reactor to reactor point out that survey questions on these topics have been interpreted differently by reactor personnel
Summary: special waste form lysimeters - arid program
International Nuclear Information System (INIS)
Skaggs, R.L.; Walter, M.B.
1987-01-01
The purpose of the Special Waste Form Lysimeters - Arid Program is to determine the performance of solidified commercial low-level waste forms using a field-scale lysimeter facility constructed for measuring the release and migration of radionuclides from the waste forms. The performance of these waste forms, as measured by radionuclide concentrations in lysimeter effluent, will be compared to that predicted by laboratory characterization of the waste forms. Waste forms being tested include nuclear power reactor waste streams that have been solidified in cement, Dow polymer, and bitumen. To conduct the field leaching experiments a lysimeter facility was built to measure leachate under actual environmental conditions. Field-scale samples of waste were buried in lysimeters equipped to measure water balance components, effluent radionuclide concentrations, and to a limited extent, radionuclide concentrations in lysimeter soil samples. The waste forms are being characterized by standard laboratory leach tests to obtain estimates of radionuclide release. These estimates will be compared to leach rates observed in the field. Adsorption studies are being conducted to determine the amount of contaminant available for transport after the release. Theoretical solubility calculations will also be performed to investigate whether common solid phases could be controlling radionuclide release. 4 references, 8 figures, 1 table
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
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.)
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
Iodine waste form summary report (FY 2007)
International Nuclear Information System (INIS)
Krumhansl, James Lee; Nenoff, Tina Maria; McMahon, Kevin A.; Gao, Huizhen; Rajan, Ashwath Natech
2007-01-01
This new program at Sandia is focused on Iodine waste form development for GNEP cycle needs. Our research has a general theme of 'Waste Forms by Design' in which we are focused on silver loaded zeolite waste forms and related metal loaded zeolites that can be validated for chosen GNEP cycle designs. With that theme, we are interested in materials flexibility for iodine feed stream and sequestration material (in a sense, the ability to develop a universal material independent on the waste stream composition). We also are designing the flexibility to work in a variety of repository or storage scenarios. This is possible by studying the structure/property relationship of existing waste forms and optimizing them to our current needs. Furthermore, by understanding the properties of the waste and the storage forms we may be able to predict their long-term behavior and stability. Finally, we are working collaboratively with the Waste Form Development Campaign to ensure materials durability and stability testing
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
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
Effects of brine migration on waste storage systems. Final report
International Nuclear Information System (INIS)
Gaffney, E.S.; Nickell, R.E.
1979-01-01
Processes which can lead to mobilization of brine adjacent to spent fuel or nuclear waste canisters and some of the thermomechanical consequences have been investigated. Velocities as high as 4 x 10 -7 m s -1 (13 m y -1 ) are calculated at the salt/canister boundary. As much as 40 liters of pure NaCl brine could accumulate around each canister during a 10-year storage period. Accumulations of bittern brines would probably be less, in the range of 2 to 5 liters. With 0.5% water, NaCl brine accumulation over a 10-year storage cycle around a spent fuel canister producing 0.6 kW of heat is expected to be less than 1 liter for centimeter-size inclusions and less than 0.5 liter for millimeter-size inclusions. For bittern brines, about 25 years would be required to accumulate 0.4 liter. The most serious mechanical consequence of brine migration would be the increased mobility of the waste canister due to pressure solution. In pressure solution enhanced deformation, the existence of a thin film of fluid either between grains or between media (such as between a canister and the salt) provides a pathway by which the salt can be redistributed leading to a marked increase in strain rates in wet rock relative to dry rock. In salt, intergranular water will probably form discontinuous layers rather than films so that they would dominate pressure solution. A mathematical model of pressure solution indicates that pressure solution will not lead to appreciable canister motions except possibly in fine grained rocks (less than 10 -4 m). In fine grained salts, details of the contact surface between the canister and the salt bed may lead to large pressure solution motions. A numerical model indicates that heat transfer in the brine layer surrounding a spent fuel canister is not conduction dominated but has a significant convective component
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
Igneous Intrusion Impacts on Waste Packages and Waste Forms
International Nuclear Information System (INIS)
P. Bernot
2004-01-01
The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and
Igneous Intrusion Impacts on Waste Packages and Waste Forms
Energy Technology Data Exchange (ETDEWEB)
P. Bernot
2004-08-16
The purpose of this model report is to assess the potential impacts of igneous intrusion on waste packages and waste forms in the emplacement drifts at the Yucca Mountain Repository. The model is based on conceptual models and includes an assessment of deleterious dynamic, thermal, hydrologic, and chemical impacts. This constitutes the waste package and waste form impacts submodel of the Total System Performance Assessment for the License Application (TSPA-LA) model assessing the impacts of a hypothetical igneous intrusion event on the repository total system performance. This submodel is carried out in accordance with Technical Work Plan for Waste Form Degradation Modeling, Testing, and Analyses in Support of SR and LA (BSC 2003a) and Total System Performance Assessment-License Application Methods and Approaches (BSC 2002a). The technical work plan is governed by the procedures of AP-SIII.10Q, Models. Any deviations from the technical work plan are documented in the TSPA-LA approach to implementing the models for waste package and waste form response during igneous intrusion is based on identification of damage zones. Zone 1 includes all emplacement drifts intruded by the basalt dike, and Zone 2 includes all other emplacement drifts in the repository that are not in Zone 1. This model report will document the following model: (1) Impacts of magma intrusion on the components of engineered barrier system (e.g., drip shields and cladding) of emplacement drifts in Zone 1, and the fate of waste forms. (2) Impacts of conducting magma heat and diffusing magma gases on the drip shields, waste packages, and cladding in the Zone 2 emplacement drifts adjacent to the intruded drifts. (3) Impacts of intrusion on Zone 1 in-drift thermal and geochemical environments, including seepage hydrochemistry. The scope of this model only includes impacts to the components stated above, and does not include impacts to other engineered barrier system (EBS) components such as the invert and
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
Fundamental Thermodynamics of Actinide-Bearing Mineral Waste Forms
International Nuclear Information System (INIS)
Williamson, Mark A.; Ebbinghaus, Bartley B.; Navrotsky, Alexandria
1999-01-01
The recent arms reduction treaties between the U.S. and Russia have resulted in inventories of plutonium in excess of current defense needs. Storage of this material poses significant, and unnecessary, risks of diversion, especially for Russia whose infrastructure for protecting these materials has been weakened since the collapse of the Soviet Union. Moreover, maintaining and protecting these materials in their current form is costly. The United States has about sixty metric tons of excess plutonium, half of which is high-purity weapon material. This high purity material will be converted into mixed oxide (MOX) fuel for use in nuclear reactors. The less pure excess plutonium does not meet the specifications for MOX fuel and will not be purified to meet the fuel specifications. Instead, it will be immobilized directly in a ceramic. The ceramic will be encased in a high level waste (HLW) glass monolith (i.e., the can-in-canister option) thus making a form that simulates the intrinsic security of spent nuclear fuel. The immobilized product will be placed in a HLW repository. To meet the repository requirements, the product must be shown to be durable for the intended storage time, the host matrix must be stable in the radiation environment, the solubility and leaching characteristics of the plutonium in the host material must be established, and optimum processing parameters must be determined for the entire compositional envelope of feed materials. In order to provide technically sound solutions to these issues, thermodynamic data are essential in developing an understanding of the chemistry and phase equilibria of the actinide-bearing mineral waste forms proposed as immobilization matrices. However, the relevant thermodynamic data (e.g., enthalpy, entropy, and heat capacity) for the ceramic forms are severely lacking and this information gap directly affects the Energy Department's ability to license the disposal matrices and methods. High-temperature solution
Waste disposal options report. Volume 2
International Nuclear Information System (INIS)
Russell, N.E.; McDonald, T.G.; Banaee, J.; Barnes, C.M.; Fish, L.W.; Losinski, S.J.; Peterson, H.K.; Sterbentz, J.W.; Wenzel, D.R.
1998-02-01
Volume 2 contains the following topical sections: estimates of feed and waste volumes, compositions, and properties; evaluation of radionuclide inventory for Zr calcine; evaluation of radionuclide inventory for Al calcine; determination of k eff for high level waste canisters in various configurations; review of ceramic silicone foam for radioactive waste disposal; epoxides for low-level radioactive waste disposal; evaluation of several neutralization cases in processing calcine and sodium-bearing waste; background information for EFEs, dose rates, watts/canister, and PE-curies; waste disposal options assumptions; update of radiation field definition and thermal generation rates for calcine process packages of various geometries-HKP-26-97; and standard criteria of candidate repositories and environmental regulations for the treatment and disposal of ICPP radioactive mixed wastes
Waste disposal options report. Volume 2
Energy Technology Data Exchange (ETDEWEB)
Russell, N.E.; McDonald, T.G.; Banaee, J.; Barnes, C.M.; Fish, L.W.; Losinski, S.J.; Peterson, H.K.; Sterbentz, J.W.; Wenzel, D.R.
1998-02-01
Volume 2 contains the following topical sections: estimates of feed and waste volumes, compositions, and properties; evaluation of radionuclide inventory for Zr calcine; evaluation of radionuclide inventory for Al calcine; determination of k{sub eff} for high level waste canisters in various configurations; review of ceramic silicone foam for radioactive waste disposal; epoxides for low-level radioactive waste disposal; evaluation of several neutralization cases in processing calcine and sodium-bearing waste; background information for EFEs, dose rates, watts/canister, and PE-curies; waste disposal options assumptions; update of radiation field definition and thermal generation rates for calcine process packages of various geometries-HKP-26-97; and standard criteria of candidate repositories and environmental regulations for the treatment and disposal of ICPP radioactive mixed wastes.
Waste forms for plutonium disposition
International Nuclear Information System (INIS)
Johnson, S.G.; O'Holleran, T.P.; Frank, S.M.; Meyer, M.K.; Hanson, M.; Staples, B.A.; Knecht, D.A.; Kong, P.C.
1997-01-01
The field of plutonium disposition is varied and of much importance, since the Department of Energy has decided on the hybrid option for disposing of the weapons materials. This consists of either placing the Pu into mixed oxide fuel for reactors or placing the material into a stable waste form such as glass. The waste form used for Pu disposition should exhibit certain qualities: (1) provide for a suitable deterrent to guard against proliferation; (2) be of minimal volume, i.e., maximize the loading; and (3) be reasonably durable under repository-like conditions. This paper will discuss several Pu waste forms that display promising characteristics
Mixed low-level waste form evaluation
International Nuclear Information System (INIS)
Pohl, P.I.; Cheng, Wu-Ching; Wheeler, T.; Waters, R.D.
1997-01-01
A scoping level evaluation of polyethylene encapsulation and vitreous waste forms for safe storage of mixed low-level waste was performed. Maximum permissible radionuclide concentrations were estimated for 15 indicator radionuclides disposed of at the Hanford and Savannah River sites with respect to protection of the groundwater and inadvertent intruder pathways. Nominal performance improvements of polyethylene and glass waste forms relative to grout are reported. These improvements in maximum permissible radionuclide concentrations depend strongly on the radionuclide of concern and pathway. Recommendations for future research include improving the current understanding of the performance of polymer waste forms, particularly macroencapsulation. To provide context to these estimates, the concentrations of radionuclides in treated DOE waste should be compared with the results of this study to determine required performance
Waste form development for a DC arc furnace
Energy Technology Data Exchange (ETDEWEB)
Feng, X.; Bloomer, P.E.; Chantaraprachoom, N.; Gong, M.; Lamar, D.A.
1996-09-01
A laboratory crucible study was conducted to develop waste forms to treat nonradioactive simulated {sup 238}Pu heterogeneous debris waste from Savannah River, metal waste from the Idaho National Engineering Laboratory (INEL), and nominal waste also from INEL using DC arc melting. The preliminary results showed that the different waste form compositions had vastly different responses for each processing effect. The reducing condition of DC arc melting had no significant effects on the durability of some waste forms while it decreased the waste form durability from 300 to 700% for other waste forms, which resulted in the failure of some TCLP tests. The right formulations of waste can benefit from devitrification and showed an increase in durability by 40%. Some formulations showed no devitrification effects while others decreased durability by 200%. Increased waste loading also affected waste form behavior, decreasing durability for one waste, increasing durability by 240% for another, and showing no effect for the third waste. All of these responses to the processing and composition variations were dictated by the fundamental glass chemistry and can be adjusted to achieve maximal waste loading, acceptable durability, and desired processing characteristics if each waste formulation is designed for the result according to the glass chemistry.
Leaching behavior of phosphate-bonded ceramic waste forms
International Nuclear Information System (INIS)
Singh, D.; Wagh, A.S.; Jeong, S.Y.; Dorf, M.
1996-04-01
Over the last few years, Argonne National Laboratory has been developing room-temperature-setting chemically bonded phosphate ceramics for solidifying and stabilizing low-level mixed wastes. This technology is crucial for stabilizing waste streams that contain volatile species and off-gas secondary waste streams generated by high-temperature treatment of such wastes. We have developed a magnesium phosphate ceramic to treat mixed wastes such as ash, salts, and cement sludges. Waste forms of surrogate waste streams were fabricated by acid-base reactions between the mixtures of magnesium oxide powders and the wastes, and phosphoric acid or acid phosphate solutions. Dense and hard ceramic waste forms are produced in this process. The principal advantage of this technology is that the contaminants are immobilized by both chemical stabilization and subsequent microencapsulation of the reaction products. This paper reports the results of durability studies conducted on waste forms made with ash waste streams spiked with hazardous and radioactive surrogates. Standard leaching tests such as ANS 16.1 and TCLP were conducted on the final waste forms. Fates of the contaminants in the final waste forms were established by electron microscopy. In addition, stability of the waste forms in aqueous environments was evaluated with long-term water-immersion tests
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.)
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
Advanced waste form and Melter development for treatment of troublesome high-level wastes
Energy Technology Data Exchange (ETDEWEB)
Marra, James [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Kim, Dong -Sang [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Maio, Vincent [Idaho National Lab. (INL), Idaho Falls, ID (United States)
2015-10-01
the as poured state and after being slowly cooled according to the canister centerline cooling (CCC) profile. Glass formulation development was also completed on other Hanford tank wastes that were identified to further challenge waste loading due to the presence of appreciable quantities (>750 g) of plutonium in the waste tanks. In addition to containing appreciable Pu quantities, the C-102 waste tank and the 244-TX waste tank contain high concentrations of aluminum and iron, respectively that will further challenge vitrification processing. Glass formulation testing also demonstrated that high waste loadings could be achieved with these tank compositions using the attributes afforded by the CCIM technology.
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
Glassy slags as novel waste forms for remediating mixed wastes with high metal contents
International Nuclear Information System (INIS)
Feng, X.; Wronkiewicz, D.J.; Bates, J.K.; Brown, N.R.; Buck, E.C.; Gong, M.; Ebert, W.L.
1994-01-01
Argonne National Laboratory (ANL) is developing a glassy slag final waste form for the remediation of low-level radioactive and mixed wastes with high metal contents. This waste form is composed of various crystalline and metal oxide phases embedded in a silicate glass phase. This work indicates that glassy slag shows promise as final waste form because (1) it has similar or better chemical durability than high-level nuclear waste (HLW) glasses, (2) it can incorporate large amounts of metal wastes, (3) it can incorporate waste streams having low contents of flux components (boron and alkalis), (4) it has less stringent processing requirements (e.g., viscosity and electric conductivity) than glass waste forms, (5) its production can require little or no purchased additives, which can result in greater reduction in waste volume and overall treatment costs. By using glassy slag waste forms, minimum additive waste stabilization approach can be applied to a much wider range of waste streams than those amenable only to glass waste forms
Diffusion from cylindrical waste forms
International Nuclear Information System (INIS)
Thomas, G.F.
1985-05-01
The diffusion of a single component material from a finite cylindrical waste form, initially containing a uniform concentration of the material, is investigated. Under the condition that the cylinder is maintained in a well-stirred bath, expressions for the fractional inventory leached and the leach rate are derived with allowance for the possible permanent immobilization of the diffusant through its decay to a stable product and/or its irreversible reaction with the waste form matrix. The usefulness of the reported results in nuclear waste disposal applications is emphasized. The results reported herein are related to those previously derived at Oak Ridge National Laboratory by Bell and Nestor. A numerical scheme involving the partial decoupling of nested infinite summations and the use of rapidly converging rational approximants is recommended for the efficient implementation of the expressions derived to obtain reliable estimates of the bulk diffusion constant and the rate constant describing the diffusant-waste form interaction from laboratory data
Alternative-waste-form evaluation for Savannah River Plant high-level waste
International Nuclear Information System (INIS)
Gould, T.H. Jr.; Crandall, J.L.
1982-01-01
Results of the waste form evaluation are summarized as: risks of human exposure are comparable and extremely small for either borosilicate glass or Synroc ceramic. Waste form properties are more than adequate for either form. The waste form decision can therefore be made on the basis of practicality and cost effectiveness. Synroc offers lower costs for transportation and emplacement. The borosilicate glass form offers the lowest total disposal cost, much simpler and less costly production, an established and proven process, lower future development costs, and an earlier startup of the DWPF
Waste Acceptance Testing of Secondary Waste Forms: Cast Stone, Ceramicrete and DuraLith
International Nuclear Information System (INIS)
Mattigod, Shas V.; Westsik, Joseph H.; Chung, Chul-Woo; Lindberg, Michael J.; Parker, Kent E.
2011-01-01
To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions has initiated secondary-waste-form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is conducting tests on four candidate waste forms to evaluate their ability to meet potential waste acceptance criteria for immobilized secondary wastes that would be placed in the IDF. All three waste forms demonstrated compressive strengths above the minimum 3.45 MPa (500 psi) set as a target for cement-based waste forms. Further, none of the waste forms showed any significant degradation in compressive strength after undergoing thermal cycling (30 cycles in a 10 day period) between -40 C and 60 C or water immersion for 90 days. The three leach test methods are intended to measure the diffusion rates of contaminants from the waste forms. Results are reported in terms of diffusion coefficients and a leachability index (LI) calculated based on the diffusion coefficients. A smaller diffusion coefficient and a larger LI are desired. The NRC, in its Waste Form Technical Position (NRC 1991), provides recommendations and guidance regarding methods to demonstrate waste stability for land disposal of radioactive waste. Included is a recommendation to conduct leach tests using the ANS 16.1 method. The resulting leachability index (LI) should be greater than 6.0. For Hanford secondary wastes, the LI > 6.0 criterion applies to sodium leached from the waste form. For technetium and iodine, higher targets of LI > 9 for Tc and LI > 11 for iodine have been set based on early waste-disposal risk and performance assessment analyses. The results of these three leach tests conducted for a total time between 11days (ASTM C1308) to 90 days (ANS 16.1) showed: (1) Technetium diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that
Waste Acceptance Testing of Secondary Waste Forms: Cast Stone, Ceramicrete and DuraLith
Energy Technology Data Exchange (ETDEWEB)
Mattigod, Shas V.; Westsik, Joseph H.; Chung, Chul-Woo; Lindberg, Michael J.; Parker, Kent E.
2011-08-12
To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions has initiated secondary-waste-form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is conducting tests on four candidate waste forms to evaluate their ability to meet potential waste acceptance criteria for immobilized secondary wastes that would be placed in the IDF. All three waste forms demonstrated compressive strengths above the minimum 3.45 MPa (500 psi) set as a target for cement-based waste forms. Further, none of the waste forms showed any significant degradation in compressive strength after undergoing thermal cycling (30 cycles in a 10 day period) between -40 C and 60 C or water immersion for 90 days. The three leach test methods are intended to measure the diffusion rates of contaminants from the waste forms. Results are reported in terms of diffusion coefficients and a leachability index (LI) calculated based on the diffusion coefficients. A smaller diffusion coefficient and a larger LI are desired. The NRC, in its Waste Form Technical Position (NRC 1991), provides recommendations and guidance regarding methods to demonstrate waste stability for land disposal of radioactive waste. Included is a recommendation to conduct leach tests using the ANS 16.1 method. The resulting leachability index (LI) should be greater than 6.0. For Hanford secondary wastes, the LI > 6.0 criterion applies to sodium leached from the waste form. For technetium and iodine, higher targets of LI > 9 for Tc and LI > 11 for iodine have been set based on early waste-disposal risk and performance assessment analyses. The results of these three leach tests conducted for a total time between 11days (ASTM C1308) to 90 days (ANS 16.1) showed: (1) Technetium diffusivity: ANSI/ANS 16.1, ASTM C1308, and EPA 1315 tests indicated that
Choices of canisters and elements for the first fuel and canister sludge shipment from K East Basin
International Nuclear Information System (INIS)
Makenas, B.J.
1996-01-01
The K East Basin contains open-top canisters with up to fourteen N Reactor fuel assemblies distributed between the two barrels of each canister. Each fuel assembly generally consists of inner and outer concentric elements fabricated from uranium metal with zirconium alloy cladding. The canisters also contain varying amounts of accumulated sludge. Retrieval of sample fuel elements and associated sludge for examination is scheduled to occur in the near future. The purpose of this document is to specify particular canisters and elements of interest as candidate sources of fuel and sludge to be shipped to laboratories
Corrosion studies on PREPP waste form
International Nuclear Information System (INIS)
Welch, J.M.; Neilson, R.M. Jr.
1984-05-01
Deformation or Failure Test and Accelerated Corrosion Test procedures were conducted to investigate the effect of formulation variables on the corrosion of oversize waste in Process Experimental Pilot Plant (PREPP) concrete waste forms. The Deformation or Failure Test did not indicate substantial waste form swelling from corrosion. The presence or absence of corrosion inhibitor was the most significant factor relative to measured half-cell potentials identified in the Accelerated Corrosion Test. However, corrosion inhibitor was determined to be only marginally beneficial. While this study produced no evidence that corrosion is of sufficient magnitude to produce serious degradation of PREPP waste forms, the need for corrosion rate testing is suggested. 11 references, 4 figures, 8 tables
Energy Technology Data Exchange (ETDEWEB)
Saslow, Sarah A.; Um, Wooyong; Russell, Renee L.
2017-08-02
This report describes the results from grout formulation and cementitious waste form qualification testing performed by Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions, LLC (WRPS). These results are part of a screening test that investigates three grout formulations proposed for wide-range treatment of different waste stream compositions expected for the Hanford Effluent Management Facility (EMF) evaporator bottom waste. This work supports the technical development need for alternative disposition paths for the EMF evaporator bottom wastes and future direct feed low-activity waste (DFLAW) operations at the Hanford Site. High-priority activities included simulant production, grout formulation, and cementitious waste form qualification testing. The work contained within this report relates to waste form development and testing, and does not directly support the 2017 Integrated Disposal Facility (IDF) performance assessment (PA). However, this work contains valuable information for use in PA maintenance past FY 2017 and future waste form development efforts. The provided results and data should be used by (1) cementitious waste form scientists to further the understanding of cementitious leach behavior of contaminants of concern (COCs), (2) decision makers interested in off-site waste form disposal, and (3) the U.S. Department of Energy, their Hanford Site contractors and stakeholders as they assess the IDF PA program at the Hanford Site. The results reported help fill existing data gaps, support final selection of a cementitious waste form for the EMF evaporator bottom waste, and improve the technical defensibility of long-term waste form risk estimates.
Glass-Ceramic Waste Forms for Uranium and Plutonium Residues Wastes - 13164
International Nuclear Information System (INIS)
Stewart, Martin W.A.; Moricca, Sam A.; Zhang, Yingjie; Day, R. Arthur; Begg, Bruce D.; Scales, Charlie R.; Maddrell, Ewan R.; Hobbs, Jeff
2013-01-01
A program of work has been undertaken to treat plutonium-residues wastes at Sellafield. These have arisen from past fuel development work and are highly variable in both physical and chemical composition. The principal radiological elements present are U and Pu, with small amounts of Th. The waste packages contain Pu in amounts that are too low to be economically recycled as fuel and too high to be disposed of as lower level Pu contaminated material. NNL and ANSTO have developed full-ceramic and glass-ceramic waste forms in which hot-isostatic pressing is used as the consolidation step to safely immobilize the waste into a form suitable for long-term disposition. We discuss development work on the glass-ceramic developed for impure waste streams, in particular the effect of variations in the waste feed chemistry glass-ceramic. The waste chemistry was categorized into actinides, impurity cations, glass formers and anions. Variations of the relative amounts of these on the properties and chemistry of the waste form were investigated and the waste form was found to be largely unaffected by these changes. This work mainly discusses the initial trials with Th and U. Later trials with larger variations and work with Pu-doped samples further confirmed the flexibility of the glass-ceramic. (authors)
International Nuclear Information System (INIS)
Takase, Hiroyasu; Benbow, S.; Grindrod, P.
1998-10-01
According to the current design of SKB, a copper overpack with a cast steel inner component will be used as the disposal canister for spent nuclear fuel. A recent study considered the case of a breach in the copper overpack, through which groundwater could enter the canister. It has pointed out that hydrogen gas generated by an anaerobic corrosion could cushion the system and reduce or eventually stop further infiltration of water into the breached canister, and thence the spent fuel. One potential pitfall in this previous study lies in the fact that it did not consider any processes which might violate the following assumptions which are essential for the gas 'cushioning': 1. Hydrogen gas accumulated in the annular gap in the canister forms a free gas phase which is stable indefinitely into future; 2. Elevated gas pressure in the canister prevents further supply of groundwater except for diffusion of vapour. In the current study we developed a set of mathematical models for the above problem and applied it to carry out an independent assessment of the long-term behaviour of the canister. A key aim in this study was to clarify whether there are any alternative processes which may affect the result obtained by the previous study by violating one of the assumptions listed above. For this purpose, a scenario development exercise was conducted. The result supported the concept described in the previous study. One exception is that possible intrusion of bentonite gel followed by its desaturation could leave paths both for the gas and water simultaneously without forming a gas cushion. This is summarised in the first part of the report. In the second part, development of mathematical models and their applications are described. The key results are: 1. The model describing behaviour of gas and pore water in the canister and the buffer material reproduced the main results of the previous study; 2. The model considering intrusion of the bentonite gel pointed out possibility
Secondary waste form testing: ceramicrete phosphate bonded ceramics
International Nuclear Information System (INIS)
Singh, D.; Ganga, R.; Gaviria, J.; Yusufoglu, Y.
2011-01-01
The cleanup activities of the Hanford tank wastes require stabilization and solidification of the secondary waste streams generated from the processing of the tank wastes. The treatment of these tank wastes to produce glass waste forms will generate secondary wastes, including routine solid wastes and liquid process effluents. Liquid wastes may include process condensates and scrubber/off-gas treatment liquids from the thermal waste treatment. The current baseline for solidification of the secondary wastes is a cement-based waste form. However, alternative secondary waste forms are being considered. In this regard, Ceramicrete technology, developed at Argonne National Laboratory, is being explored as an option to solidify and stabilize the secondary wastes. The Ceramicrete process has been demonstrated on four secondary waste formulations: baseline, cluster 1, cluster 2, and mixed waste streams. Based on the recipes provided by Pacific Northwest National Laboratory, the four waste simulants were prepared in-house. Waste forms were fabricated with three filler materials: Class C fly ash, CaSiO 3 , and Class C fly ash + slag. Optimum waste loadings were as high as 20 wt.% for the fly ash and CaSiO 3 , and 15 wt.% for fly ash + slag filler. Waste forms for physical characterizations were fabricated with no additives, hazardous contaminants, and radionuclide surrogates. Physical property characterizations (density, compressive strength, and 90-day water immersion test) showed that the waste forms were stable and durable. Compressive strengths were >2,500 psi, and the strengths remained high after the 90-day water immersion test. Fly ash and CaSiO 3 filler waste forms appeared to be superior to the waste forms with fly ash + slag as a filler. Waste form weight loss was ∼5-14 wt.% over the 90-day immersion test. The majority of the weight loss occurred during the initial phase of the immersion test, indicative of washing off of residual unreacted binder components from
Secondary waste form testing : ceramicrete phosphate bonded ceramics.
Energy Technology Data Exchange (ETDEWEB)
Singh, D.; Ganga, R.; Gaviria, J.; Yusufoglu, Y. (Nuclear Engineering Division); ( ES)
2011-06-21
The cleanup activities of the Hanford tank wastes require stabilization and solidification of the secondary waste streams generated from the processing of the tank wastes. The treatment of these tank wastes to produce glass waste forms will generate secondary wastes, including routine solid wastes and liquid process effluents. Liquid wastes may include process condensates and scrubber/off-gas treatment liquids from the thermal waste treatment. The current baseline for solidification of the secondary wastes is a cement-based waste form. However, alternative secondary waste forms are being considered. In this regard, Ceramicrete technology, developed at Argonne National Laboratory, is being explored as an option to solidify and stabilize the secondary wastes. The Ceramicrete process has been demonstrated on four secondary waste formulations: baseline, cluster 1, cluster 2, and mixed waste streams. Based on the recipes provided by Pacific Northwest National Laboratory, the four waste simulants were prepared in-house. Waste forms were fabricated with three filler materials: Class C fly ash, CaSiO{sub 3}, and Class C fly ash + slag. Optimum waste loadings were as high as 20 wt.% for the fly ash and CaSiO{sub 3}, and 15 wt.% for fly ash + slag filler. Waste forms for physical characterizations were fabricated with no additives, hazardous contaminants, and radionuclide surrogates. Physical property characterizations (density, compressive strength, and 90-day water immersion test) showed that the waste forms were stable and durable. Compressive strengths were >2,500 psi, and the strengths remained high after the 90-day water immersion test. Fly ash and CaSiO{sub 3} filler waste forms appeared to be superior to the waste forms with fly ash + slag as a filler. Waste form weight loss was {approx}5-14 wt.% over the 90-day immersion test. The majority of the weight loss occurred during the initial phase of the immersion test, indicative of washing off of residual unreacted
Development of Copper Canister through Cold Sprayed Coating Method
Energy Technology Data Exchange (ETDEWEB)
Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo; Lee, Jong Youl; Jeong, Jong Tae; Kim, Sung Ki; Cho, Dong Keun
2007-12-15
General thickness of a copper canister is 5 cm for a underground disposal application. The lower limit of a thickness is determined by a forging technology. But many experts in this area agrees that the thickness 1 cm is enough at the underground disposal for the life time of 1,000,000 years. Thus new technology is suggested for the making 1 cm thickness copper canister, that is a cold spray coating method(CSC). In this report, the CSC is examined and the technical possibility for making copper canister is measured. The overview of CSC and its characteristics are discussed. Various copper particles for the CSC are analyzed and the formed coating layers are examined to find their porosity and uniformity. A Tafa copper particle and Chang-sung copper particle are selected for making 1 cm thick test specimen. Using the CSC specimens, tensile test and XRD analysis are performed. As a corrosion evaluation, a electrochemical test such as a polarization test is done, together with humid corrosion test and chloric acid immersion test. Through the corrosion tests, it is tried to confirm that the CSC is valuable method for making a copper canister. Consequently, it is confirmed that the CSC method is very usful for making 1 cm thick copper canister. the porosity of CSC layer is very low at 0.3 in case of Tafa copper layer. In corrosion tests, the CSC layers are very stable in active environments. It is hard to say that the difference of processing method but the purity of copper is important for the corrosion rate evaluation. The CSC method is very effective method for making 1 cm thick copper canister, It is hoped that the CSC method is applied in a HLW underground disposal system in the future.
Development of Copper Canister through Cold Sprayed Coating Method
International Nuclear Information System (INIS)
Lee, Min Soo; Choi, Jong Won; Choi, Heui Joo; Lee, Jong Youl; Jeong, Jong Tae; Kim, Sung Ki; Cho, Dong Keun
2007-12-01
General thickness of a copper canister is 5 cm for a underground disposal application. The lower limit of a thickness is determined by a forging technology. But many experts in this area agrees that the thickness 1 cm is enough at the underground disposal for the life time of 1,000,000 years. Thus new technology is suggested for the making 1 cm thickness copper canister, that is a cold spray coating method(CSC). In this report, the CSC is examined and the technical possibility for making copper canister is measured. The overview of CSC and its characteristics are discussed. Various copper particles for the CSC are analyzed and the formed coating layers are examined to find their porosity and uniformity. A Tafa copper particle and Chang-sung copper particle are selected for making 1 cm thick test specimen. Using the CSC specimens, tensile test and XRD analysis are performed. As a corrosion evaluation, a electrochemical test such as a polarization test is done, together with humid corrosion test and chloric acid immersion test. Through the corrosion tests, it is tried to confirm that the CSC is valuable method for making a copper canister. Consequently, it is confirmed that the CSC method is very usful for making 1 cm thick copper canister. the porosity of CSC layer is very low at 0.3 in case of Tafa copper layer. In corrosion tests, the CSC layers are very stable in active environments. It is hard to say that the difference of processing method but the purity of copper is important for the corrosion rate evaluation. The CSC method is very effective method for making 1 cm thick copper canister, It is hoped that the CSC method is applied in a HLW underground disposal system in the future
Preliminary assessment of nine waste-form products/processes for immobilizing transuranic wastes
International Nuclear Information System (INIS)
Crisler, L.R.
1980-09-01
Nine waste-form processes for reduction of the present and projected Transuranic (TRU) waste inventory to an immobilized product have been evaluated. Product formulations, selected properties, preparation methods, technology status, problem areas needing resolution and location of current research development being pursued in the United States are discussed for each process. No definitive utility ranking is attempted due to the early stage of product/process development for TRU waste containing products and the uncertainties in the state of current knowledge of TRU waste feed compositional and quantitative makeup. Of the nine waste form products/processes included in this discussion, bitumen and cements (encapsulation agents) demonstrate the degree of flexibility necessary to immobilize the wide composition range present in the TRU waste inventory. A demonstrated process called Slagging Pyrolysis Incineration converts a varied compositional feed (municipal wastes) to a ''basalt'' like product. This process/product appears to have potential for TRU waste immobilization. The remaining waste forms (borosilicate glass, high-silica glass, glass ceramics, ''SYNROC B'' and cermets) have potential for immobilizing a smaller fraction of the TRU waste inventory than the above discussed waste forms
International Nuclear Information System (INIS)
1987-06-01
This report includes the options of horizontal and vertical emplacement, the addition of a phased repository, an additional waste form (intact spent fuel), revised geotechnical data appropriate for the Deaf Smith County site, new corrosion data for the container, and new repository design data. The waste package consists of waste form and canister within a thick-walled, low-carbon steel container surrounded by packing. The container is a hollow cylinder with a flat head welded to each end. The design concepts for the waste container or vertical and horizontal emplacement are identical. This report discusses the results of analyses of aspects of the reference waste package concept needing changes because of new data and information believed applicable to the Deaf Smith County site. Included are waste package conceptual designs or (1) the reference defense high-level waste form from the Savannah River Plant; (2) intact spent fuel with our pressurized-water-reactor or nine boiling-water-reactor assemblies per package for emplacement during Phase 1 of repository operation; and (3) spent fuel which has been disassembled and consolidated into a segmented cylindrical canister with rods from either 12 pressurized-water-reactor or 30 boiling-water-reactor assemblies per package for emplacement during Phase 2. 30 refs., 61 figs., 30 tabs
Directory of Open Access Journals (Sweden)
John Palmisano
2015-12-01
Full Text Available Canisters containing activated charcoal are commonly used in the laboratory setting to collect waste anesthetic gas (WAG. This requires the weighing of the WAG canister after each use and for investigators to maintain an accurate time log of anesthesia duration. A typical rodent anesthesia station may include the use of 3 WAG canisters; one for the anesthesia induction box, one for the operative table, and one for gas monitoring. To simplify the anesthesia breathing circuit, we have developed a “T” connector that replaces the need for having multiple WAG canisters. The “T” connector directs the waste anesthetic from multiple sources; the anesthesia induction box, operative table and gas monitor into a single WAG canister. Use of the “T” connector appears to be a safe, acceptable device that conveniently directs waste gas while improving charcoal adsorption within the canister. In addition, this device may have a positive impact on the environment with a secondary benefit of possible cost savings associated with the purchase and disposal of the hazardous waste contents.
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.)
International Nuclear Information System (INIS)
Lawroski, H.
1984-01-01
The promulgation of the amendment of 10 CFR Part 61 by the Nuclear Regulatory Commission of December 27, 1982 by Federal Register Notice with an effective date of December 27, 1983 established the criteria for licensing requirements, paragraph 60.56, contained the description to provide adequate stability of the site through the use of suitable waste forms. In May, 1983, the NRC published a final Branch Technical Position (BTP) paper on waste form. The position taken by the BTP was considerably more severe than indicated in 10 CFR Part 61. An extensive and expensive testing program was started in 1983. As an interim measure, the presently utilized solidification processes such as cement, Dow binder, Envirostone and bitumen, and the presently qualified High Integrity containers (HICs) were considered acceptable with the caveat that acceptable process control programs were being utilized. The NRC requested that topical reports for licenses be submitted. The topical reports were to contain test results to substantiate the acceptability of the waste forms. The test results to date show that the volume of wastes will have to increase to meet the position taken by the NRC in the BTP. This position will cause more waste to be generated which is contrary to the emphasis by states and others to reduce the volume of waste. The details of testing will be discussed in the paper to be presented
International Nuclear Information System (INIS)
Ross, W.A.; Lokken, R.O.; May, R.P.; Roberts, F.P.; Thornhill, R.E.; Timmerman, C.L.; Treat, R.L.; Westsik, J.H. Jr.
1982-09-01
This volume contains supporting information for the comparative assessment of the transuranic waste forms and processes summarized in Volume I. Detailed data on the characterization of the waste forms selected for the assessment, process descriptions, and cost information are provided. The purpose of this volume is to provide additional information that may be useful when using the data in Volume I and to provide greater detail on particular waste forms and processes. Volume II is divided into two sections and two appendixes. The first section provides information on the preparation of the waste form specimens used in this study and additional characterization data in support of that in Volume I. The second section includes detailed process descriptions for the eight processes evaluated. Appendix A lists the results of MCC-1 leach test and Appendix B lists additional cost data. 56 figures, 12 tables
Development of new waste form for treatment and disposal of concentrated liquid radioactive waste
International Nuclear Information System (INIS)
Kwak, Kyung Kil; Ji, Young Yong
2010-12-01
The radioactive waste form should be meet the waste acceptance criteria of national regulation and disposal site specification. We carried out a characterization of rad waste form, especially the characteristics of radioactivity, mechanical and physical-chemical properties in various rad waste forms. But asphalt products is not acceptable waste form at disposal site. Thus we are change the product materials. We select the development of the new process or new materials. The asphalt process is treatment of concentrated liquid and spent-resin and that we decide the Development of new waste form for treatment and disposal of concentrated liquid radioactive waste
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
Preparation and leaching of radioactive INEL waste forms
International Nuclear Information System (INIS)
Schuman, R.P.; Welch, J.M.; Staples, B.A.
1982-01-01
The purpose of this study is to prepare and leach test ceramic and glass waste form specimens produced from actual transuranic waste sludges and high-level waste calcines, respectively. Description of wastes, specimen fabrication, leaching procedure, analysis of leachates and results are discussed. The conclusion is that radioactive waste stored at INEL can be readily incorporated in fused ceramic and glass forms. Initial leach testing results indicate that these forms show great promise for safe long-term containment of radioactive wastes
The construction of solid waste form test facility
International Nuclear Information System (INIS)
Park, Hun Hwee; Kim, Joon Hyung; Lee, Byung Jik; Koo, Jun Mo; Kim, Jeong Guk; Jung, In Ha
1990-03-01
The solid waste form test facility (SWFTF) to test and/or evaluate the characteristics of waste forms, such as homogeniety, mechanical properties, thermal properties, waste resistance and leachability, have been constructed, and some equipments for testing actual waste forms has been purchased; radiocative monitoring system, glove box for the manipulator repair room, and uninteruppted power supply system, et al. Classifications of radioactive wastes, basic requirements and criteria to be considered during waste management were also reviewed. Some of the described items above have been standardized for the purpose of indigenigation. Therefore, safety assurance of waste forms, as well as increase in the range of participating of domestic companies in construction of further nuclear facilities could be obtained as results through constructing this facility. In the furture this facility is going to be utilized not only for the inspection of waste forms but also for the periodic decontamination for extending the life time of some expensive radiological equipments using remote handling techniques. (author)
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
Thermo-hydro-mechanical mode of canister retrieval test
International Nuclear Information System (INIS)
Zandarin, M.T.; Olivella, S.; Gens', A.; Alonso, E.E.
2010-01-01
Document available in extended abstract form only. The Canister Retrieval Tests (CRT) is a full scale in situ experiment performed by SKB at Aespoe Laboratory. The experiment involves placing a canister equipped with electrical heaters inside of a deposition hole bored in Aespoe diorite. The deposition hole is 8.55 metres deep and has a diameter of 1.76 metres. The space between canister and the hole is filled with a MX-80 bentonite buffer. The bentonite buffer was installed in form of blocks and rings of bentonite. At the top of the canister bentonite bricks occupy the volume between the canister top surface and the bottom surface of the plug. Due to the bentonite ring size there are two gaps; once between canister and buffer which was left empty and another one between buffer and rock that was filled with bentonite pellets. The top of the hole was sealed with a retaining plug composed of concrete and a steel plate. The plug was secured against heave caused by the swelling clay with nine cables anchored in the rock. An artificial pressurised saturation system was used because the supply of water from the rock was judged to be insufficient for saturating the buffer in a feasible time. A large number of instruments were installed to monitor the test as follows: - Canister - temperature and strain. - Rock mass - temperature and stress. - Retaining system - force and displacement. - Buffer - temperature, relative humidity, pore pressure and total pressure. After dismantling the tests the final dry density and water content of bentonite and pellets were measured. The comprehensive record of the Thermo-Hydro-Mechanical (THM) processes in the buffer give the possibility to investigate theoretical formulations and models, since the results of THM analyses can be checked against experimental data. As part of the European project THERESA, a 2-D axisymmetric model simulation of CRT bas been carried out. Some of the main objectives of this simulation are the study of the
Electrochemical/Pyrometallurgical Waste Stream Processing and Waste Form Fabrication
Energy Technology Data Exchange (ETDEWEB)
Steven Frank; Hwan Seo Park; Yung Zun Cho; William Ebert; Brian Riley
2015-07-01
This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration between US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and waste form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.
Equilibrium Temperature Profiles within Fission Product Waste Forms
Energy Technology Data Exchange (ETDEWEB)
Kaminski, Michael D. [Argonne National Lab. (ANL), Argonne, IL (United States)
2016-10-01
We studied waste form strategies for advanced fuel cycle schemes. Several options were considered for three waste streams with the following fission products: cesium and strontium, transition metals, and lanthanides. These three waste streams may be combined or disposed separately. The decay of several isotopes will generate heat that must be accommodated by the waste form, and this heat will affect the waste loadings. To help make an informed decision on the best option, we present computational data on the equilibrium temperature of glass waste forms containing a combination of these three streams.
Viscosity-based high temperature waste form compositions
International Nuclear Information System (INIS)
Reimann, G.A.
1994-01-01
High-temperature waste forms such as iron-enriched basalt are proposed to immobilize and stabilize a variety of low-level wastes stored at the Idaho National Engineering Laboratory. The combination of waste and soil anticipated for the waste form results in high SiO 2 + Al 2 O 3 producing a viscous melt in an arc furnace. Adding a flux such as CaO to adjust the basicity ratio (the molar ratio of basic to acid oxides) enables tapping the furnace without resorting to extreme temperatures, but adds to the waste volume. Improved characterization of wastes will permit adjusting the basicity ratio to between 0.7 and 1.0 by blending of wastes and/or changing the waste-soil ratio. This minimizes waste form volume. Also, lower pouring temperatures will decrease electrode and refractory attrition, reduce vaporization from the melt, and, with suitable flux, facilitate crystallization. Results of laboratory tests were favorable and pilot-scale melts are planned; however, samples have not yet been subjected to leach testing
ANSTO's waste forms for the 31. century
Energy Technology Data Exchange (ETDEWEB)
Vance, E.R.; Begg, B. D.; Day, R. A.; Moricca, S.; Perera, D. S.; Stewart, M. W. A.; Carter, M. L.; McGlinn, P. J.; Smith, K. L.; Walls, P. A.; Robina, M. La
2004-07-01
ANSTO waste form development for high-level radioactive waste is directed towards practical applications, particularly problematic niche wastes that do not readily lend themselves to direct vitrification. Integration of waste form chemistry and processing method is emphasised. Some longstanding misconceptions about titanate ceramics are dealt with. We have a range of titanate-bearing waste form products aimed at immobilisation of tank wastes and sludges, actinide-rich wastes, INEEL calcines and Na-bearing liquid wastes, Al-rich wastes arising from reprocessing of Al-clad fuels, Mo-rich wastes arising from reprocessing of U-Mo fuels, partitioned Cs-rich wastes, and {sup 99}Tc. Waste form production techniques cover hot isostatic and uniaxial pressing, sintering, and cold-crucible melting, and these are strongly integrated into waste form design. Speciation and leach resistance of Cs and alkalis in cementitious products and geo-polymers are being studied. Recently we have embarked on studies of candidate inert matrix fuels for Pu burning. We also have a considerable program directed at basic understanding of the waste forms in regard to crystal chemistry, dissolution behaviour in aqueous media, radiation damage effects and optimum processing techniques. (authors)
International Nuclear Information System (INIS)
McVey, D.F.; Erickson, K.L.; Seyfried, W.E. Jr.
1983-01-01
In this chapter the authors discuss the current status of heat and mass transport studies in the marine red clay sediments that are being considered as a nuclear waste isolation medium and review analytical and experimental studies. Calculations based on numerical models indicate that for a maximum allowable sediment-canister interface temperatures of 200 0 to 250 0 C, the sediment can absorb about 1.5kW initial power from waste buried 30 m in the sediment in a canister that is 3 m long and 0.3 m in diameter. The resulting fluid displacement due to convections is found to be small, less than 1 m. Laboratory studies of the geochemical effects induced by heating sediment-seawater mixtures indicate that the canister and waste form should be designed to resist a hot, relatively acidic oxidizing environment. Since the thermally altered sediment volume of about 5.5 m/sup 3/ is small relative to the sediment volume overlying the canister, the acid and oxidizing conditions should significantly affect the properties of the far field only if thermodiffusional process (Soret effect) prove to be significant. If thermodiffusional effects are important, however, near-field chemistry will differ considerably from that predicted from results of constant temperature sediment-seawater interaction experiments
Very deep borehole. Deutag's opinion on boring, canister emplacement and retrievability
Energy Technology Data Exchange (ETDEWEB)
Harrison, Tim [Well Engineering Partners BV, The Hague (Netherlands)
2000-05-01
An engineering feasibility study has been carried out to determine whether or not it is possible to drill the proposed Very Deep Borehole concept wells required by SKB for nuclear waste disposal. A conceptual well design has been proposed. All aspects of well design have been considered, including drilling tools, rig design, drilling fluids, casing design and annulus isolation. The proposed well design is for 1168.4 mm hole to be drilled to 500 m. A 1066.8 mm outer diameter (OD) casing will be run and cemented. A 1016 mm hole will be drilled to approximately 2000 m, where 914.4 mm OD casing will be run. This annulus will be sealed with bentonite slurry apart from the bottom 100 m which will be cemented. 838.2 mm hole will be drilled to a final depth of 4000 m, where 762 mm OD slotted casing will be run. All the hole sections will be drilled using a downhole hammer with foam as the drilling fluid medium. Prior to running each casing string, the hole will be displaced to mud to assist with casing running and cementing. The waste canisters will be run on a simple J-slot tool, with integral backup system in case the J-slot fails. The canisters will all be centralised. Canisters can be retrieved using the same tool as used to run them. Procedures are given for both running and retrieving. Logging and testing is recommended only in the exploratory wells, in a maximum hole size of 311.1 mm. This will require the drilling of pilot holes to enable logging and testing to take place. It is estimated that each well will take approximately 137 days to drill and case, at an estimated cost of 4.65 Meuro per well. This time and cost estimate does not include any logging, testing, pilot hole drilling or time taken to run the canisters. New technology developments to enhance the drilling process are required in recyclable foam systems, in hammer bit technology, and in the development of robust under-reamers. It is the authors conclusion that it is possible to drill the well with
Very deep borehole. Deutag's opinion on boring, canister emplacement and retrievability
International Nuclear Information System (INIS)
Harrison, Tim
2000-05-01
An engineering feasibility study has been carried out to determine whether or not it is possible to drill the proposed Very Deep Borehole concept wells required by SKB for nuclear waste disposal. A conceptual well design has been proposed. All aspects of well design have been considered, including drilling tools, rig design, drilling fluids, casing design and annulus isolation. The proposed well design is for 1168.4 mm hole to be drilled to 500 m. A 1066.8 mm outer diameter (OD) casing will be run and cemented. A 1016 mm hole will be drilled to approximately 2000 m, where 914.4 mm OD casing will be run. This annulus will be sealed with bentonite slurry apart from the bottom 100 m which will be cemented. 838.2 mm hole will be drilled to a final depth of 4000 m, where 762 mm OD slotted casing will be run. All the hole sections will be drilled using a downhole hammer with foam as the drilling fluid medium. Prior to running each casing string, the hole will be displaced to mud to assist with casing running and cementing. The waste canisters will be run on a simple J-slot tool, with integral backup system in case the J-slot fails. The canisters will all be centralised. Canisters can be retrieved using the same tool as used to run them. Procedures are given for both running and retrieving. Logging and testing is recommended only in the exploratory wells, in a maximum hole size of 311.1 mm. This will require the drilling of pilot holes to enable logging and testing to take place. It is estimated that each well will take approximately 137 days to drill and case, at an estimated cost of 4.65 Meuro per well. This time and cost estimate does not include any logging, testing, pilot hole drilling or time taken to run the canisters. New technology developments to enhance the drilling process are required in recyclable foam systems, in hammer bit technology, and in the development of robust under-reamers. It is the authors conclusion that it is possible to drill the well with
Review of radiation effects in solid-nuclear-waste forms
International Nuclear Information System (INIS)
Weber, W.J.
1981-09-01
Radiation effects on the stability of high-level nuclear waste (HLW) forms are an important consideration in the development of technology to immobilize high-level radioactive waste because such effects may significantly affect the containment of the radioactive waste. Since the required containment times are long (10 3 to 10 6 years), an understanding of the long-term cumulative effects of radiation damage on the waste forms is essential. Radiation damage of nuclear waste forms can result in changes in volume, leach rate, stored energy, structure/microstructure, and mechanical properties. Any one or combination of these changes might significantly affect the long-term stability of the nuclear waste forms. This report defines the general radiation damage problem in nuclear waste forms, describes the simulation techniques currently available for accelerated testing of nuclear waste forms, and reviews the available data on radiation effects in both glass and ceramic (primarily crystalline) waste forms. 76 references
Secondary Waste Form Down Selection Data Package – Ceramicrete
Energy Technology Data Exchange (ETDEWEB)
Cantrell, Kirk J.; Westsik, Joseph H.
2011-08-31
As part of high-level waste pretreatment and immobilized low activity waste processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed in the Integrated Disposal Facility. Currently, four waste forms are being considered for stabilization and solidification of the liquid secondary wastes. These waste forms are Cast Stone, Ceramicrete, DuraLith, and Fluidized Bed Steam Reformer. The preferred alternative will be down selected from these four waste forms. Pacific Northwest National Laboratory is developing data packages to support the down selection process. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilization and solidification of the liquid secondary wastes. The information included will be based on information available in the open literature and from data obtained from testing currently underway. This data package is for the Ceramicrete waste form. Ceramicrete is a relatively new engineering material developed at Argonne National Laboratory to treat radioactive and hazardous waste streams (e.g., Wagh 2004; Wagh et al. 1999a, 2003; Singh et al. 2000). This cement-like waste form can be used to treat solids, liquids, and sludges by chemical immobilization, microencapsulation, and/or macroencapsulation. The Ceramicrete technology is based on chemical reaction between phosphate anions and metal cations to form a strong, dense, durable, low porosity matrix that immobilizes hazardous and radioactive contaminants as insoluble phosphates and microencapsulates insoluble radioactive components and other constituents that do not form phosphates. Ceramicrete is a type of phosphate-bonded ceramic, which are also known as chemically bonded phosphate ceramics. The Ceramicrete
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
Backfill barriers for nuclear waste repositories in salt
Energy Technology Data Exchange (ETDEWEB)
Nowak, E J; Odoj, R; Merz, E [eds.
1981-06-01
Backfill materials were evaluated for containment of radionuclides, chemical modification of brine, and sensitivity to hydrothermal conditions. Experimental conditions were relevant to nuclear waste isolation in bedded salt. They were based on geologic conditions at the site of the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico, USA. Conclusions are: backfill mixtures surrounding the waste form and canister can provide a neutral or slightly acidic, potentially reducing environment, prevent convective aqueous flow, and act as an effective radionuclide migration barrier; bentonite is likely to remain hydrothermally stable but potentially sensitive to waste package interactions which could alter the pH, the ratio of dissolved ions, or the sorption properties of radionuclide species; effects of irradiation from high level waste should be investigated.
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
Extreme scenarios for nuclear waste repositories
Energy Technology Data Exchange (ETDEWEB)
Brown, M J [Harvard Univ., Cambridge, MA (USA). Div. of Applied Sciences; Crouch, E [Harvard Univ., Cambridge, MA (USA). Energy and Environmental Policy Center
1982-09-01
Two extreme scenarios for release of radioactive waste have been constructed. In the first, a volcanic eruption releases 1 km/sup 2/ of an underground nuclear waste repository, while in the second, waste enters the drinking water reservoir of a major city. With pessimistic assumptions, upper bounds on the number of cancers due to radiation are calculated. In the volcano scenario, the effects of the waste are smaller than the effects of natural radioactivity in the volcanic dust if the delay between emplacement and eruption exceeds 2000 yr. The consequences of the waste in drinking water depend on the survival time of the canisters and the rate of leaching of the nuclides from the waste matrix. For a canister life of 400 yr and a leach time of 6300 yr the cancer rate in the affected area would increase by 25%.
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
Talc-silicon glass-ceramic waste forms for immobilization of high- level calcined waste
International Nuclear Information System (INIS)
Vinjamuri, K.
1993-06-01
Talc-silicon glass-ceramic waste forms are being evaluated as candidates for immobilization of the high level calcined waste stored onsite at the Idaho Chemical Processing Plant. These glass-ceramic waste forms were prepared by hot isostatically pressing a mixture of simulated nonradioactive high level calcined waste, talc, silicon and aluminum metal additives. The waste forms were characterized for density, chemical durability, and glass and crystalline phase compositions. The results indicate improved density and chemical durability as the silicon content is increased
Processability analysis of candidate waste forms
International Nuclear Information System (INIS)
Gould, T.H. Jr.; Dunson, J.B. Jr.; Eisenberg, A.M.; Haight, H.G. Jr.; Mello, V.E.; Schuyler, R.L. III.
1982-01-01
A quantitative merit evaluation, or processability analysis, was performed to assess the relative difficulty of remote processing of Savannah River Plant high-level wastes for seven alternative waste form candidates. The reference borosilicate glass process was rated as the simplest, followed by FUETAP concrete, glass marbles in a lead matrix, high-silica glass, crystalline ceramics (SYNROC-D and tailored ceramics), and coated ceramic particles. Cost estimates for the borosilicate glass, high-silica glass, and ceramic waste form processing facilities are also reported
Immobilization in ceramic waste forms of the residues from treatment of mixed wastes
International Nuclear Information System (INIS)
Oversby, V.M.; van Konynenburg, R.A.; Glassley, W.E.; Curtis, P.G.
1993-11-01
The Environmental Restoration and Waste Management Applied Technology Program at LLNL is developing a Mixed Waste Management Facility to demonstrate treatment technologies that provide an alternative to incineration. As part of that program, we are developing final waste forms using ceramic processing methods for the immobilization of the treatment process residues. The ceramic phase assemblages are based on using Synroc D as a starting point and varying the phase assemblage to accommodate the differences in chemistry between the treatment process residues and the defense waste for which Synroc D was developed. Two basic formulations are used, one for low ash residues resulting from treatment of organic materials contaminated with RCRA metals, and one for high ash residues generated from the treatment of plastics and paper products. Treatment process residues are mixed with ceramic precursor materials, dried, calcined, formed into pellets at room temperature, and sintered at 1150 to 1200 degrees C to produce the final waste form. This paper discusses the chemical composition of the waste streams and waste forms, the phase assemblages that serve as hosts for inorganic waste elements, and the changes in waste form characteristics as a function of variation in process parameters
Corrosion aspects of high-level waste disposal in salt domes
International Nuclear Information System (INIS)
Roerbo, K.
1979-12-01
In the ELSAM/ELKRAT waste management project it is planned that the high-level waste is glassified, encapsuled in canisters and finally deposited in a deep hole drilled in a salt dome. In the present report corrosion aspects of the canisters after deposition are discussed. The chemical environment will probably be a limited amount of brine coming from brine inclusions in the surrounding salt and moving up against the temperature gradient, the temperature at the canister surface being in the range of 100-150degC. The possible types of corrosion and the expected corrosion rates for a number of potential canister materials (mild steel, austenitic and ferritic stainless steels, Ni-base alloys, copper, titanium and a few combinations of materials) are discussed. Mild steel (possibly combined with an inner layer of copper or titanium) might possibly be an appropriate choice of material for the canister. (author)
Evaluation and selection of candidate high-level waste forms
International Nuclear Information System (INIS)
1982-03-01
Seven candidate waste forms being developed under the direction of the Department of Energy's National High-Level Waste (HLW) Technology Program, were evaluated as potential media for the immobilization and geologic disposal of high-level nuclear wastes. The evaluation combined preliminary waste form evaluations conducted at DOE defense waste-sites and independent laboratories, peer review assessments, a product performance evaluation, and a processability analysis. Based on the combined results of these four inputs, two of the seven forms, borosilicate glass and a titanate based ceramic, SYNROC, were selected as the reference and alternative forms for continued development and evaluation in the National HLW Program. Both the glass and ceramic forms are viable candidates for use at each of the DOE defense waste-sites; they are also potential candidates for immobilization of commercial reprocessing wastes. This report describes the waste form screening process, and discusses each of the four major inputs considered in the selection of the two forms
Advanced waste forms from spent nuclear fuel
International Nuclear Information System (INIS)
Ackerman, J.P.; McPheeters, C.C.
1995-01-01
More than one hundred spent nuclear fuel types, having an aggregate mass of more than 5000 metric tons (2700 metric tons of heavy metal), are stored by the United States Department of Energy. This paper proposes a method for converting this wide variety of fuel types into two waste forms for geologic disposal. The method is based on a molten salt electrorefining technique that was developed for conditioning the sodium-bonded, metallic fuel from the Experimental Breeder Reactor-II (EBR-II) for geologic disposal. The electrorefining method produces two stable, optionally actinide-free, high-level waste forms: an alloy formed from stainless steel, zirconium, and noble metal fission products, and a ceramic waste form containing the reactive metal fission products. Electrorefining and its accompanying head-end process are briefly described, and methods for isolating fission products and fabricating waste forms are discussed
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
The characterization of cement waste form for final disposal of decommissioning concrete wastes
International Nuclear Information System (INIS)
Lee, Yoon-ji; Lee, Ki-Won; Min, Byung-Youn; Hwang, Doo-Seong; Moon, Jei-Kwon
2015-01-01
Highlights: • Decommissioning concrete waste recycling and disposal. • Compressive strength of cement waste form. • Characteristic of thermal resistance and leaching of cement waste form. - Abstract: In Korea, the decontamination and decommissioning of KRR-1, 2 at KAERI have been under way. The decommissioning of the KRR-2 was finished completely by 2011, whereas the decommissioning of KRR-1 is currently underway. A large quantity of slightly contaminated concrete waste has been generated from the decommissioning projects. The concrete wastes, 83ea of 200 L drums, and 41ea of 4 m 3 containers, were generated in the decommissioning projects. The conditioning of concrete waste is needed for final disposal. Concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled with a void space after concrete rubble pre-placement into 200 L drums. Thus, this research developed an optimizing mixing ratio of concrete waste, water, and cement, and evaluated the characteristics of a cement waste form to meet the requirements specified in the disposal site specific waste acceptance criteria. The results obtained from a compressive strength test, leaching test, and thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested as an optimized mixing ratio of 75:15:10. In addition, the compressive strength of the cement waste form was satisfied, including a fine powder up to a maximum of 40 wt% in concrete debris waste of about 75%. According to the scale-up test, the mixing ratio of concrete waste, water, and cement is 75:10:15, which meets the satisfied compressive strength because of an increase in the particle size in the waste
Engineering-Scale Demonstration of DuraLith and Ceramicrete Waste Forms
Energy Technology Data Exchange (ETDEWEB)
Josephson, Gary B.; Westsik, Joseph H.; Pires, Richard P.; Bickford, Jody; Foote, Martin W.
2011-09-23
To support the selection of a waste form for the liquid secondary wastes from the Hanford Waste Immobilization and Treatment Plant, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing on four candidate waste forms. Two of the candidate waste forms have not been developed to scale as the more mature waste forms. This work describes engineering-scale demonstrations conducted on Ceramicrete and DuraLith candidate waste forms. Both candidate waste forms were successfully demonstrated at an engineering scale. A preliminary conceptual design could be prepared for full-scale production of the candidate waste forms. However, both waste forms are still too immature to support a detailed design. Formulations for each candidate waste form need to be developed so that the material has a longer working time after mixing the liquid and solid constituents together. Formulations optimized based on previous lab studies did not have sufficient working time to support large-scale testing. The engineering-scale testing was successfully completed using modified formulations. Further lab development and parametric studies are needed to optimize formulations with adequate working time and assess the effects of changes in raw materials and process parameters on the final product performance. Studies on effects of mixing intensity on the initial set time of the waste forms are also needed.
Alternative High-Performance Ceramic Waste Forms
Energy Technology Data Exchange (ETDEWEB)
Sundaram, S. K. [Alfred Univ., NY (United States)
2017-02-01
This final report (M5NU-12-NY-AU # 0202-0410) summarizes the results of the project titled “Alternative High-Performance Ceramic Waste Forms,” funded in FY12 by the Nuclear Energy University Program (NEUP Project # 12-3809) being led by Alfred University in collaboration with Savannah River National Laboratory (SRNL). The overall focus of the project is to advance fundamental understanding of crystalline ceramic waste forms and to demonstrate their viability as alternative waste forms to borosilicate glasses. We processed single- and multiphase hollandite waste forms based on simulated waste streams compositions provided by SRNL based on the advanced fuel cycle initiative (AFCI) aqueous separation process developed in the Fuel Cycle Research and Development (FCR&D). For multiphase simulated waste forms, oxide and carbonate precursors were mixed together via ball milling with deionized water using zirconia media in a polyethylene jar for 2 h. The slurry was dried overnight and then separated from the media. The blended powders were then subjected to melting or spark plasma sintering (SPS) processes. Microstructural evolution and phase assemblages of these samples were studied using x-ray diffraction (XRD), scanning electron microscopy (SEM), energy dispersion analysis of x-rays (EDAX), wavelength dispersive spectrometry (WDS), transmission electron spectroscopy (TEM), selective area x-ray diffraction (SAXD), and electron backscatter diffraction (EBSD). These results showed that the processing methods have significant effect on the microstructure and thus the performance of these waste forms. The Ce substitution into zirconolite and pyrochlore materials was investigated using a combination of experimental (in situ XRD and x-ray absorption near edge structure (XANES)) and modeling techniques to study these single phases independently. In zirconolite materials, a transition from the 2M to the 4M polymorph was observed with increasing Ce content. The resulting
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
Hanford Waste Vitrification Plant Clean Air Act permit application
International Nuclear Information System (INIS)
1990-04-01
This document briefly describes the Hanford Site and provides a general overview of the Hanford Waste Vitrification Plant (HWVP). Other topics include sources of emissions, facility operating parameters, facility emissions, pollutant and radionuclide control technology and air quality. The HWVP will convert mixed wastes (high-activity radioactive and hazardous liquid wastes) to a solid vitrified form (borosilicate glass) for disposal. Mixed wastes pretreated in the Hanford Site B Plant will be pumped into double- shell tanks in the 200 East Area for interim storage. This pretreated mixed waste will be batch transferred from interim storage to the HWVP facility, where the waste will be concentrated by evaporation, treated with chemicals, and mixed with glass-forming materials. The mixture will then be continuously fed into an electrically heated glass melter. The molten glass will be poured into canisters that will be cooled, sealed, decontaminated, and stored until the vitrified product can be transferred to a geologic repository. 25 refs., 18 figs., 32 tabs
State of the art report on bituminized waste forms of radioactive wastes
International Nuclear Information System (INIS)
Kim, Tae Kook; Shon, Jong Sik; Kim, Kil Jeong; Lee, Kang Moo; Jung, In Ha
1998-03-01
In this report, research and development results on the bituminization of radioactive wastes are closely reviewed, especially those regarding waste treatment technologies, waste solidifying procedures and the characteristics of asphalt and solidified forms. A new concept of the bituminization method is suggested in this report which can improve the characteristics of solidified forms. Stable solid forms with high leach resistance, high thermal resistance and good compression strength were produced by the suggested bituminization method, in which spent polyethylene from agricultural farms was added. This report can help further research and development of improved bituminized forms of radioactive wastes that will maintain long term stabilities in disposal sites. (author). 59 refs., 19 tabs., 18 figs
Production of metal waste forms from spent fuel treatment
International Nuclear Information System (INIS)
Westphal, B.R.; Keiser, D.D.; Rigg, R.H.; Laug, D.V.
1995-01-01
Treatment of spent nuclear fuel at Argonne National Laboratory consists of a pyroprocessing scheme in which the development of suitable waste forms is being advanced. Of the two waste forms being proposed, metal and mineral, the production of the metal waste form utilizes induction melting to stabilize the waste product. Alloying of metallic nuclear materials by induction melting has long been an Argonne strength and thus, the transition to metallic waste processing seems compatible. A test program is being initiated to coalesce the production of the metal waste forms with current induction melting capabilities
Disposal criticality analysis methodology for fissile waste forms
International Nuclear Information System (INIS)
Davis, J.W.; Gottlieb, P.
1998-03-01
A general methodology has been developed to evaluate the criticality potential of the wide range of waste forms planned for geologic disposal. The range of waste forms include commercial spent fuel, high level waste, DOE spent fuel (including highly enriched), MOX using weapons grade plutonium, and immobilized plutonium. The disposal of these waste forms will be in a container with sufficiently thick corrosion resistant barriers to prevent water penetration for up to 10,000 years. The criticality control for DOE spent fuel is primarily provided by neutron absorber material incorporated into the basket holding the individual assemblies. For the immobilized plutonium, the neutron absorber material is incorporated into the waste form itself. The disposal criticality analysis methodology includes the analysis of geochemical and physical processes that can breach the waste package and affect the waste forms within. The basic purpose of the methodology is to guide the criticality control features of the waste package design, and to demonstrate that the final design meets the criticality control licensing requirements. The methodology can also be extended to the analysis of criticality consequences (primarily increased radionuclide inventory), which will support the total performance assessment for the respository
International Nuclear Information System (INIS)
Grutzeck, Michael; Jantzen, Carol M.
1999-01-01
Natural and synthetic zeolites are extremely versatile materials. They can adsorb a variety of liquids and gases, and also take part in cation exchange reactions. Zeolites are easy to synthesize from a wide variety of natural and man made materials. One combination of starting materials that exhibits a great deal of promise is a mixture of metakaolinite and/or Class F fly ash and concentrated sodium hydroxide solution. Once these ingredients are mixed and cured at elevated temperatures, they react to form a hard, dense, ceramic-like material that contains significant amounts of crystalline tectosilicates (zeolites and feldspathoids). Zeolites have the ability to sequester ions in lattice positions or within their networks of channels and voids. As such they are nearly perfect waste forms, the zeolites can host alkali, alkaline earth and a variety of higher valance cations. In addition to zeolites, it has been found that the zeolites are accompanied by an alkali aluminosilicate hydrate matrix that is a host, not only to the zeolites, but to residual amounts of insoluble hydroxide phases as well. A previous publication has established the fact that a mixture of a calcined equivalent ICPP waste (sodium aluminate/hydroxide solution containing ∼3:1 Na:Al) and fly ash and/or metakaolinite could be cured at various temperatures to produce a monolith containing Zeolite A (80 C) or Na-P1 plus hydroxy sodalite (130 C) crystals dispersed in an alkali aluminosilicate hydrate matrix. Dissolution tests have shown these materials (so-called hydroceramics) to have superior retention for alkali, alkaline earth and heavy metal ions. The zeolitization process is a simple one. Metakaolinite and/or Class F fly ash is mixed with a caustic sodium-bearing calcine and enough water to make a thick paste. The paste is transferred to a metal canister and ''soaked'' for a few hours at 70-80 C prior to steam autoclaving the sample at ∼200 C for 6-8 hours. The waste form produced in this
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)
Biodegradation testing of TMI-2 EPICOR-II waste forms
International Nuclear Information System (INIS)
Rogers, R.D.; McConnell, J.W. Jr.
1988-06-01
ASTM biodegradation tests were conducted on waste forms containing high specific activity ion exchange resins from EPICOR-II prefilters. Those tests were part of a program to test waste forms in accordance with the NRC Branch Technical Position on Waste Form. Small waste forms were manufactured using two different solidification agents, Portland Type I-II cement and vinyl ester-styrene (VES). Ion exchange material was taken from two EPICOR-II prefilters; PF-7, which contained all organic material, and PF-20, which contained organic resins and a layer of inorganic zeolites. Test results showed that the VES waste forms supported microbial growth, while cement waste forms did not support that growth. Growth was also observed adjacent to some VES waste forms. Radiation levels found in the ion exchange resins used in this study were not found to inhibit microbial growth. The extent of degradation of the waste forms could not be determined using the ASTM tests specified by the NRC Branch Technical Position on Waste Form. As a result of this work, a different testing methodology is recommended, which would provide direct verification of waste form capabilities. That methodology would evaluate solidification materials without using the ASTM procedures or subsequent compression testing. The proposed tests would provide exposure to a wide range of microbial species, use appropriately sized specimens, provide for possible use of alternate carbon sources, and extend the test length. Degradation would be determined directly by measuring metabolic activity or specimen weight loss. 16 refs., 15 figs., 3 tabs
Extreme scenarios for nuclear waste repositories
Energy Technology Data Exchange (ETDEWEB)
Brown, M J; Crouch, E
1982-09-01
Two extreme scenarios for release of radioactive waste have been constructed. In the first, a volcanic eruption releases 1 km2 of an underground nuclear waste repository, while in the second, waste enters the drinking water reservoir of a major city. With pessimistic assumptions, upper bounds on the number of cancers due to radiation are calculated. In the volcano scenario, the effects of the water are smaller than the effects of natural radioactivity in the volcanic dust if the delay between emplacement and eruption exceeds 2000 yr. The consequences of the waste in drinking water depend on the survival time of the canisters and the rate of leaching of the nuclides from the waste matrix. For a canister life of 400 yr and a leach time of 6300 yr the cancer rate in the affected area would increase by 25%.
Stress corrosion cracking of copper canisters
Energy Technology Data Exchange (ETDEWEB)
King, Fraser (Integrity Corrosion Consulting Limited (Canada)); Newman, Roger (Univ. of Toronto (Canada))
2010-12-15
films appear to be insufficiently adherent. A critical review of the surface mobility model is presented. It is argued that the formulation of the crack growth law is flawed and that, in its corrected form, predicted crack growth rates would be of the order of 10.20 m/s. Therefore, even if cracking were to occur via this mechanism, the crack velocity would be too small to lead to canister failure, even over repository timescales. Two other SCC mechanisms, the adsorption-induced dislocation emission and vacancy injection and embrittlement models, are also discussed. Although these models are still in the development stage, it is considered unlikely that they could induce cracking during the long-term anaerobic phase. Therefore, we conclude that the probability of SCC during the early aerobic period is low because of the absence of the necessary conditions for cracking and that there is no well-founded SCC mechanism that would result in cracking during the long-term anaerobic phase in the repository. Nevertheless, it is felt prudent to continue to study SCC of copper, especially under anaerobic conditions, in order to provide more support for the position that the stress corrosion cracking of copper canisters in the repository will not limit their service life
Stress corrosion cracking of copper canisters
International Nuclear Information System (INIS)
King, Fraser; Newman, Roger
2010-12-01
films appear to be insufficiently adherent. A critical review of the surface mobility model is presented. It is argued that the formulation of the crack growth law is flawed and that, in its corrected form, predicted crack growth rates would be of the order of 10.20 m/s. Therefore, even if cracking were to occur via this mechanism, the crack velocity would be too small to lead to canister failure, even over repository timescales. Two other SCC mechanisms, the adsorption-induced dislocation emission and vacancy injection and embrittlement models, are also discussed. Although these models are still in the development stage, it is considered unlikely that they could induce cracking during the long-term anaerobic phase. Therefore, we conclude that the probability of SCC during the early aerobic period is low because of the absence of the necessary conditions for cracking and that there is no well-founded SCC mechanism that would result in cracking during the long-term anaerobic phase in the repository. Nevertheless, it is felt prudent to continue to study SCC of copper, especially under anaerobic conditions, in order to provide more support for the position that the stress corrosion cracking of copper canisters in the repository will not limit their service life
CRYSTALLINE CERAMIC WASTE FORMS: REFERENCE FORMULATION REPORT
Energy Technology Data Exchange (ETDEWEB)
Brinkman, K.; Fox, K.; Marra, J.
2012-05-15
The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be successfully produced from a melting and crystallization process. The objective of this report is to explain the design of ceramic host systems culminating in a reference ceramic formulation for use in subsequent studies on process optimization and melt property data assessment in support of FY13 melter demonstration testing. The waste stream used as the basis for the development and testing is a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. In addition to the combined CS/LN/TM High Mo waste stream, variants without Mo and without Mo and Zr were also evaluated. Based on the results of fabricating and characterizing several simulated ceramic waste forms, two reference ceramic waste form compositions are recommended in this report. The first composition targets the CS/LN/TM combined waste stream with and without Mo. The second composition targets
Stabilization and disposal of Argonne-West low-level mixed wastes in ceramicrete waste forms
International Nuclear Information System (INIS)
Barber, D. B.; Singh, D.; Strain, R. V.; Tlustochowicz, M.; Wagh, A. S.
1998-01-01
The technology of room-temperature-setting phosphate ceramics or Ceramicretetrademark technology, developed at Argonne National Laboratory (ANL)-East is being used to treat and dispose of low-level mixed wastes through the Department of Energy complex. During the past year, Ceramicretetrademark technology was implemented for field application at ANL-West. Debris wastes were treated and stabilized: (a) Hg-contaminated low-level radioactive crushed light bulbs and (b) low-level radioactive Pb-lined gloves (part of the MWIR number s ign AW-W002 waste stream). In addition to hazardous metals, these wastes are contaminated with low-level fission products. Initially, bench-scale waste forms with simulated and actual waste streams were fabricated by acid-base reactions between mixtures of magnesium oxide powders and an acid phosphate solution, and the wastes. Size reduction of Pb-lined plastic glove waste was accomplished by cryofractionation. The Ceramicretetrademark process produces dense, hard ceramic waste forms. Toxicity Characteristic Leaching Procedure (TCLP) results showed excellent stabilization of both Hg and Pb in the waste forms. The principal advantage of this technology is that immobilization of contaminants is the result of both chemical stabilization and subsequent microencapsulation of the reaction products. Based on bench-scale studies, Ceramicretetrademark technology has been implemented in the fabrication of 5-gal waste forms at ANL-West. Approximately 35 kg of real waste has been treated. The TCLP is being conducted on the samples from the 5-gal waste forms. It is expected that because the waste forms pass the limits set by the EPAs Universal Treatment Standard, they will be sent to a radioactive-waste disposal facility
Leaching of nuclear power reactor wastes forms
International Nuclear Information System (INIS)
Endo, L.S.; Villalobos, J.P.; Miyamoto, H.
1986-01-01
The leaching tests for power reactor wastes carried out at IPEN/CNEN-SP are described. These waste forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. 3 years leaching results are reported, determining cesium and strontium diffusivity coefficients for boric acid waste form and ion-exchange resins. (Author) [pt
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
Naval Waste Package Design Report
International Nuclear Information System (INIS)
M.M. Lewis
2004-01-01
A design methodology for the waste packages and ancillary components, viz., the emplacement pallets and drip shields, has been developed to provide designs that satisfy the safety and operational requirements of the Yucca Mountain Project. This methodology is described in the ''Waste Package Design Methodology Report'' Mecham 2004 [DIRS 166168]. To demonstrate the practicability of this design methodology, four waste package design configurations have been selected to illustrate the application of the methodology. These four design configurations are the 21-pressurized water reactor (PWR) Absorber Plate waste package, the 44-boiling water reactor (BWR) waste package, the 5-defense high-level waste (DHLW)/United States (U.S.) Department of Energy (DOE) spent nuclear fuel (SNF) Co-disposal Short waste package, and the Naval Canistered SNF Long waste package. Also included in this demonstration is the emplacement pallet and continuous drip shield. The purpose of this report is to document how that design methodology has been applied to the waste package design configurations intended to accommodate naval canistered SNF. This demonstrates that the design methodology can be applied successfully to this waste package design configuration and support the License Application for construction of the repository
International Nuclear Information System (INIS)
Wilding, C.R.; Phillips, D.C.; Burnay, S.G.; Spindler, W.E.; Lyon, C.E.; Winter, J.A.
1991-01-01
The purpose of this programme was to determine the effects of radiation on the properties of intermediate-level waste forms relevant to their storage and disposal. It had two overall aims: to provide immediate data on the effect of radiation on important European ILW waste forms through accelerated laboratory tests; and to develop an understanding of the degradation processes so that long-term, low dose rate effects can be predicted with confidence from short-term, high dose rate experiments. The programme included cement waste forms containing inorganic wastes, organic matrix waste forms, and cement waste forms containing a substantial component of organic waste. Irradiations were carried out by external gamma sources and by the incorporation of alpha emitters, such as 238 Pu. Irradiated materials included matrix materials, simulated waste forms and real waste forms. 2 figs.; 3 tabs.; 8 refs
Thermal conductivity of multibarrier waste form components
International Nuclear Information System (INIS)
Lokken, R.O.
1981-01-01
The multiple barrier concept of radioactive waste immobilization under investigation at Pacific Northwest Laboratory (PNL) uses composite waste forms which exhibit enhanced inertness through improvements in thermal stability, mechanical strength, and leachability by the use of coatings and metal matrices. Since excessive heat may be generated by radioactive decay of the waste, the thermal conductivity of the various barriers, and more importantly of the composite, becomes an important parameter in design criteria. This report presents results of thermal conductivity measurements on 21 various glass, ceramic, metal and composite materials used in multibarrier waste forms development
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.)
International Nuclear Information System (INIS)
Moncouyoux, J.; Aure, A.; Ladirat, C.
1991-01-01
A two-year study of the degree of fracturation in full-scale high-level waste glass blocks was undertaken under contract Fl1W-0182. The project included a theoretical investigation of thermal and mechanical stresses arising in the glass and determination of the fracture-inducing stress distribution in the block, as well as an experimental study based on tomographic examinations performed at the BAM in Berlin on full-scale canisters containing glass cast at Marcoule and on water leach tests at 100 0 C to evaluate the fractured leachable surface area. The Castem code system was used to determine the temperature profiles versus time for the glass package. The mechanical study revealed the impact of removing the package from the furnace on the final stress loading. Controlled cooling of glass blocks must be extremely slow (less than 5 0 C per hour) to have a significant effect on the stress loading. Three series of tomographic experiments were carried out: two on a canister as cast, and a third after relieving the canister stresses. Static and dynamic leach tests were conducted to determine the fracturation factor of the glass blocks. Only static leaching provides satisfactory results. Fracturation factors of 10 to 15 were measured on industrial glass blocks. 50 figs.; 20 tabs.; 3 refs
International Nuclear Information System (INIS)
Kalb, P.D.; Colombo, P.
1983-01-01
The main objective of this study is to investigate new solidification agents relative to their potential application to wastes generated by advanced high volume reduction technologies, e.g., incinerator ash, dry solids, and ion exchange resins. Candidate materials selected for the solidification of these wastes include a modified sulfur cement and low-density polyethylene, neither of which are currently employed commerically for the solidification of low-level waste (LLW). As both the modified sulfur cement and the polyethylene are thermoplastic materials, a heated screw type extruder is utilized in the production of waste form samples for testing and evaluation. In this regard, work is being conducted to determine the range of conditions under which these solidification agents can be satisfactorily applied to the specific LLW streams and to provide information relevant to operating parameters and process control
Waste Isolation Pilot Plant in situ experimental program for HLW
International Nuclear Information System (INIS)
Molecke, M.A.
1977-01-01
The Waste Isolation Pilot Plant (WIPP) will be a facility to demonstrate the environmental and operational safety of storing radioactive wastes in a deep geologic bedded salt facility. The WIPP will be located in southeastern New Mexico, approximately 30 miles east of the city of Carlsbad. The major focus of the pilot plant operation involves ERDA defense related low and intermediate-level transuranic wastes. The scope of the project also specifically includes experimentation utilizing commercially generated high-level wastes, or alternatively, spent unreprocessed fuel elements. WIPP HLW experiments are being conducted in an inter-related laboratory, bench-scale, and in situ mode. This presentation focuses on the planned in situ experiments which, depending on the availability of commercially reprocessed waste plus delays in the construction schedule of the WIPP, will begin in approximately 1985. Such experiments are necessary to validate preceding laboratory results and to provide actual, total conditions of geologic storage which cannot be adequately simulated. One set of planned experiments involves emplacing bare HLW fragments into direct contact with the bedded salt environment. A second set utilizes full-size canisters of waste emplaced in the salt in the same manner as planned for a future HLW repository. The bare waste experiments will study in an accelerated manner waste-salt bed-brine interactions including matrix integrity/degradation, brine leaching, system chemistry, and potential radionuclide migration through the salt bed. Utilization of full-size canisters of HLW in situ permits us to demonstrate operational effectiveness and safety. Experiments will evaluate corrosion and compatibility interactions between the waste matrix, canister and overpack materials, getter materials, stored energy, waste buoyancy, etc. Using full size canisters also allows us to demonstrate engineered retrievability of wastes, if necessary, at the end of experimentation
Lead corrosion evaluation in high activity nuclear waste container (Argentina)
International Nuclear Information System (INIS)
Guasp, R.; Lanzani, L.; Bruzzoni, P.; Cufre, W.; Semino, C.J.
2000-01-01
This report describes a study of high activity nuclear waste canister corrosion in a deep geological disposal. In this canister design, the vitrified nuclear waste stainless steel container is shielded by a 100 mm thick lead wall. For mechanical resistance, the canister will also have a thin carbon steel external liner. Experimental and mathematical modeling studies are aimed to asses the corrosion kinetics of the carbon steel liner in first instance and then, once this liner has been corroded away, the corrosion kinetics of the main lead barrier. Being that oxygen reduction is the main cathodic reaction that supports the anodic oxidation of iron, a model is described predicting the rate of oxygen consumption in a sealed deep nuclear waste disposal vault as a result of the canister corrosion. Oxidation processes other than container corrosion, and that can account also for oxygen depletion, are not taken into consideration. Corrosion experimental studies on lead and its alloys in groundwater are also reported. These experiments are aimed to improve the corrosion resistance of commercial lead in groundwater. (author)
Special waste form lysimeters-arid. Annual report, 1985
International Nuclear Information System (INIS)
Walter, M.B.; Graham, M.J.
1985-09-01
The Special Waste Form Lysimeters-Arid program was initiated to determine typical source terms generated by commercial solidified low-level nuclear waste in an arid climate. Waste-form leaching tests are being conducted at a field facility at the Hanford site near Richland, Washington. A similar program is being conducted at a humid site. The field facility consists of 10 lysimeters placed around a central instrument caisson. The waste samples from boiling water and pressurized water reactors were emplaced in 1984, and the lysimeters are being monitored for movement of contaminants and water. Solidifying agents being tested include vinyl ester-styrene, bitumen, and cement. Laboratory leaching and geochemical modeling studies are being conducted to predict expected leach rates at the field site and to aid field-data interpretation. Small samples of the solidified waste forms were made for use in the laboratory leaching studies that include standard leach tests and leaching of solidified waste forms in soil columns. Complete chemical and radionuclide analyses are being conducted on the solid and liquid portions of the wastes. 2 refs
Special Waste Form Lysimeters-Arid: annual report 1985
International Nuclear Information System (INIS)
Walter, M.B.; Graham, M.J.
1986-01-01
The Special Waste Form Lysimeters-Arid program was initiated to determine typical source terms generated by commercial solidified low-level nuclear waste in an arid climate. Waste-form leaching tests are being conducted at a field facility at the Hanford site near Richland, Washington. A similar program is being conducted at a humid site. The field facility consists of 10 lysimeters placed around a central instrument caisson. The waste samples from boiling water and pressurized water reactors were emplaced in 1984, and the lysimeters are being monitored for movement of contaminants and water. Solidifying agents being tested include vinyl ester-styrene, bitumen, and cement. Laboratory leaching and geochemical modeling studies are being conducted to predict expected leach rates at the field site and to aid field-data interpretation. Small samples of the solidified waste forms were made for use in the laboratory leaching studies that include standard leach tests and leaching of solidified waste forms in soil columns. Complete chemical and radionuclide analyses are being conducted on the solid and liquid portions of the wastes
Interim transfer canister for consolidating nuclear fuel rods
International Nuclear Information System (INIS)
Formanek, F.J.
1987-01-01
This patent describes a canister for receiving and consolidating a group of uniformly spaced apart nuclear fuel rods, comprising: a rectangular, vertically oriented straight back panel; a pair of oppositely disposed side panels connected perpendicularly to the back panel, having a vertical straight upper portion and an inwardly tapered lower portion; a front panel opposite the back panel and connected to the side panels, having a straight vertical upper portion and inwardly tapered lower portion; whereby the back, side and front panels define a rectangular upper opening at the upper end of the canister and a generally rectangular lower opening at the other end, the lower opening having a cross-sectional area less than one-half that of the upper opening; parallel plate members spanning the canister from the front panel to the back panel, each plate spaced from the other the same uniform distance, the plates extending downwardly into the tapered portion of the canister while remaining spaced above the tapered sidewalls; first base means at the lower end of the canister, removably mounted and having an oblique orientation generally downward from the front panel to the back panel, for guiding the fuel rods to be inserted preferentially toward the lower portion of the back panel; and second base means removably mounted within the canister below first base means and oriented transversely to the longitudinal extent of the canister, for supporting the fuel rods when the first base means is removed from the canister
Evaluation and review of alternative waste forms for immobilization of high level radioactive wastes
International Nuclear Information System (INIS)
1979-01-01
Objective was to review the relative merits and potential of eleven alternative waste forms being considered for the solidification and disposal of radioactive wastes. A numerical rating of the alternative waste forms was arrived at individually by peer review panel members taking into consideration nine scientific and nine engineering parameters affecting the long-term performance and production of waste forms. A group rating for the alternative forms was achieved by averaging the individiual scores and discussing the available data base. Three final ranking lists comparing: (A) Present Scientific Merits or Least Risk for Use Today; (B) Research Priority; and (3) Present and Potential Engineering Practicality were prepared by the Panel. Each waste form in the lists is assigned a value of either (1) Top Rank, (2) Intermediate Rank, or is assigned a value of either (1) Top Rank, (2) Intermediate Rank, or (3) Bottom Rank. Relative strengths and weaknesses of the alternative waste forms and recommendations for future program directions are discussed
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.
Ceramic waste form qualification using results from witness tubes
International Nuclear Information System (INIS)
O'Holleran, T.P.; Johnson, S.G.; Bateman, K.J.
2002-01-01
A ceramic waste form has been developed to immobilize the salt waste stream from electrometallurgical treatment of spent nuclear fuel. The ceramic waste form is prepared in a hot isostatic press (HIP). The use of small, easily fabricated HIP capsules called witness tubes has been proposed as a practical way to obtain representative samples of ceramic waste form material for process monitoring, waste form qualification, and archiving. Witness tubes are filled with the same material used to fill the corresponding HIP can, and are HIPed along with the HIP can. Relevant physical, chemical, and performance (leach test) data are analyzed and compared. Differences between witness tube and HIP can materials are shown to be statistically insignificant, demonstrating that witness tubes do provide ceramic waste form material representative of the material in the corresponding HIP can.
DURABILITY TESTING OF FLUIDIZED BED STEAM REFORMER (FBSR) WASTE FORMS
International Nuclear Information System (INIS)
Jantzen, C
2006-01-01
Fluidized Bed Steam Reforming (FBSR) is being considered as a potential technology for the immobilization of a wide variety of high sodium aqueous radioactive wastes. The addition of clay and a catalyst as co-reactants converts high sodium aqueous low activity wastes (LAW) such as those existing at the Hanford and Idaho DOE sites to a granular ''mineralized'' waste form that may be made into a monolith form if necessary. Simulant Hanford and Idaho high sodium wastes were processed in a pilot scale FBSR at Science Applications International Corporation (SAIC) Science and Technology Applications Research (STAR) facility in Idaho Falls, ID. Granular mineral waste forms were made from (1) a basic Hanford Envelope A low-activity waste (LAW) simulant and (2) an acidic INL simulant commonly referred to as sodium-bearing waste (SBW). The FBSR waste forms were characterized and the durability tested via ASTM C1285 (Product Consistency Test), the Environmental Protection Agency (EPA) Toxic Characteristic Leaching Procedure (TCLP), and the Single Pass Flow Through (SPFT) test. The durability of the FBSR waste form products was tested in order to compare the measured durability to previous FBSR waste form testing on Hanford Envelope C waste forms that were made by THOR Treatment Technologies (TTT) and to compare the FBSR durability to vitreous LAW waste forms, specifically the Hanford low activity waste (LAW) glass known as the Low-activity Reference Material (LRM). The durability of the FBSR waste form is comparable to that of the LRM glass for the test responses studied
Initial specifications for nuclear waste package external dimensions and materials
International Nuclear Information System (INIS)
Gregg, D.W.; O'Neal, W.C.
1983-09-01
Initial specifications of external dimensions and materials for waste package conceptual designs are given for Defense High Level Waste (DHLW), Commercial High Level Waste (CHLW) and Spent Fuel (SF). The designs have been developed for use in a high-level waste repository sited in a tuff media in the unsaturated zone. Drawings for reference and alternative package conceptual designs are presented for each waste form for both vertical and horizontal emplacement configurations. Four metal alloys: 304L SS, 321 SS, 316L SS and Incoloy 825 are considered for the canister or overpack; 1020 carbon steel was selected for horizontal borehole liners, and a preliminary packing material selection is either compressed tuff or compressed tuff containing iron bearing smectite clay as a binder
Nuclear waste repository in basalt: preconceptual design guidelines
International Nuclear Information System (INIS)
1979-06-01
The development of the basalt waste isolation program parallels the growing need for permanent, environmentally safe, and secure means to store nuclear wastes. The repository will be located within the Columbia Plateau basalt formations where these ends can be met and radiological waste can be stored. These wastes will be stored such that the wastes may be retrieved from storage for a period after placement. After the retrieval period, the storage locations will be prepared for terminal storage. The terminal storage requirements will include decommissioning provisions. The facility boundaries will encompass no more than several square miles of land which will be above a subsurface area where the geologic makeup is primarily deep basaltic rock. The repository will receive, from an encapsulation site(s), nuclear waste in the form of canisters (not more than 18.5 feet x 16 inches in diameter) and containers (55-gallon drums). Canisters will contain spent fuel (after an interim 5-year storage period), solidified high-level wastes (HLW), or intermediate-level wastes (ILW). The containers (drums) will package the low-level transuranic wastes (LL-TRU). The storage capacity of the repository will be expanded in a time-phased program which will require that subsurface development (repository expansion) be conducted concurrently with waste storage operations. The repository will be designed to store the nuclear waste generated within the predictable future and to allow for reasonable expansion. The development and assurance of safe waste isolation is of paramount importance. All activities will be dedicated to the protection of public health and the environment. The repository will be licensed by the US Nuclear Regulatory Commission (NRC). Extensive efforts will be made to assure selection of a suitable site which will provide adequate isolation
Nuclear waste repository in basalt: preconceptual design guidelines
Energy Technology Data Exchange (ETDEWEB)
1979-06-01
The development of the basalt waste isolation program parallels the growing need for permanent, environmentally safe, and secure means to store nuclear wastes. The repository will be located within the Columbia Plateau basalt formations where these ends can be met and radiological waste can be stored. These wastes will be stored such that the wastes may be retrieved from storage for a period after placement. After the retrieval period, the storage locations will be prepared for terminal storage. The terminal storage requirements will include decommissioning provisions. The facility boundaries will encompass no more than several square miles of land which will be above a subsurface area where the geologic makeup is primarily deep basaltic rock. The repository will receive, from an encapsulation site(s), nuclear waste in the form of canisters (not more than 18.5 feet x 16 inches in diameter) and containers (55-gallon drums). Canisters will contain spent fuel (after an interim 5-year storage period), solidified high-level wastes (HLW), or intermediate-level wastes (ILW). The containers (drums) will package the low-level transuranic wastes (LL-TRU). The storage capacity of the repository will be expanded in a time-phased program which will require that subsurface development (repository expansion) be conducted concurrently with waste storage operations. The repository will be designed to store the nuclear waste generated within the predictable future and to allow for reasonable expansion. The development and assurance of safe waste isolation is of paramount importance. All activities will be dedicated to the protection of public health and the environment. The repository will be licensed by the US Nuclear Regulatory Commission (NRC). Extensive efforts will be made to assure selection of a suitable site which will provide adequate isolation.
Plan for glass waste form testing for NNWSI [Nevada Nuclear Waste Storage Investigations
International Nuclear Information System (INIS)
Aines, R.D.
1987-09-01
The purpose of glass waste form testing is to determine the rate of release of radionuclides from breached glass waste containers. This information will be used to qualify glass waste forms with respect to the release requirements. It will be the basis of the source term from glass waste for repository performance assessment modeling. This information will also serve as part of the source term in the calculation of cumulative releases after 100,000 years in the site evaluation process. It will also serve as part of the source term input for calculation of cumulative releases to the accessible environment for 10,000 years after disposal, to determine compliance with EPA regulations. This investigation will provide data to resolve information needs. Information about the waste forms which is provided by the producer will be accumulated and evaluated; the waste form will be tested, properties determined, and mechanisms of degradation determined; and models providing long-term evaluation of release rates designed and tested. 23 refs
The characterization of cement waste form for final disposal of decommissioned concrete waste
International Nuclear Information System (INIS)
Lee, K.W.; Lee, Y.J.; Hwang, D.S.; Moon, J.K.
2015-01-01
Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. In addition, 83 drums of 200 l, and 41 containers of 4 m 3 of concrete waste were generated. Conditioning of concrete waste is needed for final disposal. Concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled into a void space after concrete rubble pre-placement into 200 l drums. Thus, this research developed an optimizing mixing ratio of concrete waste, water, and cement, and evaluated the characteristics of a cement waste form to meet the requirements specified in the disposal site specific waste acceptance criteria. The results obtained from compressive strength test, leaching test, and thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10 as the optimized mixing ratio. In addition, the compressive strength of cement waste form was satisfied, including fine powder up to a maximum 40 wt% in concrete debris waste of about 75%. (authors)
Comments on 'SKB RD and D-Programme 98'. Focused on canister integrity and corrosion
International Nuclear Information System (INIS)
Bowyer, W.H.; Hermansson, H.P.
1999-04-01
According to the Act on Nuclear Activities the nuclear utilities are requested to submit a comprehensive programme for research and development every third year, aiming at the safe storage of radioactive waste produced by the nuclear power plants. The latest was published by SKB in September 1998 and is called the 'RDandD Programme 98'. The work presented in the present report was commissioned by SKI and is a result of reading the 'RDandD Programme 98' and related reports with focus on canister production, integrity and corrosion. We find that those parts of the programme often are difficult to follow owing to the lack of detail in the Programme and in one of the supporting reports. In our opinion this will make the work difficult to monitor by SKI and SKB. We also feel that the interpretation of information already available often is overoptimistic. As a consequence the difficulties ahead are understated and the programme is allowed to converge too quickly. We agree that the materials choices for both the inner and outer canisters are appropriate providing they both can be produced commercially and in a satisfactory metallurgical condition, that they can be quality assured and that no further unforeseen difficulties arise. We also agree that alternative technologies merit consideration for production of the outer canister and that alternative joining processes should be studied. We are actually concerned that greater prominence is not given to the alternatives in the programme. We believe that it should be possible to develop a satisfactory canister for disposal of high level nuclear waste according to the general method proposed by SKB and with the proposed capacity within the time-scale of the overall programme. We do not believe, however, that all the difficulties have been recognised. As a consequence of this the results to date are interpreted optimistically. We believe that progress should be subjected to more professional review within SKB and that a higher
Sampling and analysis strategies to support waste form qualification
International Nuclear Information System (INIS)
Westsik, J.H. Jr.; Pulsipher, B.A.; Eggett, D.L.; Kuhn, W.L.
1989-04-01
As part of the waste acceptance process, waste form producers will be required to (1) demonstrate that their glass waste form will meet minimum specifications, (2) show that the process can be controlled to consistently produce an acceptable waste form, and (3) provide documentation that the waste form produced meets specifications. Key to the success of these endeavors is adequate sampling and chemical and radiochemical analyses of the waste streams from the waste tanks through the process to the final glass product. This paper suggests sampling and analysis strategies for meeting specific statistical objectives of (1) detection of compositions outside specification limits, (2) prediction of final glass product composition, and (3) estimation of composition in process vessels for both reporting and guiding succeeding process steps. 2 refs., 1 fig., 3 tabs
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
The offshore disposal of radioactive waste by drilled emplacement: A feasibility study
International Nuclear Information System (INIS)
Bury, M.R.C.
1985-01-01
This book is a report, based on a study by Taylor Woodrow Construction Limited, on the overall feasibility of the disposal of high-level radioactive waste in boreholes drilled deep into the ocean bed. The work comprises an engineering appraisal of the disposal process with a view to establishing technical and operational feasibility and providing overall cost information to enable an economic assessment to be made. Contents: Summary report; Reference criteria; Drilling operation; Transfer of radioactive waste, personnel and other supplies; Handling of radioactive waste on board; Lowering strings of canisters; Emplacement and backfilling of canisters; Preliminary design of marine platform; Retrieval of flasks or canisters lost or misplaced; Variations to the features of the lowering system; Logistics of the operation; Construction cost estimate; Operational costs; Appendix
International Nuclear Information System (INIS)
Xu, Hui; Gong, Weiliang; Syltebo, Larry; Lutze, Werner; Pegg, Ian L.
2014-01-01
Highlights: • Quantitative correlations firstly established for cementitious waste forms. • Quantitative correlations firstly established for geopolymeric materials. • Ternary DuraLith geopolymer waste forms for Hanford radioactive wastes. • Extended setting times which improve workability for geopolymer waste forms. • Reduced hydration heat release from DuraLith geopolymer waste forms. - Abstract: The binary furnace slag-metakaolin DuraLith geopolymer waste form, which has been considered as one of the candidate waste forms for immobilization of certain Hanford secondary wastes (HSW) from the vitrification of nuclear wastes at the Hanford Site, Washington, was extended to a ternary fly ash-furnace slag-metakaolin system to improve workability, reduce hydration heat, and evaluate high HSW waste loading. A concentrated HSW simulant, consisting of more than 20 chemicals with a sodium concentration of 5 mol/L, was employed to prepare the alkaline activating solution. Fly ash was incorporated at up to 60 wt% into the binder materials, whereas metakaolin was kept constant at 26 wt%. The fresh waste form pastes were subjected to isothermal calorimetry and setting time measurement, and the cured samples were further characterized by compressive strength and TCLP leach tests. This study has firstly established quantitative linear relationships between both initial and final setting times and hydration heat, which were never discovered in scientific literature for any cementitious waste form or geopolymeric material. The successful establishment of the correlations between setting times and hydration heat may make it possible to efficiently design and optimize cementitious waste forms and industrial wastes based geopolymers using limited testing results
Cermet high level waste forms: a pregress report
International Nuclear Information System (INIS)
Aaron, W.S.; Quinby, T.C.; Kobisk, E.H.
1978-06-01
The fixation of high level radioactive waste from both commercial and DOE defense sources as cermets is currently under study. This waste form consists of a continuous iron-nickel base metal matrix containing small particles of fission product oxides. Preliminary evaluations of cermets fabricated from a variety of simulated wastes indicate they possess properties providing advantages over other waste forms presently being considered, namely thermal conductivity, waste loading levels, and leach resistance. This report describes the progress of this effort, to date, since its initiation in 1977
Alternate nuclear waste forms and interactions in geologic media
International Nuclear Information System (INIS)
Boatner, L.A.; Battle, G.C. Jr.
1981-04-01
The primary purposes of the conference on Alternate Nuclear Waste Forms and Interactions in Geologic Media were: First, to provide an opportunity for a review of the status of the research on some of the candidate alternative waste forms; second, to provide an opportunity for comparing the characteristics of alternate waste forms to those of glasses; and third, to stimulate increased interactions between those research groups that were engaged in a more basic approach to characterizing waste forms and those who were concerned with more applied aspects such as the processing of these materials. The motivating philosophy behind this third purpose of the conference was based on the idea that by operating from the soundest possible fundamental base for any of the candidate waste forms, hopefully any future unpleasant surprise - such as that alluded to earlier in the case of glass waste forms - could be avoided. Separate abstracts have been prepared for individual papers for inclusion in the Energy Data Base
Final report on cermet high-level waste forms
International Nuclear Information System (INIS)
Kobisk, E.H.; Quinby, T.C.; Aaron, W.S.
1981-08-01
Cermets are being developed as an alternate method for the fixation of defense and commercial high level radioactive waste in a terminal disposal form. Following initial feasibility assessments of this waste form, consisting of ceramic particles dispersed in an iron-nickel base alloy, significantly improved processing methods were developed. The characterization of cermets has continued through property determinations on samples prepared by various methods from a variety of simulated and actual high-level wastes. This report describes the status of development of the cermet waste form as it has evolved since 1977. 6 tables, 18 figures
High-level waste processing and disposal
International Nuclear Information System (INIS)
Crandall, J.L.; Krause, H.; Sombret, C.; Uematsu, K.
1984-11-01
Without reprocessing, spent LWR fuel itself is generally considered an acceptable waste form. With reprocessing, borosilicate glass canisters, have now gained general acceptance for waste immobilization. The current first choice for disposal is emplacement in an engineered structure in a mined cavern at a depth of 500-1000 meters. A variety of rock types are being investigated including basalt, clay, granite, salt, shale, and volcanic tuff. This paper gives specific coverage to the national high level waste disposal plans for France, the Federal Republic of Germany, Japan and the United States. The French nuclear program assumes prompt reprocessing of its spent fuels, and France has already constructed the AVM. Two larger borosilicate glass plants are planned for a new French reprocessing plant at La Hague. France plans to hold the glass canisters in near-surface storage for a forty to sixty year cooling period and then to place them into a mined repository. The FRG and Japan also plan reprocessing for their LWR fuels. Both are currently having some fuel reprocessed by France, but both are also planning reprocessing plants which will include waste vitrification facilities. West Germany is now constructing the PAMELA Plant at Mol, Belgium to vitrify high level reprocessing wastes at the shutdown Eurochemic Plant. Japan is now operating a vitrification mockup test facility and plans a pilot plant facility at the Tokai reprocessing plant by 1990. Both countries have active geologic repository programs. The United State program assumes little LWR fuel reprocessing and is thus primarily aimed at direct disposal of spent fuel into mined repositories. However, the US have two borosilicate glass plants under construction to vitrify existing reprocessing wastes
Treatability study of absorbent polymer waste form for mixed waste treatment
International Nuclear Information System (INIS)
Herrmann, S. D.; Lehto, M. A.; Stewart, N. A.; Croft, A. D.; Kern, P. W.
2000-01-01
A treatability study was performed to develop and characterize an absorbent polymer waste form for application to low level (LLW) and mixed low level (MLLW) aqueous wastes at Argonne National Laboratory-West (ANL-W). In this study absorbent polymers proved effective at immobilizing aqueous liquid wastes in order to meet Land Disposal Restrictions for subsurface waste disposal. Treatment of aqueous waste with absorbent polymers provides an alternative to liquid waste solidification via high-shear mixing with clays and cements. Significant advantages of absorbent polymer use over clays and cements include ease of operations and waste volume minimization. Absorbent polymers do not require high-shear mixing as do clays and cements. Granulated absorbent polymer is poured into aqueous solutions and forms a gel which passes the paint filter test as a non-liquid. Pouring versus mixing of a solidification agent not only eliminates the need for a mixing station, but also lessens exposure to personnel and the potential for spread of contamination from treatment of radioactive wastes. Waste minimization is achieved as significantly less mass addition and volume increase is required of and results from absorbent polymer use than that of clays and cements. Operational ease and waste minimization translate into overall cost savings for LLW and MLLW treatment
International Nuclear Information System (INIS)
Anon.
1985-01-01
Final disposal of radioactive waste involves preventing the waste from returning from the repository location into the biosphere by means of successively arranged containment measures known as safety barriers. In the present volume NGB 85-04 of the series of reports for Project 'Guarantee' 1985, the safety barrier system for the type C repository for high-level waste is described. The barrier parameters which are relevant for safety analysis are quantified and associated error limits and data scatter are given. The aim of the report is to give a summary documentation of the safety analysis input data and their scientific background. For secure containment of radioactive waste safety barriers are used which effectively limit the release of radioactive material from the repository (release barriers) and effectively retard the entry of the original radioactive material into the biosphere (time barriers). Safety barriers take the form of both technically constructed containment measures and the siting of the repository in suitable geological formations. The technical safety barrier system in the case of high-level waste comprises: the waste solidification matrix (borosilicate glass), massive steel canisters, encasement of the waste canisters, encasement of the waste canisters in highly compacted bentonite, sealing of vacant storage space and access routes on repository closure. The natural geological safety barriers - the host rock and overlying formations provide sufficiently long deep groundwater flow times from the repository location to the earth's surface and for additional lengthening of radionuclide migration times by means of various chemical and physical retardation mechanisms. The stability of the geological formations is so great that hydrogeological system is protected for a sufficient length of time from deterioration caused, in particular, by erosion. Observations in the final section of the report indicate that input data for the type C repository safety
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.
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.)
Processing vessel for high level radioactive wastes
International Nuclear Information System (INIS)
Maekawa, Hiromichi
1998-01-01
Upon transferring an overpack having canisters containing high level radioactive wastes sealed therein and burying it into an underground processing hole, an outer shell vessel comprising a steel plate to be fit and contained in the processing hole is formed. A bury-back layer made of dug earth and sand which had been discharged upon forming the processing hole is formed on the inner circumferential wall of the outer shell vessel. A buffer layer having a predetermined thickness is formed on the inner side of the bury-back layer, and the overpack is contained in the hollow portion surrounded by the layer. The opened upper portion of the hollow portion is covered with the buffer layer and the bury-back layer. Since the processing vessel having a shielding performance previously formed on the ground, the state of packing can be observed. In addition, since an operator can directly operates upon transportation and burying of the high level radioactive wastes, remote control is no more necessary. (T.M.)
Influence of system considerations on waste form design
International Nuclear Information System (INIS)
Bauer, A.A.; Matthews, S.C.; Peterson, R.W.
1979-01-01
The design of waste forms is constrained by waste management system considerations imposed during generation, treatment, packaging, transportation, storage, and isolation. In the isolation phase, the waste form provides one of the barriers to release in a multibarrier system that includes the natural geologic and hydrologic barriers as well as other engineered barriers
Proposed research and development plan for mixed low-level waste forms
Energy Technology Data Exchange (ETDEWEB)
O`Holleran, T.O.; Feng, X.; Kalb, P. [and others
1996-12-01
The objective of this report is to recommend a waste form program plan that addresses waste form issues for mixed low-level waste (MLLW). The report compares the suitability of proposed waste forms for immobilizing MLLW in preparation for permanent near-surface disposal and relates them to their impact on the U.S. Department of Energy`s mixed waste mission. Waste forms are classified into four categories: high-temperature waste forms, hydraulic cements, encapsulants, and specialty waste forms. Waste forms are evaluated concerning their ability to immobilize MLLW under certain test conditions established by regulatory agencies and research institutions. The tests focused mainly on leach rate and compressive strength. Results indicate that all of the waste forms considered can be tailored to give satisfactory performance immobilizing large fractions of the Department`s MLLW inventory. Final waste form selection will ultimately be determined by the interaction of other, often nontechnical factors, such as economics and politics. As a result of this report, three top-level programmatic needs have been identified: (1) a basic set of requirements for waste package performance and disposal; (2) standardized tests for determining waste form performance and suitability for disposal; and (3) engineering experience operating production-scale treatment and disposal systems for MLLW.
Proposed research and development plan for mixed low-level waste forms
International Nuclear Information System (INIS)
O'Holleran, T.O.; Feng, X.; Kalb, P.
1996-12-01
The objective of this report is to recommend a waste form program plan that addresses waste form issues for mixed low-level waste (MLLW). The report compares the suitability of proposed waste forms for immobilizing MLLW in preparation for permanent near-surface disposal and relates them to their impact on the U.S. Department of Energy's mixed waste mission. Waste forms are classified into four categories: high-temperature waste forms, hydraulic cements, encapsulants, and specialty waste forms. Waste forms are evaluated concerning their ability to immobilize MLLW under certain test conditions established by regulatory agencies and research institutions. The tests focused mainly on leach rate and compressive strength. Results indicate that all of the waste forms considered can be tailored to give satisfactory performance immobilizing large fractions of the Department's MLLW inventory. Final waste form selection will ultimately be determined by the interaction of other, often nontechnical factors, such as economics and politics. As a result of this report, three top-level programmatic needs have been identified: (1) a basic set of requirements for waste package performance and disposal; (2) standardized tests for determining waste form performance and suitability for disposal; and (3) engineering experience operating production-scale treatment and disposal systems for MLLW
Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor
International Nuclear Information System (INIS)
Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo
2016-01-01
The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated
Structural Integrity Evaluation for Damaged Fuel Canister of a Research Reactor
Energy Technology Data Exchange (ETDEWEB)
Oh, Jinho; Kwak, Jinsung; Lee, Sangjin; Lee, Jongmin; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2016-10-15
The purpose of this document is to confirm the structural integrity of damaged fuel canister through the numerical simulation. The analysis results of canister including damaged fuel are evaluated with design limits of the ASME Sec. III NF Codes and Standards. The main function of canister is to store and protect the damaged fuel assembly generated from the operation of the research reactor. The canister is classified into safety class NNS (Non-nuclear Safety) and seismic category II. The shape of the canister is designed into commercialized circular tube due to economic benefit and easy manufacturing. The damaged fuel assembly is loaded in a dedicated canister by using special tool and supported by lower block in the canister. Then it is move into the damaged fuel storage rack under safeguards arrangements. The canister is securely supported at guide plate and base plate of rack. The structural integrity evaluation for the canister is performed by using response spectrum analysis. The analysis results show that the stress intensity of the canister under the seismic loads is within the ASME Code limits. Thus, the validity of the present design of the canister has been demonstrated.
International Nuclear Information System (INIS)
Kenneth J. Bateman; Charles W. Solbrig
2006-01-01
The waste produced from processing spent fuel from the EBR II reactor must be processed into a waste form suitable for long term storage in Yucca Mountain. The method chosen produces zeolite granules mixed with glass frit, which must then be converted into a solid. This is accomplished by loading it into a can and heating to 900 C in a furnace regulated at 915 C. During heatup to 900 C, the zeolite and glass frit react and consolidate to produce a sodalite monolith. The resultant ceramic waste form (CWF) is then cooled. The waste is 52 cm in diameter and initially 300 cm long but consolidates to 150 cm long during the heating process. After cooling it is then inserted in a 5-DHLW/DOE SNF Long Canister. Without intervention, the waste takes 82 hours to heat up to 900 C in a furnace designed to geometrically fit the cylindrical waste form. This paper investigates the reduction in heating times possible with four different methods of additional heating through a center hole. The hole size is kept small to maximize the amount of CWF that is processed in a single run. A hole radius of 1.82 cm was selected which removes only 1% of the CWF. A reference computation was done with a specified inner hole surface temperature of 915 C to provide a benchmark for the amount of improvement which can be made. It showed that the heatup time can potentially be reduced to 43 hours with center hole heating. The first method, simply pouring high temperature liquid aluminum into the hole, did not produce any noticeable effect on reducing heat up times. The second method, flowing liquid aluminum through the hole, works well as long as the velocity is high enough (2.5 cm/sec) to prevent solidification of the aluminum during the initial front movement of the aluminum into the center hole. The velocity can be reduced to 1 cm/sec after the initial front has traversed the ceramic. This procedure reduces the formation time to near that of the reference case. The third method, flowing a gas
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.)
Leaching of nuclear power reactor waste forms
International Nuclear Information System (INIS)
Endo, L.S.; Villalobos, J.P.; Miyamoto, H.
1987-01-01
The leaching tests for immobilized power reactor wastes carried out at IPEN are described. These wastes forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. Three years leaching results are reported. The cesium diffuvity coefficients determined out of these results are about 1 x 10 -8 cm 2 /s for boric acid waste form and 9 x 10 -9 cm 2 /s for ion-exchange resin waste. Strontium diffusivity coefficients found are about 3 x 10 -11 cm 2 /s and 9 x 10 -11 cm 2 /s respectively. (Author) [pt
Shaft shock absorber tests for a spent fuel canister
International Nuclear Information System (INIS)
Kukkola, T.; Toermaelae, V.P.
2005-06-01
The disposal canister for spent nuclear fuel will be transferred by a lift to the repository, which is 500 m deep in the bedrock. Model tests were carried out with the objective to estimate weather feasible shock absorber can be developed against the design accident case where the canister should survive a free fall to the lift shaft. If the velocity of the canister is not controlled by air drag or by any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity in impact on water when the bottom pit of the lift well is filled with groundwater. However, the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20 m high filling to the bottom pit of the lift well by Light Expanded Clay Aggregate (LECA), gives fair impact absorption to protect the fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)
Shaft shock absorber tests for a spent fuel canister
International Nuclear Information System (INIS)
Kukkola, T.; Toermaelae, V.P.
2003-01-01
The holding canister for spent nuclear fuel will be transferred by a lift to the final disposal tunnels 500m deep in the bedrock. Model tests were carried out with an objective to estimate weather feasible shock absorbing properties can be met in a design accident case where the canister should survive a free fall due to e.g. sabotage. If the velocity of the canister is not controlled by air drag or any other deceleration means, the impact velocity may reach ultimate speed of 100m/s. The canister would retain its integrity when stricken by the surface penetration impact if the bottom pit of the lift well would be filled with groundwater. However the canister would hit the pit bottom with high velocity since the water hardly slows down the canister. The impact to the bottom of the pit should be dampened mechanically. The tests demonstrated that 20m high filling to the bottom pit of the lift well by ceramic gravel, trade mark LECA-sora, gives a fair impact absorption to protect the spent fuel canister. Presence of ground water is not harmful for impact absorption system provided that the ceramic gravel is not floating too high from the pit bottom. Almost ideal impact absorption conditions are met if the water high level does not exceed two thirds of the height of the gravel. Shaping of the bottom head of the cylindrical canister does not give meaningful advantages to the impact absorption system. The flat nose bottom head of the fuel canister gives adequate deceleration properties. (orig.)
Coated particle waste form development
International Nuclear Information System (INIS)
Oma, K.H.; Buckwalter, C.Q.; Chick, L.A.
1981-12-01
Coated particle waste forms have been developed as part of the multibarrier concept at Pacific Northwest Laboratory under the Alternative Waste Forms Program for the Department of Energy. Primary efforts were to coat simulated nuclear waste glass marbles and ceramic pellets with low-temperature pyrolytic carbon (LT-PyC) coatings via the process of chemical vapor deposition (CVD). Fluidized bed (FB) coaters, screw agitated coaters (SAC), and rotating tube coaters were used. Coating temperatures were reduced by using catalysts and plasma activation. In general, the LT-PyC coatings did not provide the expected high leach resistance as previously measured for carbon alone. The coatings were friable and often spalled off the substrate. A totally different concept, thermal spray coating, was investigated at PNL as an alternative to CVD coating. Flame spray, wire gun, and plasma gun systems were evaluated using glass, ceramic, and metallic coating materials. Metal plasma spray coatings (Al, Sn, Zn, Pb) provided a two to three orders-of-magnitude increase in chemical durability. Because the aluminum coatings were porous, the superior leach resistance must be due to either a chemical interaction or to a pH buffer effect. Because they are complex, coated waste form processes rank low in process feasibility. Of all the possible coated particle processes, plasma sprayed marbles have the best rating. Carbon coating of pellets by CVD ranked ninth when compared with ten other processes. The plasma-spray-coated marble process ranked sixth out of eleven processes
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
Special waste-form lysimeters-arid: Three-year monitoring report
International Nuclear Information System (INIS)
Jones, T.L.; Serne, R.J.; Toste, A.P.
1988-04-01
Regulations governing the disposal of commercial low-level waste require all liquid waste to be solidified before burial. Most waste must be solidified into a rigid matrix such as cement or plastic to prevent waste consolidation and site slumping after burial. These solidification processes affect the rate at which radionuclides and other solutes are released into the soil. In 1983, a program was initiated at Pacific Northwest Laboratory to study the release of waste from samples of low-level radioactive waste that had been commercially solidified. The primary method used by this program is to bury sample waste forms in field lysimeters and monitor leachate composition from the release and transport of solutes. The lysimeter facility consists of 10 lysimeters, each containing one sample of solidified waste. Five different waste forms are being tested, allowing duplicate samples of each one to be evaluated. The samples were obtained from operating nuclear power plants and are actual waste forms routinely generated at these facilities. All solidification was accomplished by commercial processes. Sample size is a partially filled 210-L drum. All containers were removed prior to burial leaving the bare waste form in contact with the lysimeter soil. 11 refs., 14 figs., 16 tabs
Preliminary waste form characteristics report Version 1.0. Revision 1
International Nuclear Information System (INIS)
Stout, R.B.; Leider, H.R.
1991-01-01
This report focuses on radioactive waste form characteristics that will be used to design a waste package and an engineered barrier system (EBS) for a suitable repository as part of the Yucca Mountain Project. The term waste form refers to irradiated reactor fuel, other high-level waste (HLW) in various physical forms, and other radioactive materials (other than HLW) which are received for emplacement in a geologic repository. Any encapsulating of stabilizing matrix is also referred to as a waste form
International Nuclear Information System (INIS)
1978-04-01
This volume, Y/OWI/TM-36/22, ''Nuclear Considerations for Repository Design,'' is one of a 23-volume series, ''Technical Support for GEIS: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-36, which supplements the ''Contribution to Draft Generic Environmental Impact Statement on Commercial Waste Management: Radioactive Waste Isolation in Geologic Formations,'' Y/OWI/TM-44. The series provides a more complete technical basis for the preconceptual designs, resource requirements, and environmental source terms associated with isolating commercial LWR wastes in underground repositories in salt, granite, shale and basalt. Wastes are considered from three fuel cycles: uranium and plutonium recycling, no recycling of spent fuel and uranium-only recycling. Included in this volume are baseline design considerations such as characteristics of canisters, drums, casks, overpacks, and shipping containers; maximum allowable and actual decay-heat levels; and canister radiation levels. Other topics include safeguard and protection considerations; occupational radiation exposure including ALARA programs; shielding of canisters, transporters and forklift trucks; monitoring considerations; mine water treatment; canister integrity; and criticality calculations
Multibarrier system preventing migration of radionuclides from radioactive waste repository
Directory of Open Access Journals (Sweden)
Olszewska Wioleta
2015-09-01
Full Text Available Safety of radioactive waste repositories operation is associated with a multibarrier system designed and constructed to isolate and contain the waste from the biosphere. Each of radioactive waste repositories is equipped with system of barriers, which reduces the possibility of release of radionuclides from the storage site. Safety systems may differ from each other depending on the type of repository. They consist of the natural geological barrier provided by host rocks of the repository and its surroundings, and an engineered barrier system (EBS. The EBS may itself comprise a variety of sub-systems or components, such as waste forms, canisters, buffers, backfills, seals and plugs. The EBS plays a major role in providing the required disposal system performance. It is assumed that the metal canisters and system of barriers adequately isolate waste from the biosphere. The evaluation of the multibarrier system is carried out after detailed tests to determine its parameters, and after analysis including mathematical modeling of migration of contaminants. To provide an assurance of safety of radioactive waste repository multibarrier system, detailed long term safety assessments are developed. Usually they comprise modeling of EBS stability, corrosion rate and radionuclide migration in near field in geosphere and biosphere. The principal goal of radionuclide migration modeling is assessment of the radionuclides release paths and rate from the repository, radionuclides concentration in geosphere in time and human exposure to ionizing radiation
Barker, Charles E.; Dallegge, Todd A.
2005-01-01
Coal desorption techniques typically use the U.S. Bureau of Mines (USBM) canister-desorption method as described by Diamond and Levine (1981), Close and Erwin (1989), Ryan and Dawson (1993), McLennan and others (1994), Mavor and Nelson (1997) and Diamond and Schatzel (1998). However, the coal desorption canister designs historically used with this method have an inherent flaw that allows a significant gas-filled headspace bubble to remain in the canister that later has to be compensated for by correcting the measured desorbed gas volume with a mathematical headspace volume correction (McLennan and others, 1994; Mavor and Nelson, 1997).
Naval Waste Package Design Sensitivity
International Nuclear Information System (INIS)
T. Schmitt
2006-01-01
The purpose of this calculation is to determine the sensitivity of the structural response of the Naval waste packages to varying inner cavity dimensions when subjected to a comer drop and tip-over from elevated surface. This calculation will also determine the sensitivity of the structural response of the Naval waste packages to the upper bound of the naval canister masses. The scope of this document is limited to reporting the calculation results in terms of through-wall stress intensities in the outer corrosion barrier. This calculation is intended for use in support of the preliminary design activities for the license application design of the Naval waste package. It examines the effects of small changes between the naval canister and the inner vessel, and in these dimensions, the Naval Long waste package and Naval Short waste package are similar. Therefore, only the Naval Long waste package is used in this calculation and is based on the proposed potential designs presented by the drawings and sketches in References 2.1.10 to 2.1.17 and 2.1.20. All conclusions are valid for both the Naval Long and Naval Short waste packages
A welding system for spent fuel canister lid
International Nuclear Information System (INIS)
Suikki, M.; Wendelin, T.
2008-06-01
The report presents a proposed welding system for spent fuel canister lids. The system is used for welding the copper lid to the copper overpack. The apparatus will be installed in the encapsulation plant. The report presents basic requirements for and implementation of the welding system, operation, service and maintenance of the equipment, as well as a cost estimate. Some aspects of the apparatus design are quite specified, but the actual detailed planning and final selection of components is not included. The report also describes actions for possible malfunction and fault conditions. Closing of the copper cylinder's lid is carried out by electron beam welding, which must be performed in vacuum. The welding system for spent fuel canister lid consists of two welding chambers, a canister docking system, an EB-welding machine with its accessories, a vacuum apparatus, as well as necessary auxiliary equipment. The system's equipment is housed in a welding room, an auxiliary system room, an operation control room, as well as mounted on the ceiling of a transfer corridor. One of the welding chambers is intended for carrying out test welding procedures and for calibration of welding parameters. The actual spent fuel canister lid welding chamber has a weldingready canister docked thereto in an airtight manner. The chamber is pumped for a vacuum, followed by closing the canister's copper lid and carrying out the lid welding process. The lid is brought into the chamber prior to docking the canister by means of a canister transfer trolley lifting gear. Lifting of the canister and rotating it during a welding process are also handled by means of the transfer trolley. The lid welding chamber houses equipment for the alignment and installation of the lid, as well as heating means for the top side of a copper overpack for ensuring a sufficient installation clearance between the lid and the overpack. The equipment not needed in the immediate vicinity of welding chambers, is
The construction of solid waste form test and inspection facility
International Nuclear Information System (INIS)
Park, Hun Hwee; Lee, Kang Moo; Jung, In Ha; Kim, Sung Hwan; Yoo, Jeong Woo; Lee, Jong Youl; Bae, Sang Min
1988-01-01
The solid waste form test and inspection facility is a facility to test and inspect the characteristics of waste forms, such as homogenity, mechanical structure, thermal behaviour, water resistance and leachability. Such kinds of characteristics in waste forms are required to meet a certain conditions for long-term storage or for final disposal of wastes. The facility will be used to evaluate safety for the disposal of wastes by test and inspection. At this moment, the efforts to search the most effective management of the radioactive wastes generated from power plants and radioisotope user are being executed by the people related to this field. Therefore, the facility becomes more significant tool because of its guidance of sucessfully converting wastes into forms to give a credit to the safety of waste disposal for managing the radioactive wastes. In addition the overall technical standards for inspecting of waste forms such as the standardized equipment and processes in the facility will be estabilished in the begining of 1990's when the project of waste management will be on the stream. Some of the items of the project have been standardized for the purpose of localization. In future, this facility will be utilized not only for the inspection of waste forms but also for the periodic decontamination apparatus by remote operation techniques. (Author)
Multibarrier waste forms. Part III: Process considerations
International Nuclear Information System (INIS)
Lokken, R.O.
1979-10-01
The multibarrier concept for the solidification and storage of radioactive waste utilizes up to three barriers to isolate radionuclides from the environment: a solidified waste inner core, an impervious coating, and a metal matrix. The coating and metal matrix give the composite waste form enhanced inertness with improvements in thermal stability, mechanical strength, and leach resistance. Preliminary process flow rates and material costs were evaluated for four multibarrier waste forms with the process complexity increasing thusly: glass marbles, uncoated supercalcine, glass-coated supercalcine, and PyC/Al 2 O 3 -coated supercalcine. This report discusses the process variables and their effect on optimization of product quality, processing simplicity, and material cost. 11 figures, 2 tables
Fuel and canister process report for the safety assessment SR-Can
International Nuclear Information System (INIS)
Werme, Lars
2006-10-01
This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The report is written by, and for, experts in the relevant scientific fields. It should though be possible for a generalist in the area of long-term safety assessments of geologic nuclear waste repositories to comprehend the contents of the report. The report is an important part of the documentation of the SR-Can project and an essential reference within the project, providing a scientifically motivated plan for the handling of geosphere processes. It is, furthermore, foreseen that the report will be essential for reviewers scrutinising the handling of geosphere issues in the SR-Can assessment. Several types of fuel will be emplaced in the repository. For the reference case with 40 years of reactor operation, the fuel quantity from boiling water reactors, BWR fuel, is estimated at 7,000 tonnes, while the quantity from pressurized water reactors, PWR fuel, is estimated at about 2,300 tonnes. In addition, 23 tonnes of mixed-oxide fuel (MOX) fuel of German origin from BWR and PWR reactors and 20 tonnes of fuel from the decommissioned heavy water reactor in Aagesta will be disposed of. To allow for future changes in the Swedish nuclear programme, the safety assessment assumes a total of 6,000 canister corresponding to 12,000 tonnes of fuel
Fuel and canister process report for the safety assessment SR-Can
Energy Technology Data Exchange (ETDEWEB)
Werme, Lars (ed.)
2006-10-15
This report documents fuel and canister processes identified as relevant to the long-term safety of a KBS-3 repository. It forms an important part of the reporting of the safety assessment SR-Can. The detailed assessment methodology, including the role of the process report in the assessment, is described in the SR-Can Main report. The report is written by, and for, experts in the relevant scientific fields. It should though be possible for a generalist in the area of long-term safety assessments of geologic nuclear waste repositories to comprehend the contents of the report. The report is an important part of the documentation of the SR-Can project and an essential reference within the project, providing a scientifically motivated plan for the handling of geosphere processes. It is, furthermore, foreseen that the report will be essential for reviewers scrutinising the handling of geosphere issues in the SR-Can assessment. Several types of fuel will be emplaced in the repository. For the reference case with 40 years of reactor operation, the fuel quantity from boiling water reactors, BWR fuel, is estimated at 7,000 tonnes, while the quantity from pressurized water reactors, PWR fuel, is estimated at about 2,300 tonnes. In addition, 23 tonnes of mixed-oxide fuel (MOX) fuel of German origin from BWR and PWR reactors and 20 tonnes of fuel from the decommissioned heavy water reactor in Aagesta will be disposed of. To allow for future changes in the Swedish nuclear programme, the safety assessment assumes a total of 6,000 canister corresponding to 12,000 tonnes of fuel.
Energy Technology Data Exchange (ETDEWEB)
Choi, Jung-Hoon, E-mail: mrchoijh@kaeri.re.kr; Eun, Hee-Chul; Lee, Tae-Kyo; Lee, Ki-Rak; Han, Seung-Youb; Jeon, Min-Ku; Park, Hwan-Seo; Ahn, Do-Hee
2017-01-15
Estimation of centerline temperature of nuclear glass waste form for each waste stream is very essential in the period of storage because the centerline temperature being over its glass transition temperature results in the increase of leaching rate of radioactive nuclides due to the devitrification of glass waste form. Here, to verify the effects of waste form diameter and transuranic element content in the rare earth waste on the centerline temperature of the waste form, the surrogate rare earth glass waste generated from pyrochemical process was immobilized with SiO{sub 2}−Al{sub 2}O{sub 3}−B{sub 2}O{sub 3} glass frit system, and thermal properties of the rare earth glass waste form were determined by thermomechanical analysis and thermal conductivity analysis. The estimation of centerline temperature was carried out using the experimental thermal data and steady-state conduction equation in a long and solid cylinder type waste form. It was revealed that thermal stability of waste form in case of 0.3 m diameter was not affected by the TRU content even in the case of 80% TRU recovery ratio in the electrowinning process, meaning that the waste form of 0.3 m diameter is thermally stable due to the low centerline temperature relative to its glass transition temperature of the rare earth glass waste form.
Economics of defense high level waste management in the United States
International Nuclear Information System (INIS)
McDonell, W.R.
1987-01-01
Life-cycle costs of defense waste disposal, as presented in the foregoing sections, are summarized. Expressed as incremental costs per canister of waste deposited in a Federal geologic repository and per gallon of decontaminated salt solution immobilized in onsite concrete vaults, the tabulated values provide a measure of waste management costs relatively independent of the inventories of waste processed. Total values are about $350,000 per glass waste canister processed and $4.68 per gallon of decontaminated salt immobilized. These costs do not generally include contributions of fixed charges, such as capital costs, except in the case of transport and repository charges for which the quantities of waste handled determine allocation of fixed costs included in the fee assessments. 14 refs., 2 figs., 3 tabs
Energy Technology Data Exchange (ETDEWEB)
Cantrell, Kirk J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Westsik, Joseph H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Serne, R Jeffrey [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Cozzi, Alex D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)
2016-05-16
A review of the most up-to-date and relevant data currently available was conducted to develop a set of recommended values for use in the Integrated Disposal Facility (IDF) performance assessment (PA) to model contaminant release from a cementitious waste form for aqueous wastes treated at the Hanford Effluent Treatment Facility (ETF). This data package relies primarily upon recent data collected on Cast Stone formulations fabricated with simulants of low-activity waste (LAW) and liquid secondary wastes expected to be produced at Hanford. These data were supplemented, when necessary, with data developed for saltstone (a similar grout waste form used at the Savannah River Site). Work is currently underway to collect data on cementitious waste forms that are similar to Cast Stone and saltstone but are tailored to the characteristics of ETF-treated liquid secondary wastes. Recommended values for key parameters to conduct PA modeling of contaminant release from ETF-treated liquid waste are provided.
Process Design Concepts for Stabilization of High Level Waste Calcine
Energy Technology Data Exchange (ETDEWEB)
T. R. Thomas; A. K. Herbst
2005-06-01
The current baseline assumption is that packaging ¡§as is¡¨ and direct disposal of high level waste (HLW) calcine in a Monitored Geologic Repository will be allowed. The fall back position is to develop a stabilized waste form for the HLW calcine, that will meet repository waste acceptance criteria currently in place, in case regulatory initiatives are unsuccessful. A decision between direct disposal or a stabilization alternative is anticipated by June 2006. The purposes of this Engineering Design File (EDF) are to provide a pre-conceptual design on three low temperature processes under development for stabilization of high level waste calcine (i.e., the grout, hydroceramic grout, and iron phosphate ceramic processes) and to support a down selection among the three candidates. The key assumptions for the pre-conceptual design assessment are that a) a waste treatment plant would operate over eight years for 200 days a year, b) a design processing rate of 3.67 m3/day or 4670 kg/day of HLW calcine would be needed, and c) the performance of waste form would remove the HLW calcine from the hazardous waste category, and d) the waste form loadings would range from about 21-25 wt% calcine. The conclusions of this EDF study are that: (a) To date, the grout formulation appears to be the best candidate stabilizer among the three being tested for HLW calcine and appears to be the easiest to mix, pour, and cure. (b) Only minor differences would exist between the process steps of the grout and hydroceramic grout stabilization processes. If temperature control of the mixer at about 80„aC is required, it would add a major level of complexity to the iron phosphate stabilization process. (c) It is too early in the development program to determine which stabilizer will produce the minimum amount of stabilized waste form for the entire HLW inventory, but the volume is assumed to be within the range of 12,250 to 14,470 m3. (d) The stacked vessel height of the hot process vessels
Phosphate bonded ceramics as candidate final-waste-form materials
International Nuclear Information System (INIS)
Singh, D.; Wagh, A.S.; Cunnane, J.; Sutaria, M.; Kurokawa, S.; Mayberry, J.
1994-04-01
Room-temperature setting phosphate-bonded ceramics were studied as candidate materials for stabilization of DOE low-level problem mixed wastes which cannot be treated by other established stabilization techniques. Phosphates of Mg, Mg-Na, Al and Zr were studied to stabilize ash surrogate waste containing RCRA metals as nitrates and RCRA organics. We show that for a typical loading of 35 wt.% of the ash waste, the phosphate ceramics pass the TCLP test. The waste forms have high compression strength exceeding ASTM recommendations for final waste forms. Detailed X-ray diffraction studies and differential thermal analyses of the waste forms show evidence of chemical reaction of the waste with phosphoric acid and the host matrix. The SEM studies show evidence of physical bonding. The excellent performance in the leaching tests is attributed to a chemical solidification and physical as well as chemical bonding of ash wastes in these phosphate ceramics
System for handling and storing radioactive waste
Anderson, John K.; Lindemann, Paul E.
1984-01-01
A system and method for handling and storing spent reactor fuel and other solid radioactive waste, including canisters to contain the elements of solid waste, storage racks to hold a plurality of such canisters, storage bays to store these racks in isolation by means of shielded doors in the bays. This system also includes means for remotely positioning the racks in the bays and an access tunnel within which the remotely operated means is located to position a rack in a selected bay. The modular type of these bays will facilitate the construction of additional bays and access tunnel extension.
Materials considerations relative to multibarrier waste isolation
International Nuclear Information System (INIS)
McCoy, H.E.; Griess, J.C.
1981-07-01
The environmental conditions associated with the storage of radioactive wastes are reviewed, and the corrosion of potential waste containment materials under these conditions is evaluated. The desired service life of about 1000 years is beyond the time period for which existing corrosion data can be extrapolated with certainty; however, titanium alloys seem to offer the most promise. The mechanical requirements for canisters and overpacks are considered and several candidate materials are selected. Designs for a canister and an overpack have been developed, and these are used to estimate the costs for three possible materials of construction
International Nuclear Information System (INIS)
Nakamura, Haruto; Tashiro, Shingo
1988-10-01
Researches on high-level waste management at the High Level Waste Management Laboratory and the Waste Safety Testing Facility Operation Division of the Japan Atomic Energy Research Institute in the fiscal year of 1987 are reviewed in the three sections of the report. The topics are as follows: 1) On performance and durability of waste forms and engineered barrier materials, accelerated alpha radiation stability of glass form and Synroc has been investigated and stress corrosion cracking of canister materials was examined under simulated conditions. 2) Sorption of 237 Np on granite samples and behavior of iron during weathering of granites were studied with respect to safety evaluation for geological disposal. 3) Actual waste was transported from the Tokai Reprocessing Plant and hot operation using the actual waste was initiated at WASTEF. (author)
International Nuclear Information System (INIS)
1977-09-01
The objective of this document is to provide information or alternatives that are being considered for the long-term management of defense high-level radioactive waste stored at Hanford in underground tanks and in stainless steel-lined concrete basins. For purposes of basic programmatic decision making, four major alternatives based on disposal location are considered. The steps leading to placement of the waste in the following locations are illustrated: existing waste tanks; onsite engineered surface facilities; onsite geologic repository; and offsite geologic repository. The four major disposal alternatives are expanded into 27 alternative plans by considering: (1) variations in the final form of the high-level fraction (with radionuclide removal) to include glass, concrete, and powder; (2) variations in the final form of the dehydrated waste product to include glass, calcined clay, and powder; and (3) variations in the treatment and handling of encapsulated waste to include packaging of capsules in canisters and conversion of the strontium fluoride and cesium chloride to glass; canisters stored in sealed casks on the surface are disposed of in a surface vault after the radionuclides have decayed sufficiently to avoid a heat-transfer problem. A description of the technology, a preliminary risk assessment, and preliminary cost estimates for each of these 27 plans are presented. The technology required to implement any of the 27 alternative plans has not been developed to the point where any plan can be considered completely technically sound and feasible
Low-level radioactive waste form qualification testing
International Nuclear Information System (INIS)
Sohal, M.S.; Akers, D.W.
1998-06-01
This report summarizes activities that have already been completed as well as yet to be performed by the Idaho National Engineering and Environmental Laboratory (INEEL) to develop a plan to quantify the behavior of radioactive low-level waste forms. It briefly describes the status of various tasks, including DOE approval of the proposed work, several regulatory and environmental related documents, tests to qualify the waste form, preliminary schedule, and approximate cost. It is anticipated that INEEL and Brookhaven National Laboratory will perform the majority of the tests. For some tests, services of other testing organizations may be used. It should take approximately nine months to provide the final report on the results of tests on a waste form prepared for qualification. It is anticipated that the overall cost of the waste quantifying service is approximately $150,000. The following tests are planned: compression, thermal cycling, irradiation, biodegradation, leaching, immersion, free-standing liquid tests, and full-scale testing
Low-level radioactive waste form qualification testing
Energy Technology Data Exchange (ETDEWEB)
Sohal, M.S.; Akers, D.W.
1998-06-01
This report summarizes activities that have already been completed as well as yet to be performed by the Idaho National Engineering and Environmental Laboratory (INEEL) to develop a plan to quantify the behavior of radioactive low-level waste forms. It briefly describes the status of various tasks, including DOE approval of the proposed work, several regulatory and environmental related documents, tests to qualify the waste form, preliminary schedule, and approximate cost. It is anticipated that INEEL and Brookhaven National Laboratory will perform the majority of the tests. For some tests, services of other testing organizations may be used. It should take approximately nine months to provide the final report on the results of tests on a waste form prepared for qualification. It is anticipated that the overall cost of the waste quantifying service is approximately $150,000. The following tests are planned: compression, thermal cycling, irradiation, biodegradation, leaching, immersion, free-standing liquid tests, and full-scale testing.
Corrosion experience in nuclear waste processing at Battelle Northwest
International Nuclear Information System (INIS)
Slate, S.C.; Maness, R.F.
1976-11-01
Emphasis is on corrosion as related to waste storage canister. Most work has been done in support of the In-Can Melter (ICM) vitrification system. It is assumed that the canister goes through the ICM process and is then stored in a water basin. The most severe corrosion effect seen is oxidation of stainless steel (SS) surfaces in contact with gases containing oxygen during processing. The processing temperature is near 1100 0 C and furnace atmosphere, used until now, has been air with unrestricted flow to the furnace. The oxidation rate at 1100 0 C is 15.8 g/cm 2 for 304L SS. Techniques for eliminating this corrosion currently being investigated include the use of different materials, such as Inconel 601, and the use of an inert cover gas. Corrosion due to the waste melt is not as rapid as the air oxidation. This effect has been studied extensively in connection with the development of a metallic crucible melter at Battelle. Data are available on the corrosion rates of several waste compositions in contact with various materials. Long-term compatibility tests between the melt and the metal have been run; it was found the corrosion rates due to the melt or its vapor do not pose a serious problem to the waste canister. However, these rates are high enough to preclude the practical use of a metallic melter. Interim water storage of the canister may be a problem if proper corrective measurements are not taken.The canister may be susceptible to stress corrosion cracking (SCC) because it will be sensitized to some extent and it will be nearly stressed to yield. The most favorable solution to SCC involves minimizing canister sensitization and stress plus providing good water quality control. It has been recommended to keep the chlorine ion concentration below 1 ppM and the pH above 10. At these conditions no failures of 304L are predicted due to SCC. It is concluded that corrosion of a canister used during the In-Can Melter process and interim storage can be controlled
Crystalline Ceramic Waste Forms: Comparison Of Reference Process For Ceramic Waste Form Fabrication
Energy Technology Data Exchange (ETDEWEB)
Brinkman, K. S. [Savannah River National Laboratory; Marra, J. C. [Savannah River National Laboratory; Amoroso, J. [Savannah River National Laboratory; Tang, M. [Los Alamos National Laboratory
2013-08-22
The research conducted in this work package is aimed at taking advantage of the long term thermodynamic stability of crystalline ceramics to create more durable waste forms (as compared to high level waste glass) in order to reduce the reliance on engineered and natural barrier systems. Durable ceramic waste forms that incorporate a wide range of radionuclides have the potential to broaden the available disposal options and to lower the storage and disposal costs associated with advanced fuel cycles. Assemblages of several titanate phases have been successfully demonstrated to incorporate radioactive waste elements, and the multiphase nature of these materials allows them to accommodate variation in the waste composition. Recent work has shown that they can be produced from a melting and crystallization process. The objective of this report is to explore the phase formation and microstructural differences between lab scale melt processing in varying gas environments with alternative densification processes such as Hot Pressing (HP) and Spark Plasma Sintering (SPS). The waste stream used as the basis for the development and testing is a simulant derived from a combination of the projected Cs/Sr separated stream, the Trivalent Actinide - Lanthanide Separation by Phosphorous reagent Extraction from Aqueous Komplexes (TALSPEAK) waste stream consisting of lanthanide fission products, the transition metal fission product waste stream resulting from the transuranic extraction (TRUEX) process, and a high molybdenum concentration with relatively low noble metal concentrations. Melt processing as well as solid state sintering routes SPS and HP demonstrated the formation of the targeted phases; however differences in microstructure and elemental partitioning were observed. In SPS and HP samples, hollandite, pervoskite/pyrochlore, zirconolite, metallic alloy and TiO{sub 2} and Al{sub 2}O{sub 3} were observed distributed in a network of fine grains with small residual pores
Characterization of low and medium level radioactive waste forms
International Nuclear Information System (INIS)
Sambell, R.A.J.
1983-01-01
The work reported was carried out during the first year of the Commission of the European Community's programme on the characterization of low and medium level waste forms. Ten reference waste forms plus others of special national interest have been identified covering PWR, BWR, GCR and reprocessing wastes. The immobilising media include the three main matrices: cement, polymers and bitumen, and a glass. Characterization is viewed as one input to quality assurance of the waste form and covers: waste-matrix compatibility, radiation effects, leaching, microbiological attack, shrinkage and swelling, ageing processes and thermal effects. The aim is a balanced programme of comparative data, predictive modelling and an undserstanding of basic mechanisms
NNWSI waste form testing at Argonne National Laboratory
International Nuclear Information System (INIS)
Bates, J.K.; Gerding, T.J.; Abrajano, T.A. Jr.; Ebert, W.L.; Mazer, J.J.
1988-11-01
The Nevada Nuclear Waste Storage Investigation (NNWSI) Project is investigating the tuff beds of Yucca Mountain, Nevada, as a potential location for a high-level radioactive waste repository. As part of the waste package development portion of this project, experiments are being performed by the Chemical Technology Division of Argonne National Laboratory to study the behavior of the waste form under anticipated repository conditions. These experiments include the development and performance of a test to measure waste form behavior in unsaturated conditions and the performance of experiments designed to study the behavior of waste package components in an irradiated environment. Previous reports document developments in these areas through 1986. This report summarizes progress during the period January--June 1987, 19 refs., 17 figs., 20 tabs
Waste package project quarterly report, April 1, 1995--June 30, 1995
International Nuclear Information System (INIS)
Ladkany, S.G.
1995-01-01
Research activities spanned multidisciplinary areas: structural and stress analysis of the container including nonlinear yield and damage assessment and structural stability studies of the dynamic behavior of the steel multipurpose container under normal and accidental handling conditions; nuclear fission criticality studies in the canisters; investigation of novel canister design concepts and corrosion studies; heat transfer studies of the waste canisters and in adjacent rock drifts; fluid flow in porous media and radionuclide transport in near field rock; stresses and stability of the rock formations resulting from thermal loading of fuel elements and multi-tunnel concept being analyzed; faulted rock tunnel model; and three link steel robot for remote handling of high level nuclear waste
Waste Form Features, Events, and Processes
International Nuclear Information System (INIS)
R. Schreiner
2004-01-01
The purpose of this report is to evaluate and document the inclusion or exclusion of the waste form features, events and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment for License Application (TSPA-LA). A screening decision, either Included or Excluded, is given for each FEP along with the technical bases for screening decisions. This information is required by the Nuclear Regulatory Commission (NRC) in 10 CFR 63.114 (d, e, and f) [DIRS 156605]. The FEPs addressed in this report deal with the issues related to the degradation and potential failure of the waste form and the migration of the waste form colloids. For included FEPs, this analysis summarizes the implementation of the FEP in TSPA-LA, (i.e., how the FEP is included). For excluded FEPs, this analysis provides the technical bases for exclusion from TSPA-LA (i.e., why the FEP is excluded). This revision addresses the TSPA-LA FEP list (DTN: MO0407SEPFEPLA.000 [DIRS 170760]). The primary purpose of this report is to identify and document the analyses and resolution of the features, events, and processes (FEPs) associated with the waste form performance in the repository. Forty FEPs were identified that are associated with the waste form performance. This report has been prepared to document the screening methodology used in the process of FEP inclusion and exclusion. The analyses documented in this report are for the license application (LA) base case design (BSC 2004 [DIRS 168489]). In this design, a drip shield is placed over the waste package and no backfill is placed over the drip shield (BSC 2004 [DIRS 168489]). Each FEP may include one or more specific issues that are collectively described by a FEP name and a FEP description. The FEP description may encompass a single feature, process or event, or a few closely related or coupled processes if the entire FEP can be addressed by a single specific screening argument or TSPA-LA disposition. The FEPs are
Waste Form Features, Events, and Processes
Energy Technology Data Exchange (ETDEWEB)
R. Schreiner
2004-10-27
The purpose of this report is to evaluate and document the inclusion or exclusion of the waste form features, events and processes (FEPs) with respect to modeling used to support the Total System Performance Assessment for License Application (TSPA-LA). A screening decision, either Included or Excluded, is given for each FEP along with the technical bases for screening decisions. This information is required by the Nuclear Regulatory Commission (NRC) in 10 CFR 63.114 (d, e, and f) [DIRS 156605]. The FEPs addressed in this report deal with the issues related to the degradation and potential failure of the waste form and the migration of the waste form colloids. For included FEPs, this analysis summarizes the implementation of the FEP in TSPA-LA, (i.e., how the FEP is included). For excluded FEPs, this analysis provides the technical bases for exclusion from TSPA-LA (i.e., why the FEP is excluded). This revision addresses the TSPA-LA FEP list (DTN: MO0407SEPFEPLA.000 [DIRS 170760]). The primary purpose of this report is to identify and document the analyses and resolution of the features, events, and processes (FEPs) associated with the waste form performance in the repository. Forty FEPs were identified that are associated with the waste form performance. This report has been prepared to document the screening methodology used in the process of FEP inclusion and exclusion. The analyses documented in this report are for the license application (LA) base case design (BSC 2004 [DIRS 168489]). In this design, a drip shield is placed over the waste package and no backfill is placed over the drip shield (BSC 2004 [DIRS 168489]). Each FEP may include one or more specific issues that are collectively described by a FEP name and a FEP description. The FEP description may encompass a single feature, process or event, or a few closely related or coupled processes if the entire FEP can be addressed by a single specific screening argument or TSPA-LA disposition. The FEPs are
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
Determining leach rates of monolithic waste forms
International Nuclear Information System (INIS)
Gilliam, T.M.; Dole, L.R.
1986-01-01
The ANS 16.1 Leach Procedure provides a conservative means of predicting long-term release from monolithic waste forms, offering a simple and relatively quick means of determining effective solid diffusion coefficients. As presented here, these coefficients can be used in a simple model to predict maximum release rates or be used in more complex site-specific models to predict actual site performance. For waste forms that pass the structural integrity test, this model also allows the prediction of EP-Tox leachate concentrations from these coefficients. Thus, the results of the ANS 16.1 Leach Procedure provide a powerful tool that can be used to predict the waste concentration limits in order to comply with the EP-Toxicity criteria for characteristically nonhazardous waste. 12 refs., 3 figs
Defense Remote Handled Transuranic Waste Cost/Schedule Optimization Study
International Nuclear Information System (INIS)
Pierce, G.D.; Wolaver, R.W.; Carson, P.H.
1986-11-01
The purpose of this study is to provide the DOE information with which it can establish the most efficient program for the long management and disposal, in the Waste Isolation Pilot Plant (WIPP), of remote handled (RH) transuranic (TRU) waste. To fulfill this purpose, a comprehensive review of waste characteristics, existing and projected waste inventories, processing and transportation options, and WIPP requirements was made. Cost differences between waste management alternatives were analyzed and compared to an established baseline. The result of this study is an information package that DOE can use as the basis for policy decisions. As part of this study, a comprehensive list of alternatives for each element of the baseline was developed and reviewed with the sites. The principle conclusions of the study follow. A single processing facility for RH TRU waste is both necessary and sufficient. The RH TRU processing facility should be located at Oak Ridge National Laboratory (ORNL). Shielding of RH TRU to contact handled levels is not an economic alternative in general, but is an acceptable alternative for specific waste streams. Compaction is only cost effective at the ORNL processing facility, with a possible exception at Hanford for small compaction of paint cans of newly generated glovebox waste. It is more cost effective to ship certified waste to WIPP in 55-gal drums than in canisters, assuming a suitable drum cask becomes available. Some waste forms cannot be packaged in drums, a canister/shielded cask capability is also required. To achieve the desired disposal rate, the ORNL processing facility must be operational by 1996. Implementing the conclusions of this study can save approximately $110 million, compared to the baseline, in facility, transportation, and interim storage costs through the year 2013. 10 figs., 28 tabs
Analyses of atmospheric radon 222 / canisters exposed by Greenpeace in Niger (Arlit / Akokan sector)
International Nuclear Information System (INIS)
Chareyron, B.
2010-01-01
The companies SOMAIR and COMINAK, subsidiaries of the AREVA group, are mining uranium deposits in northern Niger. In the course of a field mission carried out in November 2009, a Greenpeace International team deposited detectors (canisters of activated charcoal) to measure radon 222, a radioactive gas formed by the decay of the radium 226 present in the uranium ore. This report includes the results of the analysis of the activated charcoal canisters conducted in CRIIRAD's laboratory, and a brief commentary on the interpretation of the results. (authors)
Application of PCT to the EBR II ceramic waste form
International Nuclear Information System (INIS)
Ebert, W. L.; Lewis, M. A.; Johnson, S. G.
2002-01-01
We are evaluating the use of the Product Consistency Test (PCT) developed to monitor the consistency of borosilicate glass waste forms for application to the multiphase ceramic waste form (CWF) that will be used to immobilize waste salts generated during the electrometallurgical conditioning of spent sodium-bonded nuclear fuel from the Experimental Breeder Reactor No. 2 (EBR II). The CWF is a multiphase waste form comprised of about 70% sodalite, 25% borosilicate glass binder, and small amounts of halite and oxide inclusions. It must be qualified for disposal as a non-standard high-level waste (HLW) form. One of the requirements in the DOE Waste Acceptance System Requirements Document (WASRD) for HLW waste forms is that the consistency of the waste forms be monitored.[1] Use of the PCT is being considered for the CWF because of the similarities of the dissolution behaviors of both the sodalite and glass binder phases in the CWF to borosilicate HLW glasses. This paper provides (1) a summary of the approach taken in selecting a consistency test for CWF production and (2) results of tests conducted to measure the precision and sensitivity of the PCT conducted with simulated CWF
Geotechnical engineering for ocean waste disposal. An introduction
Lee, Homa J.; Demars, Kenneth R.; Chaney, Ronald C.; ,
1990-01-01
As members of multidisciplinary teams, geotechnical engineers apply quantitative knowledge about the behavior of earth materials toward designing systems for disposing of wastes in the oceans and monitoring waste disposal sites. In dredge material disposal, geotechnical engineers assist in selecting disposal equipment, predict stable characteristics of dredge mounds, design mound caps, and predict erodibility of the material. In canister disposal, geotechnical engineers assist in specifying canister configurations, predict penetration depths into the seafloor, and predict and monitor canister performance following emplacement. With sewage outfalls, geotechnical engineers design foundation and anchor elements, estimate scour potential around the outfalls, and determine the stability of deposits made up of discharged material. With landfills, geotechnical engineers evaluate the stability and erodibility of margins and estimate settlement and cracking of the landfill mass. Geotechnical engineers also consider the influence that pollutants have on the engineering behavior of marine sediment and the extent to which changes in behavior affect the performance of structures founded on the sediment. In each of these roles, careful application of geotechnical engineering principles can contribute toward more efficient and environmentally safe waste disposal operations.