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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

Jantzen, C.M.

1992-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-30

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

Corrosion impact of reductant on DWPF and downstream facilities

Energy Technology Data Exchange (ETDEWEB)

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

2014-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

RECENT PROCESS IMPROVEMENTS TO INCREASE HLW THROUGHPUT AT THE DWPF

International Nuclear Information System (INIS)

Herman, C

2007-01-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1988-01-01

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

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

Relaxation of the lower frit loading constraint for DWPF process control

International Nuclear Information System (INIS)

Brown, K.G.

2000-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1993-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-06-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

species and to verify the performance of materials in the key process vessels as well as downstream vessels and processes such as the evaporator where heating is occurring. The following testing would provide data for establishing the viability of these components: Electrochemical testing - evaluate the corrosion rate and susceptibility to localized corrosion within the SRAT, SME, OGCT, Quencher and Evaporator. Testing would be conducted at operational temperatures in simulants with ranges of glycolic acid, iron, chloride, sulfate, mercury, and antifoaming agents; Hot-wall testing – evaluate the corrosion under heat transfer conditions to simulate those for heating coils and evaporator coil surfaces. Testing would be at nominal chemistries with concentration of glycolic acid, chloride, sulfate and mercury at high expected concentrations. Some tests would be performed with antifoaming agents; Melter coupon testing – evaluate the performance of alloy 690 in melter feeds containing glycolic acid. This testing would be conducted as part of the melter flammability testing; and, Polymer testing – evaluate changes in polymer properties in immersion testing with DWPF simulants to provide product-specific data for service life evaluation and analyze the Hansen solubility parameters for relevant polymers in glycolic vs. formic acid. During this literature review process, the difficulties associated with measuring the liquid level in formic acid tanks were revealed. A test is recommended to resolve this issue prior to the introduction of glycolic acid into the DWPF. This testing would evaluate the feasibility of using ultrasonic inspection techniques to determine liquid level and other desirable attributes of glycolic acid in DWPF storage tanks and related equipment.

DWPF remotable television and cell lighting facilities

International Nuclear Information System (INIS)

Heckendorn, F.M. II.

1984-01-01

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

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

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

International Nuclear Information System (INIS)

Ha, B.C.

1994-01-01

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

Conditions for precipitation of copper phases in DWPF waste glass

International Nuclear Information System (INIS)

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

1993-01-01

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

DWPF Simulant CPC Studies For SB8

Energy Technology Data Exchange (ETDEWEB)

Newell, J. D.

2013-09-25

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

High level radioactive waste management facility design criteria

International Nuclear Information System (INIS)

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

1993-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-30

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

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

International Nuclear Information System (INIS)

Amoroso, J.

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-09-18

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

Liquidus Temperature Data for DWPF Glass

International Nuclear Information System (INIS)

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

1999-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-01-01

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

Analysis of the DWPF glass pouring system using neural networks

International Nuclear Information System (INIS)

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

1997-01-01

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

DWPF MATERIALS EVALUATION SUMMARY REPORT

Energy Technology Data Exchange (ETDEWEB)

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

1996-09-12

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

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

International Nuclear Information System (INIS)

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

1988-01-01

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

The DWPF waste form qualification program

International Nuclear Information System (INIS)

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

1994-01-01

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

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

International Nuclear Information System (INIS)

Harris, S.P.

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-05-11

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

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Glass sampling program during DWPF Integrated Cold Runs

International Nuclear Information System (INIS)

Plodinec, M.J.

1990-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-07-12

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-11-20

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Program plan: DWPF/HLWDP stirred Melter Program Plan

International Nuclear Information System (INIS)

Smith, M.E.

1994-01-01

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

Pilot scale processing of simulated Savannah River Site high level radioactive waste

International Nuclear Information System (INIS)

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

1991-01-01

The Savannah River Laboratory operates the Integrated DWPF Melter System (IDMS), which is a pilot-scale test facility used in support of the start-up and operation of the US Department of Energy's Defense Waste Processing Facility (DWPF). Specifically, the IDMS is used in the evaluation of the DWPF melter and its associated feed preparation and offgass treatment systems. This article provides a general overview of some of the test work which has been conducted in the IDMS facility. The chemistry associated with the chemical treatment of the sludge (via formic acid adjustment) is discussed. Operating experiences with simulated sludge containing high levels of nitrite, mercury, and noble metals are summarized

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-30

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-30

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

The corrosion behavior of DWPF glasses

International Nuclear Information System (INIS)

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

1995-01-01

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

DWPF process control

International Nuclear Information System (INIS)

Heckendoin, F.M. II

1983-01-01

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

High level nuclear wastes

International Nuclear Information System (INIS)

Lopez Perez, B.

1987-01-01

The transformations involved in the nuclear fuels during the burn-up at the power nuclear reactors for burn-up levels of 33.000 MWd/th are considered. Graphs and data on the radioactivity variation with the cooling time and heat power of the irradiated fuel are presented. Likewise, the cycle of the fuel in light water reactors is presented and the alternatives for the nuclear waste management are discussed. A brief description of the management of the spent fuel as a high level nuclear waste is shown, explaining the reprocessing and giving data about the fission products and their radioactivities, which must be considered on the vitrification processes. On the final storage of the nuclear waste into depth geological burials, both alternatives are coincident. The countries supporting the reprocessing are indicated and the Spanish programm defined in the Plan Energetico Nacional (PEN) is shortly reviewed. (author) 8 figs., 4 tabs

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

International Nuclear Information System (INIS)

1993-09-01

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

Evaluation of a high-level waste radiological maintenance facility

International Nuclear Information System (INIS)

Collins, K.J.

1998-01-01

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

DWPF Recycle Evaporator Simulant Tests

International Nuclear Information System (INIS)

Stone, M

2005-01-01

aluminum oxide during the evaporation process. The following recommendations were made: Recycle from the DWTT should be metered in slowly to the ''typical'' recycle streams to avoid spikes in solids content to allow consistent processing and avoid process upsets. Additional studies should be conducted to determine acceptable volume ratios for the HEME dissolution and decontamination solutions in the evaporator feed. Dow Corning 2210 antifoam should be evaluated for use to control foaming. Additional tests are required to determine the concentration of antifoam required to prevent foaming during startup, the frequency of antifoam additions required to control foaming during steady state processing, and the ability of the antifoam to control foam over a range of potential feed compositions. This evaluation should also include evaluation of the degradation of the antifoam and impact on the silicon and TOC content of the condensate. The caustic HEME dissolution recycle stream should be neutralized to at least pH of 7 prior to blending with the acidic recycle streams. Dow Corning 2210 should be used during the evaporation testing using the radioactive recycle samples received from DWPF. Evaluation of additional antifoam candidates should be conducted as a backup for Dow Corning 2210. A camera and/or foam detection instrument should be included in the evaporator design to allow monitoring of the foaming behavior during operation. The potential for foam formation and high solids content should be considered during the design of the evaporator vessel

DWPF liquid sample station: Status of equipment development

International Nuclear Information System (INIS)

Caplan, J.R.

1987-01-01

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

SME Acceptability Determination For DWPF Process Control (U)

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-12

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

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

International Nuclear Information System (INIS)

Shanahan, K.L.

1992-02-01

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

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

International Nuclear Information System (INIS)

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

1984-01-01

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

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

International Nuclear Information System (INIS)

Edwards, T.B.

1995-01-01

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

DWPF PCCS version 2.0 test case

International Nuclear Information System (INIS)

Brown, K.G.; Pickett, M.A.

1992-01-01

To verify the operation of the Product Composition Control System (PCCS), a test case specific to DWPF operation was developed. The values and parameters necessary to demonstrate proper DWPF product composition control have been determined and are presented in this paper. If this control information (i.e., for transfers and analyses) is entered into the PCCS as illustrated in this paper, and the results obtained correspond to the independently-generated results, it can safely be said that the PCCS is operating correctly and can thus be used to control the DWPF. The independent results for this test case will be generated and enumerated in a future report. This test case was constructed along the lines of the normal DWPF operation. Many essential parameters are internal to the PCCS (e.g., property constraint and variance information) and can only be manipulated by personnel knowledgeable of the Symbolics reg-sign hardware and software. The validity of these parameters will rely on induction from observed PCCS results. Key process control values are entered into the PCCS as they would during normal operation. Examples of the screens used to input specific process control information are provided. These inputs should be entered into the PCCS database, and the results generated should be checked against the independent, computed results to confirm the validity of the PCCS

Spray nozzle pattern test for the DWPF HEME Task QA Plan

International Nuclear Information System (INIS)

Lee, L.

1991-01-01

The DWPF melter off-gas systems have two High Efficiency Mist Eliminators (HEME) upstream of the High-Efficiency Particulates Air filters (HEPA) to remove fine mists and particulates from the off-gas. To have an acceptable filter life and an efficient operation, an air atomized water is spray on the HEME. The water spray keeps the HEME wet and dissolves the soluble particulates and enhances and HEME efficiency. DWPF Technical asked SRL to determine the conditions which will give satisfactory atomization and distribution of water so that the HEME will operate efficiently. The purpose of this document is to identify, QA controls to be applied in the pursuit of this task (WSRC-RP-91-1151)

Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-29

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

Defining And Characterizing Sample Representativeness For DWPF Melter Feed Samples

International Nuclear Information System (INIS)

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

2013-01-01

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

Impact of Alkali Source on Vitrification of SRS High Level Waste

International Nuclear Information System (INIS)

LAMBERT, D. P.; MILLER, D. H.; PEELER, D. K.; SMITH, M. E.; STONE, M. E.

2005-01-01

The Defense Waste Processing Facility (DWPF) Savannah River Site is currently immobilizing high level nuclear waste sludge by vitrification in borosilicate glass. The processing strategy involves blending a large batch of sludge into a feed tank, washing the sludge to reduce the amount of soluble species, then processing the large ''sludge batch'' through the DWPF. Each sludge batch is tested by the Savannah River National Laboratory (SRNL) using simulants and tests with samples of the radioactive waste to ''qualify'' the batch prior to processing in the DWPF. The DWPF pretreats the sludge by first acidifying the sludge with nitric and formic acid. The ratio of nitric to formic acid is adjusted as required to target a final glass composition that is slightly reducing (the target is for ∼20% of the iron to have a valence of two in the glass). The formic acid reduces the mercury in the feed to elemental mercury which is steam stripped from the feed. After a concentration step, the glass former (glass frit) is added as a 50 wt% slurry and the batch is concentrated to approximately 50 wt% solids. The feed slurry is then fed to a joule heated melter maintained at 1150 C. The glass must meet both processing (e.g., viscosity and liquidus temperature) and product performance (e.g., durability) constraints The alkali content of the final waste glass is a critical parameter that affects key glass properties (such as durability) as well as the processing characteristics of the waste sludge during the pretreatment and vitrification processes. Increasing the alkali content of the glass has been shown to improve the production rate of the DWPF, but the total alkali in the final glass is limited by constraints on glass durability and viscosity. Two sources of alkali contribute to the final alkali content of the glass: sodium salts in the waste supernate and sodium and lithium oxides in the glass frit added during pretreatment processes. Sodium salts in the waste supernate can

Development of an ASTM standard glass durability test, the Product Consistency Test (PCT), for high level radioactive waste glass

International Nuclear Information System (INIS)

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

1994-01-01

The nation's first, and the world's largest, facility to immobilize high-level nuclear waste in durable borosilicate glass has started operation at the Savannah River Site (SRS) in Aiken, South Carolina. The product specifications on the glass wasteform produced in the Defense Waste Processing Facility (DWPF) required extensive characterization of the glass product before actual production began and for continued characterization during production. To aid in this characterization, a glass durability (leach) test was needed that was easily reproducible, could be performed remotely on highly radioactive samples, and could yield results rapidly. Several standard leach tests were examined with a variety of test configurations. Using existing tests as a starting point, the DWPF Product Consistency Test (PCT was developed in which crushed glass samples are exposed to 90 ± 2 degree C deionized water for seven days. Based on extensive testing, including a seven-laboratory round robin and confirmatory testing with radioactive samples, the PCT is very reproducible, yields reliable results rapidly, and can be performed in shielded cell facilities with radioactive samples

Bounding estimate of DWPF mercury emissions

International Nuclear Information System (INIS)

Jacobs, R.A.

1992-01-01

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

Materials Science of High-Level Nuclear Waste Immobilization

International Nuclear Information System (INIS)

Weber, William J.; Navrotsky, Alexandra; Stefanovsky, S. V.; Vance, E. R.; Vernaz, Etienne Y.

2009-01-01

With the increasing demand for the development of more nuclear power comes the responsibility to address the technical challenges of immobilizing high-level nuclear wastes in stable solid forms for interim storage or disposition in geologic repositories. The immobilization of high-level nuclear wastes has been an active area of research and development for over 50 years. Borosilicate glasses and complex ceramic composites have been developed to meet many technical challenges and current needs, although regulatory issues, which vary widely from country to country, have yet to be resolved. Cooperative international programs to develop advanced proliferation-resistant nuclear technologies to close the nuclear fuel cycle and increase the efficiency of nuclear energy production might create new separation waste streams that could demand new concepts and materials for nuclear waste immobilization. This article reviews the current state-of-the-art understanding regarding the materials science of glasses and ceramics for the immobilization of high-level nuclear waste and excess nuclear materials and discusses approaches to address new waste streams

Task technical plan: DWPF air permit/dispersion modeling

International Nuclear Information System (INIS)

Lambert, D.P.

1993-01-01

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

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

International Nuclear Information System (INIS)

Edwards, T.B.

2000-01-01

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

Process technology for vitrification of defense high-level waste at the Savannah River Plant

International Nuclear Information System (INIS)

Boersma, M.D.

1984-01-01

Vitrification in borosilicate glass is now the leading worldwide process for immobilizing high-level radioactive waste. Each vitrification project, however, has its unique mission and technical challenges. The Defense Waste Vitrification Facility (DWPF) now under construction at the Savannah River Plant will concentrate and vitrify a large amount of relatively low-power alkaline waste. Process research and development for the DWPF have produced significant advances in remote chemical operations, glass melting, off-gas treatment, slurry handling, decontamination, and welding. 6 references, 1 figure, 5 tables

Leaching TC-99 from DWPF glass in simulated geologic repository groundwaters

International Nuclear Information System (INIS)

Bibler, N.E.; Jurgensen, A.R.

1986-01-01

The purpose was to determine if DWPF glass in geologic groundwaters would immobilize Tc-99 as well as it does other elements. A previous study (using a borosilicate glass of a very different composition from DWPF glass) indicated that Tc-99 leached rapidly from the glass suggesting that glass may not be a good matrix for immobilizing Tc-99. It was suggested that the Tc-99 had migrated to vesicles in the glass while the glass was still molten. To determine if borosilicate glass was a good immobilizing matrix for Tc-99, this study was performed using DWPF glass. The leaching of Tc-99 was compared to other elements in the glass. It was shown that rapid leaching will not occur with SRP glass. The leach rate for Tc-99 was nearly identical to that for B, a matrix element in the glass. Another objective was to compare the release of Tc-99 under oxidizing and reducing conditions with other elements in the glass. In the tests described here, even though the glass was dissolving more under reducing conditions as a result of abnormally high pH values, less Tc-99 appeared in solution

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

New glass material oxidation and dissolution system facility: Direct conversion of surplus fissile materials, spent nuclear fuel, and other material to high-level-waste glass. Storage and disposition of weapons-usable fissile materials programmatic environmental impact statement data report: Predecisional draft

International Nuclear Information System (INIS)

Forsberg, C.W.; Elam, K.R.; Reich, W.J.

1995-01-01

With the end of the Cold War, countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. It has been recommended that these surplus fissile materials (SFMs) be processed so that they are no more accessible than plutonium in spent nuclear fuel (SNF). This SNF standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. This report provides for the PEIS the necessary input data on a new method for the disposition of SFMs: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptunium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal. The primary form of this SNF is Hanford-N SNF with preirradiation uranium enrichments between 0.95 and 1.08%. The final product is a plutonium, low-enriched-uranium, HLW, borosilicate glass for disposition in a geological repository. The proposed conversion process is the Glass Material Oxidation and Dissolution System (GMODS), which is a new process. The initial analysis of the GMODS process indicates that a MODS facility for this application would be similar in size and environmental impact to the Defense Waste Processing Facility (DWPF) at the Savannah River Site. Because of this, the detailed information available on DWPF was used as the basis for much of the GMODS input into the SFMs PEIS

Parametric testing of a DWPF glass

International Nuclear Information System (INIS)

Bazan, F.; Rego, J.

1985-03-01

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

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

International Nuclear Information System (INIS)

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

1996-10-01

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

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

Storage of High Level Nuclear Waste in Germany

Directory of Open Access Journals (Sweden)

Dietmar P. F. Möller

2007-01-01

Full Text Available Nuclear energy is very often used to generate electricity. But first the energy must be released from atoms what can be done in two ways: nuclear fusion and nuclear fission. Nuclear power plants use nuclear fission to produce electrical energy. The electrical energy generated in nuclear power plants does not produce polluting combustion gases but a renewable energy, an important fact that could play a key role helping to reduce global greenhouse gas emissions and tackling global warming especially as the electricity energy demand rises in the years ahead. This could be assumed as an ideal win-win situation, but the reverse site of the medal is that the production of high-level nuclear waste outweighs this advantage. Hence the paper attempt to highlight the possible state-of-art concepts for the safe and sustaining storage of high-level nuclear waste in Germany.

Glass formulation requirements for DWPF coupled operations using crystalline silicotitanates

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-06

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-30

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

Bounding estimate of DWPF mercury emissions

International Nuclear Information System (INIS)

Jacobs, R.A.

1993-01-01

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

DWPF waste glass Product Composition Control System

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Maximum total organic carbon limit for DWPF melter feed

International Nuclear Information System (INIS)

Choi, A.S.

1995-01-01

DWPF recently decided to control the potential flammability of melter off-gas by limiting the total carbon content in the melter feed and maintaining adequate conditions for combustion in the melter plenum. With this new strategy, all the LFL analyzers and associated interlocks and alarms were removed from both the primary and backup melter off-gas systems. Subsequently, D. Iverson of DWPF- T ampersand E requested that SRTC determine the maximum allowable total organic carbon (TOC) content in the melter feed which can be implemented as part of the Process Requirements for melter feed preparation (PR-S04). The maximum TOC limit thus determined in this study was about 24,000 ppm on an aqueous slurry basis. At the TOC levels below this, the peak concentration of combustible components in the quenched off-gas will not exceed 60 percent of the LFL during off-gas surges of magnitudes up to three times nominal, provided that the melter plenum temperature and the air purge rate to the BUFC are monitored and controlled above 650 degrees C and 220 lb/hr, respectively. Appropriate interlocks should discontinue the feeding when one or both of these conditions are not met. Both the magnitude and duration of an off-gas surge have a major impact on the maximum TOC limit, since they directly affect the melter plenum temperature and combustion. Although the data obtained during recent DWPF melter startup tests showed that the peak magnitude of a surge can be greater than three times nominal, the observed duration was considerably shorter, on the order of several seconds. The long surge duration assumed in this study has a greater impact on the plenum temperature than the peak magnitude, thus making the maximum TOC estimate conservative. Two models were used to make the necessary calculations to determine the TOC limit

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-09-25

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Freeze and restart of the DWPF Scale Glass Melter

International Nuclear Information System (INIS)

Choi, A.S.

1989-01-01

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

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

International Nuclear Information System (INIS)

Sullivan, I.K.; Bignell, D.

1994-01-01

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

Qualification of a Vitrified High Level Waste Product to Support Used Nuclear Fuel Recycling in the US

International Nuclear Information System (INIS)

Murray, P.; Bailly, F.; Strachan, D.; Senentz, G.; Veyer, C.

2009-01-01

As part of the Department of Energy (DOE) Global Nuclear Energy Partnership (GNEP), AREVA formed the International Nuclear Recycling Alliance (INRA) consisting of recognized world-leading companies in the area of used nuclear fuel (UNF) recycling,. The INRA team, consisting of AREVA, Mitsubishi Heavy Industries (MHI), Japan Nuclear Fuel Ltd (JNFL), Batelle Memorial Institute (BMI), URS Washington Division and Babcock and Wilcox (B and W), prepared a pre-conceptual design for an upgradable engineering-scale recycling plant with a nominal through put of 800 tHM/y. The pre-conceptual design of this leading-edge facility was based upon the extensive experience of the INRA team in recycling plant design and real world 'lessons learned' from actually building, commissioning, and operating recycling facilities in both France and Japan. The conceptual flowsheet, based upon the COEX TM separations process, separates the useful products for recycling into new fuel and sentences all the remaining fission products and minor actinides (MA) to the high level waste, (HLW) for vitrification. The proposed vitrified waste product will be similar to that currently produced in recycling plants in France. This wasteform has been qualified in France by conducting extensive studies and demonstrations. In the US, the qualification of vitrified glass products has been conducted by the US National Laboratories for the Defence Waste Processing Facility (DWPF), the West Valley Demonstration Plant (WVDP), and the Waste Treatment Plant (WTP). The vitrified waste product produced by recycling is sufficiently different from these current waste forms to warrant additional trials and studies. In this paper we review the differences in the vitrified waste forms previously qualified in the US with that produced from recycling of UNF in France. The lessons learned from qualifying a vitrified waste form in Europe is compared to the current US process for vitrified waste qualification including waste

Yield Stress Reduction of DWPF Melter Feed Slurries

International Nuclear Information System (INIS)

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

2007-01-01

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

Radioactive demonstration of DWPF product control strategy

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

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

1991-01-01

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

ISOLOK VALVE ACCEPTANCE TESTING FOR DWPF SME SAMPLING PROCESS

Energy Technology Data Exchange (ETDEWEB)

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

2011-12-05

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-17

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

Disposal of high level nuclear wastes: Thermodynamic equilibrium and environment ethics

Institute of Scientific and Technical Information of China (English)

RANA Mukhtar Ahmed

2009-01-01

Contamination of soil, water or air, due to a failure of containment or disposal of high level nuclear wastes, can potentially cause serious hazards to the environment or human health. Essential elements of the environment and radioactivity dangers to it are illustrated. Issues of high level nuclear waste disposal are discussed with a focus on thermodynamic equilibrium and environment ethics. Major aspects of the issues are analyzed and described briefly to build a perception of risks involved and ethical implications. Nuclear waste containment repository should be as close as possible to thermodynamic equilibrium. A clear demonstration about safety aspects of nuclear waste management is required in gaining public and political confidence in any possible scheme of permanent disposal. Disposal of high level nuclear waste offers a spectrum of environment connected challenges and a long term future of nuclear power depends on the environment friendly solution of the problem of nuclear wastes.

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

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

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

1994-06-01

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

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

International Nuclear Information System (INIS)

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

1992-06-01

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

Disposal of high level nuclear wastes: thermodynamic equilibrium and environment ethics

International Nuclear Information System (INIS)

Rana, M.A.

2009-01-01

Contamination of soil, water or air, due to a failure of containment or disposal of high level nuclear wastes, can potentially cause serious hazards to the environment or human health. Essential elements of the environment and radioactivity dangers to it are illustrated. Issues of high level nuclear waste disposal are discussed with a focus on thermodynamic equilibrium and environment ethics. Major aspects of the issues are analyzed and described briefly to build a perception of risks involved and ethical implications. Nuclear waste containment repository should be as close as possible to thermodynamic equilibrium. A clear demonstration about safety aspects of nuclear waste management is required in gaining public and political confidence in any possible scheme of permanent disposal. Disposal of high level nuclear waste offers a spectrum of environment connected challenges and a long term future of nuclear power depends on the environment friendly solution of the problem of nuclear wastes. (authors)

Crystallization In High Level Waste (HLW) Glass Melters: Operational Experience From The Savannah River Site

Energy Technology Data Exchange (ETDEWEB)

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

2014-02-27

processing strategy for the Hanford Tank Waste Treatment and Immobilization Plant (WTP). The basis of this alternative approach is an empirical model predicting the crystal accumulation in the WTP glass discharge riser and melter bottom as a function of glass composition, time, and temperature. When coupled with an associated operating limit (e.g., the maximum tolerable thickness of an accumulated layer of crystals), this model could then be integrated into the process control algorithms to formulate crystal tolerant high level waste (HLW) glasses targeting higher waste loadings while still meeting process related limits and melter lifetime expectancies. This report provides a review of the scaled melter testing that was completed in support of the Defense Waste Processing Facility (DWPF) melter. Testing with scaled melters provided the data to define the DWPF operating limits to avoid bulk (volume) crystallization in the un-agitated DWPF melter and provided the data to distinguish between spinels generated by K-3 refractory corrosion versus spinels that precipitated from the HLW glass melt pool. This report includes a review of the crystallization observed with the scaled melters and the full scale DWPF melters (DWPF Melter 1 and DWPF Melter 2). Examples of actual DWPF melter attainment with Melter 2 are given. The intent is to provide an overview of lessons learned, including some example data, that can be used to advance the development and implementation of an empirical model and operating limit for crystal accumulation for WTP. Operation of the first and second (current) DWPF melters has demonstrated that the strategy of using a liquidus temperature predictive model combined with a 100 °C offset from the normal melter operating temperature of 1150 °C (i.e., the predicted liquidus temperature (TL) of the glass must be 1050 °C or less) has been successful in preventing any detrimental accumulation of spinel in the DWPF melt pool, and spinel has not been

Integrated DWPF Melter System (IDMS) campaign report: The first two noble metals operations

International Nuclear Information System (INIS)

Hutson, N.D.; Zamecnik, J.R.; Smith, M.E.; Miller, D.H.; Ritter, J.A.

1991-01-01

The Integrated DWPF Melter System (IDMS) is designed and constructed to provide an engineering-scale representation of the DWPF melter and its associated feed preparation and off-gas systems. The facility is the first pilot-scale melter system capable of processing mercury, and flowsheet levels of halides and noble metals. In order to characterize the processing of noble metals (Pd, Rh, Ru, and Ag) on a large scale, the IDMS will be operated batchstyle for at least nine feed preparation cycles. The first two of these operations are complete. The major observation to date occurred during the second run when significant amounts of hydrogen were evolved during the feed preparation cycle. The runs were conducted between June 7, 1990 and March 8, 1991. This time period included nearly six months of ''fix-up'' time when forced air purges were installed on the SRAT MFT and other feed preparation vessels to allow continued noble metals experimentation

Materials performance in a high-level radioactive waste vitrification system

International Nuclear Information System (INIS)

Imrich, K.J.; Chandler, G.T.

1996-01-01

The Defense Waste Processing Facility (DWPF) is a Department of Energy Facility designed to vitrify highly radioactive waste. An extensive materials evaluation program has been completed on key components in the DWPF after twelve months of operation using nonradioactive simulated wastes. Results of the visual inspections of the feed preparation system indicate that the system components, which were fabricated from Hastelloy C-276, should achieve their design lives. Significant erosion was observed on agitator blades that process glass frit slurries; however, design modifications should mitigate the erosion. Visual inspections of the DWPF melter top head and off gas components, which were fabricated from Inconel 690, indicated that varying degrees of degradation occurred. Most of the components will perform satisfactorily for their two year design life. The components that suffered significant attack were the borescopes, primary film cooler brush, and feed tubes. Changes in the operation of the film cooler brush and design modifications to the feed tubes and borescopes is expected to extend their service lives to two years. A program to investigate new high temperature engineered materials and alloys with improved oxidation and high temperature corrosion resistance will be initiated

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-06-12

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

High-level nuclear waste disposal

International Nuclear Information System (INIS)

Burkholder, H.C.

1985-01-01

The meeting was timely because many countries had begun their site selection processes and their engineering designs were becoming well-defined. The technology of nuclear waste disposal was maturing, and the institutional issues arising from the implementation of that technology were being confronted. Accordingly, the program was structured to consider both the technical and institutional aspects of the subject. The meeting started with a review of the status of the disposal programs in eight countries and three international nuclear waste management organizations. These invited presentations allowed listeners to understand the similarities and differences among the various national approaches to solving this very international problem. Then seven invited presentations describing nuclear waste disposal from different perspectives were made. These included: legal and judicial, electric utility, state governor, ethical, and technical perspectives. These invited presentations uncovered several issues that may need to be resolved before high-level nuclear wastes can be emplaced in a geologic repository in the United States. Finally, there were sixty-six contributed technical presentations organized in ten sessions around six general topics: site characterization and selection, repository design and in-situ testing, package design and testing, disposal system performance, disposal and storage system cost, and disposal in the overall waste management system context. These contributed presentations provided listeners with the results of recent applied RandD in each of the subject areas

Solidification of Savannah River Plant high-level waste

International Nuclear Information System (INIS)

Maher, R.; Shafranek, L.F.; Stevens, W.R. III.

1983-01-01

The Department of Energy, in accord with recommendations from the Du Pont Company, has started construction of a Defense Waste Processing Facility (DWPF) at the Savannah River Plant. The facility should be completed by the end of 1988, and full-scale operation should begin in 1990. This facility will immobilize in borosilicate glass the large quantity of high-level radioactive waste now stored at the plant plus the waste to be generated from continued chemical reprocessing operations. The existing wastes at the Savannah River Plant will be completely converted by about 2010. 21 figures

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

International Nuclear Information System (INIS)

CHANG, ROBERT

2006-01-01

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

Managing the high level waste nuclear regulatory commission licensing process

International Nuclear Information System (INIS)

Baskin, K.P.

1992-01-01

This paper reports that the process for obtaining Nuclear Regulatory Commission permits for the high level waste storage facility is basically the same process commercial nuclear power plants followed to obtain construction permits and operating licenses for their facilities. Therefore, the experience from licensing commercial reactors can be applied to the high level waste facility. Proper management of the licensing process will be the key to the successful project. The management of the licensing process was categorized into four areas as follows: responsibility, organization, communication and documentation. Drawing on experience from nuclear power plant licensing and basic management principles, the management requirement for successfully accomplishing the project goals are discussed

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Answers to your questions on high-level nuclear waste

International Nuclear Information System (INIS)

1987-11-01

This booklet contains answers to frequently asked questions about high-level nuclear wastes. Written for the layperson, the document contains basic information on the hazards of radiation, the Nuclear Waste Management Program, the proposed geologic repository, the proposed monitored retrievable storage facility, risk assessment, and public participation in the program

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

International Nuclear Information System (INIS)

Cornman, W.R.

1981-12-01

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

IMPACT OF IRRADIATION AND THERMAL AGING ON DWPF SIMULATED SLUDGE PROPERTIES

International Nuclear Information System (INIS)

Eibling, R; Michael Stone, M

2006-01-01

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

A truck cask design for shipping defense high-level waste

International Nuclear Information System (INIS)

Madsen, M.M.; Zimmer, A.

1985-01-01

The Defense High-Level Waste (DHLW) cask is a Type B packaging currently under development by the U.S. Department of Energy (DOE). This truck cask has been designed to initially transport borosilicate glass waste from the Defense Waste Processing Facility (DWPF) to the Waste Isolation Pilot Plant (WIPP). Specific program activities include designing, testing, certifying, and fabricating a prototype legal-weight truck cask system. The design includes such state-of-the-art features as integral impact limiters and remote handling features. A replaceable shielding liner provides the flexibility for shipping a wide range of waste types and activity levels

Translating DWPF design criteria into an engineered facility design

International Nuclear Information System (INIS)

Kemp, J.B.

1986-01-01

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

Statistical analysis of the DWPF prototypic sampler

International Nuclear Information System (INIS)

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

1991-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-08

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

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

International Nuclear Information System (INIS)

Ginsberg, T.

1994-04-01

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

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

International Nuclear Information System (INIS)

Jantzen, C.; Laurinat, J.

2011-01-01

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

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

International Nuclear Information System (INIS)

Schumacher, R.F.

2001-01-01

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

US program for the immobilization of high-level nuclear wastes

International Nuclear Information System (INIS)

Crandall, J.L.

1979-01-01

A program has been developed for long-term management of high-level nuclear waste. The Savannah River Operations Office of the US Department of Energy is acting as the lead office for this program with technical advice from the E.I. du Pont de Nemours and Company. The purpose of the long-term program is to immobilize the DOE high-level waste in forms that act as highly efficient barriers against radionuclide release to the disposal site and to provide technology for similar treatment of commercial high-level waste in case reprocessing of commercial nuclear fuels is ever resumed. Descriptions of existing DOE and commercial wastes, program strategy, program expenditures, development of waste forms, evaluation and selection of waste forms, regulatory aspects of waste form selection, project schedules, and cost estimates for immobilization facilities are discussed

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-10-01

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

Impact of Spherical Frit Beads on Simulated DWPF Slurries

International Nuclear Information System (INIS)

SMITH, MICHAEL

2005-01-01

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

Estimation of Total Error in DWPF Reported Radionuclide Inventories

International Nuclear Information System (INIS)

Edwards, T.B.

1995-01-01

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

High-level nuclear waste disposal: Ethical considerations

International Nuclear Information System (INIS)

Maxey, M.N.

1985-01-01

Popular skepticism about, and moral objections to, recent legislation providing for the management and permanent disposal of high-level radioactive wastes have derived their credibility from two major sources: government procrastination in enacting waste disposal program, reinforcing public perceptions of their unprecedented danger and the inflated rhetoric and pretensions to professional omnicompetence of influential scientists with nuclear expertise. Ethical considerations not only can but must provide a mediating framework for the resolution of such a polarized political controversy. Implicit in moral objections to proposals for permanent nuclear waste disposal are concerns about three ethical principles: fairness to individuals, equitable protection among diverse social groups, and informed consent through due process and participation

Immobilisation of high level nuclear reactor wastes in SYNROC

Energy Technology Data Exchange (ETDEWEB)

Ringwood, A E; Kesson, S E; Ware, N G; Hibberson, W; Major, A [Australian National Univ., Canberra. Inst. of Advanced Studies

1979-03-15

It is stated that the elements occurring in high-level nuclear reactor wastes can be safely immobilised by incorporating them within the crystal lattices of the constituent minerals of a synthetic rock (SYNROC). The preferred form of SYNROC can accept up to 20% of high level waste calcine to form dilute solid solutions. The constituent minerals, or close structural analogues, have survived in a wide range of geochemical environments for periods of 20 to 2,000 Myr whilst immobilising the same elements present in nuclear wastes. SYNROC is unaffected by leaching for 24 hours in pure water or 10 wt % NaCl solution at high temperatures and pressure whereas borosilicate glasses completely decompose in a few hours in much less severe hydrothermal conditions. The combination of these leaching results with the geological evidence of long-term stability indicates that SYNROC would be vastly superior to glass in its capacity to safely immobilise nuclear wastes, when buried in a suitable geological repository. A dense, compact, mechanically strong form of SYNROC suitable for geological disposal can be produced by a process as economical as that which incorporates radioactive waste in borosilicate glasses.

Hydrogen generation during treatment of simulated high-level radioactive waste with formic acid

International Nuclear Information System (INIS)

Ritter, J.A.; Zamecnik, J.R.; Hsu, C.W.

1992-01-01

The Integrated Defense Waste Processing Facility (DWPF) Melter System (IDMS), operated by the Savannah River Laboratory, is a one-fifth scale pilot facility used in support of the start-up and operation of the Department of Energy's DWPF. Five IDMS runs determined the effect of the presence of noble metals in HLW sludge on the H 2 generation rate during the preparation of melter feed with formic acid. Overall, the results clearly showed that H 2 generation in the DWPF SRAT could, at times, exceed the lower flammable limit of H 2 in air (4 vol%), depending on such factors as offgas generation and air inleakage of the DWPF vessels. Therefore, the installation of a forced air purge system and H 2 monitors were recommended to the DWPF to control the generation of H 2 during melter feed preparation by fuel dilution

Management of the high-level nuclear power facilities

International Nuclear Information System (INIS)

Preda, Marin

2003-05-01

This thesis approaches current issues in the management of the high power nuclear facilities and as such it appears to be important particularly for nuclear power plant operation topics. Of special interest are the failure events entailing possible catastrophic situations. The contents is structured onto ten chapters. The first chapter describes the operation regimes of the nuclear high power facilities. Highlighted here are the thesis scope and the original features of the work. The second chapter deals with operational policies developed in order to ensure the preventive maintenance of the nuclear installations. Also managing structures are described devoted to practical warranting the equipment safety function of non-classical power stations. In the third chapter cases of nuclear accidents are analyzed especially stressing the probabilistic risk and the operation regimes having in view the elimination of catastrophic events. In the fourth and fifth chapters the control of nuclear radiation emission is treated focusing the quality issue of nuclear installations required to avoid hazardous effects at level of nuclear reactor operation stage. At the same time set of operational measures is given here for preventing risks, catastrophes and chaotic situations. The chapter five presents both theoretical and practical approaches of the nuclear reactor core management concerning particularly the fuel testing, the water primary system and the quality of the involved equipment. In the sixth and seventh chapters issues of risk-quality correlations are approached as well as the structure of expert systems for monitoring the operational regimes of nuclear facilities. The efficiency of the power systems with nuclear injection is discussed and some original ideas developed in this work are evidenced in the eighth and ninth chapters. Presented are here both the operational principles and models of raising the efficiency of the interconnected nuclear stations and prices' policy

Remotely replaceable jumpers and embedded wiring for the DWPF

International Nuclear Information System (INIS)

Heckendorn, F.M. II.

1984-01-01

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

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Lot No. 1 of Frit 202 for DWPF cold runs

International Nuclear Information System (INIS)

Schumacher, R.F.

1993-01-01

This report was prepared at the end of 1992 and summarizes the evaluation of the first lot sample of DWPF Frit 202 from Cataphote Inc. Publication of this report was delayed until the results from the carbon analyses could be included. To avoid confusion the frit specifications presented in this report were those available at the end of 1992. The specifications were slightly modified early in 1993. The frit was received and evaluated for moisture, particle size distribution, organic-inorganic carbon and chemical composition. Moisture content and particle size distribution were determined on a representative sample at SRTC. These properties were within the DWPF specifications for Frit 202. A representative sample was submitted to Corning Engineering Laboratory Services for chemical analyses. The sample was split and two dissolutions prepared. Each dissolution was analyzed on two separate days. The results indicate that there is a high probability (>95%) that the silica content of this frit is below the specification limit of 77.0 ± 1.0 wt %. The average of the four analyzed values was 75.1 wt % with a standard deviation of 0.28 wt %. All other oxides were within the elliptical two sigma limits. Control standard frit samples were submitted and analyzed at the same time and the results were very similar to previous analyses of these materials

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

International Nuclear Information System (INIS)

Edwards, Thomas

2005-01-01

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

Nuclear level density

International Nuclear Information System (INIS)

Cardoso Junior, J.L.

1982-10-01

Experimental data show that the number of nuclear states increases rapidly with increasing excitation energy. The properties of highly excited nuclei are important for many nuclear reactions, mainly those that go via processes of the compound nucleus type. In this case, it is sufficient to know the statistical properties of the nuclear levels. First of them is the function of nuclear levels density. Several theoretical models which describe the level density are presented. The statistical mechanics and a quantum mechanics formalisms as well as semi-empirical results are analysed and discussed. (Author) [pt

76 FR 35137 - Vulnerability and Threat Information for Facilities Storing Spent Nuclear Fuel and High-Level...

Science.gov (United States)

2011-06-16

... High-Level Radioactive Waste AGENCY: U.S. Nuclear Regulatory Commission. ACTION: Public meeting... Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C Waste,'' and 73... Spent Nuclear Fuel (SNF) and High-Level Radioactive Waste (HLW) storage facilities. The draft regulatory...

DWPF Melter Off-Gas Flammability Assessment for Sludge Batch 9

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-11

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

Final disposal of high levels waste and spent nuclear fuel

International Nuclear Information System (INIS)

Gelin, R.

1984-05-01

Foreign and international activities on the final disposal of high-level waste and spent nuclear fuel have been reviewed. A considerable research effort is devoted to development of acceptable disposal options. The different technical concepts presently under study are described in the report. Numerous studies have been made in many countries of the potential risks to future generations from radioactive wastes in underground disposal repositories. In the report the safety assessment studies and existing performance criteria for geological disposal are briefly discussed. The studies that are being made in Canada, the United States, France and Switzerland are the most interesting for Sweden as these countries also are considering disposal into crystalline rocks. The overall time-tables in different countries for realisation of the final disposal are rather similar. Normally actual large-scale disposal operations for high-level wastes are not foreseen until after year 2000. In the United States the Congress recently passed the important Nuclear Waste Policy Act. It gives a rather firm timetable for site-selection and construction of nuclear waste disposal facilities. According to this act the first repository for disposal of commercial high-level waste must be in operation not later than in January 1998. (Author)

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

Status of the high-level nuclear waste disposal program in Japan

International Nuclear Information System (INIS)

Uematsu, K.

1985-01-01

The Japan Atomic Energy Commission (JAEC) initiated a high-level radioactive waste disposal program in 1976. Since then, the Advisory Committee on Radioactive Waste Management of JAEC has revised the program twice. The latest revision was issued in 1984. The committee recommended a four-phase program and the last phase calls for the beginning of emplacement of the high-level nuclear waste into a selected repository in the Year 2000. The first phase is already completed, and the second phase of this decade calls for the selection of a candidate disposal site and the conducting of the RandD of waste disposal in an underground research laboratory and in a hot test facility. This paper covers the current status of the high-level nuclear waste disposal program in Japan

What are Spent Nuclear Fuel and High-Level Radioactive Waste?

International Nuclear Information System (INIS)

2002-01-01

Spent nuclear fuel and high-level radioactive waste are materials from nuclear power plants and government defense programs. These materials contain highly radioactive elements, such as cesium, strontium, technetium, and neptunium. Some of these elements will remain radioactive for a few years, while others will be radioactive for millions of years. Exposure to such radioactive materials can cause human health problems. Scientists worldwide agree that the safest way to manage these materials is to dispose of them deep underground in what is called a geologic repository

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-04-26

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

Development Of Remote Hanford Connector Gasket Replacement Tooling For DWPF

International Nuclear Information System (INIS)

Krementz, D.; Coughlin, Jeffrey

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-09-26

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

High Level Waste (HLW) Processing Experience with Increased Waste Loading

International Nuclear Information System (INIS)

JANTZEN, CAROL

2004-01-01

The Defense Waste Processing Facility (DWPF) Engineering requested characterization of glass samples that were taken after the second melter had been operational for about 5 months. After the new melter had been installed, the waste loading had been increased to about 38 weight percentage after a new quasicrystalline liquidus model had been implemented. The DWPF had also switched from processing with refractory Frit 200 to a more fluid Frit 320. The samples were taken after DWPF observed very rapid buildup of deposits in the upper pour spout bore and on the pour spout insert while processing the high waste loading feedstock. These samples were evaluated using various analytical techniques to determine the cause of the crystallization. The pour stream sample was homogeneous, amorphous, and representative of the feed batch from which it was derived. Chemical analysis of the pour stream sample indicated that a waste loading of 38.5 weight per cent had been achieved. The data analysis indicated that surface crystallization, induced by temperature and oxygen fugacity gradients in the pour spout, caused surface crystallization to occur in the spout and on the insert at the higher waste loadings even though there was no crystallization in the pour stream

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

International Nuclear Information System (INIS)

Edwards, T. B.

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-08

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

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

International Nuclear Information System (INIS)

Plodinec, M.J.

1993-01-01

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

DWPF integrated cold runs revised technical bases for precipitate hydrolysis

International Nuclear Information System (INIS)

Landon, L.F.

1992-01-01

The report defines new precipitate hydrolysis process operating parameters for DWPF Chemical runs assuming the precipitate feed simulants to be processed reflect the decision to implement a final wash of the tetraphenylborate slurry before transfer to DWPF (i.e. the Late Wash Facility). Control of the nitrite content of the tetraphenylborate slurry to 0.01M or less has eliminated the need for hydroxylamine nitrate (HAN) during hydrolysis. Consequently, the oxidant nitrous oxide will not be generated. However, nitric oxide (NO) is expected to be generated (reaction of formic acid with nitrite) and some fraction of the NO can be expected to be oxidized to nitrogen dioxide. The rate of NO generation with low nitrite feed has not been quantified at this time nor is the extent to which the NO is oxidized to NO 2 known. A mass spectrometer is being installed in the Precipitate Hydrolysis Experimental Facility (PHEF) which will enable the NO generation rate to be defined as well as the extent to which the NO is oxidized to NO 2 . There is some undocumented data available for C 6 H 6 /NO and C 6 H 6 /NO 2 with N 2 as the diluent but no similar data for CO 2 . Development of test data in the required time frame is not possible. However, MOC's will be estimated for benzene/NO/NO 2 /CO 2 gas mixtures (the MOC is expected to be approximately 60% less than for the HAN process). Once these data are obtained, and NO/NO 2 concentration profiles are obtained from PHEF hydrolysis process demonstrations, a flammability control strategy for the DWPF Salt Processing Cell will be developed. Implementation of the HAN process purge strategy upon startup of the SPC with the late wash process would be conservative

DWPF GLASS BEADS AND GLASS FRIT TRANSPORT DEMONSTRATION

Energy Technology Data Exchange (ETDEWEB)

Adamson, D; Bradley Pickenheim, B

2008-11-24

DWPF is considering replacing irregularly shaped glass frit with spherical glass beads in the Slurry Mix Evaporator (SME) process to decrease the yield stress of the melter feed (a non-Newtonian Bingham Plastic). Pilot-scale testing was conducted on spherical glass beads and glass frit to determine how well the glass beads would transfer when compared to the glass frit. Process Engineering Development designed and constructed the test apparatus to aid in the understanding and impacts that spherical glass beads may have on the existing DWPF Frit Transfer System. Testing was conducted to determine if the lines would plug with the glass beads and the glass frit slurry and what is required to unplug the lines. The flow loop consisted of vertical and horizontal runs of clear PVC piping, similar in geometry to the existing system. Two different batches of glass slurry were tested: a batch of 50 wt% spherical glass beads and a batch of 50 wt% glass frit in process water. No chemicals such as formic acid was used in slurry, only water and glass formers. The glass beads used for this testing were commercially available borosilicate glass of mesh size -100+200. The glass frit was Frit 418 obtained from DWPF and is nominally -45+200 mesh. The spherical glass beads did not have a negative impact on the frit transfer system. The transferring of the spherical glass beads was much easier than the glass frit. It was difficult to create a plug with glass bead slurry in the pilot transfer system. When a small plug occurred from setting overnight with the spherical glass beads, the plug was easy to displace using only the pump. In the case of creating a man made plug in a vertical line, by filling the line with spherical glass beads and allowing the slurry to settle for days, the plug was easy to remove by using flush water. The glass frit proved to be much more difficult to transfer when compared to the spherical glass beads. The glass frit impacted the transfer system to the point

DWPF GLASS BEADS AND GLASS FRIT TRANSPORT DEMONSTRATION

International Nuclear Information System (INIS)

Adamson, D.; Pickenheim, Bradley

2008-01-01

DWPF is considering replacing irregularly shaped glass frit with spherical glass beads in the Slurry Mix Evaporator (SME) process to decrease the yield stress of the melter feed (a non-Newtonian Bingham Plastic). Pilot-scale testing was conducted on spherical glass beads and glass frit to determine how well the glass beads would transfer when compared to the glass frit. Process Engineering Development designed and constructed the test apparatus to aid in the understanding and impacts that spherical glass beads may have on the existing DWPF Frit Transfer System. Testing was conducted to determine if the lines would plug with the glass beads and the glass frit slurry and what is required to unplug the lines. The flow loop consisted of vertical and horizontal runs of clear PVC piping, similar in geometry to the existing system. Two different batches of glass slurry were tested: a batch of 50 wt% spherical glass beads and a batch of 50 wt% glass frit in process water. No chemicals such as formic acid was used in slurry, only water and glass formers. The glass beads used for this testing were commercially available borosilicate glass of mesh size -100+200. The glass frit was Frit 418 obtained from DWPF and is nominally -45+200 mesh. The spherical glass beads did not have a negative impact on the frit transfer system. The transferring of the spherical glass beads was much easier than the glass frit. It was difficult to create a plug with glass bead slurry in the pilot transfer system. When a small plug occurred from setting overnight with the spherical glass beads, the plug was easy to displace using only the pump. In the case of creating a man made plug in a vertical line, by filling the line with spherical glass beads and allowing the slurry to settle for days, the plug was easy to remove by using flush water. The glass frit proved to be much more difficult to transfer when compared to the spherical glass beads. The glass frit impacted the transfer system to the point

Maximum total organic carbon limits at different DWPF melter feed maters (U)

International Nuclear Information System (INIS)

Choi, A.S.

1996-01-01

The document presents information on the maximum total organic carbon (TOC) limits that are allowable in the DWPF melter feed without forming a potentially flammable vapor in the off-gas system were determined at feed rates varying from 0.7 to 1.5 GPM. At the maximum TOC levels predicted, the peak concentration of combustible gases in the quenched off-gas will not exceed 60 percent of the lower flammable limit during a 3X off-gas surge, provided that the indicated melter vapor space temperature and the total air supply to the melter are maintained. All the necessary calculations for this study were made using the 4-stage cold cap model and the melter off-gas dynamics model. A high-degree of conservatism was included in the calculational bases and assumptions. As a result, the proposed correlations are believed to by conservative enough to be used for the melter off-gas flammability control purposes

Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Boatner, L.A.; Sales, B.C.

1989-01-01

This patent describes lead-iron phosphate glasses containing a high level of Fe 2 O 3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90 0 C, with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10 2 to 10 3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe 2 O 3 in forming the lead-iron phosphate glass is critical. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Formation rate of ammonium nitrate in the off-gas line of SRAT and SME in DWPF

International Nuclear Information System (INIS)

Lee, L.

1992-01-01

A mathematical model for the formation rate of ammonium nitrate in the off-gas line of the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mixed Evaporator (SME) in DWPF has been developed. The formation rate of ammonium nitrate in the off-gas line depends on pH, temperature, volume and total concentration of ammonia and ammonium ion. Based on a typical SRAT and SME cycle in DWPF, this model predicts the SRAT contributes about 50 lbs of ammonium nitrate while SME contributes about 60 lbs of ammonium nitrate to the off-gas line

Lead-iron phosphate glass as a containment medium for the disposal of high-level nuclear wastes

Science.gov (United States)

Boatner, L.A.; Sales, B.C.

1984-04-11

Disclosed are lead-iron phosphate glasses containing a high level of Fe/sub 2/O/sub 3/ for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste

Characterization of high level nuclear waste glass samples following extended melter idling

Energy Technology Data Exchange (ETDEWEB)

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

2015-06-16

The Savannah River Site Defense Waste Processing Facility (DWPF) melter was recently idled with glass remaining in the melt pool and riser for approximately three months. This situation presented a unique opportunity to collect and analyze glass samples since outages of this duration are uncommon. The objective of this study was to obtain insight into the potential for crystal formation in the glass resulting from an extended idling period. The results will be used to support development of a crystal-tolerant approach for operation of the high-level waste melter at the Hanford Tank Waste Treatment and Immobilization Plant (WTP). Two glass pour stream samples were collected from DWPF when the melter was restarted after idling for three months. The samples did not contain crystallization that was detectible by X-ray diffraction. Electron microscopy identified occasional spinel and noble metal crystals of no practical significance. Occasional platinum particles were observed by microscopy as an artifact of the sample collection method. Reduction/oxidation measurements showed that the pour stream glasses were fully oxidized, which was expected after the extended idling period. Chemical analysis of the pour stream glasses revealed slight differences in the concentrations of some oxides relative to analyses of the melter feed composition prior to the idling period. While these differences may be within the analytical error of the laboratories, the trends indicate that there may have been some amount of volatility associated with some of the glass components, and that there may have been interaction of the glass with the refractory components of the melter. These changes in composition, although small, can be attributed to the idling of the melter for an extended period. The changes in glass composition resulted in a 70-100 °C increase in the predicted spinel liquidus temperature (TL) for the pour stream glass samples relative to the analysis of the melter feed prior to

Future health physics prospects in high-level nuclear waste management

International Nuclear Information System (INIS)

Waite, D.A.; Mayberry, J.J.

1986-01-01

The objective of this presentation is to provide an overview of health physics activities anticipated to be required at a high-level nuclear waste repository and to project the numbers of health physics personnel expected to be required to carry out these activities. Health physics personnel receiving consideration in the projections include the health physics manager, shift supervisors, area supervisors, health physicists, and technologists. Phases of the repository addressed are construction, operation, retrieval, and decommissioning. Specific topics discussed in the process of developing the projections are: (a) the basic features of a geologic repository, (b) the staffing requirements of such a repository, (c) health physics involvement in repository operations, and (d) the anticipated schedule for operation of repositories in the United States. A quantitative assessment of future health physics prospects in high-level nuclear waste management is included

PLUTONIUM SOLUBILITY IN HIGH-LEVEL WASTE ALKALI BOROSILICATE GLASS

Energy Technology Data Exchange (ETDEWEB)

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

2011-01-04

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

Plutonium Solubility In High-Level Waste Alkali Borosilicate Glass

International Nuclear Information System (INIS)

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

2011-01-01

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

Examination Of Sulfur Measurements In DWPF Sludge Slurry And SRAT Product Materials

International Nuclear Information System (INIS)

Bannochie, C. J.; Wiedenman, B. J.

2012-01-01

Savannah River National Laboratory (SRNL) was asked to re-sample the received SB7b WAPS material for wt. % solids, perform an aqua regia digestion and analyze the digested material by inductively coupled plasma - atomic emission spectroscopy (ICP-AES), as well as re-examine the supernate by ICP-AES. The new analyses were requested in order to provide confidence that the initial analytical subsample was representative of the Tank 40 sample received and to replicate the S results obtained on the initial subsample collected. The ICP-AES analyses for S were examined with both axial and radial detection of the sulfur ICP-AES spectroscopic emission lines to ascertain if there was any significant difference in the reported results. The outcome of this second subsample of the Tank 40 WAPS material is the first subject of this report. After examination of the data from the new subsample of the SB7b WAPS material, a team of DWPF and SRNL staff looked for ways to address the question of whether there was in fact insoluble S that was not being accounted for by ion chromatography (IC) analysis. The question of how much S is reaching the melter was thought best addressed by examining a DWPF Slurry Mix Evaporator (SME) Product sample, but the significant dilution of sludge material, containing the S species in question, that results from frit addition was believed to add additional uncertainty to the S analysis of SME Product material. At the time of these discussions it was believed that all S present in a Sludge Receipt and Adjustment Tank (SRAT) Receipt sample would be converted to sulfate during the course of the SRAT cycle. A SRAT Product sample would not have the S dilution effect resulting from frit addition, and hence, it was decided that a DWPF SRAT Product sample would be obtained and submitted to SRNL for digestion and sample preparation followed by a round-robin analysis of the prepared samples by the DWPF Laboratory, F/H Laboratories, and SRNL for S and sulfate. The

An instrumentation and control philosophy for high-level nuclear waste processing facilities

International Nuclear Information System (INIS)

Weigle, D.H.

1990-01-01

The purpose of this paper is to present an instrumentation and control philosophy which may be applied to high-level nuclear waste processing facilities. This philosophy describes the recommended criteria for automatic/manual control, remote/local control, remote/local display, diagnostic instrumentation, interlocks, alarm levels, and redundancy. Due to the hazardous nature of the process constituents of a high-level nuclear waste processing facility, it is imperative that safety and control features required for accident-free operation and maintenance be incorporated. A well-instrumented and controlled process, while initially more expensive in capital and design costs, is generally safer and less expensive to operate. When the long term cost savings of a well designed process is coupled with the high savings enjoyed by accident avoidance, the benefits far outweigh the initial capital and design costs

Statistical process control support during Defense Waste Processing Facility chemical runs

International Nuclear Information System (INIS)

Brown, K.G.

1994-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Crawford, C.; Bannochie, C.

2014-05-12

, XRD and SEM) in support of the Salt IPT chemistry team. The overall conclusions from analyses performed in this study are that the PRFT slurry consists of 0.61 Wt.% insoluble MST solids suspended in a 0.77 M [Na+] caustic solution containing various anions such as nitrate, nitrite, sulfate, carbonate and oxalate. The corresponding measured sulfur level in the PRFT slurry, a critical element for determining how much of the PRFT slurry gets blended into the SRAT, is 0.437 Wt.% TS. The PRFT slurry does not contain insoluble oxalates nor significant quantities of high activity sludge solids. The lack of sludge solids has been alluded to by the Salt IPT chemistry team in citing that the mixing pump has been removed from Tank 49H, the feed tank to ARP-MCU, thus allowing the sludge solids to settle out. The PRFT aqueous slurry from DWPF was found to contain 5.96 Wt.% total dried solids. Of these total dried solids, relatively low levels of insoluble solids (0.61 Wt.%) were measured. The densities of both the filtrate and slurry were 1.05 g/mL. Particle size distribution of the PRFT solids in filtered caustic simulant and XRD analysis of washed/dried PRFT solids indicate that the PRFT slurry contains a bimodal distribution of particles in the range of 1 and 6 μm and that the particles contain sodium titanium oxide hydroxide Na2Ti2O4(OH)2 crystalline material as determined by XRD. These data are in excellent agreement with similar data obtained from laboratory sampling of vendor supplied MST. Scanning Electron Microscopy (SEM) combined with Energy Dispersive X-ray Spectroscopy (EDS) analysis of washed/dried PRFT solids shows the particles to be like previous MST analyses consisting of irregular shaped micron-sized solids consisting primarily of Na and Ti. Thermogravimetric analysis of the washed and unwashed PRFT solids shows that the washed solids are very similar to MST solids. The TGA mass loss signal for the unwashed solids shows similar features to TGA performed on

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

International Nuclear Information System (INIS)

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

1983-01-01

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

Status of commercial nuclear high-level waste disposal. Special report

International Nuclear Information System (INIS)

Dau, G.J.; Williams, R.F.

1976-09-01

The results of this review, presented in the form of a functional description of high level waste management system, shows that technology is available to dispose of nuclear waste safely by several different processes. The most attractive alternative in terms of available technology and shortness of time to demonstrate it at commercial scale is a system that converts the waste to a solid by immobilizing the radioactive elements in a glass matrix. Brief comments are also given on international efforts in high level waste management and advanced disposal concepts

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

International Nuclear Information System (INIS)

Holtzscheiter, E.W.

1994-01-01

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

Control of high-level radioactive waste-glass melters

International Nuclear Information System (INIS)

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

1990-01-01

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

Reduction and resource recycling of high-level radioactive wastes through nuclear transmutation with PHITS code

International Nuclear Information System (INIS)

Fujita, Reiko

2017-01-01

In the ImPACT program of the Cabinet Office, programs are underway to reduce long-lived fission products (LLFP) contained in high-level radioactive waste through nuclear transmutation, or to recycle/utilize useful nuclear species. This paper outlines this program and describes recent achievements. This program consists of five projects: (1) separation/recovery technology, (2) acquisition of nuclear transmutation data, (3) nuclear reaction theory model and simulation, (4) novel nuclear reaction control and development of elemental technology, and (5) discussions on process concept. The project (1) develops a technology for dissolving vitrified solid, a technology for recovering LLFP from high-level waste liquid, and a technology for separating odd and even lasers. Project (2) acquires the new nuclear reaction data of Pd-107, Zr-93, Se-79, and Cs-135 using RIKEN's RIBF or JAEA's J-PARC. Project (3) improves new nuclear reaction theory and structural model using the nuclear reaction data measured in (2), improves/upgrades nuclear reaction simulation code PHITS, and proposes a promising nuclear transmutation pathway. Project (4) develops an accelerator that realizes the proposed transmutation route and its elemental technology. Project (5) performs the conceptual design of the process to realize (1) to (4), and constructs the scenario of reducing/utilizing high-level radioactive waste to realize this design. (A.O.)

Lead-iron phosphate glass: a stable storage medium for high-level nuclear waste

International Nuclear Information System (INIS)

Sales, B.C.; Boatner, L.A.

1984-01-01

Results are presented which show that lead-iron phosphate glasses are a promising new waste form for the safe immobilization of both high-level defense and high-level commercial radioactive waste. Relative to the borosilicate nuclear waste glasses that are currently the ''reference'' waste form for the long-term disposal of nuclear waste, lead-iron phosphate glasses have several distinct advantages: (1) an aqueous corrosion rate that is about 1000 times lower, (2) a processing temperature that is 100 0 to 250 0 C lower and, (3) a much lower melt viscosity in the temperature range from 800 0 to 1000 0 C. Most significantly, the lead-iron phosphate waste form can be processed using a technology similar to that developed for borosilicate nuclear waste glasses

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

International Nuclear Information System (INIS)

Schumacker, R.F.

1992-01-01

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

Lead iron phosphate glass as a containment medium for disposal of high-level nuclear waste

Science.gov (United States)

Boatner, Lynn A.; Sales, Brian C.

1989-01-01

Lead-iron phosphate glasses containing a high level of Fe.sub.2 O.sub.3 for use as a storage medium for high-level radioactive nuclear waste. By combining lead-iron phosphate glass with various types of simulated high-level nuclear waste, a highly corrosion resistant, homogeneous, easily processed glass can be formed. For corroding solutions at 90.degree. C., with solution pH values in the range between 5 and 9, the corrosion rate of the lead-iron phosphate nuclear waste glass is at least 10.sup.2 to 10.sup.3 times lower than the corrosion rate of a comparable borosilicate nuclear waste glass. The presence of Fe.sub.2 O.sub.3 in forming the lead-iron phosphate glass is critical. Lead-iron phosphate nuclear waste glass can be prepared at temperatures as low as 800.degree. C., since they exhibit very low melt viscosities in the 800.degree. to 1050.degree. C. temperature range. These waste-loaded glasses do not readily devitrify at temperatures as high as 550.degree. C. and are not adversely affected by large doses of gamma radiation in H.sub.2 O at 135.degree. C. The lead-iron phosphate waste glasses can be prepared with minimal modification of the technology developed for processing borosilicate glass nuclear wasteforms.

Geology of high-level nuclear waste disposal

International Nuclear Information System (INIS)

Roxburgh, I.S.

1988-01-01

The concept of geological disposal is set out by describing the major rock types in terms of their ability to isolate high-level nuclear waste. The advantages and problems posed by particular rock formations are explored and the design and construction of geological repositories is considered, along with the methods used to estimate their safety. It gives special consideration to the use of sea-covered rock and sediment as well as the on-land situation. Throughout the book the various principles and problems inherent in geological disposal are explained and illustrated by reference to a multitude of European and North American case studies, backed up by a large number of tables, figures and an extensive bibliography

Robotics and remote handling concepts for disposal of high-level nuclear waste

International Nuclear Information System (INIS)

McAffee, Douglas; Raczka, Norman; Schwartztrauber, Keith

1997-01-01

This paper summarizes preliminary remote handling and robotic concepts being developed as part of the US Department of Energy's (DOE) Yucca Mountain Project. The DOE is currently evaluating the Yucca Mountain Nevada site for suitability as a possible underground geologic repository for the disposal of high level nuclear waste. The current advanced conceptual design calls for the disposal of more than 12,000 high level nuclear waste packages within a 225 km underground network of tunnels and emplacement drifts. Many of the waste packages may weigh as much as 66 tonnes and measure 1.8 m in diameter and 5.6 m long. The waste packages will emit significant levels of radiation and heat. Therefore, remote handling is a cornerstone of the repository design and operating concepts. This paper discusses potential applications areas for robotics and remote handling technologies within the subsurface repository. It also summarizes the findings of a preliminary technology survey which reviewed available robotic and remote handling technologies developed within the nuclear, mining, rail and industrial robotics and automation industries, and at national laboratories, universities, and related research institutions and government agencies

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

International Nuclear Information System (INIS)

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

1991-01-01

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

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

Organics Characterization Of DWPF Alternative Reductant Simulants, Glycolic Acid, And Antifoam 747

International Nuclear Information System (INIS)

White, T. L.; Wiedenman, B. J.; Lambert, D. P.; Crump, S. L.; Fondeur, F. F.; Papathanassiu, A. E.; Kot, W. K.; Pegg, I. L.

2013-01-01

The present study examines the fate of glycolic acid and other organics added in the Chemical Processing Cell (CPC) of the Defense Waste Processing Facility (DWPF) as part of the glycolic alternate flowsheet. Adoption of this flowsheet is expected to provide certain benefits in terms of a reduction in the processing time, a decrease in hydrogen generation, simplification of chemical storage and handling issues, and an improvement in the processing characteristics of the waste stream including an increase in the amount of nitrate allowed in the CPC process. Understanding the fate of organics in this flowsheet is imperative because tank farm waste processed in the CPC is eventually immobilized by vitrification; thus, the type and amount of organics present in the melter feed may affect optimal melt processing and the quality of the final glass product as well as alter flammability calculations on the DWPF melter off gas. To evaluate the fate of the organic compounds added as the part of the glycolic flowsheet, mainly glycolic acid and antifoam 747, samples of simulated waste that was processed using the DWPF CPC protocol for tank farm sludge feed were generated and analyzed for organic compounds using a variety of analytical techniques at the Savannah River National Laboratory (SRNL). These techniques included Ion Chromatography (IC), Gas Chromatography-Mass Spectrometry (GC-MS), Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES), and Nuclear Magnetic Resonance (NMR) Spectroscopy. A set of samples were also sent to the Catholic University of America Vitreous State Laboratory (VSL) for analysis by NMR Spectroscopy at the University of Maryland, College Park. Analytical methods developed and executed at SRNL collectively showed that glycolic acid was the most prevalent organic compound in the supernatants of Slurry Mix Evaporator (SME) products examined. Furthermore, the studies suggested that commercially available glycolic acid contained minor amounts

Organics Characterization Of DWPF Alternative Reductant Simulants, Glycolic Acid, And Antifoam 747

Energy Technology Data Exchange (ETDEWEB)

White, T. L. [Savannah River Site (SRS), Aiken, SC (United States); Wiedenman, B. J. [Savannah River Site (SRS), Aiken, SC (United States); Lambert, D. P. [Savannah River Site (SRS), Aiken, SC (United States); Crump, S. L. [Savannah River Site (SRS), Aiken, SC (United States); Fondeur, F. F. [Savannah River Site (SRS), Aiken, SC (United States); Papathanassiu, A. E. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States); Kot, W. K. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States); Pegg, I. L. [Catholic University of America Vitreous State Laboratory, Washington, DC (United States)

2013-10-01

The present study examines the fate of glycolic acid and other organics added in the Chemical Processing Cell (CPC) of the Defense Waste Processing Facility (DWPF) as part of the glycolic alternate flowsheet. Adoption of this flowsheet is expected to provide certain benefits in terms of a reduction in the processing time, a decrease in hydrogen generation, simplification of chemical storage and handling issues, and an improvement in the processing characteristics of the waste stream including an increase in the amount of nitrate allowed in the CPC process. Understanding the fate of organics in this flowsheet is imperative because tank farm waste processed in the CPC is eventually immobilized by vitrification; thus, the type and amount of organics present in the melter feed may affect optimal melt processing and the quality of the final glass product as well as alter flammability calculations on the DWPF melter off gas. To evaluate the fate of the organic compounds added as the part of the glycolic flowsheet, mainly glycolic acid and antifoam 747, samples of simulated waste that was processed using the DWPF CPC protocol for tank farm sludge feed were generated and analyzed for organic compounds using a variety of analytical techniques at the Savannah River National Laboratory (SRNL). These techniques included Ion Chromatography (IC), Gas Chromatography-Mass Spectrometry (GC-MS), Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES), and Nuclear Magnetic Resonance (NMR) Spectroscopy. A set of samples were also sent to the Catholic University of America Vitreous State Laboratory (VSL) for analysis by NMR Spectroscopy at the University of Maryland, College Park. Analytical methods developed and executed at SRNL collectively showed that glycolic acid was the most prevalent organic compound in the supernatants of Slurry Mix Evaporator (SME) products examined. Furthermore, the studies suggested that commercially available glycolic acid contained minor amounts

DWPF SB6 Initial CPC Flowsheet Testing SB6-1 TO SB6-4L Tests Of SB6-A And SB6-B Simulants

International Nuclear Information System (INIS)

Lambert, D.; Pickenheim, B.; Best, D.

2009-01-01

The Defense Waste Processing Facility (DWPF) will transition from Sludge Batch 5 (SB5) processing to Sludge Batch 6 (SB6) processing in late fiscal year 2010. Tests were conducted using non-radioactive simulants of the expected SB6 composition to determine the impact of varying the acid stoichiometry during the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) processes. The work was conducted to meet the Technical Task Request (TTR) HLW/DWPF/TTR-2008-0043, Rev.0 and followed the guidelines of a Task Technical and Quality Assurance Plan (TT and QAP). The flowsheet studies are performed to evaluate the potential chemical processing issues, hydrogen generation rates, and process slurry rheological properties as a function of acid stoichiometry. These studies were conducted with the estimated SB6 composition at the time of the study. This composition assumed a blend of 101,085 kg of Tank 4 insoluble solids and 179,000 kg of Tank 12 insoluble solids. The current plans are to subject Tank 12 sludge to aluminum dissolution. Liquid Waste Operations assumed that 75% of the aluminum would be dissolved during this process. After dissolution and blending of Tank 4 sludge slurry, plans included washing the contents of Tank 51 to ∼1M Na. After the completion of washing, the plan assumes that 40 inches on Tank 40 slurry would remain for blending with the qualified SB6 material. There are several parameters that are noteworthy concerning SB6 sludge: (1) This is the second batch DWPF will be processing that contains sludge that has had a significant fraction of aluminum removed through aluminum dissolution; (2) The sludge is high in mercury, but the projected concentration is lower than SB5; (3) The sludge is high in noble metals, but the projected concentrations are lower than SB5; and(4) The sludge is high in U and Pu - components that are not added in sludge simulants. Six DWPF process simulations were completed in 4-L laboratory-scale equipment using

German Spent Nuclear Fuel Legacy: Characteristics and High-Level Waste Management Issues

Directory of Open Access Journals (Sweden)

A. Schwenk-Ferrero

2013-01-01

Full Text Available Germany is phasing-out the utilization of nuclear energy until 2022. Currently, nine light water reactors of originally nineteen are still connected to the grid. All power plants generate high-level nuclear waste like spent uranium or mixed uranium-plutonium dioxide fuel which has to be properly managed. Moreover, vitrified high-level waste containing minor actinides, fission products, and traces of plutonium reprocessing loses produced by reprocessing facilities has to be disposed of. In the paper, the assessments of German spent fuel legacy (heavy metal content and the nuclide composition of this inventory have been done. The methodology used applies advanced nuclear fuel cycle simulation techniques in order to reproduce the operation of the German nuclear power plants from 1969 till 2022. NFCSim code developed by LANL was adopted for this purpose. It was estimated that ~10,300 tonnes of unreprocessed nuclear spent fuel will be generated until the shut-down of the ultimate German reactor. This inventory will contain ~131 tonnes of plutonium, ~21 tonnes of minor actinides, and 440 tonnes of fission products. Apart from this, ca.215 tonnes of vitrified HLW will be present. As fission products and transuranium elements remain radioactive from 104 to 106 years, the characteristics of spent fuel legacy over this period are estimated, and their impacts on decay storage and final repository are discussed.

The scope and nature of the problem of high level nuclear waste disposal

International Nuclear Information System (INIS)

Jennekens, J.

1981-09-01

The disposal of high level nuclear waste poses a challenge to the Canadian technical and scientific communities, but a much greater challenge to government and industry leaders who must convince the public that the so-called 'problem' can be resolved by a pragmatic approach utilizing existing skills and knowledge. This paper outlines the objectives of radioactive waste management, the quantities of high level waste expected to be produced by the Canadian nuclear power program, the regulatory process which will apply and the government initiatives which have been and will be taken to ensure that the health, safety, security, and environmental interests of the public will be protected. (author)

Criticality assessment of initial operations at the Defense Waste Processing Facility

International Nuclear Information System (INIS)

Ha, B.C.; Williamson, T.G.

1993-01-01

At the Savannah River Site (SRS), high level radioactive wastes will be immobilized into borosilicate glass for long term storage and eventual disposal. Since the waste feed streams contain uranium and plutonium, the Defense Waste Processing Facility (DWPF) process has been evaluated to ensure that a subcritical condition is maintained. It was determined that the risk of nuclear criticality in the DWPF during initial, sludge-only operations is minimal due to the dilute concentration of fissile material in the sludge combined with excess neutron absorbers

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-09-14

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

Burst Test Qualification Analysis of DWPF Canister-Plug Weld

International Nuclear Information System (INIS)

Gupta, N.K.; Gong, Chung.

1995-02-01

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

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

International Nuclear Information System (INIS)

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

1981-01-01

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

Transport of Spent Nuclear Fuels, High and Intermediate Level Wastes: A Continuous Challenge

International Nuclear Information System (INIS)

Otton, C.; Blachet, L.

2009-01-01

For more than 45 years TN International has been involved in the radioactive materials transportation field. Since the beginning the used nuclear fuel transportation has been its core business. During all these years TN International, now part of AREVA, has been able to anticipate and fulfil the needs for new transport or storage casks design to fit the nuclear industry evolutions. A whole fleet of casks able to transport all the materials of the nuclear fuel cycle has been developed. In this presentation we will focus on the casks for the spent fuel, high level waste and intermediate level waste transportation. Answering to the constant evolution of the nuclear industry transport needs is a challenge that TN International faces routinely. Concerning the spent nuclear fuel transportation, TN International has developed in the early 80's a fleet of TN12 type casks fitted with several types of baskets able to safely transport all the spent fuel from the nuclear power plant or the research laboratories to AREVA La Hague plant. The current challenge is the design of a new transport cask generation taking into account the needs of the industry for the next 30 years. The replacement of the TN12 cask generation is to be scheduled as the regulations have changed and the fuel characteristics have evolved. The new generation of casks will take into account all the technical evolutions made during the TN12 thirty years of use. MOX spent fuel has now its dedicated cask: the TN112 which certificate of approval has been obtained in July 2008. This cask is able to transport 12 MOX spent fuel elements with a short cooling time. The first loading of the cask has been performed in 2008 in the EDF nuclear power plant of Saint-Laurent-des-Eaux. Concerning the high level waste such as the La Hague vitrified residues a whole fleet of casks has been developed such as the TN 28 VT dedicated to transport, the TN81 and TN85 dedicated to transport and storage. These casks have permitted the

DETERMINATION OF REPORTABLE RADIONUCLIDES FOR DWPF SLUDGE BATCH 4 MACROBATCH 5

International Nuclear Information System (INIS)

Bannochie, C; Ned Bibler, N; David Diprete, D

2008-01-01

The Waste Acceptance Product Specifications (WAPS)1 1.2 require that 'The Producer shall report the inventory of radionuclides (in Curies) that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115'. As part of the strategy to meet WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type, all radionuclides (with half-lives greater than 10 years) that have concentrations greater than 0.01 percent of the total inventory from time of production through the 1100 year period from 2015 through 3115. The initial listing of radionuclides to be included is based on the design-basis glass as identified in the Waste Form Compliance Plan (WCP)2 and Waste Form Qualification Report (WQR)3. However, it is required that this list be expanded if other radionuclides with half-lives greater than 10 years are identified that may meet the greater than 0.01% criterion for Curie content. Specification 1.6 of the WAPS, International Atomic Energy Agency (IAEA) Safeguards Reporting for High Level Waste (HLW), requires that the ratio by weights of the following uranium and plutonium isotopes be reported: U-233, U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, and Pu-242. Therefore, the complete set of reportable radionuclides must also include this set of U and Pu isotopes. The DWPF is receiving radioactive sludge slurry from HLW Tank 40. The radioactive sludge slurry in Tank 40 is a blend of the previous contents of Tank 40 (Sludge Batch 3) and the sludge that was transferred to Tank 40 from Tank 51. The blend of sludge from Tank 51 and Tank 40 defines Sludge Batch 4 (also referred to as Macrobatch 5 (MB5)). This report develops the list of reportable radionuclides and associated activities and determines the radionuclide activities as a function of time. The DWPF will use this list and the activities as one of

The Product Composition Control System at Savannah River: The statistical process control algorithm

International Nuclear Information System (INIS)

Brown, K.G.

1993-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) in Aiken, South Carolina, will be used to immobilize the approximately 130 million liters of high-level nuclear waste currently stored at the site in 51 carbon steel tanks. Waste handling operations separate this waste into highly radioactive insoluble sludge and precipitate and less radioactive water soluble salts. (In a separate facility, the soluble salts are disposed of as low-level waste in a mixture of cement, slag, and flyash.) In DWPF, precipitate (PHA) is blended with insoluble sludge and ground glass tit to produce melter feed slurry which is continuously fed to the DWPF melter. The melter produces a molten borosilicate glass which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in a geologic repository. The repository requires that the glass wasteform be resistant to leaching by underground water that might contact it. In addition, there are processing constraints on melt viscosity, liquidus temperature, and waste solubility

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Siting high-level nuclear waste repositories: A progress report for Rhode Island

International Nuclear Information System (INIS)

Frohlich, R.K.; Vild, B.F.

1986-03-01

In this booklet, we will not try to argue the pros and cons of nuclear power or weapons production. We will focus instead on the issue of nuclear waste disposal. With the passage of the Nuclear Waste Policy Act (NWPA) of 1982, the US Congress and the President charged federal and state regulators with the responsibility of settling that issue by the end of this century - with extensive public involvement. This booklet, now in its second printing, is designed to explain the nature of ''high-level'' nuclear waste, the essential criteria for its safe and permanent disposal, and Rhode Island's participation in the federal repository program. It has been funded from a USDOE grant derived from a utility-financed Nuclear Waste Fund established under the NWPA. 17 refs., 10 figs., 2 tabs

Level density from realistic nuclear potentials

International Nuclear Information System (INIS)

Calboreanu, A.

2006-01-01

Nuclear level density of some nuclei is calculated using a realistic set of single particle states (sps). These states are derived from the parameterization of nuclear potentials that describe the observed sps over a large number of nuclei. This approach has the advantage that one can infer level density for nuclei that are inaccessible for a direct study, but are very important in astrophysical processes such as those close to the drip lines. Level densities at high excitation energies are very sensitive to the actual set of sps. The fact that the sps spectrum is finite has extraordinary consequences upon nuclear reaction yields due to the leveling-off of the level density at extremely high excitation energies wrongly attributed so far to other nuclear effects. Single-particle level density parameter a parameter is extracted by fitting the calculated densities to the standard Bethe formula

Using electrochemical separation to reduce the volume of high-level nuclear waste

International Nuclear Information System (INIS)

Slater, S.A.; Gay, E.C.

1998-01-01

Argonne National Laboratory (ANL) has developed an electrochemical separation technique called electrorefining that will treat a variety of metallic spent nuclear fuel and reduce the volume of high-level nuclear waste that requires disposal. As part of that effort, ANL has developed a high throughput electrorefiner (HTER) that has a transport rate approximately three times faster than electrorefiners previously developed at ANL. This higher rate is due to the higher electrode surface area, a shorter transport path, and more efficient mixing, which leads to smaller boundary layers about the electrodes. This higher throughput makes electrorefining an attractive option in treating Department of Energy spent nuclear fuels. Experiments have been done to characterize the HTER, and a simulant metallic fuel has been successfully treated. The HTER design and experimental results is discussed

Symposium on the development of nuclear waste policy: Siting the high-level nuclear waste repository

International Nuclear Information System (INIS)

Pijawka, K.D.; Mushkatel, A.H.

1991-01-01

The Nuclear Waste Policy Act of 1982 (NWPA) attempted to formulate a viable national policy for managing the disposal of high-level nuclear wastes. The NWPA authorized the selection of two repository sites: the first to be constructed in the West and a second site developed in the eastern United States. A detailed process for site selection was outlined in the NWPA. In addition, the NWPA authorized open-quotes the development of a waste transportation system; required the Department of Energy (DOE) to submit a proposal to construct a facility for monitored retrievable storage (MRS) after conducting a study of the need for, and feasibility of such a facility; and required the President to evaluate the use of the repositories ... for the disposal of high-level waste resulting from defense activitiesclose quotes (DOE, 1988, p. 1). A series of provisions granting oversight participation to states and Indian tribes, as well as a compensation package for the ultimate host state were also included. Responsibility for implementing the NWPA was assigned to DOE

Parity dependence of the nuclear level density at high excitation

International Nuclear Information System (INIS)

Rao, B.V.; Agrawal, H.M.

1995-01-01

The basic underlying assumption ρ(l+1, J)=ρ(l, J) in the level density function ρ(U, J, π) has been checked on the basis of high quality data available on individual resonance parameters (E 0 , Γ n , J π ) for s- and p-wave neutrons in contrast to the earlier analysis where information about p-wave resonance parameters was meagre. The missing level estimator based on the partial integration over a Porter-Thomas distribution of neutron reduced widths and the Dyson-Mehta Δ 3 statistic for the level spacing have been used to ascertain that the s- and p-wave resonance level spacings D(0) and D(1) are not in error because of spurious and missing levels. The present work does not validate the tacit assumption ρ(l+1, J)=ρ(l, J) and confirms that the level density depends upon parity at high excitation. The possible implications of the parity dependence of the level density on the results of statistical model calculations of nuclear reaction cross sections as well as on pre-compound emission have been emphasized. (orig.)

Defense waste processing facility startup progress report

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

Herman, C

2006-01-01

The Actinide Removal Process (ARP) facility and the Modular Caustic Side Solvent Extraction Unit (MCU) are scheduled to begin processing salt waste in fiscal year 2007. A portion of the streams generated in the salt processing facilities will be transferred to the Defense Waste Processing Facility (DWPF) to be incorporated in the glass matrix. Before the streams are introduced, a combination of impact analyses and research and development studies must be performed to quantify the impacts on DWPF processing. The Process Science and Engineering (PS and E) section of the Savannah River National Laboratory (SRNL) was requested via Technical Task Request (TTR) HLW/DWPF/TTR-2004-0031 to evaluate the impacts on DWPF processing. Simulant Chemical Process Cell (CPC) flowsheet studies have been performed using previous composition and projected volume estimates for the ARP sludge/monosodium titanate (MST) stream. Due to changes in the flammability control strategy for DWPF for salt processing, the incorporation strategy for ARP has changed and additional ARP flowsheet tests were necessary to validate the new processing strategy. The last round of ARP testing included the incorporation of the MCU stream and identified potential processing issues with the MCU solvent. The identified issues included the potential carry-over and accumulation of the MCU solvent components in the CPC condensers and in the recycle stream to the Tank Farm. Therefore, DWPF requested SRNL to perform additional MCU flowsheet studies to better quantify the organic distribution in the CPC vessels. The previous MCU testing used a Sludge Batch 4 (SB4) simulant since it was anticipated that both of these facilities would begin salt processing during SB4 processing. The same sludge simulant recipe was used in this round of ARP and MCU testing to minimize the number of changes between the two phases of testing so a better comparison could be made. ARP and MCU stream simulants were made for this phase of

Development of anodic stripping voltametry for the determination of palladium in high level nuclear waste

Energy Technology Data Exchange (ETDEWEB)

Bhardwaj, T. K. [North Carolina State University, Raleigh (United States); Sharma, H. S.; Affarwal, S. K. [Bhabha Atomic Research Centre, Mumbai (India); Jain, P. C. [Meerut College, Meerut (India)

2012-12-15

Deposition potential, deposition time, square wave frequency, rotation speed of the rotating disc electrode, and palladium concentration were studied on a Glassy Carbon Electrode (GCE) in 0.01M HCl for the determination of palladium in High Level Nuclear Waste (HLNW) by anodic stripping voltammetry. Experimental conditions were optimized for the determination of palladium at two different, 10-8 and 10-7 M, levels. Error and standard deviation of this method were under 1% for all palladium standard solutions. The developed technique was successfully applied as a subsidiary method for the determination of palladium in simulated high level nuclear waste with very good precision and high accuracy (under 1 % error and standard deviation).

Fair rules for siting a high-level nuclear waste repository

International Nuclear Information System (INIS)

Easterling, D.

1992-01-01

Geologic repositories are designed to resolve the ever-growing problem of high-level nuclear waste, but these facilities invite intense local opposition due to the perceived severity of the risks and the possibility of stigma effects. This analysis examines whether the perceived fairness of the siting process affects local residents' support for hosting a repository. In particular, a survey of 1,001 Nevada residents is used to test the hypothesis that an individual's willingness to accept a local repository will increase if he or she is convinced that this is the safest disposal option available. A logistic analysis indicates that beliefs regarding relative suitability have an independent effect on the acceptability of a local repository (i.e., Yucca Mountain). The article then considers the question of how to implement an optimizing strategy for siting facilities, comparing an idealized strategy against the original Nuclear Waste Policy Act (NWPA) of 1982 and the Amendments Act of 1987. Although choosing the safest site seems as if it could enhance public acceptance of the repository program, there is currently little prospect of identifying the best option to the high-level waste problem and, as a results, little chance of gaining the public support that is necessary to promote a successful siting outcome. 81 refs., 1 fig., 5 tabs

Dynamic nuclear polarization at high Landau levels in a quantum point contact

Science.gov (United States)

Fauzi, M. H.; Noorhidayati, A.; Sahdan, M. F.; Sato, K.; Nagase, K.; Hirayama, Y.

2018-05-01

We demonstrate a way to polarize and detect nuclear spin in a gate-defined quantum point contact operating at high Landau levels. Resistively detected nuclear magnetic resonance (RDNMR) can be achieved up to the fifth Landau level and at a magnetic field lower than 1 T. We are able to retain the RDNMR signals in a condition where the spin degeneracy of the first one-dimensional (1D) subband is still preserved. Furthermore, the effects of orbital motion on the first 1D subband can be made smaller than those due to electrostatic confinement. This developed RDNMR technique is a promising means to study electronic states in a quantum point contact near zero magnetic field.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-30

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

Should the U.S. proceed to consider licensing deep geological disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Curtiss, J.R.

1993-01-01

The United States, as well as other countries facing the question of how to handle high-level nuclear waste, has decided that the most appropriate means of disposal is in a deep geologic repository. In recent years, the Radioactive Waste Management Committee of the Nuclear Energy Agency has developed several position papers on the technical achievability of deep geologic disposal, thus demonstrating the serious consideration of deep geologic disposal in the international community. The Committee has not, as yet, formally endorsed disposal in a deep geologic repository as the preferred method of handling high-level nuclear waste. The United States, on the other hand, has studied the various methods of disposing of high-level nuclear waste, and has determined that deep geologic disposal is the method that should be developed. The purpose of this paper is to present a review of the United States' decision on selecting deep geologic disposal as the preferred method of addressing the high-level waste problem. It presents a short history of the steps taken by the U.S. in determining what method to use, discusses the NRC's waste Confidence Decision, and provides information on other issues in the U.S. program such as reconsideration of the final disposal standard and the growing inventory of spent fuel in storage

Chem I Supplement. Chemistry Related to Isolation of High-Level Nuclear Waste.

Science.gov (United States)

Hoffman, Darleane C.; Choppin, Gregory R.

1986-01-01

Discusses some of the problems associated with the safe disposal of high-level nuclear wastes. Describes several waste disposal plans developed by various nations. Outlines the multiple-barrier concept of isolation in deep geological questions associated with the implementation of such a method. (TW)

Mineral-modeled ceramics for long-term storage of high-level nuclear wastes

International Nuclear Information System (INIS)

Vance, E.R.

1980-01-01

Over the past ten years, Penn State's Materials Research Laboratory has done extensive work on mineral-modeled ceramics for high-level nuclear waste storage. These ceramics are composed of several mineral analogues that form a monolithic polycrystalline aggregate. Mineral-modeling can be made in a similar fashion to nuclear waste glasses, and their naturally occurring analogues are known to last millions, and even billions, of years in hot, wet conditions. It is believed that such ceramics could reduce dispersal of radionuclides by leaching to a minimum

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

Integration of SWPF into the DWPF Flowsheet: Gap Analysis and Test Matrix Development

Energy Technology Data Exchange (ETDEWEB)

Peeler, D. K. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2014-12-10

Based on Revision 19 of the High Level Waste (HLW) System Plan, it is anticipated that the Salt Waste Processing Facility (SWPF) will be integrated into the Defense Waste Processing Facility (DWPF) flowsheet in October 2018 (or with Sludge Batch 11 (SB11)). Given that, Savannah River Remediation (SRR) has requested a technical basis be developed that validates the current Product Composition Control System (PCCS) models for use during the processing of the SWPF-based coupled flowsheet or that leads to the refinements of or modifications to the models that are needed so that the models may be used during the processing of the SWPF-based coupled flowsheet. To support this objective, Savannah River National Laboratory (SRNL) has completed three key interim activities prior to validation of the current or development of refined PCCS models over the anticipated glass composition region for SWPF processing. These three key activities include: (1) defining the glass compositional region over which SWPF is anticipated to be processed, (2) comparing the current PCCS model validation ranges to the SWPF glass compositional region from which compositional gaps can be identified, and (3) developing a test matrix to cover the compositional gaps.

Status of the United States' high-level nuclear waste disposal program

International Nuclear Information System (INIS)

Rusche, B.

1985-01-01

The Nuclear Waste Policy Act of 1982 is a remarkable piece of legislation in that there is general agreement on its key provisions. Nevertheless, this is a program intended to span more than a century, with some choices by Congress, states, Indian tribes and the nuclear power industry yet to be made. The crafters of the Act clearly recognized this. And further, the crafters recognized ''. . .that. . .state, Indian tribe and public participation in the planning and development of repositories is essential in order to promote public confidence in the safety of disposal of such waste and spent fuel . . . High-level radioactive waste and spent nuclear fuel have become major subjects of public concern, and appropriate precautions must be taken to ensure that such waste and spent fuel do not adversely affect the public health and safety and the environment for this or future generations

Risk perception on management of nuclear high-level and transuranic waste storage

Energy Technology Data Exchange (ETDEWEB)

Dees, Lawrence A. [Colorado Christian Univ., Lakewood, CO (United States)

1994-08-15

The Department of Energy`s program for disposing of nuclear High-Level Waste (HLW) and transuranic (TRU) waste has been impeded by overwhelming political opposition fueled by public perceptions of actual risk. Analysis of these perceptions shows them to be deeply rooted in images of fear and dread that have been present since the discovery of radioactivity. The development and use of nuclear weapons linked these images to reality and the mishandling of radioactive waste from the nations military weapons facilities has contributed toward creating a state of distrust that cannot be erased quickly or easily. In addition, the analysis indicates that even the highly educated technical community is not well informed on the latest technology involved with nuclear HLW and TRU waste disposal. It is not surprising then, that the general public feels uncomfortable with DOE`s management plans for with nuclear HLW and TRU waste disposal. Postponing the permanent geologic repository and use of Monitored Retrievable Storage (MRS) would provide the time necessary for difficult social and political issues to be resolved. It would also allow time for the public to become better educated if DOE chooses to become proactive.

Risk perception on management of nuclear high-level and transuranic waste storage

International Nuclear Information System (INIS)

Dees, L.A.

1994-01-01

The Department of Energy's program for disposing of nuclear High-Level Waste (HLW) and transuranic (TRU) waste has been impeded by overwhelming political opposition fueled by public perceptions of actual risk. Analysis of these perceptions shows them to be deeply rooted in images of fear and dread that have been present since the discovery of radioactivity. The development and use of nuclear weapons linked these images to reality and the mishandling of radioactive waste from the nations military weapons facilities has contributed toward creating a state of distrust that cannot be erased quickly or easily. In addition, the analysis indicates that even the highly educated technical community is not well informed on the latest technology involved with nuclear HLW and TRU waste disposal. It is not surprising then, that the general public feels uncomfortable with DOE's management plans for with nuclear HLW and TRU waste disposal. Postponing the permanent geologic repository and use of Monitored Retrievable Storage (MRS) would provide the time necessary for difficult social and political issues to be resolved. It would also allow time for the public to become better educated if DOE chooses to become proactive

DWPF Melter No.2 Prototype Bus Bar Test Report

International Nuclear Information System (INIS)

Gordon, J.

2003-01-01

Characterization and performance testing of a prototype DWPF Melter No.2 Dome Heater Bus Bar are described. The prototype bus bar was designed to address the design features of the existing system which may have contributed to water leaks on Melter No.1. Performance testing of the prototype revealed significant improvement over the existing design in reduction of both bus bar and heater connection maximum temperature, while characterization revealed a few minor design and manufacturing flaws in the bar. The prototype is recommended as an improvement over the existing design. Recommendations are also made in the area of quality control to ensure that critical design requirements are met

Erosion/corrosion concerns in feed preparation systems at the Defense Waste Processing Facility

International Nuclear Information System (INIS)

Gee, J.T.; Chandler, C.T.; Daugherty, W.L.; Imrich, K.J.; Jenkins, C.F.

1997-01-01

The Savannah River Site (SRS) has been operating a nuclear fuel cycle since the 1950's to produce nuclear materials in support of the national defense effort. The Department of Energy authorized the construction of the Defense Waste Processing Facility (DWPF) to immobilize the high level radioactive waste resulting from these processes as a durable borosilicate glass. The DWPF, after having undergone extensive testing, has been approved for operations and is currently immobilizing radioactive waste. To ensure reliability of the DWPF remote canyon processing equipment, a materials evaluation program was performed prior to radioactive operations to determine to what extent erosion/corrosion would impact design life of equipment. The program consisted of performing pre-service baseline inspections on critical equipment and follow-up inspections after completion of DWPF cold chemical demonstration runs. Non-destructive examination (NDE) techniques were used to assess erosion/corrosion as well as evaluation of corrosion coupon racks. These results were used to arrive at predicted equipment life for selected feed preparation equipment. It was concluded with the exception of the coil and agitator for the slurry mix evaporator (SME), which are exposed to erosive glass frit particles, all of the equipment should meet its design life

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

Numerical investigation of high level nuclear waste disposal in deep anisotropic geologic repositories

KAUST Repository

Salama, Amgad; El Amin, Mohamed F.; Sun, Shuyu

2015-01-01

One of the techniques that have been proposed to dispose high level nuclear waste (HLW) has been to bury them in deep geologic formations, which offer relatively enough space to accommodate the large volume of HLW accumulated over the years since

Nuclear safety of extended sludge processing on tank 42 and 51 sludge (DWPF sludge feed batch one)

International Nuclear Information System (INIS)

Clemons, J.S.

1993-01-01

The sludge in tanks 42 and 51 is to be washed with inhibited water to remove soluble salts and combined in tank 51 in preparation for feed to DWPF. Since these tanks contain uranium and plutonium, the process of washing must be evaluated to ensure subcriticality is maintained. When the sludge is washed, inhibited water is added, the tank contents are slurried and allowed to settle. The sludge wash water is then decanted to the evaporator feed tank where it is fed to the evaporator to reduce the volume. The resulting evaporator concentrate is sent to a salt tank where it cools and forms crystallized salt cake. This salt cake will later be dissolved, processed in ITP and sent to Z-Area. This report evaluates the supernate and sludge during washing, the impact on the evaporator during concentration of decanted wash water, and the salt tank where the concentrated supernate is deposited. The conclusions generated in this report are specific to the sludge currently contained in tanks 42 and 51

Separation of aromatic precipitates from simulated high level radioactive waste by hydrolysis, evaporation and liquid-liquid extraction

International Nuclear Information System (INIS)

Young, S.R.; Shah, H.B.; Carter, J.T.

1991-01-01

The Defense Waste Processing Facility (DWPF) at the SRS will be the United States' first facility to process High Level radioactive Waste (HLW) into a borosilicate glass matrix. The removal of aromatic precipitates by hydrolysis, evaporation and liquid-liquid extraction will be a key step in the processing of the HLW. This step, titled the Precipitate Hydrolysis Process, has been demonstrated by the Savannah River Laboratory with the Precipitate Hydrolysis Experimental Facility (PHEF). The mission of the PHEF is to demonstrate processing of simulated high level radioactive waste which contains tetraphenylborate precipitates and nitrite. Reduction of nitrite by hydroxylamine nitrate and hydrolysis of the tetraphenylborate by formic acid is discussed. Gaseous production, which is primarily benzene, nitrous oxide and carbon dioxide, has been quantified. Production of high-boiling organic compounds and the accumulation of these organic compounds within the process are addressed

The product composition control system at Savannah River: Statistical process control algorithm

International Nuclear Information System (INIS)

Brown, K.G.

1994-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will be used to immobilize the approximately 130 million liters of high-level nuclear waste currently stored at the site in 51 carbon steel tanks. Waste handling operations separate this waste into highly radioactive insoluble sludge and precipitate and less radioactive water soluble salts. In DWPF, precipitate (PHA) is blended with insoluble sludge and ground glass frit to produce melter feed slurry which is continuously fed to the DWPF melter. The melter produces a molten borosilicate glass which is poured into stainless steel canisters for cooling and, ultimately, shipment to and storage in an geologic repository. Described here is the Product Composition Control System (PCCS) process control algorithm. The PCCS is the amalgam of computer hardware and software intended to ensure that the melt will be processable and that the glass wasteform produced will be acceptable. Within PCCS, the Statistical Process Control (SPC) Algorithm is the means which guides control of the DWPF process. The SPC Algorithm is necessary to control the multivariate DWPF process in the face of uncertainties arising from the process, its feeds, sampling, modeling, and measurement systems. This article describes the functions performed by the SPC Algorithm, characterization of DWPF prior to making product, accounting for prediction uncertainty, accounting for measurement uncertainty, monitoring a SME batch, incorporating process information, and advantages of the algorithm. 9 refs., 6 figs

A method for assay of special nuclear material in high level liquid waste streams

International Nuclear Information System (INIS)

Venkata Subramani, C.R.; Swaminathan, K.; Asuvathraman, R.; Kutty, K.V.G.

2003-01-01

The assay of special nuclear material in the high level liquid waste streams assumes importance as this is the first stage in the extraction cycle and considerable losses of plutonium could occur here. This stream contains all the fission products as also the minor actinides and hence normal nuclear techniques cannot be used without prior separation of the special nuclear material. This paper presents the preliminary results carried out using wavelength dispersive x-ray fluorescence as part of the developmental efforts to assay SNM in these streams by instrumental techniques. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-11-10

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

MAR Assessments Of The High Level Waste System Plan Revision 16

International Nuclear Information System (INIS)

Peeler, D.; Edwards, T.

2011-01-01

High-level waste (HLW) throughput (i.e., the amount of waste processed per unit of time) is primarily a function of two critical parameters: waste loading (WL) and melt rate. For the Defense Waste Processing Facility (DWPF), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). Significant increases in waste throughput have been achieved at DWPF since initial radioactive operations began in 1996. Key technical and operational initiatives that supported increased waste throughput included improvements in facility attainment, the Chemical Processing Cell (CPC) flowsheet, process control models and frit formulations. As a result of these key initiatives, DWPF increased WLs from a nominal 28% for Sludge Batch 2 (SB2) to ∼34 to 38% for SB3 through SB6 while maintaining or slightly improving canister fill times. Although considerable improvements in waste throughput have been obtained, future contractual waste loading targets are nominally 40%, while canister production rates are also expected to increase (to a rate of 325 to 400 canisters per year). Although implementation of bubblers have made a positive impact on increasing melt rate for recent sludge batches targeting WLs in the mid30s, higher WLs will ultimately make the feeds to DWPF more challenging to process. Savannah River Remediation (SRR) recently requested the Savannah River National Laboratory (SRNL) to perform a paper study assessment using future sludge projections to evaluate whether the current Process Composition Control System (PCCS) algorithms would provide projected operating windows to allow future contractual WL targets to be met. More specifically, the objective of this study was to evaluate future sludge batch projections (based on Revision 16 of the HLW Systems Plan) with respect to projected operating windows using current PCCS models and associated constraints. Based on the assessments, the waste loading interval over

Erosion Modeling Analysis For Modified DWPF SME Tank

International Nuclear Information System (INIS)

LEE, SI

2004-01-01

In support of an erosion evaluation for the modified cooling coil guide and its supporting structure in the DWPF SME vessel, a computational model was developed to identify potential sites of high erosion using the same methodology established by previous work. The erosion mechanism identified in the previous work was applied to the evaluation of high erosion locations representative of the actual flow process in the modified coil guide of the SME vessel, abrasive erosion which occurs by high wall shear of viscous liquid. The results show that primary locations of the highest erosion due to the abrasive wall erosion are at the leading edge of the guide, external surface of the insert plate, the tank floor next to the insert plate of the coil guide support, and the upstream lead-in plate. The present modeling results show a good comparison between the original and the modified cases in terms of high erosion sites, as well as the degree of erosion and the calculated shear stress. Wall she ar of the tank floor is reduced by about 30 per cent because of the new coil support plate. Calculations for the impeller speed lower than 103 rpm in the SME showed similar erosion patterns but significantly reduced wall shear stresses and reduced overall erosion. Comparisons of the 103 rpm results with SME measurements indicated that no significant erosion of the tank floor in the SME is to be expected. Thus, it is recommended that the agitator speed of SME does not exceed 103 rpm

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

International Nuclear Information System (INIS)

Jantzen, C.M.

1992-01-01

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

Supplemental environmental impact statement - defense waste processing facility

International Nuclear Information System (INIS)

1994-11-01

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

Four themes that underlie the high-level nuclear waste management program

International Nuclear Information System (INIS)

Sprecher, W.M.

1989-01-01

In 1982, after years of deliberation and in response to mounting pressures from environmental, industrial, and other groups, the US Congress enacted the Nuclear Waste Policy Act (NWPA) of 1982, which was signed into law by the President in January 1983. That legislation signified a major milestone in the nation's management of high-level nuclear waste, since it represented a consensus among the nation's lawmakers to tackle a problem that had evaded solution for decades. Implementation of the NWPA has proven to be exceedingly difficult, as attested by the discord generated by the US Department of Energy's (DOE's) geologic repository and monitored retrievable storage (MRS) facility siting activities. The vision that motivated the crafters of the 1982 act became blurred as opposition to the law increased. After many hearings that underscored the public's concern with the waste management program, the Congress enacted the Nuclear Waste Policy Amendments Act of 1987 (Amendments Act), which steamlined and focused the program, while establishing three independent bodies: the MRS Review Commission, the Nuclear Waste Technical Review Board, and the Office of the Nuclear Waste Negotiator. Yet, even as the program evolves, several themes characterizing the nation's effort to solve the waste management problem continue to prevail. The first of these themes has to do with social consciousness, and the others that follow deal with technical leadership, public involvement and risk perceptions, and program conservatism

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-05-07

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-08-28

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

IMPACTS OF SMALL COLUMN ION EXCHANGE STREAMS ON DWPF GLASS FORMULATION: KT01, KT02, KT03, AND KT04-SERIES GLASS COMPOSITIONS

Energy Technology Data Exchange (ETDEWEB)

Fox, K.; Edwards, T.

2010-11-01

Four series of glass compositions were selected, fabricated, and characterized as part of a study to determine the impacts of the addition of Crystalline Silicotitanate (CST) and Monosodium Titanate (MST) from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) glass waste form and the applicability of the DWPF process control models. The KT01 and KT02-series of glasses were chosen to allow for the identification of the influence of the concentrations of major components of the glass on the retention of TiO{sub 2}. The KT03 series of glasses was chosen to allow for the identification of these influences when higher Nb{sub 2}O{sub 5} and ZrO{sub 2} concentrations are included along with TiO2. The KT04 series of glasses was chosen to investigate the properties and performance of glasses based on the best available projections of actual compositions to be processed at the DWPF (i.e., future sludge batches including the SCIX streams).

Remote process cell mercury transfer pumps for DWPF

International Nuclear Information System (INIS)

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

1986-01-01

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

Tentative job analysis for a high-level, fixed-site, nuclear security officer

International Nuclear Information System (INIS)

Adams, K.G.; Trujillo, A.A.

1977-10-01

A tentative job analysis for a high-level, fixed-site, nuclear security officer is presented. The primary objective of the report is to provide a framework for evaluating the functions of a security officer in physical protection systems. Several job requirements related to duties, basic skills, personal contacts, supervision, working conditions, and decision making are presented. Individual character traits desirable in security officers are described

Hydrogen generation and foaming during tests in the GFPS simulating DWPF operations with Tank 42 sludge and CST

Energy Technology Data Exchange (ETDEWEB)

Koopman, D.C.

1999-12-08

This report summarizes the pilot-scale research requested by the salt disposition team to examine the effect of crystalline silicotitanate (CST) resin with adsorbed noble metals on the maximum hydrogen generation rate produced during the DWPF melter feed preparation processes.

Hydrogen generation and foaming during tests in the GFPS simulating DWPF operations with Tank 42 sludge and CST

International Nuclear Information System (INIS)

Koopman, D.C.

1999-01-01

This report summarizes the pilot-scale research requested by the salt disposition team to examine the effect of crystalline silicotitanate (CST) resin with adsorbed noble metals on the maximum hydrogen generation rate produced during the DWPF melter feed preparation processes

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

International Nuclear Information System (INIS)

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

1982-04-01

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

Workshop on the role of natural analogs in geologic disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Murphy, W.M.; Kovach, L.A.

1995-01-01

A workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste (HLW) was held in San Antonio, Texas, on July 22-25, 1991. It was sponsored by the US Nuclear Regulatory Commission (NRC) and the Center for Nuclear Waste Regulatory Analyses (CNWRA). Invitations to the workshop were extended to a large number of individuals with a variety of technical and professional interests related to geologic disposal of nuclear waste and natural analog studies. The objective of the workshop was to examine the role of natural analog studies in performance assessment, site characterization, and prioritization of research related to geologic disposal of HLW

Nuclear Level Densities

International Nuclear Information System (INIS)

Grimes, S.M.

2005-01-01

Recent research in the area of nuclear level densities is reviewed. The current interest in nuclear astrophysics and in structure of nuclei off of the line of stability has led to the development of radioactive beam facilities with larger machines currently being planned. Nuclear level densities for the systems used to produce the radioactive beams influence substantially the production rates of these beams. The modification of level-density parameters near the drip lines would also affect nucleosynthesis rates and abundances

Treatment of simulated high-level radioactive waste with formic acid: Bench-scale study on hydrogen evolution

International Nuclear Information System (INIS)

Hsu, C.L.W.; Ritter, J.A.

1996-01-01

At the Savannah River Site, the Defense Waste Processing Facility (DWPF) was constructed to vitrify high-level radioactive liquid waste in borosilicate glass for permanent storage. Formic acid, which serves as both an acid and a reducing agent, is used to treat the washed alkaline sludge during melter feed preparation primarily to improve the processability of the feed and to reduce mercury to its zero state for steam stripping. The high-level sludge is composed of many transition metal hydroxides. Among them, there are small quantities of platinum group metals. During the treatment of simulated sludge with formic acid, significant amounts of hydrogen were generated when the platinum group metals were included in the sludge. Apparently the noble metals in the sludge were reduced to their zero states and caused formic acid to decompose catalytically into hydrogen and carbon dioxide, usually with an induction period. The production of hydrogen gas presented the DWPF with a safety issue. Therefore, the objective of this research was to gain a fundamental understanding of what controlled the hydrogen evolution so that a practical solution to the safety issue could be obtained. A bench-scale parametric study revealed the following: increasing the amount of formic acid added to the sludge increased the hydrogen generation rate dramatically; once the catalysts were activated, the hydrogen generation rate decreased significantly with a lowering of the temperature of the sludge; the relative catalytic activities of the noble metals in the sludge decreased in the following order: rhodium > ruthenium much-gt palladium; ammonium ions were generated catalytically from the reaction between formic acid and nitrate; and when present, the noble metals caused higher upward drifts of the sludge pH

The nuclear physical method for high pressure steam manifold water level gauging and its error

International Nuclear Information System (INIS)

Li Nianzu; Li Beicheng; Jia Shengming

1993-10-01

A new method, which is non-contact on measured water level, for measuring high pressure steam manifold water level with nuclear detection technique is introduced. This method overcomes the inherent drawback of previous water level gauges based on other principles. This method can realize full range real time monitoring on the continuous water level of high pressure steam manifold from the start to full load of boiler, and the actual value of water level can be obtained. The measuring errors were analysed on site. Errors from practical operation in Tianjin Junliangcheng Power Plant and in laboratory are also presented

Canada's high-level nuclear waste disposal concept and its evaluation process

International Nuclear Information System (INIS)

Sheng, Grant; Shemilt, L.W.

2004-01-01

The concept of disposing high-level nuclear waste in granitic rocks in the Canadian Shield, developed by Atomic Energy of Canada Limited (AECL), is anticipated to undergo a national public review within two years. The disposal concept, its documentation, and its process of evaluation, including the role of the public, government and the scientific/engineering community, are summarized. A Technical Advisory Committee (TAC) has provided external peer review of the Program since 1979 and its findings are published in annual reports which are publicly available. (author)

The socioeconomic impacts of high-level nuclear waste facility siting

International Nuclear Information System (INIS)

Murdock, S.H.; Leistritz, F.L.; Hamm, R.R.

1987-01-01

High-level nuclear waste repositories will be located in sparsely settled rural areas in the U.S. These projects will significantly effect the economic, demographic, public service, fiscal, and social (the socioeconomic) dimensions of those rural areas. This paper examines some of the potential socioeconomic impacts and the characteristics of mitigation programs necessary, if these impacts are to be addressed. Both standard impacts, those resulting from the fact that--like many other large-scale developments--repositories will involve a substantial number of new workers and residents (relative to the size of existing communities) and special impacts, those resulting from the fact that repositories store radioactive materials, are examined

High performance gamma measurements of equipment retrieved from Hanford high-level nuclear waste tanks

International Nuclear Information System (INIS)

Troyer, G.L.

1997-01-01

The cleanup of high level defense nuclear waste at the Hanford site presents several progressive challenges. Among these is the removal and disposal of various components from buried active waste tanks to allow new equipment insertion or hazards mitigation. A unique automated retrieval system at the tank provides for retrieval, high pressure washing, inventory measurement, and containment for disposal. Key to the inventory measurement is a three detector HPGe high performance gamma spectroscopy system capable of recovering data at up to 90% saturation (200,000 counts per second). Data recovery is based on a unique embedded electronic pulser and specialized software to report the inventory. Each of the detectors have different shielding specified through Monte Carlo simulation with the MCNP program. This shielding provides performance over a dynamic range of eight orders of magnitude. System description, calibration issues and operational experiences are discussed

High performance gamma measurements of equipment retrieved from Hanford high-level nuclear waste tanks

Energy Technology Data Exchange (ETDEWEB)

Troyer, G.L.

1997-03-17

The cleanup of high level defense nuclear waste at the Hanford site presents several progressive challenges. Among these is the removal and disposal of various components from buried active waste tanks to allow new equipment insertion or hazards mitigation. A unique automated retrieval system at the tank provides for retrieval, high pressure washing, inventory measurement, and containment for disposal. Key to the inventory measurement is a three detector HPGe high performance gamma spectroscopy system capable of recovering data at up to 90% saturation (200,000 counts per second). Data recovery is based on a unique embedded electronic pulser and specialized software to report the inventory. Each of the detectors have different shielding specified through Monte Carlo simulation with the MCNP program. This shielding provides performance over a dynamic range of eight orders of magnitude. System description, calibration issues and operational experiences are discussed.

Mathematical modelling of heat production in deep geological repository of high-level nuclear waste

International Nuclear Information System (INIS)

Kovanda, O.

2017-01-01

Waste produced by nuclear industry requires special handling. Currently, there is a research taking place, focused at possibilities of nuclear waste storage in deep geological repositories, hosted in stable geological environment. The high-level nuclear waste produces significant amount of heat for a long time, which can affect either environment outside of or within the repository in a negative way. Therefore to reduce risks, it is desirable to know the principles of such heat production, which can be achieved using mathematical modeling. This thesis comes up with a general model of heat production-time dependency, dependable on initial composition of the waste. To be able to model real situations, output of this thesis needs to be utilized in an IT solution. (authors)

Sorption of strontium on uranyl peroxide: implications for a high-level nuclear waste repository.

Science.gov (United States)

Sureda, Rosa; Martínez-Lladó, Xavier; Rovira, Miquel; de Pablo, Joan; Casas, Ignasi; Giménez, Javier

2010-09-15

Strontium-90 is considered the most important radioactive isotope in the environment and one of the most frequently occurring radionuclides in groundwaters at nuclear facilities. The uranyl peroxide studtite (UO2O2 . 4H2O) has been observed to be formed in spent nuclear fuel leaching experiments and seems to have a relatively high sorption capacity for some radionuclides. In this work, the sorption of strontium onto studtite is studied as a function of time, strontium concentration in solution and pH. The main results obtained are (a) sorption is relatively fast although slower than for cesium; (b) strontium seems to be sorbed via a monolayer coverage of the studtite surface, (c) sorption has a strong dependence on ionic strength, is negligible at acidic pH, and increases at neutral to alkaline pH (almost 100% of the strontium in solution is sorbed above pH 10). These results point to uranium secondary solid phase formation on the spent nuclear fuel as an important mechanism for strontium retention in a high-level nuclear waste repository (HLNW). Copyright 2010 Elsevier B.V. All rights reserved.

Instrument reliability for high-level nuclear-waste-repository applications

International Nuclear Information System (INIS)

Rogue, F.; Binnall, E.P.; Armantrout, G.A.

1983-01-01

Reliable instrumentation will be needed to evaluate the characteristics of proposed high-level nuclear-wasted-repository sites and to monitor the performance of selected sites during the operational period and into repository closure. A study has been done to assess the reliability of instruments used in Department of Energy (DOE) waste repository related experiments and in other similar geological applications. The study included experiences with geotechnical, hydrological, geochemical, environmental, and radiological instrumentation and associated data acquisition equipment. Though this paper includes some findings on the reliability of instruments in each of these categories, the emphasis is on experiences with geotechnical instrumentation in hostile repository-type environments. We review the failure modes, rates, and mechanisms, along with manufacturers modifications and design changes to enhance and improve instrument performance; and include recommendations on areas where further improvements are needed

United States Program on Spent Nuclear Fuel and High-Level Radioactive Waste Management

International Nuclear Information System (INIS)

Stewart, L.

2004-01-01

The President signed the Congressional Joint Resolution on July 23, 2002, that designated the Yucca Mountain site for a proposed geologic repository to dispose of the nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The United States (U.S.) Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) is currently focusing its efforts on submitting a license application to the U.S. Nuclear Regulatory Commission (NRC) in December 2004 for construction of the proposed repository. The legislative framework underpinning the U.S. repository program is the basis for its continuity and success. The repository development program has significantly benefited from international collaborations with other nations in the Americas

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

International Nuclear Information System (INIS)

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

1980-01-01

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

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

International Nuclear Information System (INIS)

Johnson, F.

2012-01-01

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

Management of commercial high-level nuclear waste in the United States

International Nuclear Information System (INIS)

Mann, S.A.

1981-01-01

The Nuclear Waste Policy Act of 1982 (Act), enacted by the 97th Congress in December and signed into law on Jan 7, 1983, by President Reagan, brings a whole new perspective to the nation's nuclear waste management effort. An elaborate set of near-term requirements and actions have to be accomplished within the first 180 days of this Act. As an initial step, Secretary of Energy Donald P. Hodel has established a Nuclear Waste Policy Act (NWPA) project office. The director of the NWPA project office, Robert L. Morgan, is responsible for the department's initial activities to implement the Act until the Office of Civilian Radioactive Waste Management, established by Section 304 of the Act, is activated. The Act requires major efforts in two primary areas: disposal and storage of spent fuel and high-level waste. Efforts in the former area are to be financed by fees collected from utilities and placed into a Nuclear Waste Fund for disposal services. The Act provides for federal storage of up to 1900 tons of spent fuel for those utilities that the Nuclear Regulatory Commission (NRC) determines cannot reasonably provide sufficient additional on-site storage. This federal storage is to be financed through utility-user fees that are placed into an Interim Storage Fund. The Act also provides for cooperative research, development, and demonstration activities at utility sites and federal sites. These activities are to be jointly funded by the utilities and the federal government. Lastly, there are generic research and development (RandD) activities in the spent fuel area that are funded from general appropriations. Mandated milestones have been established by the Act in the areas of gologic repository, fund management, monitored retrievable storage, and spent fuel storage

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

International Nuclear Information System (INIS)

1991-05-01

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

Spent nuclear fuel and high level radioactive waste transportation. White paper

International Nuclear Information System (INIS)

1985-06-01

The High-Level Radioactive Waste Committee of the Western Interstate Energy Board has been involved in a year-long cooperative project with the US Department of Energy (DOE) to develop an information base on the transportation of spent nuclear fuel and high-level radioactive waste (HLW) so that western states can be constructive and informed participants in the repository program under the Nuclear Waste Policy Act (NWPA). The historical safety record of transportation of HLW and spent fuel is excellent; no release of these radioactive materials has ever occurred during transportation. Projected shipments under the NWPA will, however, greatly exceed current shipments in the US. For example, over the past five years, 119 metric tons of civilian spent fuel have been shipped in this country, while shipments to the first and second repository are each expected to peak at 3000 metric tons per year. The Committee believes that the successful development and operation of a national HLW/spent fuel transportation system can best be accomplished through an open process based on the common sense approach of taking all reasonable measures to minimize public risk and performing whatever actions are reasonably required to promote public acceptance. Therefore, the Committee recommends that the Department of Energy further the goals of the NWPA by developing a Comprehensive Transportation Plan which adopts a systematic, comprehensive, and integrated approach to resolving all spent fuel and HLW transportation issues in a timely manner. The suggested scope of such a plan is discussed in this White paper. Many of the suggested elements of such a plan are similar to those being developed by the Department of energy for inclusion in the Department's Transportation Institutional Plan

Preliminary Evaluation Of DWPF Impacts Of Boric Acid Use In Cesium Strip FOR SWPF And MCU

International Nuclear Information System (INIS)

Stone, M.

2010-01-01

A new solvent system is being evaluated for use in the Modular Caustic-Side Solvent Extraction Unit (MCU) and in the Salt Waste Processing Facility (SWPF). The new system includes the option to replace the current dilute nitric acid strip solution with boric acid. To support this effort, the impact of using 0.01M, 0.1M, 0.25M and 0.5M boric acid in place of 0.001M nitric acid was evaluated for impacts on the DWPF facility. The evaluation only covered the impacts of boric acid in the strip effluent and does not address the other changes in solvents (i.e., the new extractant, called MaxCalix, or the new suppressor, guanidine). Boric acid additions may lead to increased hydrogen generation during the SRAT and SME cycles as well as change the rheological properties of the feed. The boron in the strip effluent will impact glass composition and could require each SME batch to be trimmed with boric acid to account for any changes in the boron from strip effluent additions. Addition of boron with the strip effluent will require changes in the frit composition and could lead to changes in melt behavior. The severity of the impacts from the boric acid additions is dependent on the amount of boric acid added by the strip effluent. The use of 0.1M or higher concentrations of boric acid in the strip effluent was found to significantly impact DWPF operations while the impact of 0.01M boric acid is expected to be relatively minor. Experimental testing is required to resolve the issues identified during the preliminary evaluation. The issues to be addressed by the testing are: (1) Impact on SRAT acid addition and hydrogen generation; (2) Impact on melter feed rheology; (3) Impact on glass composition control; (4) Impact on frit production; and (5) Impact on melter offgas. A new solvent system is being evaluated for use in the Modular Caustic-Side Solvent Extraction Unit (MCU) and in the Salt Waste Processing Facility (SWPF). The new system includes the option to replace the

Should high-level nuclear waste be disposed of at geographically dispersed sites?

International Nuclear Information System (INIS)

Bassett, G.W. Jr.

1992-01-01

Consideration of the technical feasibility of Yucca Mountain in Nevada as the site for a high-level nuclear waste repository has led to an intense debate regarding the economic, social, and political impacts of the repository. Impediments to the siting process mean that the nuclear waste problem is being resolved by adhering to the status quo, in which nuclear waste is stored at scattered sites near major population centers. To assess the merits of alternative siting strategies--including both the permanent repository and the status quo- we consider the variables that would be included in a model designed to select (1) the optimal number of disposal facilities, (2) the types of facilities (e.g., permanent repository or monitored retrievable facility), and (3) the geographic location of storage sites. The objective function in the model is an all-inclusive measure of social cost. The intent of the exercise is not to demonstrate the superiority of any single disposal strategy; uncertainties preclude a conclusive proof of optimality for any of the disposal options. Instead, we want to assess the sensitivity of a variety of proposed solutions to variations in the physical, economic, political, and social variables that influence a siting strategy

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

International Nuclear Information System (INIS)

Harbour, J.R.

1995-01-01

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

The systems approach to spent fuel and high level nuclear waste management

International Nuclear Information System (INIS)

Norman, N.A.

1987-01-01

Based on the prior successful experience in the application of systems engineering techniques to large, complex, multi-participant programs the U.S. DOE has elected to apply the systems engineering process to the U.S. High Level Nuclear Waste Management Program (HLNWMP). Features of the systems engineering process as it applies to the U.S. HLNWMP and the importance of the component interdependencies are descirbed. Four principle components of the U.S. HLW Managment System are also given. They are the reactors or defense sources, transportation system, interium storoge modes, and the final repository. (Huang)

High-level waste immobilization program: an overview

International Nuclear Information System (INIS)

Bonner, W.R.

1979-09-01

The High-Level Waste Immobilization Program is providing technology to allow safe, affordable immobilization and disposal of nuclear waste. Waste forms and processes are being developed on a schedule consistent with national needs for immobilization of high-level wastes stored at Savannah River, Hanford, Idaho National Engineering Laboratory, and West Valley, New York. This technology is directly applicable to high-level wastes from potential reprocessing of spent nuclear fuel. The program is removing one more obstacle previously seen as a potential restriction on the use and further development of nuclear power, and is thus meeting a critical technological need within the national objective of energy independence

DWPF nitric-glycolic flowsheet chemical process cell chemistry. Part 1

Energy Technology Data Exchange (ETDEWEB)

Zamecnik, J. R. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Edwards, T. B. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

2016-02-01

The conversions of nitrite to nitrate, the destruction of glycolate, and the conversion of glycolate to formate and oxalate were modeled for the Nitric-Glycolic flowsheet using data from Chemical Process Cell (CPC) simulant runs conducted by SRNL from 2011 to 2015. The goal of this work was to develop empirical correlations for these variables versus measureable variables from the chemical process so that these quantities could be predicted a-priori from the sludge composition and measurable processing variables. The need for these predictions arises from the need to predict the REDuction/OXidation (REDOX) state of the glass from the Defense Waste Processing Facility (DWPF) melter. This report summarizes the initial work on these correlations based on the aforementioned data. Further refinement of the models as additional data is collected is recommended.

High-level nuclear waste borosilicate glass: A compendium of characteristics

International Nuclear Information System (INIS)

Cunnane, J.C.; Bates, J.K.; Ebert, W.L.; Feng, X.; Mazer, J.J.; Wronkiewicz, D.J.; Sproull, J.; Bourcier, W.L.; McGrail, B.P.

1992-01-01

With the imminent startup, in the United States, of facilities for vitrification of high-level nuclear waste, a document has been prepared that compiles the scientific basis for understanding the alteration of the waste glass products under the range of service conditions to which they may be exposed during storage, transportation, and eventual geologic disposal. A summary of selected parts of the content of this document is provided. Waste glass alterations in a geologic repository may include corrosion of the glass network due to groundwater and/or water vapor contact. Experimental testing results are described and interpreted in terms of the underlying chemical reactions and physical processes involved. The status of mechanistic modeling, which can be used for long-term predictions, is described and the remaining uncertainties associated with long-term simulations are summarized

Preliminary analysis of projected construction employment effects of building the defense waste processing facility at the Savannah River Plant

International Nuclear Information System (INIS)

Garey, R.B.; Blair, L.M.; Craig, R.L.; Stevenson, W.

1981-09-01

This study estimates the probable effects of constructing the DWPF on the surrounding labor markets. Analyses are based on data from the local and regional labor markets, information from experts on local construction activities, information on the labor requirements of the Vogtle Power Plant (two nuclear reactors) being built by Georgia Power Company in Burke County, Georgia, and an econometric model of the construction labor market, based on several surveys of workers building three Tennessee Valley Authority nuclear power plants. The results of this study are reported in three parts. In Part I, completed in May 1980, we describe the 1979 (base year) employment levels within the local and regional labor markets surrounding SRP, from which most DWPF construction workers are likely to be drawn. In Part II, completed in June 1980, we define the four local sources of construction employment that will compete for craftsmen when DWPF is built. Also in Part II, most of the projected impacts of the DWPF reference immobilization alternative (one of several alternatives that may be chosen) are reported. Several construction schedules and labor demand scenarios for the reference alternative are considered. In Part III, completed in January 1981, most of the estimated impacts of the DWPF alternative referred to as the staged process alternative are reported. Several construction schedules and labor demand scenarios for this alternative are considered

Glass: a candidate engineered material for management of high level nuclear waste

International Nuclear Information System (INIS)

Mishra, R.K.; Kaushik, C.P.

2011-01-01

While the commercial importance of glass is generally recognized, a few people are aware of extremely wide range of glass formulations that can be made and of the versatility of this engineered material. Some of the recent developments in the field of glass leading to various technological applications include glass fiber reinforcement of cement to give new building materials, substrates for microelectronics circuitry in form of semiconducting glasses, nuclear waste immobilization and specific medical applications. The present paper covers fundamental understanding of glass structure and its application for immobilization of high level radioactive liquid waste. High level radioactive liquid waste (HLW) arising during reprocessing of spent fuel are immobilized in sodium borosilicate glass matrix developed indigenously. Glass compositions are modified according to the composition of HLW to meet the criteria of desirable properties in terms. These glass matrices have been characterized for different properties like homogeneity, chemical durability, thermal stability and radiation stability. (author)

Basic approach to the disposal of low level radioactive waste generated from nuclear reactors containing comparatively high radioactivity

International Nuclear Information System (INIS)

Moriyama, Yoshinori

1998-01-01

Low level radioactive wastes (LLW) generated from nuclear reactors are classified into three categories: LLW containing comparatively high radioactivity; low level radioactive waste; very low level radioactive waste. Spent control rods, part of ion exchange resin and parts of core internals are examples of LLW containing comparatively high radioactivity. The Advisory Committee of Atomic Energy Commission published the report 'Basic Approach to the Disposal of LLW from Nuclear Reactors Containing Comparatively High Radioactivity' in October 1998. The main points of the proposed concept of disposal are as follows: dispose of underground deep enough not be disturb common land use (e.g. 50 to 100 m deep); dispose of underground where radionuclides migrate very slowly; dispose of with artificial engineered barrier which has the same function as the concrete pit; control human activities such as land use of disposal site for a few hundreds years. (author)

High-level waste solidification system for the Western New York Nuclear Service Center

International Nuclear Information System (INIS)

Carrell, J.R.; Holton, L.K.; Siemens, D.H.

1982-01-01

A preconceptual design for a waste conditioning and solidification system for the immobilization of the high-level liquid wastes (HLLW) stored at the Western New York Nuclear Service Center (WNYNSC), West Valley, New York was completed in 1981. The preconceptual design was conducted as part of the Department of Energy's (DOE) West Valley Demonstration Project, which requires a waste management demonstration at the WNYNSC. This paper summarizes the bases, assumptions, results and conclusions of the preconceptual design study

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

International Nuclear Information System (INIS)

McDonell, W.R.

1982-04-01

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

Mechanical stability of bentonite buffer system for high level nuclear waste

Energy Technology Data Exchange (ETDEWEB)

Lempinen, A. [Helsinki Univ. of Technology, Espoo (Finland). Lab. of Theoretical and Applied Mechanics

1998-05-01

According to present plans, high level nuclear waste in Finland is going to be disposed of in bedrock at a depth of several hundred metres. The spent fuel containers will be placed in boreholes drilled in the floors of deposition tunnels with engineered clay buffer, which is made of bentonite blocks. The tunnels will be filled with a mixture of bentonite and crushed rock. For stability calculations a thermomechanical model for compressed bentonite is needed. In the study a thermomechanically consistent model for reversible processes for swelling clays is presented. Preliminary calculations were performed and they show that uncertainty in material parameter values causes significantly different results. Therefore, measurements that are consistent with the model are needed 12 refs.

Mechanical stability of bentonite buffer system for high level nuclear waste

International Nuclear Information System (INIS)

Lempinen, A.

1998-05-01

According to present plans, high level nuclear waste in Finland is going to be disposed of in bedrock at a depth of several hundred metres. The spent fuel containers will be placed in boreholes drilled in the floors of deposition tunnels with engineered clay buffer, which is made of bentonite blocks. The tunnels will be filled with a mixture of bentonite and crushed rock. For stability calculations a thermomechanical model for compressed bentonite is needed. In the study a thermomechanically consistent model for reversible processes for swelling clays is presented. Preliminary calculations were performed and they show that uncertainty in material parameter values causes significantly different results. Therefore, measurements that are consistent with the model are needed

Site investigations for final disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Aeikaes, T.; Laine, T.

1982-12-01

Research concerning disposal of high-level nuclear waste of the Industrial Power Company Ltd has focused on deep underground disposal in Finnish precambrian bedrock. The present target is to have a repository for high-level waste in operation by 2020. Selection of the repository site is based on site investigations. In addition to geosciences, selection of appropriate site includes many branches of studies; engineering, safety analysis, ecology, transport, demography etc. The investigations required for site selection for high-level waste have been arranged in a sequence of four phases. The aim of the phases is that investigations become more and more detailed as the selection process continues. Phase I of the investigations is the characterization of potential areas. This comprises establishment of criteria for site selection and identification of areas that meet selection criteria. Objective of these studies is to determine areas for phase II field investigations. The studies are largely made by reviewing existing data and remote-sensing techniques. Phase II field investigations will be undertaken between 1986-1992. The number of potential candidates for repository site is reduced to few preferred areas by preceeding generic study. The site selection process culminates in phase III in site confirmation studies carried out at 2...3 most suitable sites during 1992-2010. This is then followed by phase IV, which comprises very detailed investigations at the selected site. An alternative for these investigations is to undertake them by using pilot shaft and drifts. Active development is taking place in all phases concerning investigation methods, criteria, parameters, data processing and modelling. The applicability of the various investigation methods and techniques is tested in a deep borehole in phase I. The co-operation with countries with similar geological conditions makes it possible to compare results obtained by different techniques

Legal and judicial perspectives on the disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Muntzing, L.M.

1985-01-01

Nuclear Waste Policy Act of 1982 creates a maze of legal requirements that are complex and subject to differing interpretations. The intervention of the courts should be anticipated as varying interests dispute the correct path through the maze to high-level waste disposal. A significant number of legal issues and unsettled questions exist that will need to be resolved. The challenge will be to minimize the delays that legal conflicts can produce. This can be helped by resolving uncertainties and conflicts early before they are on the critical path or negotiating resolutions, normally a difficult and costly endeavor

Rheological Characterization of Unusual DWPF Slurry Samples

International Nuclear Information System (INIS)

Koopman, D. C.

2005-01-01

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

Improved polyphase ceramic form for high-level defense nuclear waste

International Nuclear Information System (INIS)

Harker, A.B.; Morgan, P.E.D.; Clarke, D.R.; Flintoff, J.J.; Shaw, T.M.

1983-01-01

An improved ceramic nuclear waste form and fabrication process have been developed using simulated Savannah River Plant defense high-level waste compositions. The waste form provides flexibility with respect to processing conditions while exhibiting superior resistance to ground water leaching than other currently proposed forms. The ceramic, consolidated by hot-isostatic pressing at 1040 0 C and 10,000 psi, is composed of six major phases, nepheline, zirconolite, a murataite-type cubic phase, magnetite-type spinel, a magnetoplumbite solid solution, and perovskite. The waste form provides multiple crystal lattice sites for the waste elements, minimizes amorphous intergranular material, and can accommodate waste loadings in excess of 60 wt %. The fabrication of the ceramic can be accomplished with existing manufacturing technology and eliminates the effects of radionuclide volatilization and off-gas induced corrosion experienced with the molten processes for vitreous form production

Thermo-mechanical analysis of high level nuclear wastes in granite

International Nuclear Information System (INIS)

Millard, A.; Guri, G.; Raimbault, M.

1991-01-01

In order to appraise the safety of a storage of high level nuclear wastes in rock masses, it is necessary to assess, among other features, the thermo-mechanical behaviour of the host rock for long periods (thousands of years). In France, four different media are considered as potential host rocks: granite, shale, salt, clay. The present paper is devoted to some analysis of a generic storage configuration in granite. The case of a rock mass without any major fault has been considered. The granite is modelled by means of an elastic fracturing model (no tension type). The results obtained show that some fissures, induced by the heat generation, develop mainly above the repository. The opening of the fissures, within the frame of the adopted hypothesis, have not a strong influence on the rock mass, as a geological barrier for the radionuclides. (author)

Process for solidifying high-level nuclear waste

Science.gov (United States)

Ross, Wayne A.

1978-01-01

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

On-site storage of high level nuclear waste: Attitudes and perceptions of local residents

International Nuclear Information System (INIS)

Bassett, G.W. Jr.; Jenkins-Smith, H.C.; Silva, C.

1996-01-01

No public policy issue has been as difficult as high-level nuclear waste. Debates continue regarding Yucca Mountain as a disposal site, and - more generally - the appropriateness of geologic disposal and the need to act quickly. Previous research has focused on possible social, political, and economic consequences of a facility in Nevada. Impacts have been predicted to be potentially large and to emanate mainly from stigmatization of the region due to increased perceptions of risk. Analogous impacts from leaving waste at power plants have been either ignored or assumed to be negligible. This paper presents survey results on attitudes of residents in three countries where nuclear waste is currently stored. Topics include perceived risk, knowledge of nuclear waste and radiation, and impacts on jobs, tourism, and housing values from leaving waste on site. Results are similar to what has been reported for Nevada; the public is concerned about possible adverse effects from on-site storage of waste. 24 refs., 7 figs., 5 tabs

Can Sisyphus succeed? Getting U.S. high-level nuclear waste into a geological repository.

Science.gov (United States)

North, D Warner

2013-01-01

The U.S. government has the obligation of managing the high-level radioactive waste from its defense activities and also, under existing law, from civilian nuclear power generation. This obligation is not being met. The January 2012 Final Report from the Blue Ribbon Commission on America's Nuclear Future provides commendable guidance but little that is new. The author, who served on the federal Nuclear Waste Technical Review Board from 1989 to 1994 and subsequently on the Board on Radioactive Waste Management of the National Research Council from 1994 to 1999, provides a perspective both on the Commission's recommendations and a potential path toward progress in meeting the federal obligation. By analogy to Sisyphus of Greek mythology, our nation needs to find a way to roll the rock to the top of the hill and have it stay there, rather than continuing to roll back down again. © 2012 Society for Risk Analysis.

On-site storage of high level nuclear waste: attitudes and perceptions of local residents.

Science.gov (United States)

Bassett, G W; Jenkins-Smith, H C; Silva, C

1996-06-01

No public policy issue has been as difficult as high-level nuclear waste. Debates continue regarding Yucca Mountain as a disposal site, and-more generally-the appropriateness of geologic disposal and the need to act quickly. Previous research has focused on possible social, political, and economic consequences of a facility in Nevada. Impacts have been predicted to be potentially large and to emanate mainly from stigmatization of the region due to increased perceptions of risk. Analogous impacts from leaving waste at power plants have been either ignored or assumed to be negligible. This paper presents survey results on attitudes of residents in three counties where nuclear waste is currently stored. Topics include perceived risk, knowledge of nuclear waste and radiation, and impacts on jobs, tourism, and housing values from leaving waste on site. Results are similar to what has been reported for Nevada; the public is concerned about possible adverse effects from on-site storage of waste.

Status of nuclear fuel reprocessing, spent fuel storage, and high-level waste disposal. Nuclear Fuel Cycle Committee, California Energy Resources Conservation and Development Commission. Draft report

International Nuclear Information System (INIS)

Anon.

1978-01-01

An analysis of the current status of technologies and issues in the major portions of the back-end of the nuclear fuel cycle is presented. The discussion on nuclear fuel reprocessing covers the reprocessing requirement, reprocessing technology assessment, technology for operation of reprocessing plants, and approval of reprocessing plants. The chapter devoted to spent fuel storage covers the spent fuel storge problem, the legislative response, options for maintaining full core discharge capacity, prospective availability of alterntive storage options, and the outlook for California. The existence of a demonstrated, developed high-level waste disposal technology is reviewed. Recommendations for Federal programs on high-level waste disposal are made

Retrievability of high level waste and spent nuclear fuel. Proceedings of an international seminar

International Nuclear Information System (INIS)

2000-12-01

The possibility of retrieving spent nuclear fuel or reprocessing high-level radioactive wastes placed in geological repositories is an issue that has attracted increased attention during the past few years, not only among technical experts but also among politicians at different levels, environmental organisations and other interested representatives of the public. This publication contains the presented invited papers, an edited record of the discussions and some concluding remarks. The seminar addressed a wide range of aspects of retrievability including technical options; public acceptance; ethical aspects; long term monitoring and cost considerations; safety and regulatory aspects. Each of the presented papers was indexed separately

Retrievability of high level waste and spent nuclear fuel. Proceedings of an international seminar

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-12-01

The possibility of retrieving spent nuclear fuel or reprocessing high-level radioactive wastes placed in geological repositories is an issue that has attracted increased attention during the past few years, not only among technical experts but also among politicians at different levels, environmental organisations and other interested representatives of the public. This publication contains the presented invited papers, an edited record of the discussions and some concluding remarks. The seminar addressed a wide range of aspects of retrievability including technical options; public acceptance; ethical aspects; long term monitoring and cost considerations; safety and regulatory aspects. Each of the presented papers was indexed separately.

Long-term high-level waste technology program

International Nuclear Information System (INIS)

1980-04-01

The Department of Energy (DOE) is conducting a comprehensive program to isolate all US nuclear wastes from the human environment. The DOE Office of Nuclear Energy - Waste (NEW) has full responsibility for managing the high-level wastes resulting from defense activities and additional responsiblity for providing the technology to manage existing commercial high-level wastes and any that may be generated in one of several alternative fuel cycles. Responsibilities of the Three Divisions of DOE-NEW are shown. This strategy document presents the research and development plan of the Division of Waste Products for long-term immobilization of the high-level radioactive wastes resulting from chemical processing of nuclear reactor fuels and targets. These high-level wastes contain more than 99% of the residual radionuclides produced in the fuels and targets during reactor operations. They include essentially all the fission products and most of the actinides that were not recovered for use

Review of Statistical Analyses Resulting from Performance of HLDWD-DWPF-005

International Nuclear Information System (INIS)

Beck, R.S.

1997-01-01

The Engineering Department at the Defense Waste Processing Facility (DWPF) has reviewed two reports from the Statistical Consulting Section (SCS) involving the statistical analysis of test results for analysis of small sample inserts (references 1 ampersand 2). The test results cover two proposed analytical methods, a room temperature hydrofluoric acid preparation (Cold Chem) and a sodium peroxide/sodium hydroxide fusion modified for insert samples (Modified Fusion). The reports support implementation of the proposed small sample containers and analytical methods at DWPF. Hydragard sampler valve performance was typical of previous results (reference 3). Using an element from each major feed stream. lithium from the frit and iron from the sludge, the sampler was determined to deliver a uniform mixture in either sample container.The lithium to iron ratios were equivalent for the standard 15 ml vial and the 3 ml insert.The proposed method provide equivalent analyses as compared to the current methods. The biases associated with the proposed methods on a vitrified basis are less than 5% for major elements. The sum of oxides for the proposed method compares favorably with the sum of oxides for the conventional methods. However, the average sum of oxides for the Cold Chem method was 94.3% which is below the minimum required recovery of 95%. Both proposed methods, cold Chem and Modified Fusion, will be required at first to provide an accurate analysis which will routinely meet the 95% and 105% average sum of oxides limit for Product Composition Control System (PCCS).Issued to be resolved during phased implementation are as follows: (1) Determine calcine/vitrification factor for radioactive feed; (2) Evaluate covariance matrix change against process operating ranges to determine optimum sample size; (3) Evaluate sources for low sum of oxides; and (4) Improve remote operability of production versions of equipment and instruments for installation in 221-S.The specifics of

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

International Nuclear Information System (INIS)

1983-06-01

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

Ecological risk assessment of deep geological disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Hart, D.R.; Lush, D.L.; Acton, D.W.

1993-01-01

Contaminant fate and transport models, radiological dosimetry models, chemical dose-response models and population dynamic models were used to estimate ecological risks to moose and brook trout populations arising from a proposed high-level nuclear waste repository. Risks from potential contaminant releases were compared with risks from physical habitat alteration in constructing a repository and service community, and with risks from increased hunting and fish pressure in the area. For a reference environment typical of a proposed location somewhere in the Canadian Shield, preliminary results suggest that the population consequences of contaminant release will be minor relative to those of habitat alteration and natural resource use

High Sodium Simulant Testing To Support SB8 Sludge Preparation

International Nuclear Information System (INIS)

Newell, J. D.

2012-01-01

Scoping studies were completed for high sodium simulant SRAT/SME cycles to determine any impact to CPC processing. Two SRAT/SME cycles were performed with simulant having sodium supernate concentration of 1.9M at 130% and 100% of the Koopman Minimum Acid requirement. Both of these failed to meet DWPF processing objectives related to nitrite destruction and hydrogen generation. Another set of SRAT/SME cycles were performed with simulant having a sodium supernate concentration of 1.6M at 130%, 125%, 110%, and 100% of the Koopman Minimum Acid requirement. Only the run at 110% met DWPF processing objectives. Neither simulant had a stoichiometric factor window of 30% between nitrite destruction and excessive hydrogen generation. Based on the 2M-110 results it was anticipated that the 2.5M stoichiometric window for processing would likely be smaller than from 110-130%, since it appeared that it would be necessary to increase the KMA factor by at least 10% above the minimum calculated requirement to achieve nitrite destruction due to the high oxalate content. The 2.5M-130 run exceeded the DWPF hydrogen limits in both the SRAT and SME cycle. Therefore, testing of this wash endpoint was halted. This wash endpoint with this minimum acid requirement and mercury-noble metal concentration profile appears to be something DWPF should not process due to an overly narrow window of stoichiometry. The 2M case was potentially processable in DWPF, but modifications would likely be needed in DWPF such as occasionally accepting SRAT batches with undestroyed nitrite for further acid addition and reprocessing, running near the bottom of the as yet ill-defined window of allowable stoichiometric factors, potentially extending the SRAT cycle to burn off unreacted formic acid before transferring to the SME cycle, and eliminating formic acid additions in the frit slurry

An optimized approach towards the treatment of high level liquid waste in the nuclear cycle

International Nuclear Information System (INIS)

Maio, V.; Todd, T.; Law, J.; Roach, J.; Sabharwall, P.

2006-01-01

Full text: One key long-standing issue that must be overcome to realize the successful growth of nuclear power is an economical, politically acceptable, stakeholder-compatible, and technically feasible resolution pertaining to the safe treatment and disposal of high-level liquid radioactive waste (HLLW). In addition to spent nuclear reactor fuel, HLLW poses a unique challenge in regard to environmental and security concerns, since future scenarios for a next generation of domestic and commercialized nuclear fuel cycle infrastructures must include reprocessing - the primary source of HLLW-to ensure the cost effectiveness of nuclear power as well as mitigate any threats as related to proliferation. Past attempts to immobilize HLLW - generated by both the weapons complex and the commercial power sector-have been plagued by an inability to convince the public and some technical peer reviewers that any proposed geological disposal sites (e.g., Yucca Mountain) can accommodate and contain the HLLW for a period of geological time equivalent to ten fold the radiological half-life of the longest lived of the actinides remaining after reprocessing. The paper explores combined equipment and chemical processing approaches for advancing and economizing the immobilization of high level liquid waste to ensure its long term durability, its decoupling from the unknown behavior of the repository over long geological time periods, and its economical formulation as required for the nuclear fuel cycle of the future. One approach involves the investigation of crystalline based waste forms as opposed to the glass/amorphous based waste forms, and how recent developments in crystalline forms show promise in sequestering the long lived actinides for over tens of millions of years. Another approach -compatible with the first- involves the use of an alternative melter technology-the Cold Crucible Induction Melter (CCIM)- to overcome the engineering material problems of Joule Heated Meters (JHM

Water level monitoring device in nuclear reactor

International Nuclear Information System (INIS)

Miura, Kiyohide; Otake, Tomohiro.

1988-01-01

Purpose: To monitor the water level in a pressure vessel of BWR type nuclear reactors at high accuracy by improving the compensation functions. Constitution: In the conventional water level monitor in a nuclear reactor, if the pressure vessel is displaced by the change of the pressure in the reactor or the temperature of the reactor water, the relative level of the reference water head in a condensation vessel is changed to cause deviation between the actual water level and the indicated water level to reduce the monitoring accuracy. According to the invention, means for detecting the position of the reference water head and means for detection the position in the condensation vessel are disposed to the pressure vessel. Then, relative positional change between the condensation vessel and the reference water head is calculated based on detection sinals from both of the means. The water level is compensated and calculated by water level calculation means based on the relative positional change, water level signals from the level gage and the pressure signals from the pressure gage. As a result, if the pressure vessel is displaced due to the change of the temperature or pressure, it is possible to measure the reactor water level accurately thereby remakably improve the reliability for the water level control in the nuclear reactor. (Horiuchi, T.)

Seismic design ampersand analysis considerations for high level nuclear waste repositories

International Nuclear Information System (INIS)

Hossain, Q.A.

1993-01-01

A high level nuclear waste repository, like the one at Nevada's Yucca Mountain that is being investigated for site suitability, will have some unique seismic design and analysis considerations. These are discussed, and a design philosophy that can rationally account for the unique performance objectives of such facilities is presented. A case is made for the use of DOE's performance goal-based seismic design and evaluation methodology that is based on a hybrid open-quotes deterministicclose quotes and open-quotes probabilisticclose quotes concept. How and to what extent this methodology should be modified to adopt it for a potential site like Yucca Mountain is also outlined. Finally, the issue of designing for seismic fault rupture is discussed briefly, and the desirability of using the proposed seismic design philosophy in fault rupture evaluation is described

Seismic design and analysis considerations for high level nuclear waste repositories

International Nuclear Information System (INIS)

Hossain, Q.A.

1993-01-01

A high level nuclear waste repository, like the one at Nevada's Yucca Mountain that is being investigated for site suitability, will have some unique seismic design and analysis considerations. These are discussed, and a design philosophy that can rationally account for the unique performance objectives of such facilities is presented. A case is made for the use of DOE's performance goal-based seismic design and evaluation methodology that is based on a hybrid ''deterministic'' and ''probabilistic'' concept. How and to what extent this methodology should be modified to adopt it for a potential site like Yucca Mountain is also outlined. Finally, the issue of designing for seismic fault rupture is discussed briefly, and the desirability of using the proposed seismic design philosophy in fault rupture evaluation is described

Disposal of high level and intermediate level radioactive wastes

International Nuclear Information System (INIS)

Flowers, R.H.

1991-01-01

The waste products from the nuclear industry are relatively small in volume. Apart from a few minor gaseous and liquid waste streams, containing readily dispersible elements of low radiotoxicity, all these products are processed into stable solid packages for disposal in underground repositories. Because the volumes are small, and because radioactive wastes are latecomers on the industrial scene, a whole new industry with a world-wide technological infrastructure has grown up alongside the nuclear power industry to carry out the waste processing and disposal to very high standards. Some of the technical approaches used, and the Regulatory controls which have been developed, will undoubtedly find application in the future to the management of non-radioactive toxic wastes. The repository site outlined would contain even high-level radioactive wastes and spent fuels being contained without significant radiation dose rates to the public. Water pathway dose rates are likely to be lowest for vitrified high-level wastes with spent PWR fuel and intermediate level wastes being somewhat higher. (author)

Current high-level waste solidification technology

International Nuclear Information System (INIS)

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

1976-01-01

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

Steam stripping of polycyclic aromatics from simulated high-level radioactive waste

International Nuclear Information System (INIS)

Lambert, D.P.; Shah, H.B.; Young, S.R.; Edwards, R.E.; Carter, J.T.

1992-01-01

The Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS) will be the United States' first facility to process High Level radioactive Waste (HLW) into a borosilicate glass matrix. The removal of aromatic precipitates by hydrolysis, evaporation, liquid-liquid extraction and decantation will be a key step in the processing of the HLW. This step, titled the Precipitate Hydrolysis Process, has been demonstrated by the Savannah River Technology Center with the Precipitate Hydrolysis Experimental Facility (PHEF). The mission of the PHEF is to demonstrate processing of simulated high level radioactive waste which contains tetraphenylborate precipitates and nitrite. Aqueous washing or nitrite destruction is used to reduce nitrite. Formic acid with a copper catalyst is used to hydrolyze tetraphenylborate (TPB). The primary offgases are benzene, carbon dioxide, nitrous oxide, and nitric oxide. Hydrolysis of TPB in the presence of nitrite results in the production of polycyclic aromatics and aromatic amines (referred as high boiling organics) such as biphenyl, diphenylamine, terphenyls etc. The decanter separates the organic (benzene) and aqueous phase, but the high boiling organic separation is difficult. This paper focuses on the evaluation of the operating strategies, including steam stripping, to maximize the removal of the high boiling organics from the aqueous stream. Two areas were investigated, (1) a stream stripping comparison of the late wash flowsheet to the HAN flowsheet and (2) the extraction performance of the original decanter to the new decanter. The focus of both studies was to minimize the high boiling organic content of the Precipitate Hydrolysis Aqueous (PHA) product in order to minimize downstream impacts caused by organic deposition

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

Energy Technology Data Exchange (ETDEWEB)

Stefanovsky, Sergey; Marra, James; Lebedev, Vladimir

2014-01-13

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

Implementation of flowsheet change to minimize hydrogen and ammonia generation during chemical processing of high level waste in the Defense Waste Processing Facility

Energy Technology Data Exchange (ETDEWEB)

Lambert, Dan P. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Woodham, Wesley H. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Williams, Matthew S. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Newell, J. David [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Luther, Michelle C. [Auburn Univ., AL (United States); Brandenburg, Clayton H. [Univ.of South Carolina, Columbia, SC (United States)

2016-09-27

Testing was completed to develop a chemical processing flowsheet for the Defense Waste Processing Facility (DWPF), designed to vitrify and stabilize high level radioactive waste. DWPF processing uses a reducing acid (formic acid) and an oxidizing acid (nitric acid) to rheologically thin the slurry and complete the necessary acid base and reduction reactions (primarily mercury and manganese). Formic acid reduces mercuric oxide to elemental mercury, allowing the mercury to be removed during the boiling phase of processing through steam stripping. In runs with active catalysts, formic acid can decompose to hydrogen and nitrate can be reduced to ammonia, both flammable gases, due to rhodium and ruthenium catalysis. Replacement of formic acid with glycolic acid eliminates the generation of rhodium- and ruthenium-catalyzed hydrogen and ammonia. In addition, mercury reduction is still effective with glycolic acid. Hydrogen, ammonia and mercury are discussed in the body of the report. Ten abbreviated tests were completed to develop the operating window for implementation of the flowsheet and determine the impact of changes in acid stoichiometry and the blend of nitric and glycolic acid as it impacts various processing variables over a wide processing region. Three full-length 4-L lab-scale simulations demonstrated the viability of the flowsheet under planned operating conditions. The flowsheet is planned for implementation in early 2017.

Fabrication and closure development of corrosion resistant containers for Nevada's Yucca Mountain high-level nuclear waste repository

International Nuclear Information System (INIS)

Russell, E.W.; Nelson, T.A.; Domian, H.A.; LaCount, D.F.; Robitz, E.S.; Stein, K.O.

1989-11-01

US Congress and the President have determined that the Yucca Mountain site in Nevada is to be characterized to determine its suitability for construction of the first US high-level nuclear waste repository. Work in connection with this site is carried out within the Yucca Mountain Project (YMP). Lawrence Livermore National Laboratory (LLNL) has the responsibility for designing, developing, and projecting the performance of the waste package for the permanent storage of high-level nuclear waste. Babcock ampersand Wilcox (B ampersand W) is involved with the YMP as a subcontractor to LLNL. B ampersand W's role is to recommend and demonstrate a method for fabricating the metallic waste container and a method for performing the final closure of the container after it has been filled with waste. Various fabrication and closure methods are under consideration for the production of containers. This paper presents progress to date in identifying and evaluating the candidate manufacturing processes. 2 refs., 2 figs., 4 tabs

Bridging nuclear safety, security and safeguards at geological disposl of high level radioactive waste and spent nuclear fuel

International Nuclear Information System (INIS)

Niemeyer, Irmgard; Deissmann, Guido; Bosbach, Dirk

2016-01-01

Findings and recommendations: • Further R&D needed to identify concepts, methods and technologies that would be best suited for the holistic consideration of safety, security and safeguards provisions of geological disposal. • 3S ‘toolbox’, including concepts, methods and technologies for: ■ material accountancy, ■ measurement techniques for spent fuel verification, ■ containment and surveillance, ■ analysis of open source information, ■ environmental sampling and monitoring, ■ continuity of knowledge, ■ design implications. •: Bridging safety, security and safeguards in research funding and research activities related to geological disposal of high-level radioactive waste and spent nuclear fuel.

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in an cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rack. These waste containers are vertically emplaced in the borehole 300 meters below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3--4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions

Nuclear level density parameter 's dependence on angular momentum

International Nuclear Information System (INIS)

Aggarwal, Mamta; Kailas, S.

2009-01-01

Nuclear level densities represent a very important ingredient in the statistical Model calculations of nuclear reaction cross sections and help to understand the microscopic features of the excited nuclei. Most of the earlier experimental nuclear level density measurements are confined to low excitation energy and low spin region. A recent experimental investigation of nuclear level densities in high excitation energy and angular momentum domain with some interesting results on inverse level density parameter's dependence on angular momentum in the region around Z=50 has motivated us to study and analyse these experimental results in a microscopic theoretical framework. In the experiment, heavy ion fusion reactions are used to populate the excited and rotating nuclei and measured the α particle evaporation spectra in coincidence with ray multiplicity. Residual nuclei are in the range of Z R 48-55 with excitation energy range 30 to 40 MeV and angular momentum in 10 to 25. The inverse level density parameter K is found to be in the range of 9.0 - 10.5 with some exceptions

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

Energy Technology Data Exchange (ETDEWEB)

Shafer, A.

2010-05-05

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

Synroc - a multiphase ceramic for high level nuclear waste immobilisation

International Nuclear Information System (INIS)

Reeve, K.D.; Vance, E.R.; Hart, K.P.; Smith, K.L.; Lumpkin, G.R.; Mercer, D.J.

1992-01-01

Many natural minerals - particularly titanates - are very durable geochemically, having survived for millions of years with very little alteration. Moreover, some of these minerals have quantitatively retained radioactive elements and their daughter products over this time. The Synroc concept mimics nature by providing an all-titanate synthetic mineral phase assemblage to immobilise high level waste (HLW) from nuclear fuel reprocessing operations for safe geological disposal. In principle, many chemically hazardous inorganic wastes arising from industry could also be immobilised in highly durable ceramics and disposed of geologically, but in practice the cost structure of most industries is such that lower cost waste management solutions - for example, the development of reusable by-products or the use of cements rather than ceramics - have to be devised. In many thousands of aqueous leach tests at ANSTO, mostly at 70-90 deg C, Synroc has been shown to be exceptionally durable. The emphases of the recent ANSTO program have been on tailoring of the Synroc composition to varying HLW compositions, leach testing of Synroc containing radioactive transuranic actinides, study of leaching mechanisms by SEM and TEM, and the development and costing of a conceptual fully active Synroc fabrication plant design. A summary of recent results on these topics will be presented. 29 refs., 4 figs

Environmental program planning for the proposed high-level nuclear waste repository at Yucca Mountain, Nevada

International Nuclear Information System (INIS)

1987-08-01

This report was prepared to illustrate the policy and actions that the State of Nevada believe are required to assure that the quality of the environment is adequately considered during the course of the DOE work at the proposed high-level nuclear waste repository at Yucca Mountain. The report describes the DOE environmental program and the studies planned by NWPO to reflect the State's position toward environmental protection. 41 refs., 2 figs., 11 tabs

Disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Glasby, G.P.

1977-01-01

Although controversy surrounding the possible introduction of nuclear power into New Zealand has raised many points including radiation hazards, reactor safety, capital costs, sources of uranium and earthquake risks on the one hand versus energy conservation and alternative sources of energy on the other, one problem remains paramount and is of global significance - the storage and dumping of the high-level radioactive wastes of the reactor core. The generation of abundant supplies of energy now in return for the storage of these long-lived highly radioactive wastes has been dubbed the so-called Faustian bargain. This article discusses the growth of the nuclear industry and its implications to high-level waste disposal particularly in the deep-sea bed. (auth.)

Other-than-high-level waste

International Nuclear Information System (INIS)

Bray, G.R.

1976-01-01

The main emphasis of the work in the area of partitioning transuranic elements from waste has been in the area of high-level liquid waste. But there are ''other-than-high-level wastes'' generated by the back end of the nuclear fuel cycle that are both large in volume and contaminated with significant quantities of transuranic elements. The combined volume of these other wastes is approximately 50 times that of the solidified high-level waste. These other wastes also contain up to 75% of the transuranic elements associated with waste generated by the back end of the fuel cycle. Therefore, any detailed evaluation of partitioning as a viable waste management option must address both high-level wastes and ''other-than-high-level wastes.''

High-level radioactive waste repositories site selection plan

International Nuclear Information System (INIS)

Castanon, A.; Recreo, F.

1985-01-01

A general vision of the high level nuclear waste (HLNW) and/or nuclear spent fuel facilities site selection processes is given, according to the main international nuclear safety regulatory organisms quidelines and the experience from those countries which have reached a larger development of their national nuclear programs. (author)

Natural geochemical analogues of the near field of high-level nuclear waste repositories

International Nuclear Information System (INIS)

Apps, J.A.

1995-01-01

United States practice has been to design high-level nuclear waste (HLW) geological repositories with waste densities sufficiently high that repository temperatures surrounding the waste will exceed 100 degrees C and could reach 250 degrees C. Basalt and devitrified vitroclastic tuff are among the host rocks considered for waste emplacement. Near-field repository thermal behavior and chemical alteration in such rocks is expected to be similar to that observed in many geothermal systems. Therefore, the predictive modeling required for performance assessment studies of the near field could be validated and calibrated using geothermal systems as natural analogues. Examples are given which demonstrate the need for refinement of the thermodynamic databases used in geochemical modeling of near-field natural analogues and the extent to which present models can predict conditions in geothermal fields

Computer-based control of nuclear power information systems at international level

International Nuclear Information System (INIS)

Boniface, Ekechukwu; Okonkwo, Obi

2011-01-01

In most highly industrialized countries of the world information plays major role in anti-nuclear campaign. Information and discussions on nuclear power need critical and objective analysis before the structured information presentation to the public to avoid bias anti-nuclear information on one side and neglect of great risk in nuclear power. This research is developing a computer-based information system for the control of nuclear power at international level. The system is to provide easy and fast information highways for the followings: (1) Low Regulatory dose and activity limit as level of high danger for individuals and public. (2) Provision of relevant technical or scientific education among the information carriers in the nuclear power countries. The research is on fact oriented investigation about radioactivity. It also deals with fact oriented education about nuclear accidents and safety. A standard procedure for dissemination of latest findings using technical and scientific experts in nuclear technology is developed. The information highway clearly analyzes the factual information about radiation risk and nuclear energy. Radiation cannot be removed from our environment. The necessity of radiation utilizations defines nuclear energy as two-edge sword. It is therefore, possible to use computer-based information system in projecting and dissemination of expert knowledge about nuclear technology positively and also to use it in directing the public on the safety and control of the nuclear energy. The computer-based information highway for nuclear energy technology is to assist in scientific research and technological development at international level. (author)

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Comprehensive data base of high-level nuclear waste glasses: September 1987 status report: Volume 1, Discussion and glass durability data

International Nuclear Information System (INIS)

Kindle, C.H.; Kreiter, M.R.

1987-12-01

The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is assembling a comprehensive data base (CDB) of experimental data collected for high-level nuclear waste package components. Data collected throughout the world are included in the data base; current emphasis is on waste glasses and their properties. The goal is to provide a data base of properties and compositions and an analysis of dominant property trends as a function of composition. This data base is a resource that nuclear waste producers, disposers, and regulators can use to compare properties of a particular high-level nuclear waste glass product with the properties of other glasses of similar compositions. Researchers may use the data base to guide experimental tests to fill gaps in the available knowledge or to refine empirical models. The data are incorporated into a computerized data base that will allow the data to be extracted based on, for example, glass composition or test duration. 3 figs

Overall review strategy for the Nuclear Regulatory Commission's High-Level Waste Repository Program

International Nuclear Information System (INIS)

Johnson, R.L.

1994-11-01

The Overall Review Strategy gives general guidance to the Nuclear Regulatory Commission staff for conducting it's license application and pre-license application reviews. These reviews are in support of the Commission's construction authorization decision for a geologic repository for the disposal of high-level radioactive waste. Objectives and strategies are defined that focus the staff's reviews on determining compliance with requirements of 10 CFR Part 60. These strategies define how the staff prioritizes its reviews on those key technical uncertainties considered to be most important to repository performance. Strategies also give guidance for developing, in an integrated way, the License Application Review Plan together with supporting performance assessments, analyses, and research

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

Peer review for high-level nuclear waste repositories: Generic technical position

International Nuclear Information System (INIS)

Altman, W.D.; Donnelly, J.P.; Kennedy, J.E.

1988-02-01

This document provides guidance on the use of the peer review process in the high-level nuclear waste repository program. The applicant must demonstrate in the license application that the applicable health, safety, and environmental regulations in 10 CFR Part 60 have been met. Confidence in the data used to support the license application is obtained through a quality assurance (AQ) program. Peer reviews may be used as part of the QA actions necessary to provide adequate confidence in the work being reviewed. Because of several unique conditions inherent to the geologic repository program, expert judgment will need to be utilized in assessing the adequacy of work. Peer reviews are a mechanism by which these judgments may be made. This document provides guidance on areas where a peer review is appropriate, the acceptability of peers, and the conduct and documentation of a peer review

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

International Nuclear Information System (INIS)

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

1983-11-01

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

Public opposition to the siting of the high-level nuclear waste repository: The importance of trust

International Nuclear Information System (INIS)

Pijawka, K.D.; Mushkatel, A.H.

1991-01-01

This paper examines several dimensions of public opposition to the proposed siting of the high-level nuclear waste repository at Yucca Mountain. In order to provide a context for the public's views of the repository in metropolitan Clark County, both governmental studies of the repository siting process are analyzed, as well as elements of the Nuclear Waste Policy Act. This analysis suggests that one potentially key component of the public's opposition to the siting, as well as their perceptions of risk of the facility, may be the result of a lack of trust in the Department of Energy. Empirical analysis of survey data collected in Nevada in 1988 confirms the strong relationship between political trust and repository risk perceptions

Attitudes and opposition in siting a high level nuclear waste repository

International Nuclear Information System (INIS)

Sjoeberg, L.; Viklund, M.; Truedsson, J.

1998-09-01

In Sweden, the Swedish Nuclear Fuel and Waste Management Company (SKB) handles all issues concerning nuclear waste, including the siting process, in which the final outcome is intended to be a repository for high level nuclear waste placed deep down in bedrock. The main objective of the siting process is to find a host community fulfilling two important conditions: the safety demands have been met and agreements with the municipality can be accomplished. Only in such municipalities, so-called feasibility studies will be conducted. After conducting general studies in the whole country, SKB, in October 1992, sent letters with information about the intended feasibility studies to all Swedish municipalities. As a result, feasibility studies are or have been considered - and in some cases also been conducted - in eleven Swedish municipalities up until 1998. These are the municipalities where the attitudes and opposition towards a feasibility study, and possibly a final repository, are studied. The discussion can be divided into three main parts: Management of the siting process; Inherent 'chaotic' processes and/or factors and risk perception. It is argued that two important problems could have been avoided at least partly: The citizens in many municipalities were uncertain of the relationship between a feasibility study and a final repository, and in many municipalities the citizens were afraid that the Government could overrule the municipal veto. Because of these fears, a common argument among the opponents of a feasibility study was: 'to be sure of not receiving a final repository, we say no to a feasibility study'. Some inherent factors, more or less prevalent in the municipalities as well as in society in general, may also partly explain the outcome of the siting process. The municipalities in which the debate has been heated, and where public support has been more difficult to reach, share some common characteristics. Esp. in the municipalities in the north of

Attitudes and opposition in siting a high level nuclear waste repository

Energy Technology Data Exchange (ETDEWEB)

Sjoeberg, L.; Viklund, M.; Truedsson, J

1998-09-01

In Sweden, the Swedish Nuclear Fuel and Waste Management Company (SKB) handles all issues concerning nuclear waste, including the siting process, in which the final outcome is intended to be a repository for high level nuclear waste placed deep down in bedrock. The main objective of the siting process is to find a host community fulfilling two important conditions: the safety demands have been met and agreements with the municipality can be accomplished. Only in such municipalities, so-called feasibility studies will be conducted. After conducting general studies in the whole country, SKB, in October 1992, sent letters with information about the intended feasibility studies to all Swedish municipalities. As a result, feasibility studies are or have been considered - and in some cases also been conducted - in eleven Swedish municipalities up until 1998. These are the municipalities where the attitudes and opposition towards a feasibility study, and possibly a final repository, are studied. The discussion can be divided into three main parts: Management of the siting process; Inherent `chaotic` processes and/or factors and risk perception. It is argued that two important problems could have been avoided at least partly: The citizens in many municipalities were uncertain of the relationship between a feasibility study and a final repository, and in many municipalities the citizens were afraid that the Government could overrule the municipal veto. Because of these fears, a common argument among the opponents of a feasibility study was: `to be sure of not receiving a final repository, we say no to a feasibility study`. Some inherent factors, more or less prevalent in the municipalities as well as in society in general, may also partly explain the outcome of the siting process. The municipalities in which the debate has been heated, and where public support has been more difficult to reach, share some common characteristics. Esp. in the municipalities in the north of

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

MATRIX 2 RESULTS OF THE FY07 ENHANCED DOE HIGH-LEVEL WASTE MELTER THROUGHPUT STUDIES AT SRNL

Energy Technology Data Exchange (ETDEWEB)

Raszewski, F; Tommy Edwards, T; David Peeler, D

2008-10-23

High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this study was to generate supplemental validation data that could be used to determine the applicability of the current liquidus temperature (TL) model to expanded DWPF glass composition regions of interest based on higher WLs. Two specific flowsheets were used in this study to provide such insight: (1) Higher WL glasses (45 and 50%) based on future sludge batches that have (and have not) undergone the Al-dissolution process. (2) Coupled operations supported by the Salt Waste Processing Facility (SWPF), which increase the TiO{sub 2} concentration in glass to greater than 2 wt%. Glasses were also selected to address technical issues associated with Al{sub 2}O{sub 3} solubility, nepheline formation, and homogeneity issues for coupled operations. A test matrix of 28 glass compositions was developed to provide insight into these issues. The glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), TL measurement and the Product Consistency Test (PCT). The results of this study are summarized below: (1) TiO{sub 2} concentrations up to {approx} 3.5 wt% were retained in DWPF type glasses, where retention is defined as the absence of crystalline TiO{sub 2} (i.e., unreacted or undissolved) in the as-fabricated glasses. Although this TiO{sub 2} content does not bound the projected SWPF high output flowsheet (up to 6 wt% TiO{sub 2} may be required in glass), these data demonstrate the potential for increasing the TiO{sub 2} limit in glass above the current limit of 2 wt

Study of nuclear level densities for exotic nuclei

International Nuclear Information System (INIS)

Nasrabadi, M. N.; Sepiani, M.

2012-01-01

Nuclear level density is one of the properties of nuclei with widespread applications in astrophysics and nuclear medicine. Since there has been little experimental and theoretical research on the study of nuclei which are far from stability line, studying nuclear level density for these nuclei is of crucial importance. Also, as nuclear level density is an important input for nuclear research codes, hence studying the methods for calculation of this parameter is essential. Besides introducing various methods and models for calculating nuclear level density for practical applications, we used exact spectra distribution (SPDM) for determining nuclear level density of two neutron and proton enriched exotic nuclei with the same mass number.

Siting the nation's first high-level nuclear waste repository: Social impacts for Utah

International Nuclear Information System (INIS)

Olshansky, S.J.

1986-01-01

The siting of a high-level nuclear waste repository in the United States has been an issue of great controversy, particulary for the states under consideration. In addition to concerns expressed about the geological stability of the proposed sites, numerous social issues have been raised by the general public--most of which have been addressed by the draft environmental impact statements. Among the social impacts raised by the Department of Engery and the general public, those receiving the greatest attention were the potential influence of the repository on local economics, tourism, and the health status of the local residents. The major issues of interest in the present study include 1) the effects of respondent knowledge of nuclear waste disposal issues on opinions of health effects and tourism, particularly as they are affected by visitation patterns, and 2) the effects of occupation and education (in particular) on knowledge of nuclear waste disposal issues and opinions on technical and non-technical aspects of siting the repository. Preliminary results indicate that only about 40 percent of the respondents have visited the national parks in southeastern Utah, but over 70 percent feel they are informed about the issues associated with siting the repository. Over 60 percent of the respondents were very concerned about the possible negative effect the repository could have on jobs, tourism, health effects, and environmental quality. Cross-tabulations indicate that the respondents self rating on knowledge of nuclear disposal issues has a statistically significant influence on responses to socioeconomic issues, yet the same self rating scale is significantly influenced by the frequency with which respondents have visited the national parks in southeastern Utah

Water level control for a nuclear steam generator

International Nuclear Information System (INIS)

Wen Tan

2011-01-01

Research highlights: → A water level control system for a nuclear steam generator (SG) is proposed. → The parameters of the control system are directly related to those of the plant model thus scheduling is easy to implement in practice. → The proposed gain-scheduled controller can achieve good performance at both low and high power levels. - Abstract: A water level control system for a nuclear steam generator (SG) is proposed. The control system consists of a feedback controller and a feedforward controller. The feedback controller is of first order, the feedforward controller is of second order, and parameters of the two controllers are directly related to the parameters of plant model thus scheduling is easy to implement in practice. Robustness and performance of the feedback and the feedforward controllers are analyzed in details and tuning of the two parameters of the controllers are discussed. Comparisons among a single robust controller, a multi-model controller and a gain-scheduled controller are studied. It is shown that the proposed gain-scheduled controller can achieve good performance at both low and high power levels.

DWPF coupled feed flowsheet material balance with batch one sludge and copper nitrate catalyst

Energy Technology Data Exchange (ETDEWEB)

Choi, A.S.

1993-09-28

The SRTC has formally transmitted a recommendation to DWPF to replace copper formate with copper nitrate as the catalyst form during precipitate hydrolysis [1]. The SRTC was subsequently requested to formally document the technical bases for the recommendation. A memorandum was issued on August 23, 1993 detailing the activities (and responsible individuals) necessary to address the impact of this change in catalyst form on process compatibility, safety, processibility environmental impact and product glass quality [2]. One of the activities identified was the preparation of a material balance in which copper nitrate is substituted for copper formate and the identification of key comparisons between this material balance and the current Batch 1 sludge -- Late Wash material balance [3].

International high-level radioactive waste repositories

International Nuclear Information System (INIS)

Lin, W.

1996-01-01

Although nuclear technologies benefit everyone, the associated nuclear wastes are a widespread and rapidly growing problem. Nuclear power plants are in operation in 25 countries, and are under construction in others. Developing countries are hungry for electricity to promote economic growth; industrialized countries are eager to export nuclear technologies and equipment. These two ingredients, combined with the rapid shrinkage of worldwide fossil fuel reserves, will increase the utilization of nuclear power. All countries utilizing nuclear power produce at least a few tens of tons of spent fuel per year. That spent fuel (and reprocessing products, if any) constitutes high-level nuclear waste. Toxicity, long half-life, and immunity to chemical degradation make such waste an almost permanent threat to human beings. This report discusses the advantages of utilizing repositories for disposal of nuclear wastes

Albedo Neutron Dosimetry in a Deep Geological Disposal Repository for High-Level Nuclear Waste.

Science.gov (United States)

Pang, Bo; Becker, Frank

2017-04-28

Albedo neutron dosemeter is the German official personal neutron dosemeter in mixed radiation fields where neutrons contribute to personal dose. In deep geological repositories for high-level nuclear waste, where neutrons can dominate the radiation field, it is of interest to investigate the performance of albedo neutron dosemeter in such facilities. In this study, the deep geological repository is represented by a shielding cask loaded with spent nuclear fuel placed inside a rock salt emplacement drift. Due to the backscattering of neutrons in the drift, issues concerning calibration of the dosemeter arise. Field-specific calibration of the albedo neutron dosemeter was hence performed with Monte Carlo simulations. In order to assess the applicability of the albedo neutron dosemeter in a deep geological repository over a long time scale, spent nuclear fuel with different ages of 50, 100 and 500 years were investigated. It was found out, that the neutron radiation field in a deep geological repository can be assigned to the application area 'N1' of the albedo neutron dosemeter, which is typical in reactors and accelerators with heavy shielding. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Unintentional stoppages of trucks hauling high-level nuclear waste in Nevada

International Nuclear Information System (INIS)

Glickman, T.S.

1991-01-01

Any unintentional stoppage of trucks hauling high-level nuclear waste in Nevada could intensify public concerns about the negative impacts of the Yucca Mountain Project. This paper provides estimates of the expected number and frequency of such stoppages under several alternative scenarios for the routing and volume of the anticipated truck shipments. At least twice as many truck stoppages are expected to occur on freeways as on non-freeways, and at least twice as many are expected to occur in rural areas as in other areas. The expected frequency of such stoppages in the state is estimated to range from about 3 per year to one every 3 1/2 years. About one in every 4 of the stoppages is expected to take place in Las Vegas and 3 out of every 4 are expected to be non-accidents such as mechanical disablements. The minimum duration of the stoppages is expected to be on the order of half an hour for disablements and an hour for accidents

Transmutation of high level nuclear waste in an accelerator driven system: towards a demonstration device of industrial interest (EUROTRANS)

International Nuclear Information System (INIS)

Knebel, Joachim U.; Ait Abderrahim, Hamid; Caron-Carles, Marylise

2010-01-01

The Integrated Project EUROTRANS (EURopean Research Programme for the TRANSmutation of High Level Nuclear Waste in an Accelerator Driven System) within the ongoing EURATOM 6th Framework Programme (FP6) is devoted to the study of transmutation of high-level waste from nuclear power plants. The work is focused on transmutation in an Accelerator Driven System (ADS). The objective of EUROTRANS is the assessment of the design and the feasibility of an industrial ADS prototype dedicated to transmutation. The necessary R and D results in the areas of accelerator components, fuel development, structural materials, thermal-hydraulics, heavy liquid metal technology and nuclear data will be made available, together with the experimental demonstration of the ADS component coupling. The outcome of this work will allow to provide a reasonably reliable assessment of technological feasibility and a cost estimate for ADS based transmutation, and to possibly decide on the detailed design of an experimental ADS and its construction in the future. EUROTRANS is integrating activities of 51 participants from 16 countries, within the industry (10 participants), the national research centres (20) and 17 universities. 16 universities are collectively represented by ENEN (European Nuclear Education Network). EUROTRANS is the continuation of the three FP5 Clusters FUETRA, BASTRA and TESTRA together with the PDS-XADS Project. It is a five-year project which started in April 2005

The precautionary principle and high-level nuclear waste policy

International Nuclear Information System (INIS)

Frishman, S.

1999-01-01

The 'Precautionary Principle' has grown from the broadening observation that there is compelling evidence that damage to humans and the world-wide environment is of such a magnitude and seriousness that new principles for conducting human activities are necessary. One of the various statements of the Precautionary Principle is: when an activity raises threats of harm to human health or the environment, precautionary measures should be taken even if some cause and effect relationships are not fully established scientifically. The use of a precautionary principle was a significant recommendation emerging from the 1992 United Nations Conference on Environment and Development, held in Rio de Janeiro, Brazil, and it is gaining acceptance in discussions ranging from global warming to activities that affect the marine environment, and far beyond. In the US high-level nuclear waste policy, there is a growing trend on the part of geologic repository proponents and regulators to shift the required safety evaluation from a deterministic analysis of natural and engineered barriers and their interactions to risk assessments and total system waste containment and isolation performance assessment. This is largely a result of the realisation that scientific 'proof' of safety cannot be demonstrated to the level repository proponents have led the American public to expect. Therefore, they are now developing other methods in an attempt to effectively lower the repository safety expectations of the public. Implicit in this shift in demonstration of 'proof' is that levels of uncertainty far larger than those generally taken as scientifically acceptable must be accepted in repository safety, simply because greater certainty is either too costly, in time and money, or impossible to achieve at the potential Yucca Mountain repository site. In the context of the Precautionary Principle, the repository proponent must bear the burden of providing 'Acceptable' proof, established by an open

High level nuclear waste repository in salt: Sealing systems status and planning report: Draft report

International Nuclear Information System (INIS)

1985-09-01

This report documents the initial conceptual design studies for a repository sealing system for a high-level nuclear waste repository in salt. The first step in the initial design studies was to review the current design level, termed schematic designs. This review identified practicality of construction and development of a design methodology as two key issues for the conceptual design. These two issues were then investigated during the initial design studies for seal system materials, seal placement, backfill emplacement, and a testing and monitoring plan. The results of these studies have been used to develop a program plan for completion of the sealing system conceptual design. 60 refs., 26 figs., 18 tabs

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory

International Nuclear Information System (INIS)

Rechard, R.P.

1993-12-01

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency's Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories

Some legal aspects on high level radioactive waste disposal in Japan

International Nuclear Information System (INIS)

Tanabe, Tomoyuki

1997-01-01

In Japan, it is considered to be an urgent problem to prepare the system for the research and execution of high level radioactive waste disposal. Under what regulation scheme the disposal should be done has not been sufficiently examined. In this research, the examination was carried out on the legal aspects of high level radioactive waste disposal as follows. First, the current legislation on the disposal in Japan was analyzed, and it was made clear that high level radioactive waste disposal has not been stipulated clearly. Next, on the legal choices which are conceivable on the way the legislation for high level radioactive waste disposal should be, from the aspects of applying the law on regulating nuclear reactors and others, applying the law on nuclear power damage reparation, and industrialization by changing the government ordinances, those were arranged in six choices, and the examination was carried out for each choice from the viewpoints of the relation with the base stipulation for waste-burying business, the speciality of high level radioactive waste disposal as compared with other actions of nuclear power business, the coordination with existing nuclear power of nuclear power business, the coordination with existing nuclear power law system and the formation of national consensus. In this research, it was shown that the execution of high level radioactive waste disposal as the business based on the separate legislation is the realistic choice. (K.I.)

High-level radioactive waste disposal type and theoretical analyses

International Nuclear Information System (INIS)

Lu Yingfa; Wu Yanchun; Luo Xianqi; Cui Yujun

2006-01-01

Study of high-level radioactive waste disposal is necessary for the nuclear electrical development; the determination of nuclear waste depository type is one of importance safety. Based on the high-level radioactive disposal type, the relative research subjects are proposed, then the fundamental research characteristics of nuclear waste disposition, for instance: mechanical and hydraulic properties of rock mass, saturated and unsaturated seepage, chemical behaviors, behavior of special soil, and gas behavior, etc. are introduced, the relative coupling equations are suggested, and a one dimensional result is proposed. (authors)

STATE OF THE ART OF DRILLING LARGE DIAMETER BOREHOLES FOR DEPOSITION OF HIGH LEVEL WASTE AND SPENT NUCLEAR FUEL

Directory of Open Access Journals (Sweden)

Trpimir Kujundžić

2012-07-01

Full Text Available Deep geological disposal is internationally recognized as the safest and most sustainable option for the long-term management of high-level radioactive waste. Mainly, clay rock, salt rock and crystalline rock are being considered as possible host rocks. Different geological environment in different countries led to the various repository concepts. Main feature of the most matured repository concept is that canisters with spent nuclear fuel are emplaced in vertical or horizontal large diameter deposition holes. Drilling technology of the deposition holes depends on repository concept and geological and geomechanical characteristics of the rock. The deposition holes are mechanically excavated since drill & blast is not a possible method due to requirements on final geometry like surface roughness etc. Different methods of drilling large diameter boreholes for deposition of high-level waste and spent nuclear fuel are described. Comparison of methods is made considering performance and particularities in technology.

Direct conversion of surplus fissile materials, spent nuclear fuel, and other materials to high-level-waste glass

International Nuclear Information System (INIS)

Forsberg, C.W.; Elam, K.R.

1995-01-01

With the end of the cold war the United States, Russia, and other countries have excess plutonium and other materials from the reductions in inventories of nuclear weapons. The United States Academy of Sciences (NAS) has recommended that these surplus fissile materials (SFMs) be processed so they are no more accessible than plutonium in spent nuclear fuel (SNF). This spent fuel standard, if adopted worldwide, would prevent rapid recovery of SFMs for the manufacture of nuclear weapons. The NAS recommended investigation of three sets of options for disposition of SFMs while meeting the spent fuel standard: (1) incorporate SFMs with highly radioactive materials and dispose of as waste, (2) partly burn the SFMs in reactors with conversion of the SFMs to SNF for disposal, and (3) dispose of the SFMs in deep boreholes. The US Government is investigating these options for SFM disposition. A new method for the disposition of SFMs is described herein: the simultaneous conversion of SFMs, SNF, and other highly radioactive materials into high-level-waste (HLW) glass. The SFMs include plutonium, neptinium, americium, and 233 U. The primary SFM is plutonium. The preferred SNF is degraded SNF, which may require processing before it can be accepted by a geological repository for disposal

Technologies for destruction of long-lived radionuclides in high-level nuclear waste - overview and requirements

International Nuclear Information System (INIS)

Arthur, E.D.

1993-01-01

A major issue surrounding current nuclear power generation is the management and disposal of long-lived, high-level waste (HLW). The planned and scientifically acceptable destination for this waste is in deep underground, geologically stable, repositories. However, public concerns surrounding such disposal of long-lived nuclear wastes and other issues such as proliferation and safety negatively affect the potential role that nuclear power can play in meeting current and future national energy needs. This paper and this topical session on nuclear waste minimization, management, and remediation focus on two nuclear systems and their associated technologies that have the potential to address concerns surrounding long-lived radionuclides in HLW. Both systems offer technology applicable to HLW from current light water reactors (LWRs). In addition, these systems represent advanced nuclear power concepts that have important features associated with integrated management of wastes long-term fuel supplies, and enhanced safety. The first system is the integral fast reactor (IFR) concept. This system incorporates a metal-fueled fast reactor coupled with chemical separations based on pyroprocessing to produce power while burning long-lived actinide waste. The IFR applications include the burning of actinides from current LWR spent fuel and energy production in a breeder environment. The second concept, accelerator transmutation of waste (ATW), is based on an accelerator-induced intense source of thermal neutrons and is aimed at the destruction of long-lived actinides and fission products. This concept can be applied to long-lived radionuclides in spent-fuel HLW as well as a future fission power source built around use of natural thorium or uranium as fuels coupled with concurrent waste destruction

Technologies for destruction of long-lived radionuclides in high-level nuclear waste: Overview and requirements

International Nuclear Information System (INIS)

Arthur, E.D.

1993-01-01

This paper, and this topical session on Nuclear Waste Minimization, Management and Remediation, focuses on two nuclear systems, and their associated technologies, that have the potential to address concerns surrounding long-lived radionuclides in high-level waste. Both systems offer technology applicable to HLW from present light-water reactors (LWR). Additionally these systems represent advanced nuclear power concepts that have important features associated with integrated management of wastes, long-term fuel supplies, and enhanced safety. The first system is the Integral Fast Reactor (IFR) concept. This system incorporates a metal-fueled fast reactor coupled with chemical separations based on pyroprocessing to produce power while simultaneously burning long-lived actinide waste. IFR applications include burning of actinides from current LWR spent fuel and energy production in a breeder environment. The second concept, Accelerator Transmutation of Waste (ATW), is based upon an accelerator-induced intense source of thermal neutrons and is aimed at destruction of long-lived actinides and fission products. This concept can be applied to long-lived radionuclides in spent fuel HLW as well as a future fission power source built around use of natural thorium or uranium as fuels coupled with concurrent waste destruction

Spent nuclear fuel project high-level information management plan

Energy Technology Data Exchange (ETDEWEB)

Main, G.C.

1996-09-13

This document presents the results of the Spent Nuclear Fuel Project (SNFP) Information Management Planning Project (IMPP), a short-term project that identified information management (IM) issues and opportunities within the SNFP and outlined a high-level plan to address them. This high-level plan for the SNMFP IM focuses on specific examples from within the SNFP. The plan`s recommendations can be characterized in several ways. Some recommendations address specific challenges that the SNFP faces. Others form the basis for making smooth transitions in several important IM areas. Still others identify areas where further study and planning are indicated. The team`s knowledge of developments in the IM industry and at the Hanford Site were crucial in deciding where to recommend that the SNFP act and where they should wait for Site plans to be made. Because of the fast pace of the SNFP and demands on SNFP staff, input and interaction were primarily between the IMPP team and members of the SNFP Information Management Steering Committee (IMSC). Key input to the IMPP came from a workshop where IMSC members and their delegates developed a set of draft IM principles. These principles, described in Section 2, became the foundation for the recommendations found in the transition plan outlined in Section 5. Availability of SNFP staff was limited, so project documents were used as a basis for much of the work. The team, realizing that the status of the project and the environment are continually changing, tried to keep abreast of major developments since those documents were generated. To the extent possible, the information contained in this document is current as of the end of fiscal year (FY) 1995. Programs and organizations on the Hanford Site as a whole are trying to maximize their return on IM investments. They are coordinating IM activities and trying to leverage existing capabilities. However, the SNFP cannot just rely on Sitewide activities to meet its IM requirements

Retrievability of high-level nuclear waste from geologic repositories - Regulatory and rock mechanics/design considerations

International Nuclear Information System (INIS)

Tanious, N.S.; Nataraja, M.S.; Daemen, J.J.K.

1987-01-01

Retrievability of nuclear waste from high-level geologic repositories is one of the performance objectives identified in 10CFR60 (Code of Federal Regulations, 1985). 10CFR60.111 states that the geologic repository operations area shall be designed to preserve the option of waste retrieval. In designing the repository operations area, rock mechanics considerations play a major role especially in evaluating the feasibility of retrieval operations. This paper discusses generic considerations affecting retrievability as they relate to repository design, construction, and operation, with emphasis on regulatory and rock mechanics aspects

Methods of Disposing Of High-Level Radioactive Waste: A Review

International Nuclear Information System (INIS)

Abumurade, K.

2002-01-01

High level nuclear waste from both commercial reactors and defense industry presents a difficult problem to the scientific community as well as the public. The solutions to this problem is still debatable both technically and ethically. There are few methods proposed for disposing of high level waste. Each method has its own advantages and disadvantages. However, the very deep underground geologic repository is the best choice for disposing of high-level radioactive wastes. The cost benefit equation of nuclear power production and its waste is discussed. However, the public should be educated about this matter to minimize the gap between them and the nuclear power community including scientists industry, and governments. (Author) 15 refs., 4 tabs., 1 fig

Treatment technologies for non-high-level wastes (USA)

International Nuclear Information System (INIS)

Cooley, C.R.; Clark, D.E.

1976-06-01

Non-high-level waste arising from operations at nuclear reactors, fuel fabrication facilities, and reprocessing facilities can be treated using one of several technical alternatives prior to storage. Each alternative and the associated experience and status of development are summarized. The technology for treating non-high-level wastes is generally available for industrial use. Improved techniques applicable to the commercial nuclear fuel cycle are being developed and demonstrated to reduce the volume of waste and to immobilize it for storage. 36 figures, 59 references

Studies of Corrosion Resistant Materials Being Considered for High-Level Nuclear Waste Containment in Yucca Mountain Relevant Environments

International Nuclear Information System (INIS)

McCright, R.D.; Ilevbare, G.; Estill, J.; Rebak, R.

2001-01-01

Containment of spent nuclear fuel and vitrified forms of high level nuclear waste require use of materials that are highly corrosion resistant to all of the anticipated environmental scenarios that can occur in a geological repository. Ni-Cr-Mo Alloy 22 (UNS N60622) is proposed for the corrosion resistant outer barrier of a two-layer waste package container at the potential repository site at Yucca Mountain. A range of water compositions that may contact the outer barrier is under consideration, and a testing program is underway to characterize the forms of corrosion and to quantify the corrosion rates. Results from the testing support models for long term prediction of the performance of the container. Results obtained to date indicate a very low general corrosion rate for Alloy 22 and very high resistance to all forms of localized and environmentally assisted cracking in environments tested to date

The high level vibration test program

International Nuclear Information System (INIS)

Hofmayer, C.H.; Curreri, J.R.; Park, Y.J.; Kato, W.Y.; Kawakami, S.

1989-01-01

As part of cooperative agreements between the US and Japan, tests have been performed on the seismic vibration table at the Tadotsu Engineering Laboratory of Nuclear Power Engineering Test Center (NUPEC) in Japan. The objective of the test program was to use the NUPEC vibration table to drive large diameter nuclear power piping to substantial plastic strain with an earthquake excitation and to compare the results with state-of-the-art analysis of the problem. The test model was subjected to a maximum acceleration well beyond what nuclear power plants are designed to withstand. A modified earthquake excitation was applied and the excitation level was increased carefully to minimize the cumulative fatigue damage due to the intermediate level excitations. Since the piping was pressurized, and the high level earthquake excitation was repeated several times, it was possible to investigate the effects of ratchetting and fatigue as well. Elastic and inelastic seismic response behavior of the test model was measured in a number of test runs with an increasing excitation input level up to the limit of the vibration table. In the maximum input condition, large dynamic plastic strains were obtained in the piping. Crack initiation was detected following the second maximum excitation run. Crack growth was carefully monitored during the next two additional maximum excitation runs. The final test resulted in a maximum crack depth of approximately 94% of the wall thickness. The HLVT (high level vibration test) program has enhanced understanding of the behavior of piping systems under severe earthquake loading. As in other tests to failure of piping components, it has demonstrated significant seismic margin in nuclear power plant piping

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

Californium-252 neutron activation analysis of high-level processed nuclear tank waste

International Nuclear Information System (INIS)

Troyer, G.L.; Purcell, M.A.

2000-01-01

The basis for production assessment of the vitrification of Hanford nuclear fuel reprocessing wastes will be high-precision measurements of the elemental sodium content. However, the chemical analysis of both radioactive and nonradioactive components in nuclear waste can be challenged by high radiation dose rates. The dose rates compromise many analytical techniques as well as pose personnel dosimetry risks. In many cases, reduction of dose rates through dilution compromises the precision and sensitivity for certain key components. The use of neutron activation analysis (NAA) provides a method of analysis that avoids the need for dilutions or extensive sample preparation. These waste materials also contain trace quantities of fissionable isotopes, which, through neutron activation, can be estimated by delayed neutron counting of fissioned fragments

Geology of high-level nuclear waste disposal: an introduction

International Nuclear Information System (INIS)

Roxbugh, I.S.

1987-01-01

Hazardous waste is produced by the nuclear fuel cycle from mining and milling of uranium ore, refinement and enrichment, reactor use, and during reprocessing of spent fuel. Waste can be classified according to origin, physical state, and levels of radioactivity and radiotoxicity. The method of the long-term waste disposal is based on the degree of the hazard and the length of time (1000 years to millions of years) for the waste to become safe. The International Atomic Energy Agency (IAEA) has classified radioactive waste into five categories (I-V) based on the amount of radioactivity and heat output of the waste. The text is concerned mainly with the two most hazardous categories (I and II). Disposal at various geological sites using proven mining, engineering, and deep drilling techniques has been proposed and studied. An ideal geological repository would have (1) minimum ground water movement, (2) geochemical and mineralogical properties to retard or immobilize the effects of the nuclear waste from reaching the biosphere, (3) thermochemical properties to allow for heat loading without damage, and (4) structural strength for the operational period. Types of geological environments (both undersea and on land) include evaporites, crystalline rocks, and argillaceous deposits. European and North American case histories are described, and there is a glossary and an extensive list of references in this concise review

Nuclear systems of level measurement

International Nuclear Information System (INIS)

Lara, A.J.; Cabrera, M.J.

1992-01-01

In the industry there are processes in which is necessary to maintain the products level controlled which are handled for their transformation. The majority of such processes and by the operation conditions, they do not admit measure systems of level of invasive type then the application of nuclear techniques for level measurement results a big aid in these cases, since all the system installation is situated beyond frontiers of vessels that contain the product for measuring. In the Department of Nuclear Technology Applications of Mexican Petroleum Institute was developed a level measurement system by gamma rays transmission which operates in the Low Density Polyethylene plant of Petrochemical Complex Escolin at Poza Rica, Veracruz, Mexico. (Author)

Analysis of mercury in simulated nuclear waste

International Nuclear Information System (INIS)

Policke, T.A.; Johnson, L.C.; Best, D.R.

1991-01-01

Mercury, Hg, is a non-radioactive component in the High Level Waste at the Savannah River Site (SRS). Thus, it is a component of the Defense Waste Processing Facility's (DWPF) process streams. It is present because mercuric nitrate (Hg(NO 3 ) 2 ) is used to dissolve spent fuel rods. Since mercury halides are extremely corrosive, especially at elevated temperatures such as those seen in a melter (1150 degrees C), its concentration throughout the process needs to be monitored so that it is at an acceptable level prior to reaching the melter off-gas system. The Hg can be found in condensates and sludge feeds and throughout the process and process lines, i.e., at any sampling point. The different samples types that require Hg determinations in the process streams are: (1) sludges, which may be basic or acidic and may or may not include aromatic organics, (2) slurries, which are sludges with frit and will always contain organics (formate and aromatics), and (3) condensates, from feed prep and melter off-gas locations. The condensates are aqueous and the mercury may exist as a complex mixture of halides, oxides, and metal, with levels between 10 and 100 ppm. The mercury in the sludges and slurries can be Hg 0 , Hg +1 , or Hg +2 , with levels between 200 and 3000 ppm, depending upon the location, both time and position, of sampling. For DWPF, both total and soluble Hg concentrations need to be determined. The text below describes how these determinations are being made by the Defense Waste Processing Technology (DWPT) Analytical Laboratory at the Savannah River Site. Both flame atomic absorption (FAA) and cold vapor atomic (CVAA) measurements are discussed. Also, the problems encountered in the steps toward measuring HG in these samples types of condensates and sludges are discussed along with their solutions

Biological ramifications of the subseabed disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Gomez, L.S.; Hessler, R.R.; Jackson, D.W.; Marietta, M.G.; Smith, K.L. Jr.; Talbert, D.M.; Yayanos, A.A.

1980-01-01

The primary goal of the US Subseabed Disposal Program (SDP) is to assess the technical and environmental feasibility of disposing of high-level nuclear waste in deep-sea sediments. The subseabed biology program is charged with assessing possible ecosystem effects of radionuclides as well as possible health effects to man from radionuclides which may be released in the deep sea and transported to the ocean surface. Current biological investigations are attempting to determine benthic community structure; benthic community metabolism; the biology of deep-sea mobile scavengers; the faunal composition of midwater nekton; rates of microbial processes, and the radiation sensitivity of deep-sea organisms. Existing models of the dispersal of radionuclides in the deep sea have not considered many of the possible biological mechanisms which may influence the movement of radionuclides. Therefore, a multi-compartment foodweb model is being developed which considers both biological and physical influences on radionuclide transport. This model will allow parametric studies to be made of the impact on the ocean environment and on man of potential releases of radionuclides

Biological ramifications of the subseabed disposal of high-level nuclear waste

International Nuclear Information System (INIS)

Gomez, L.S.; Hessler, R.R.; Jackson, D.W.; Marietta, M.G.; Smith, K.L. Jr.; Talbert, D.M.; Yayanos, A.A.

1980-05-01

The primary goal of the US Subseabed Disposal Program (SDP) is to assess the technical and environmental feasibility of disposing of high-level nuclear waste in deep-sea sediments. The subseabed biology program is charged with assessing possible ecosystem effects of radionuclides as well as possible health effects to man from radionuclides which may be released in the deep sea and transported to the ocean surface. Current biological investigations are attempting to determine benthic community structure; benthic community metabolism; the biology of deep-sea mobile scavengers; the faunal composition of midwater nekton; rates of microbial processes; and the radiation sensitivity of deep-sea organisms. Existing models of the dispersal of radionuclides in the deep sea have not considered many of the possible biological mechanisms which may influence the movement of radionuclides. Therefore, a multi-compartment foodweb model is being developed which considers both biological and physical influences on radionuclide transport. This model will allow parametric studies to be made of the impact on the ocean environment and on man of potential releases of radionuclides

Enhanced nuclear level decay in hot dense plasmas

International Nuclear Information System (INIS)

Gosselin, G.; Morel, P.

2004-01-01

A model of nuclear level decay in a plasma environment is described. Nuclear excitation and decay by photon processes, nuclear excitation by electron capture, and decay by internal conversion are taken into account. The electrons in the plasma are described by a relativistic average atom model for the bound electrons and by a relativistic Thomas-Fermi-Dirac model for the free electrons. Nuclear decay of isomeric level may be enhanced through an intermediate level lying above the isomer. An enhanced nuclear decay rate may occur for temperatures far below the excitation energy of the transition to the intermediate level. In most cases, the enhancement factor may reach several decades

Preliminary Analysis of Species Partitioning in the DWPF Melter

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-15

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

Nuclear power: levels of safety

International Nuclear Information System (INIS)

Lidsky, L.M.

1988-01-01

The rise and fall of the nuclear power industry in the United States is a well-documented story with enough socio-technological conflict to fill dozens of scholarly, and not so scholarly, books. Whatever the reasons for the situation we are now in, and no matter how we apportion the blame, the ultimate choice of whether to use nuclear power in this country is made by the utilities and by the public. Their choices are, finally, based on some form of risk-benefit analysis. Such analysis is done in well-documented and apparently logical form by the utilities and in a rather more inchoate but not necessarily less accurate form by the public. Nuclear power has failed in the United States because both the real and perceived risks outweigh the potential benefits. The national decision not to rely upon nuclear power in its present form is not an irrational one. A wide ranging public balancing of risk and benefit requires a classification of risk which is clear and believable for the public to be able to assess the risks associated with given technological structures. The qualitative four-level safety ladder provides such a framework. Nuclear reactors have been designed which fit clearly and demonstrably into each of the possible qualitative safety levels. Surprisingly, it appears that safer may also mean cheaper. The intellectual and technical prerequisites are in hand for an important national decision. Deployment of a qualitatively different second generation of nuclear reactors can have important benefits for the United States. Surprisingly, it may well be the nuclear establishment itself, with enormous investments of money and pride in the existing nuclear systems, that rejects second generation reactors. It may be that we will not have a second generation of reactors until the first generation of nuclear engineers and nuclear power advocates has retired

The Defense Waste Processing Facility, from vision to reality

International Nuclear Information System (INIS)

Randall, C.T.

2000-01-01

When the Savannah River Plant began operation in the early 1950's producing nuclear materials for the National defense, liquid, highly radioactive waste was generated as a by-product. Since that time the waste has been stored in large, carbon steel tanks that are buried underground. In 1960 one of the tanks developed a leak, and before recovery measures could be taken, about 25-gallons of radioactive salt solution had overflowed the secondary liner and seeped into the soil surrounding the tank. Significant improvements to the tanks were made, but constant surveillance was still required. Thus, the opinion began forming that storage of the mobile, highly radioactive waste in tanks was not a responsible long-term practice. So in the late 1960's the Savannah River Laboratory began research to find a suitable long-term solution to the waste disposal problem. Several alternative waste forms were evaluated, and in 1972 the first Savannah River waste was vitrified on a laboratory scale. By the mid-1970's, the DuPont Company, prime contractor at the Savannah River Plant, began to develop a vision of constructing America's first vitrification plant to immobilize the high level radioactive waste in borosilicate glass. This vision was later championed by DuPont in the form of a vitrification plant called the Defense Waste Processing Facility (DWPF). Today, the DWPF processes Savannah River High Level Waste sludge turning it into a solid, durable waste form of borosilicate glass. The DWPF is the world's largest vitrification facility. It was brought to reality through over 25-years of research and 13-years of careful construction, tests, and reviews at a cost of approximately $3 billion dollars

Thirty Years of Social Science Research on High-Level Nuclear Waste: Achievements and Future Challenges

International Nuclear Information System (INIS)

Solomon, Barry D.; Andren, Mats; Strandberg, Urban

2010-09-01

Research on high-level nuclear waste management has focused on technical and scientific issues since the U.S. National Academy of Sciences first studied the problem in the mid 1950s and recommended long-term disposal in deep salt formations. In this review, we trace the development of the problem's definition and its associated research since socioeconomic, political and policy issues were first given consideration and nuclear waste management became recognized as more than a technical issue. Three time periods are identified. First, from the mid 1970s to early 1980s, initial research explored institutional dimensions of nuclear waste, including ethics. The second period began in the early 1980s with a concerted effort to solve the problem and site nuclear waste repositories, and ended in the mid 1990s with minimal progress in the U.S. and general stalemate in Asia and Europe (with the notable exception of Sweden). This phase accelerated research on risk perception and stigma of nuclear waste, and elevated a focus on public trust. Great attention was given to repository siting conflicts, while minimal attention was placed on ethics, equity, political systems, and public participation. The last period, since the mid 1990s, has been characterized by continuing political stalemate and increased attention to public participation, political systems and international solutions. Questions of ethics have been given renewed attention, while research on risk perceptions and siting conflicts continues. We frame these periods in a broader context of the shifting role of applied social scientists. The paper concludes with a general discussion of this research area and prospects for future research

Thirty Years of Social Science Research on High-Level Nuclear Waste: Achievements and Future Challenges

Energy Technology Data Exchange (ETDEWEB)

Solomon, Barry D. (Dept. of Social Sciences, Michigan Technological Univ., Houghton (United States)), e-mail: bdsolomo@mtu.edu; Andren, Mats; Strandberg, Urban (Center for Public Sector Research, Univ. of Goeteborg, Goeteborg (Sweden))

2010-09-15

Research on high-level nuclear waste management has focused on technical and scientific issues since the U.S. National Academy of Sciences first studied the problem in the mid 1950s and recommended long-term disposal in deep salt formations. In this review, we trace the development of the problem's definition and its associated research since socioeconomic, political and policy issues were first given consideration and nuclear waste management became recognized as more than a technical issue. Three time periods are identified. First, from the mid 1970s to early 1980s, initial research explored institutional dimensions of nuclear waste, including ethics. The second period began in the early 1980s with a concerted effort to solve the problem and site nuclear waste repositories, and ended in the mid 1990s with minimal progress in the U.S. and general stalemate in Asia and Europe (with the notable exception of Sweden). This phase accelerated research on risk perception and stigma of nuclear waste, and elevated a focus on public trust. Great attention was given to repository siting conflicts, while minimal attention was placed on ethics, equity, political systems, and public participation. The last period, since the mid 1990s, has been characterized by continuing political stalemate and increased attention to public participation, political systems and international solutions. Questions of ethics have been given renewed attention, while research on risk perceptions and siting conflicts continues. We frame these periods in a broader context of the shifting role of applied social scientists. The paper concludes with a general discussion of this research area and prospects for future research

Third report, development of site suitability and design performance data base for a high level nuclear waste repository

International Nuclear Information System (INIS)

1978-05-01

This report presents the results of mining and geotechnical studies provided as support for the development of a technical data base suitable for the generation of a regulatory framework governing high-level nuclear waste disposal: flow path models for basalts, salt domes, and crystalline rocks; simplistic model evaluations; loss of administrative control; repository conceptual designs; repository design considerations; and design performance criteria

Combinatorial nuclear level density by a Monte Carlo method

International Nuclear Information System (INIS)

Cerf, N.

1994-01-01

We present a new combinatorial method for the calculation of the nuclear level density. It is based on a Monte Carlo technique, in order to avoid a direct counting procedure which is generally impracticable for high-A nuclei. The Monte Carlo simulation, making use of the Metropolis sampling scheme, allows a computationally fast estimate of the level density for many fermion systems in large shell model spaces. We emphasize the advantages of this Monte Carlo approach, particularly concerning the prediction of the spin and parity distributions of the excited states,and compare our results with those derived from a traditional combinatorial or a statistical method. Such a Monte Carlo technique seems very promising to determine accurate level densities in a large energy range for nuclear reaction calculations

Survey of cognition on nuclear and radiation in Beijing high school students

International Nuclear Information System (INIS)

Wang Chao; He Jianrong; Zhu Xiayang; Yang Guoliang; Cong Huiling; Hu Qinfang

2014-01-01

Objective: To explore cognition level on nuclear and radiation in Beijing high school students, which may provide evidence for promoting science popularization on nuclear and radiation. Methods: Questionnaire-based survey was conducted in Beijing high school students, randomized cluster sampling was used to recruit study participants. Demographic information was collected, and cognition level on nuclear and radiation was evaluated by questionnaire. Results: A total of 1029 pieces of eligible questionnaires were collected. The correct rate for answering common sense about nuclear and radiation was 58%, with score of boys significantly higher than that of girls (t = 4.131, P < 0.05). About subjective cognition of nuclear and radiation knowledge, 87 (8.5%) indicated 'quite clear', 779 (75.7%) indicated 'know a little', 163 (15.8%) indicated 'know nothing'. There was significant difference in score of common sense about nuclear and radiation among people with various subjective cognition level of nuclear and radiation (J-T = 8.279, P < 0.05). There was a linear correlation between support degree for nuclear power and subjective cognition level of nuclear and radiation (r = 0.161, P < 0.05). There was significant difference in score of common sense about nuclear and radiation among people with various support degree for nuclear power (J-T = 7.508, P < 0.05), whereas those who had got high scores tended to support nuclear power to a higher degree. Conclutions: Students knew little about knowledge on nuclear and radiation. It is necessary to strengthen propaganda and education on nuclear and radiation, which may help enhance the students' comprehensive quality, and sustainable expansion of nuclear power more support in the long run. (authors)

Overview of high-level waste management accomplishments

International Nuclear Information System (INIS)

Lawroski, H.; Berreth, J.R.; Freeby, W.A.

1980-01-01

Storage of power reactor spent fuel is necessary at present because of the lack of reprocessing operations particularly in the U.S. By considering the above solidification and storage scenario, there is more than reasonable assurance that acceptable, stable, low heat generation rate, solidified waste can be produced, and safely disposed. The public perception of no waste disposal solutions is being exploited by detractors of nuclear power application. The inability to even point to one overall system demonstration lends credibility to the negative assertions. By delaying the gathering of on-line information to qualify repository sites, and to implement a demonstration, the actions of the nuclear power detractors are self serving in that they can continue to point out there is no demonstration of satisfactory high-level waste disposal. By maintaining the liquid and solidified high-level waste in secure above ground storage until acceptable decay heat generation rates are achieved, by producing a compatible, high integrity, solid waste form, by providing a second or even third barrier as a compound container and by inserting the enclosed waste form in a qualified repository with spacing to assure moderately low temperature disposal conditions, there appears to be no technical reason for not progressing further with the disposal of high-level wastes and needed implementation of the complete nuclear power fuel cycle

Chemical durability of borosilicate glasses containing simulated high-level nuclear wastes, 1

International Nuclear Information System (INIS)

Hara, Shigeo; Terai, Ryohei; Yamanaka, Hiroshi

1983-01-01

The Soxhlet-type leaching test apparatus has been developed to evaluate the chemical durability of some borosilicate glasses containing simulated High-Level nuclear Wastes, HLW. After the leaching over the temperature range of 50 0 -95 0 C, the weight loss of specimens with time was determined on both the samples of blocks and grains, and various components dissolved into water were analyzed by atomic absorption and colorimetry technique. It was found that Soxhlet-type test method was more useful than JIS test method, because the specimens in Soxhlet type apparatus were forced always to react with pure water and the mechanism of leaching could be evaluate accurately. The chemical durability of commercial glasses decreases generally with increasing of alkali contents in glasses. In the case of these borosilicate glasses containing HLW, however, the leachability was apparently independent on the alkali contents because of the complexity of these glass compositions. The variation of leaching rate with temperature suggests that dissolution mechanism changes with temperature. (author)

Nuclear high-spin data for A = 174, 176 and 184

Energy Technology Data Exchange (ETDEWEB)

Junde, Huo [Jilin Univ. (China). Dept. of Physics

1996-06-01

Nuclear high-spin data are important in the frontier areas of nuclear structure physics. The information on A = 174, 176 and 184 mass chains from various reaction experiments together with their adopted high-spin levels and gamma transition properties are presented and discussed. High-spin data for A = 174, 176 and 184 mass chains were evaluated in 1995.

Perceived risk, stigma, and potential economic impacts of a high-level nuclear waste repository in Nevada

International Nuclear Information System (INIS)

Slovic, P.; Layman, M.; Kraus, N.N.; Chalmers, J.; Gesel, G.; Flynn, J.

1989-07-01

This paper describes a program of research designed to assess the potential impacts of a high-level nuclear waste repository at Yucca Mountain, Nevada, upon tourism, retirement and job-related migration, and business development in Las Vegas and the state. Adverse economic impacts may be expected to result from two related social processes. One has to do with perceptions of risk and socially amplified reactions to ''unfortunate events'' associated with the repository (major and minor accidents, discoveries of radiation releases, evidence of mismanagement, attempts to sabotage or disrupt the facility, etc.). The second process that may trigger significant adverse impacts is that of stigmatization. The conceptual underpinnings of risk perception, social amplification, and stigmatization are discussed in this paper and empirical data are presented to demonstrate how nuclear images associated with Las Vegas and the State of Nevada might trigger adverse effects on tourism, migration, and business development

Joint project to implement a training course/seminar for high-level decision making officals on nuclear policy between Korea and IAEA

International Nuclear Information System (INIS)

Lee, E. J.; Suh, S. H.; Kim, S. H. and others

2001-09-01

Through this project, the KAERI has published a english text book titled 'Nuclear Power project: Policy and Korean Experience' with the cooperation with the Agency. Korean experts, who have more than 30 years experiences in nuclear policy and nuclear power project management, involved to prepare the contents. Experts, who are working for Department of Nuclear Energy and Department of Technical Cooperation in the Agency, reviewed the contents. The KAERI has provided a training course/seminar for a high-level delegation of nuclear policy decision makers, which is consisted of 3 deputy ministers and 3 general directors in Viet Nam Communist Party, Assembly, Government Departments, upon the agreement between Vietnam Atomic Energy Commission (VAEC) and MOST at Hanoi in November 2000. The KAERI decided to provide the IAEA regional training course for middle level managers of nuclear policy and project management in RCA member states in November 2001 and 2002 under the cooperation with the Agency and the Korea International Cooperation Agency (KOICA). The training course curricula, a english text book will be utilized as basic technical materials to promote the national nuclear cooperation program with the developing countries, that are considering to introduce the first or a new nuclear power plant, such as China, Indonesia, Bangladesh, Morocco, Egypt, Ukraine, etc. Through the project, the bilateral nuclear cooperation between Viet Nam and Korea in the field of nuclear manpower development has improved practically. The KAERI has decided to provide on-the job-training program for Vietnamese experts in the field of nuclear policy, safety analysis, thermal-hydraulic analysis, etc under the 2000 KISTEP Scientist Exchange Program

Joint project to implement a training course/seminar for high-level decision making officals on nuclear policy between Korea and IAEA

Energy Technology Data Exchange (ETDEWEB)

Lee, E. J.; Suh, S. H.; Kim, S. H. and others

2001-09-01

Through this project, the KAERI has published a english text book titled 'Nuclear Power project: Policy and Korean Experience' with the cooperation with the Agency. Korean experts, who have more than 30 years experiences in nuclear policy and nuclear power project management, involved to prepare the contents. Experts, who are working for Department of Nuclear Energy and Department of Technical Cooperation in the Agency, reviewed the contents. The KAERI has provided a training course/seminar for a high-level delegation of nuclear policy decision makers, which is consisted of 3 deputy ministers and 3 general directors in Viet Nam Communist Party, Assembly, Government Departments, upon the agreement between Vietnam Atomic Energy Commission (VAEC) and MOST at Hanoi in November 2000. The KAERI decided to provide the IAEA regional training course for middle level managers of nuclear policy and project management in RCA member states in November 2001 and 2002 under the cooperation with the Agency and the Korea International Cooperation Agency (KOICA). The training course curricula, a english text book will be utilized as basic technical materials to promote the national nuclear cooperation program with the developing countries, that are considering to introduce the first or a new nuclear power plant, such as China, Indonesia, Bangladesh, Morocco, Egypt, Ukraine, etc. Through the project, the bilateral nuclear cooperation between Viet Nam and Korea in the field of nuclear manpower development has improved practically. The KAERI has decided to provide on-the job-training program for Vietnamese experts in the field of nuclear policy, safety analysis, thermal-hydraulic analysis, etc under the 2000 KISTEP Scientist Exchange Program.

'Level-level correlation and absorption in nuclear reactions'

International Nuclear Information System (INIS)

Hussein, M.S.

Level-level correlation (LLC) in nuclear reactions is discussed in general and it is shown that in the presence of LLC, N sub(μ) = Σ/g μa/ 2 > divided by gamma μ T tilde, where T tilde is the average absorption in the eigen channels [pt

Treatment of High-Level Waste Arising from Pyrochemical Processes

International Nuclear Information System (INIS)

Lizin, A.A.; Kormilitsyn, M.V.; Osipenko, A.G.; Tomilin, S.V.; Lavrinovich, Yu.G.

2013-01-01

JSC “SSC RIAR” has been performing research and development activities in support of closed fuel cycle of fast reactor since the middle of 1960s. Fuel cycle involves fabrication and reprocessing of spent nuclear fuel (SNF) using pyrochemical methods of reprocessing in molten alkali metal chlorides. At present pyrochemical methods of SNF reprocessing in molten chlorides has reached such a level in their development that makes it possible to compare their competitiveness with classic aqueous methods. Their comparative advantage lies in high safety, compactness, high protectability as to nonproliferation of nuclear materials, and reduction of high level waste volume

West Valley high-level nuclear waste glass development: a statistically designed mixture study

Energy Technology Data Exchange (ETDEWEB)

Chick, L.A.; Bowen, W.M.; Lokken, R.O.; Wald, J.W.; Bunnell, L.R.; Strachan, D.M.

1984-10-01

The first full-scale conversion of high-level commercial nuclear wastes to glass in the United States will be conducted at West Valley, New York, by West Valley Nuclear Services Company, Inc. (WVNS), for the US Department of Energy. Pacific Northwest Laboratory (PNL) is supporting WVNS in the design of the glass-making process and the chemical formulation of the glass. This report describes the statistically designed study performed by PNL to develop the glass composition recommended for use at West Valley. The recommended glass contains 28 wt% waste, as limited by process requirements. The waste loading and the silica content (45 wt%) are similar to those in previously developed waste glasses; however, the new formulation contains more calcium and less boron. A series of tests verified that the increased calcium results in improved chemical durability and does not adversely affect the other modeled properties. The optimization study assessed the effects of seven oxide components on glass properties. Over 100 melts combining the seven components into a wide variety of statistically chosen compositions were tested. Viscosity, electrical conductivity, thermal expansion, crystallinity, and chemical durability were measured and empirically modeled as a function of the glass composition. The mathematical models were then used to predict the optimum formulation. This glass was tested and adjusted to arrive at the final composition recommended for use at West Valley. 56 references, 49 figures, 18 tables.

High-level radioactive waste in Canada. Background paper

International Nuclear Information System (INIS)

Fawcett, R.

1993-11-01

The disposal of radioactive waste is one of the most challenging environmental problems facing Canada today. Since the Second World War, when Canadian scientists first started to investigate nuclear reactions, there has been a steady accumulation of such waste. Research reactors built in the early postwar years produced small amounts of radioactive material but the volume grew steadily as the nuclear power reactors constructed during the 1960s and 1970s began to spawn used fuel bundles. Although this radioactive refuse has been safely stored for the short term, no permanent disposal system has yet been fully developed and implemented. Canada is not alone in this regard. A large number of countries use nuclear power reactors but none has yet put in place a method for the long-term disposal of the radioactive waste. Scientists and engineers throughout the world are investigating different possibilities; however, enormous difficulties remain. In Canada, used fuel bundles from nuclear reactors are defined as high-level waste; all other waste created at different stages in the nuclear fuel cycle is classified as low-level. Although disposal of low-level waste is an important issue, it is a more tractable problem than the disposal of high-level waste, on which this paper will concentrate. The paper discusses the nuclear fuel waste management program in Canada, where a long-term disposal plan has been under development by scientists and engineers over the past 15 years, but will not be completed for some time. Also discussed are responses to the program by parliamentary committees and aboriginal and environmental groups, and the work in the area being conducted in other countries. (author). 1 tab

High-level radioactive waste in Canada. Background paper

Energy Technology Data Exchange (ETDEWEB)

Fawcett, R [Library of Parliament, Ottawa, ON (Canada). Science and Technology Div.

1993-11-01

The disposal of radioactive waste is one of the most challenging environmental problems facing Canada today. Since the Second World War, when Canadian scientists first started to investigate nuclear reactions, there has been a steady accumulation of such waste. Research reactors built in the early postwar years produced small amounts of radioactive material but the volume grew steadily as the nuclear power reactors constructed during the 1960s and 1970s began to spawn used fuel bundles. Although this radioactive refuse has been safely stored for the short term, no permanent disposal system has yet been fully developed and implemented. Canada is not alone in this regard. A large number of countries use nuclear power reactors but none has yet put in place a method for the long-term disposal of the radioactive waste. Scientists and engineers throughout the world are investigating different possibilities; however, enormous difficulties remain. In Canada, used fuel bundles from nuclear reactors are defined as high-level waste; all other waste created at different stages in the nuclear fuel cycle is classified as low-level. Although disposal of low-level waste is an important issue, it is a more tractable problem than the disposal of high-level waste, on which this paper will concentrate. The paper discusses the nuclear fuel waste management program in Canada, where a long-term disposal plan has been under development by scientists and engineers over the past 15 years, but will not be completed for some time. Also discussed are responses to the program by parliamentary committees and aboriginal and environmental groups, and the work in the area being conducted in other countries. (author). 1 tab.

Progress in evaluation of radionuclide geochemical information developed by DOE high-level nuclear waste repository site projects

International Nuclear Information System (INIS)

Meyer, R.E.; Arnold, W.D.; O'Kelley, G.D.; Case, F.I.; Land, J.F.

1989-08-01

Information that is being developed by projects within the Department of Energy (DOE) pertinent to the potential geochemical behavior of radionuclides at candidate sites for a high-level radioactive waste repository is being evaluated by Oak Ridge National Laboratory (ORNL) for the Nuclear Regulatory Commission (NRC). During this report period, all experiments were conducted with tuff from the proposed high-level nuclear waste site at Yucca Mountain, Nevada. The principal emphasis in this report period was on column studies of migration of uranium and technetium in water from well J-13 at the Yucca Mountain site. Columns 1 cm in diameter and about 5 cm long were constructed and carefully packed with ground tuff. The characteristics of the columns were tested by determination of elution curves of tritium and TcO 4 - . Elution peaks obtained in past studies with uranium were asymmetrical and the shapes were often complex, observations that suggested irreversibilities in the sorption reaction. To try to understand these observations, the effects of flow rate and temperature on uranium migration were studied in detail. Sorption ratios calculated from the elution peaks became larger as the flow rate decreased and as the temperature increased. These observations support the conclusion that the sorption of uranium is kinetically hindered. To confirm this, batch sorption ratio experiments were completed for uranium as a function of time for a variety of conditions

Ceramic Single Phase High-Level Nuclear Waste Forms: Hollandite, Perovskite, and Pyrochlore

Science.gov (United States)

Vetter, M.; Wang, J.

2017-12-01

The lack of viable options for the safe, reliable, and long-term storage of nuclear waste is one of the primary roadblocks of nuclear energy's sustainable future. The method being researched is the incorporation and immobilization of harmful radionuclides (Cs, Sr, Actinides, and Lanthanides) into the structure of glasses and ceramics. Borosilicate glasses are the main waste form that is accepted and used by today's nuclear industry, but they aren't the most efficient in terms of waste loading, and durability is still not fully understood. Synroc-phase ceramics (i.e. hollandite, perovskite, pyrochlore, zirconolite) have many attractive qualities that glass waste forms do not: high waste loading, moderate thermal expansion and conductivity, high chemical durability, and high radiation stability. The only downside to ceramics is that they are more complex to process than glass. New compositions can be discovered by using an Artificial Neural Network (ANN) to have more options to optimize the composition, loading for performance by analyzing the non-linear relationships between ionic radii, electronegativity, channel size, and a mineral's ability to incorporate radionuclides into its structure. Cesium can be incorporated into hollandite's A-site, while pyrochlore and perovskite can incorporate actinides and lanthanides into their A-site. The ANN is used to predict new compositions based on hollandite's channel size, as well as the A-O bond distances of pyrochlore and perovskite, and determine which ions can be incorporated. These new compositions will provide more options for more experiments to potentially improve chemical and thermodynamic properties, as well as increased waste loading capabilities.

Development of geopolymers as candidate materials for low/intermediate level highly alkaline nuclear waste

International Nuclear Information System (INIS)

Perera, D.S.; Vance, E.R.; Kiyama, S.; Aly, Z.; Yee, P.

2006-01-01

Full text: Geopolymers have been studied for many years as a possible improvement on cement in respect of compressive strength, resistance to fire, heat and acidity, and as a medium for the encapsulation of hazardous or low/ intermediate level radioactive waste. They are made by adding aluminosilicates to concentrated alkali solutions and the application of heat at 0 Cfor subsequent polymerisation. In this work we studied them as suitable candidate materials to incorporate NaOH/NaA10 2 containing waste with low levels of Cs, Sr and Nd. Geopolymers were produced by incorporating the highly alkaline solution as part of the composition with added metakaolinite, fumed silica and extra NaOH, such that the overall geopolymer composition was of molar ratios Si/Al = 2 and Na/Al = 1. The simulated waste contained Na2SO 4 , therefore Ba(OH) 2 was also added to precipitate the SO 4 x 2 as BaSO 4 . Three geoplymers of the same composition containing simulated wastes were leach tested in triplicate after heating at 400 0 Cfor 1 h (to remove -98% of free and interstitial water) under the PCT-B test protocol at 90 0 Cfor 7 days and their results are listed in Table 1. The Cs, Sr and Nd normalised leach rates were low. The Na leach rate was ∼ 4 g/L thus passing the PCT-B test protocol value of 13.5 g/L for EA glass. The X-ray diffraction and scanning electron microscopy showed that BaS04 did precipitate, however all the S did not appear to have precipitated. The ANSI/ANS-16.1-2003 test was carried out on the above geopolymer composition for 5 days. The ANSI Leachability Index D (diffusivity of 10''cm sec'') for the elements released are listed in Table 2. A Portland cement was also tested for comparison and the Leachability index values are 11, 8 and 10 for Al, Na and Ca respectively. Both passed the test protocol insofar as they were > 6. Geopolymers thus passing the tests for high level nuclear waste glass (PCT-B) and for low level nuclear waste (ANSI) show promising potential

A report on high-level nuclear waste transportation: Prepared pursuant to assembly concurrent resolution No. 8 of the 1987 Nevada Legislature

International Nuclear Information System (INIS)

1988-12-01

This report has been prepared by the staff of the State of Nevada Agency for Nuclear Projects/Nuclear Waste Project Office (NWPO) in response to Assembly Concurrent Resolution No. 8 (ACR 8), passed by the Nevada State Legislature in 1987. ACR 8 directed the NWPO, in cooperation with affected local governments and the Legislative committee on High-Level Radioactive Waste, to prepare this report which scrutinizes the US Department of Energy's (DOE) plans for transportation of high-level radioactive waste to the proposed yucca Mountain repository, which reviews the regulatory structure under which shipments to a repository would be made and which presents NWPO's plans for addressing high-level radioactive waste transportation issues. The report is divided into three major sections. Section 1.0 provides a review of DOE's statutory requirements, its repository transportation program and plans, the major policy, programmatic, technical and institutional issues and specific areas of concern for the State of Nevada. Section 2.0 contains a description of the current federal, state and tribal transportation regulatory environment within which nuclear waste is shipped and a discussion of regulatory issues which must be resolved in order for the State to minimize risks and adverse impacts to its citizens. Section 3.0 contains the NWPO plan for the study and management of repository-related transportation. The plan addresses four areas, including policy and program management, regulatory studies, technical reviews and studies and institutional relationships. A fourth section provides recommendations for consideration by State and local officials which would assist the State in meeting the objectives of the plan

A report on high-level nuclear waste transportation: Prepared pursuant to assembly concurrent resolution No. 8 of the 1987 Nevada Legislature

Energy Technology Data Exchange (ETDEWEB)

NONE

1988-12-01

This report has been prepared by the staff of the State of Nevada Agency for Nuclear Projects/Nuclear Waste Project Office (NWPO) in response to Assembly Concurrent Resolution No. 8 (ACR 8), passed by the Nevada State Legislature in 1987. ACR 8 directed the NWPO, in cooperation with affected local governments and the Legislative committee on High-Level Radioactive Waste, to prepare this report which scrutinizes the US Department of Energy`s (DOE) plans for transportation of high-level radioactive waste to the proposed yucca Mountain repository, which reviews the regulatory structure under which shipments to a repository would be made and which presents NWPO`s plans for addressing high-level radioactive waste transportation issues. The report is divided into three major sections. Section 1.0 provides a review of DOE`s statutory requirements, its repository transportation program and plans, the major policy, programmatic, technical and institutional issues and specific areas of concern for the State of Nevada. Section 2.0 contains a description of the current federal, state and tribal transportation regulatory environment within which nuclear waste is shipped and a discussion of regulatory issues which must be resolved in order for the State to minimize risks and adverse impacts to its citizens. Section 3.0 contains the NWPO plan for the study and management of repository-related transportation. The plan addresses four areas, including policy and program management, regulatory studies, technical reviews and studies and institutional relationships. A fourth section provides recommendations for consideration by State and local officials which would assist the State in meeting the objectives of the plan.

Workshop on the role of natural analogs in geologic disposal of high-level nuclear waste: Proceedings

International Nuclear Information System (INIS)

Kovach, L.A.; Murphy, W.M.

1995-09-01

A Workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste was held in San Antonio, Texas on July 22--25, 1991. The proceedings comprise seventeen papers submitted by participants at the workshop. A series of papers addresses the relation of natural analog studies to the regulation, performance assessment, and licensing of a geologic repository. Applications of reasoning by analogy are illustrated in papers on the role of natural analogs in studies of earthquakes, petroleum, and mineral exploration. A summary is provided of a recently completed, internationally coordinated natural analog study at Pocos de Caldas, Brazil. Papers also cover problems and applications of natural analog studies in four technical areas of nuclear waste management-. waste form and waste package, near-field processes and environment, far-field processes and environment, and volcanism and tectonics. Summaries of working group deliberations in these four technical areas provide reviews and proposals for natural analog applications. Individual papers have been cataloged separately

Workshop on the role of natural analogs in geologic disposal of high-level nuclear waste: Proceedings

Energy Technology Data Exchange (ETDEWEB)

Kovach, L.A. [ed.] [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Regulatory Applications; Murphy, W.M. [ed.] [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

1995-09-01

A Workshop on the Role of Natural Analogs in Geologic Disposal of High-Level Nuclear Waste was held in San Antonio, Texas on July 22--25, 1991. The proceedings comprise seventeen papers submitted by participants at the workshop. A series of papers addresses the relation of natural analog studies to the regulation, performance assessment, and licensing of a geologic repository. Applications of reasoning by analogy are illustrated in papers on the role of natural analogs in studies of earthquakes, petroleum, and mineral exploration. A summary is provided of a recently completed, internationally coordinated natural analog study at Pocos de Caldas, Brazil. Papers also cover problems and applications of natural analog studies in four technical areas of nuclear waste management-. waste form and waste package, near-field processes and environment, far-field processes and environment, and volcanism and tectonics. Summaries of working group deliberations in these four technical areas provide reviews and proposals for natural analog applications. Individual papers have been cataloged separately.

Statistical interpretation of low energy nuclear level schemes

Energy Technology Data Exchange (ETDEWEB)

Egidy, T von; Schmidt, H H; Behkami, A N

1988-01-01

Nuclear level schemes and neutron resonance spacings yield information on level densities and level spacing distributions. A total of 75 nuclear level schemes with 1761 levels and known spins and parities was investigated. The A-dependence of level density parameters is discussed. The spacing distributions of levels near the groundstate indicate transitional character between regular and chaotic properties while chaos dominates near the neutron binding energy.

High levels of nuclear MYC protein predict the presence of MYC rearrangement in diffuse large B-cell lymphoma

DEFF Research Database (Denmark)

Green, Tina Marie; Nielsen, Ole; de Stricker, Karin

2012-01-01

, and quantitative real-time polymerase chain reaction (QRT-PCR). Overall, 15% of the cases had an MYC break. QRT-PCR analysis of MYC expression showed that 72% of DLBCLs with an MYC break had aberrantly high or low levels of MYC transcript. Excluding the cases with aberrantly low MYC expression, we found...... a significant positive correlation between levels of MYC transcripts and MYC tumor cells; however, QRT-PCR is not readily applicable as a screening tool. Immunohistochemically, all tumors showed a nuclear staining pattern that was simple to evaluate. The percentage of MYC lymphoma cells correlated closely...

Separation processes for high-level radioactive waste treatment

International Nuclear Information System (INIS)

Sutherland, D.G.

1992-11-01

During World War II, production of nuclear materials in the United States for national defense, high-level waste (HLW) was generated as a byproduct. Since that time, further quantities of HLW radionuclides have been generated by continued nuclear materials production, research, and the commercial nuclear power program. In this paper HLW is defined as the highly radioactive material resulting from the processing of spent nuclear fuel. The HLW is the liquid waste generated during the recovery of uranium and plutonium in a fuel processing plant that generally contains more than 99% of the nonvolatile fission products produced during reactor operation. Since this paper deals with waste separation processes, spent reactor fuel elements that have not been dissolved and further processed are excluded

Proposed classification scheme for high-level and other radioactive wastes

International Nuclear Information System (INIS)

Kocher, D.C.; Croff, A.G.

1986-01-01

The Nuclear Waste Policy Act (NWPA) of 1982 defines high-level radioactive waste (HLW) as: (A) the highly radioactive material resulting from the reprocessing of spent nuclear fuel....that contains fission products in sufficient concentrations; and (B) other highly radioactive material that the Commission....determines....requires permanent isolation. This paper presents a generally applicable quantitative definition of HLW that addresses the description in paragraph (B). The approach also results in definitions of other waste classes, i.e., transuranic (TRU) and low-level waste (LLW). A basic waste classification scheme results from the quantitative definitions

Intermediate and high energy nuclear reactions at the hadronic structural level

Energy Technology Data Exchange (ETDEWEB)

Slowinski, B [Institute of Physics, Warsaw, University of Technology, Poland, Institute of Atomic Energy, Swierk, (Poland)

1997-12-31

Form tens of MeV to several hundred of GeV is stretched out quite a large interval of energy when the interaction between hadrons (for instance, pion/nucleon-nucleus and nucleus-nucleus reactions) can be described by the considerably simplified way with still acceptable accuracy. This happens because in this energy region hadrons (i.e. pions, nucleons etc.) remain quasiparticles of nuclear matter mostly without revealing any internal structure, their de Broglie`s wavelength is much shorter as compared to the average intranuclear nucleon`s distance, and the energy transfers in the reaction are, on the average, significantly greater than the binding energy of nucleons inside nuclei. Consequently an approach to the analysis of these phenomena based on simple geometric and probabilistic considerations is justifiable, especially for many practical purposes, in particular, for shielding and dosimetric estimations, material behaviour prediction, as well as for the approximate evaluation of electronuclear breeding effects in different composites of target materials, for nuclear passivation problems and so on. In this work basic physical reasons of such a simplified picture of intermediate and high energy nuclear reactions are presented. The most usual phenomenological models of hadronic multiple emission/production and recent results of the cascade evaporation type models, are also discussed. 2 figs.

Environmental information document defense waste processing facility

International Nuclear Information System (INIS)

1981-07-01

This report documents the impact analysis of a proposed Defense Waste Processing Facility (DWPF) for immobilizing high-level waste currently being stored on an interim basis at the Savannah River Plant (SRP). The DWPF will process the waste into a form suitable for shipment to and disposal in a federal repository. The DWPF will convert the high-level waste into: a leach-resistant form containing above 99.9% of all the radioactivity, and a residue of slightly contaminated salt. The document describes the SRP site and environs, including population, land and water uses; surface and subsurface soils and waters; meteorology; and ecology. A conceptual integrated facility for concurrently producing glass waste and saltcrete is described, and the environmental effects of constructing and operating the facility are presented. Alternative sites and waste disposal options are addressed. Also environmental consultations and permits are discussed

Conceptualization of a hypothetical high-level nuclear waste repository site in unsaturated, fractured tuff

International Nuclear Information System (INIS)

Parsons, A.M.; Olague, N.E.; Gallegos, D.P.

1991-01-01

Under the sponsorship of the US Nuclear Regulatory Commission (NRC), Sandia National Laboratories (SNL) is developing a performance assessment methodology for the analysis of long-term disposal and isolation of high-level nuclear wastes (HLW) in alternative geologic media. As part of this exercise, SNL created a conceptualization of ground-water flow and radionuclide transport in the far field of a hypothetical HLW repository site located in unsaturated, fractured tuff formations. This study provides a foundation for the development of conceptual mathematical, and numerical models to be used in this performance assessment methodology. This conceptualization is site specific in terms of geometry, the regional ground-water flow system, stratigraphy, and structure in that these are based on information from Yucca Mountain located on the Nevada Test Site. However, in terms of processes in unsaturated, fractured, porous media, the model is generic. This report also provides a review and evaluation of previously proposed conceptual models of unsaturated and saturated flow and solute transport. This report provides a qualitative description of a hypothetical HLW repository site in fractured tuff. However, evaluation of the current knowledge of flow and transport at Yucca Mountain does not yield a single conceptual model. Instead, multiple conceptual models are possible given the existing information

Conceptualization of a hypothetical high-level nuclear waste repository site in unsaturated, fractured tuff

Energy Technology Data Exchange (ETDEWEB)

Parsons, A.M.; Olague, N.E.; Gallegos, D.P. [Sandia National Labs., Albuquerque, NM (USA)

1991-01-01

Under the sponsorship of the US Nuclear Regulatory Commission (NRC), Sandia National Laboratories (SNL) is developing a performance assessment methodology for the analysis of long-term disposal and isolation of high-level nuclear wastes (HLW) in alternative geologic media. As part of this exercise, SNL created a conceptualization of ground-water flow and radionuclide transport in the far field of a hypothetical HLW repository site located in unsaturated, fractured tuff formations. This study provides a foundation for the development of conceptual mathematical, and numerical models to be used in this performance assessment methodology. This conceptualization is site specific in terms of geometry, the regional ground-water flow system, stratigraphy, and structure in that these are based on information from Yucca Mountain located on the Nevada Test Site. However, in terms of processes in unsaturated, fractured, porous media, the model is generic. This report also provides a review and evaluation of previously proposed conceptual models of unsaturated and saturated flow and solute transport. This report provides a qualitative description of a hypothetical HLW repository site in fractured tuff. However, evaluation of the current knowledge of flow and transport at Yucca Mountain does not yield a single conceptual model. Instead, multiple conceptual models are possible given the existing information.

REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

Energy Technology Data Exchange (ETDEWEB)

W. Lee Poe, Jr

1998-10-01

In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are the same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site data was weighted to develop the regional site, and provides the weighted data used in the CSAR analysis. It is

REGIONAL BINNING FOR CONTINUED STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTES

International Nuclear Information System (INIS)

W. Lee Poe, Jr.

1998-01-01

In the Continued Storage Analysis Report (CSAR) (Reference 1), DOE decided to analyze the environmental consequences of continuing to store the commercial spent nuclear fuel (SNF) at 72 commercial nuclear power sites and DOE-owned spent nuclear fuel and high-level waste at five Department of Energy sites by region rather than by individual site. This analysis assumes that three commercial facilities pairs--Salem and Hope Creek, Fitzpatrick and Nine-Mile Point, and Dresden and Moms--share common storage due to their proximity to each other. The five regions selected for this analysis are shown on Figure 1. Regions 1, 2, and 3 are the same as those used by the Nuclear Regulatory Commission in their regulatory oversight of commercial power reactors. NRC Region 4 was subdivided into two regions to more appropriately define the two different climates that exist in NRC Region 4. A single hypothetical site in each region was assumed to store all the SNF and HLW in that region. Such a site does not exist and has no geographic location but is a mathematical construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect appropriate inventory, facility and material degradation, and radionuclide transport, the waste inventories, engineered barriers, and environmental conditions for the hypothetical sites were developed from data for each of the existing sites within the given region. Weighting criteria to account for the amount and types of SNF and HLW at each site were used in the development of the environmental data for the regional site, such that the results of the analyses for the hypothetical site were representative of the sum of the results of each actual site if they had been modeled independently. This report defines the actual site data used in development of this hypothetical site, shows how the individual site data was weighted to develop the regional site, and provides the weighted data used in the CSAR analysis. It is

Geochemistry research planning for the underground storage of high-level nuclear waste

International Nuclear Information System (INIS)

Apps, J.A.

1983-09-01

This report is a preliminary attempt to plan a comprehensive program of geochemistry research aimed at resolving problems connected with the underground storage of high-level nuclear waste. The problems and research needs were identified in a companion report to this one. The research needs were taken as a point of departure and developed into a series of proposed projects with estimated manpowers and durations. The scope of the proposed research is based on consideration of an underground repository as a multiple barrier system. However, the program logic and organization reflect conventional strategies for resolving technological problems. The projects were scheduled and the duration of the program, critical path projects and distribution of manpower determined for both full and minimal programs. The proposed research was then compared with ongoing research within DOE, NRC and elsewhere to identify omissions in current research. Various options were considered for altering the scope of the program, and hence its cost and effectiveness. Finally, recommendations were made for dealing with omissions and uncertainties arising from program implementation. 11 references, 6 figures, 4 tables

The management of used nuclear fuel and high level nuclear waste in Europe. A report for the Lovar Foundation for the Protection of the Environment

International Nuclear Information System (INIS)

Saunders, P.; Clark, S.; Francis, D.

1994-01-01

purpose of this report is to provide an explanation of the options available for the management of used nuclear fuels and highly radioactive waste. It is hoped that these explanations are presented in an unbiased way that is understandable to readers without a scientific background. A bibliography is, however, provided for those who wish to check the contents of the report, where possible, in peer reviewed literature sources. It has not been part of our remit to draw environmental conclusions on the content of the report, or to criticise the policies of governments or argue for or against any particular way of dealing with nuclear waste. The full report is divided into five chapters and three appendices. Following the introduction, Chapter 2 contains a description of the origins, characteristics and regulation of used nuclear fuel; the options for used fuel management are examined in Chapter 3; Chapter 4 contains a description of the environmental consequences of used fuel waste management; and Chapter 5 provides a comparison of the problems associates with radioactive waste compared with those of chemically toxic waste. The appendices contain detail on the properties of radioactive materials, radiation protection, and the ways in which European countries have dealt with spent fuel and high-level radioactive waste

Guidelines on the scope, content, and use of comprehensive risk assessment in the management of high-level nuclear waste transportation

International Nuclear Information System (INIS)

Golding, D.; White, A.

1990-12-01

This report discusses the scope of risk assessment strategies in the management of the transport of high-level radioactive wastes. In spite of the shortcomings of probabilistic risk assessment(PRA), the Transportation Needs Assessment recommended this as the preferred methodology to assess the risks of high level nuclear waste (HLNW) transportation. A PRA also will need to heed the lessons learned from the development and application of PRA elsewhere, such as in the nuclear power industry. A set of guidelines will aid this endeavor by outlining the appropriate scope, content, and use of a risk assessment which is more responsive to the uncertainties, human-technical interactions, social forces, and iterative relationship with risk management strategies, than traditional PRAS. This more expansive definition, which encompasses but is not totally reliant on rigorous data requirements and quantitative probability estimates, we term Comprehensive Risk Assessment (CRA) Guidelines will be developed in three areas: the limitations of existing methodologies and suggested modifications; CRA as part of a flexible, effective, adaptive risk management system for HLNW transportation; and, the use of CRA in risk communication

Evidence for dawsonite in Hanford high-level nuclear waste tanks.

Science.gov (United States)

Reynolds, Jacob G; Cooke, Gary A; Herting, Daniel L; Warrant, R Wade

2012-03-30

Gibbsite [Al(OH)(3)] and boehmite (AlOOH) have long been assumed to be the most prevalent aluminum-bearing minerals in Hanford high-level nuclear waste sludge. The present study shows that dawsonite [NaAl(OH)(2)CO(3)] is also a common aluminum-bearing phase in tanks containing high total inorganic carbon (TIC) concentrations and (relatively) low dissolved free hydroxide concentrations. Tank samples were probed for dawsonite by X-ray Diffraction (XRD), Scanning Electron Microscopy with Energy Dispersive Spectrometry (SEM-EDS) and Polarized Light Optical Microscopy. Dawsonite was conclusively identified in four of six tanks studied. In a fifth tank (AN-102), the dawsonite identification was less conclusive because it was only observed as a Na-Al bearing phase with SEM-EDS. Four of the five tank samples with dawsonite also had solid phase Na(2)CO(3) · H(2)O. The one tank without observable dawsonite (Tank C-103) had the lowest TIC content of any of the six tanks. The amount of TIC in Tank C-103 was insufficient to convert most of the aluminum to dawsonite (Al:TIC mol ratio of 20:1). The rest of the tank samples had much lower Al:TIC ratios (between 2:1 and 0.5:1) than Tank C-103. One tank (AZ-102) initially had dawsonite, but dawsonite was not observed in samples taken 15 months after NaOH was added to the tank surface. When NaOH was added to a laboratory sample of waste from Tank AZ-102, the ratio of aluminum to TIC in solution was consistent with the dissolution of dawsonite. The presence of dawsonite in these tanks is of significance because of the large amount of OH(-) consumed by dawsonite dissolution, an effect confirmed with AZ-102 samples. Copyright © 2012 Elsevier B.V. All rights reserved.

Type GQS-1 high pressure steam manifold water level monitoring system

International Nuclear Information System (INIS)

Li Nianzu; Li Beicheng; Jia Shengming

1993-10-01

The GQS-1 high pressure steam manifold water level monitoring system is an advanced nuclear gauge that is suitable for on-line detecting and monitor in high pressure steam manifold water level. The physical variable of water level is transformed into electrical pulses by the nuclear sensor. A computer is equipped for data acquisition, analysis and processing and the results are displayed on a 14 inch color monitor. In addition, a 4 ∼ 20 mA output current is used for the recording and regulation of water level. The main application of this gauge is for on-line measurement of high pressure steam manifold water level in fossil-fired power plant and other industries

A conceptual subsurface facility design for a high-level nuclear waste repository at Yucca Mountain

International Nuclear Information System (INIS)

McKenzie, D.G., III; Bhattacharyya, K.K.; Segrest, A.M.

1996-01-01

The US Department of Energy is responsible for the design, construction, operation and closure of a repository in which to permanently dispose of the nation's high level nuclear waste. In addition to the objective of safely isolating the waste inventory, the repository must provide a safe working environment for its workforce, and protect the public. The conceptual design for this facility is currently being developed. Tunnel Boring Machine will be used to excavate 228 kilometers of tunneling to construct the facility over a 30 year period. The excavation operations will be physically separated from the waste emplacement operations, and each operation will have its own dedicated ventilation system. The facility is being designed to remain open for 150 years

Anticipated Degradation Modes of Metallic Engineered Barriers for High-Level Nuclear Waste Repositories

Science.gov (United States)

Rodríguez, Martín A.

2014-03-01

Metallic engineered barriers must provide a period of absolute containment to high-level radioactive waste in geological repositories. Candidate materials include copper alloys, carbon steels, stainless steels, nickel alloys, and titanium alloys. The national programs of nuclear waste management have to identify and assess the anticipated degradation modes of the selected materials in the corresponding repository environment, which evolves in time. Commonly assessed degradation modes include general corrosion, localized corrosion, stress-corrosion cracking, hydrogen-assisted cracking, and microbiologically influenced corrosion. Laboratory testing and modeling in metallurgical and environmental conditions of similar and higher aggressiveness than those expected in service conditions are used to evaluate the corrosion resistance of the materials. This review focuses on the anticipated degradation modes of the selected or reference materials as corrosion-resistant barriers in nuclear repositories. These degradation modes depend not only on the selected alloy but also on the near-field environment. The evolution of the near-field environment varies for saturated and unsaturated repositories considering backfilled and unbackfilled conditions. In saturated repositories, localized corrosion and stress-corrosion cracking may occur in the initial aerobic stage, while general corrosion and hydrogen-assisted cracking are the main degradation modes in the anaerobic stage. Unsaturated repositories would provide an oxidizing environment during the entire repository lifetime. Microbiologically influenced corrosion may be avoided or minimized by selecting an appropriate backfill material. Radiation effects are negligible provided that a thick-walled container or an inner shielding container is used.

Defense Waste Processing Facility Process Simulation Package Life Cycle

International Nuclear Information System (INIS)

Reuter, K.

1991-01-01

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

Waste acceptance product specifications for vitrified high-level waste forms

International Nuclear Information System (INIS)

Applewhite-Ramsey, A.; Sproull, J.F.

1994-01-01

The Department of Energy (DOE) Office of Environmental Restoration and Waste Management (EM) has developed Waste Acceptance Product Specifications (EM-WAPS). The EM-WAPS will be the basis for defining product acceptance criteria compatible with the requirements of the Civilian Radioactive Waste Management System (CRWMS). The relationship between the EM-WAPS and the CRWMS Systems Requirements document (WA-SRD) will be discussed. The impact of the EM-WAPS on the Savannah River Sit (SRS) Defense Waste Processing Facility's (DWPF) Waste Acceptance Program, Waste Qualification Run planning, and startup schedule will also be reported. 14 refs., 2 tabs

Determination of Reportable Radionuclides for DWPF Sludge Batch 2 (Macro Batch 3)

International Nuclear Information System (INIS)

Bibler, N.E.

2002-01-01

The Waste Acceptance Product Specifications (WAPS) 1.2 require that ''The Producer shall report the inventory of radionuclides (in Curies) that have half-lives longer than 10 years and that are, or will be, present in concentrations greater than 0.05 percent of the total inventory for each waste type indexed to the years 2015 and 3115''. As part of the strategy to meet WAPS 1.2, the Defense Waste Processing Facility (DWPF) will report for each waste type, all radionuclides (with half-lives greater than 10 years) that have concentrations greater than 0.01 percent of the total inventory from time of production through the 1100 year period from 2015 through 3115. The initial listing of radionuclides to be included is based on the design-basis glass as identified in the Waste Form Compliance Plan (WCP) and Waste Form Qualification Report (WQR). However, it is required that this list be expanded if other radionuclides with half-lives greater than 10 years are identified that meet the greater than 0.01 percent criterion for Curie content

ONDRAF/NIRAS and high-level radioactive waste management in Belgium

International Nuclear Information System (INIS)

Decamps, F.

1993-01-01

The National Agency for Radioactive Waste and Enriched Fissile Materials, ONDRAF/NIRAS, is a public body with legal personality in charge of managing all radioactive waste on Belgian territory, regardless of its origin and source. It is also entrusted with tasks related to the management of enriched fissile materials, plutonium containing materials and used or unused nuclear fuel, and with certain aspects of the dismantling of closed down nuclear facilities. High-level radioactive waste management comprises essentially and for the time being the storage of high-level liquid waste produced by the former EUROCHEMIC reprocessing plant and of high-level and very high-level heat producing waste resulting from the reprocessing in France of Belgian spent fuel, as well as research and development (R and D) with regard to geological disposal in clay of this waste type

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 2: Appendices

Energy Technology Data Exchange (ETDEWEB)

Rechard, R.P. [ed.

1993-12-01

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste, as mandated by the Nuclear Waste Policy Act of 1982. The waste forms comprised about 700 metric tons of initial heavy metal (or equivalent units) stored at the INEL: graphite spent fuel, experimental low enriched and highly enriched spent fuel, and high-level waste generated during reprocessing of some spent fuel. Five different waste treatment options were studied; in the analysis, the options and resulting waste forms were analyzed separately and in combination as five waste disposal groups. When the waste forms were studied in combination, the repository was assumed to also contain vitrified high-level waste from three DOE sites for a common basis of comparison and to simulate the impact of the INEL waste forms on a moderate-sized repository, The performance of the waste form was assessed within the context of a whole disposal system, using the U.S. Environmental Protection Agency`s Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 40 CFR 191, promulgated in 1985. Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

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

Nuclear waste. DOE's program to prepare high-level radioactive waste for final disposal

International Nuclear Information System (INIS)

Bannerman, Carl J.; Owens, Ronald M.; Dowd, Leonard L.; Herndobler, Christopher S.; Purvine, Nancy R.; Stenersen, Stanley G.

1989-11-01

years later than the schedule established in early 1984, and the cost could be about $1.1 billion, more than double the 1984 cost estimate. DOE has plans for immobilization facilities at the other two Sites, but unresolved issues could affect the reliability of current cost and schedule estimates; the Hanford facility, currently in the design phase, has an estimated immobilization completion date of 2008, but this date assumes that Hanford's defense mission nuclear processing activities will end in the mid 1990s and only the waste stored in Hanford's double-shell tanks will be immobilized; the INEL facility is currently in such an early planning phase that DOE has not yet selected the waste immobilization technology that it will use. The waste may be transformed into a glass-ceramic or other material instead of being vitrified. DOE expects to make this decision in 1993. Section 1 contains an overview of DOE's high-level waste immobilization program. Sections 2 through 5 contain more detailed information about each of the four projects

SPEEDUP simulation of liquid waste batch processing. Revision 1

International Nuclear Information System (INIS)

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

1994-01-01

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

High system-safety level of nuclear power stations

International Nuclear Information System (INIS)

Lutz, H.R.

1976-01-01

A bluntly worded disquisition contrasting the incidence of death and harm to persons in the chemical industry with the low hazards in nuclear power stations. Quotes conclusions from a U.S. accident study that the risk from 100 large power stations is 100 times smaller than from chlorine manufacture and transport. The enclosure of a reactor in a safety container, the well understood effects of radioactivity on man, and the ease of measuring leakage well below safe limits, are safety features which he considers were not matched in the products and plant of the Seveso factory which suffered disaster. Questions the usefulness of warnings about nuclear dangers when chemical dangers are much greater and road dangers very much greater still. (R.W.S.)

Radioactive waste isolation in salt: peer review of the Office of Nuclear Waste Isolation's report on Functional Design Criteria for a Repository for High-Level Radioactive Waste

International Nuclear Information System (INIS)

Hambley, D.F.; Russell, J.E.; Busch, J.S.; Harrison, W.; Edgar, D.E.; Tisue, M.W.

1984-08-01

This report summarizes Argonne's review of the Office of Nuclear Waste Isolation's (ONWI's) draft report entitled Functional Design Criteria for High-Level Nuclear Waste Repository in Salt, dated January 23, 1984. Recommendations are given for improving the ONWI draft report

Miscibility Evaluation Of The Next Generation Solvent With Polymers Currently Used At DWPF, MCU, And Saltstone

Energy Technology Data Exchange (ETDEWEB)

Fondeur, F. F.

2013-04-17

The Office of Waste Processing, within the Office of Technology Innovation and Development, funded the development of an enhanced Caustic-Side Solvent Extraction (CSSX) solvent for deployment at the Savannah River Site for removal of cesium from High Level Waste. This effort lead to the development of the Next Generation Solvent (NGS) with Tris (3,7-dimethyl octyl) guanidine (TiDG). The first deployment target for the NGS solvent is within the Modular CSSX Unit (MCU). Deployment of a new chemical within an existing facility requires verification that the new chemical components are compatible with the installed equipment. In the instance of a new organic solvent, the primary focus is on compatibility of the solvent with organic polymers used in the affected facility. This report provides the calculated data from exposing these polymers to the Next Generation Solvent. An assessment of the dimensional stability of polymers known to be used or present in the MCU, Defense Waste Processing Facility (DWPF), and Saltstone facilities that will be exposed to the NGS showed that TiDG could selectively affect the elastomers and some thermoplastics to varying extents, but the typical use of these polymers in a confined geometry will likely prevent the NGS from impacting component performance. The polymers identified as of primary concern include Grafoil® (flexible graphite), Tefzel®, Isolast®, ethylene-propylene-diene monomer (EPDM) rubber, nitrile-butadiene rubber (NBR), styrene-butadiene rubber (SBR), ultra high molecular weight polyethylene (UHMWPE), and fluorocarbon rubber (FKM). Certain polymers like NBR and EPDM were found to interact mildly with NGS but their calculated swelling and the confined geometry will impede interaction with NGS. In addition, it was found that Vellumoid (cellulose fibers-reinforced glycerin and protein) may leach protein and Polyvinyl Chloride (PVC) may leach plasticizer (such as Bis-Ethylhexyl-Phthalates) into the NGS solvent. Either case

Effects of long-term exposure of tuffs to high-level nuclear waste-repository conditions. Preliminary report

International Nuclear Information System (INIS)

Blacic, J.; Carter, J.; Halleck, P.; Johnson, P.; Shankland, T.; Andersen, R.; Spicochi, K.; Heller, A.

1982-02-01

Tests have been performed to explore the effects of extended exposure of tuffs from the southwestern portion of the Nevada Test Site to temperatures and pressures similar to those that will be encountered in a high-level nuclear waste repository. Tuff samples ranging from highly welded, nonzeolitized to unwelded, highly zeolitized varieties were subjected to temperatures of 80, 120, and 180 0 C; confining pressures of 9.7 and 19.7 MPa; and water-pore pressures of 0.5 to 19.7 MPa for durations of 2 to 6 months. The following basic properties were measured before and after exposure and compared: tensile strength, uniaxial compressive strength, grain density, porosity, mineralogy, permeability, thermal expansion, and thermal conductivity. Depending on rock type and exposure conditions, significant changes in ambient tensile strength, compressive strength, grain density, and porosity were measured. Mineralogic examination, permeability, and thermal property measurements remain to be completed

NOBLE METAL CHEMISTRY AND HYDROGEN GENERATION DURING SIMULATED DWPF MELTER FEED PREPARATION

Energy Technology Data Exchange (ETDEWEB)

Koopman, D

2008-06-25

Simulations of the Defense Waste Processing Facility (DWPF) Chemical Processing Cell vessels were performed with the primary purpose of producing melter feeds for the beaded frit program plus obtaining samples of simulated slurries containing high concentrations of noble metals for off-site analytical studies for the hydrogen program. Eight pairs of 22-L simulations were performed of the Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles. These sixteen simulations did not contain mercury. Six pairs were trimmed with a single noble metal (Ag, Pd, Rh, or Ru). One pair had all four noble metals, and one pair had no noble metals. One supporting 4-L simulation was completed with Ru and Hg. Several other 4-L supporting tests with mercury have not yet been performed. This report covers the calculations performed on SRNL analytical and process data related to the noble metals and hydrogen generation. It was originally envisioned as a supporting document for the off-site analytical studies. Significant new findings were made, and many previous hypotheses and findings were given additional support as summarized below. The timing of hydrogen generation events was reproduced very well within each of the eight pairs of runs, e.g. the onset of hydrogen, peak in hydrogen, etc. occurred at nearly identical times. Peak generation rates and total SRAT masses of CO{sub 2} and oxides of nitrogen were reproduced well. Comparable measures for hydrogen were reproduced with more variability, but still reasonably well. The extent of the reproducibility of the results validates the conclusions that were drawn from the data.

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-11-13

The principal objective of the work described in this Final Report is to develop and identify glass frit compositions for a specified DWPF high-aluminum based sludge waste stream that maximizes waste loading while maintaining high production rate for the waste composition provided by ORP/SRS. This was accomplished through a combination of crucible-scale, vertical gradient furnace, and confirmation tests on the DM100 melter system. The DM100-BL unit was selected for these tests. The DM100-BL was used for previous tests on HLW glass compositions that were used to support subsequent tests on the HLW Pilot Melter. It was also used to process compositions with waste loadings limited by aluminum, bismuth, and chromium, to investigate the volatility of cesium and technetium during the vitrification of an HLW AZ-102 composition, to process glass formulations at compositional and property extremes, and to investigate crystal settling on a composition that exhibited one percent crystals at 963{degrees}C (i.e., close to the WTP limit). The same melter was selected for the present tests in order to maintain comparisons between the previously collected data. The tests provide information on melter processing characteristics and off-gas data, including formation of secondary phases and partitioning. Specific objectives for the melter tests are as follows: Determine maximum glass production rates without bubbling for a simulated SRS Sludge Batch 19 (SB19). Demonstrate a feed rate equivalent to 1125 kg/m{sup 2}/day glass production using melt pool bubbling. Process a high waste loading glass composition with the simulated SRS SB19 waste and measure the quality of the glass product. Determine the effect of argon as a bubbling gas on waste processing and the glass product including feed processing rate, glass redox, melter emissions, etc.. Determine differences in feed processing and glass characteristics for SRS SB19 waste simulated by the co-precipitated and direct

Management of high level nuclear waste - the nordic approach

International Nuclear Information System (INIS)

Engstrom, S.; Aikas, T.

2000-01-01

Both the Swedish and the Finnish nuclear waste programmes are aimed at disposal of encapsulated spent nuclear fuel into the crystalline bedrock. In both countries research and development work have been performed since the 1970's. The focus of the programme in both countries is now shifting to practical demonstration of encapsulation technology. In parallel a site-selection programme is being carried out. Finland has selected a site at Eurajoki and is currently waiting for the Government to agree to the choice of the site. In Sweden, at least two sites will be selected by year 2001 with the goal, after performed drillings, to select one of them around 2008. Site selection for the deep repository is probably the most difficult and most sensitive part of the whole programme. The repository will be sited at a suitable place in Sweden respectively Finland where high safety requirements will be met with the consent of the concerned municipality. If there is a Nordic approach to tackle this issue that would probably be: - A stepwise approach in which the disposal is implemented in gradually each step having a decision making stage leading to a commitment of various parties involved to the following stage. -A total transparency of the work performed and the decision making process. - A genuine will from the industry to establish a dialogue with the public in the involved communities. - A will to take the time and the patience necessary to establish a constructive working relationship with the communities participating in the site selection. (authors)

Is Yucca Mountain a long-term solution for disposing of US spent nuclear fuel and high-level radioactive waste?

Science.gov (United States)

Thorne, M C

2012-06-01

On 26 January 2012, the Blue Ribbon Commission on America's Nuclear Future released a report addressing, amongst other matters, options for the managing and disposal of high-level waste and spent fuel. The Blue Ribbon Commission was not chartered as a siting commission. Accordingly, it did not evaluate Yucca Mountain or any other location as a potential site for the storage or disposal of spent nuclear fuel and high-level waste. Nevertheless, if the Commission's recommendations are followed, it is clear that any future proposals to develop a repository at Yucca Mountain would require an extended period of consultation with local communities, tribes and the State of Nevada. Furthermore, there would be a need to develop generally applicable regulations for disposal of spent fuel and high-level radioactive waste, so that the Yucca Mountain site could be properly compared with alternative sites that would be expected to be identified in the initial phase of the site-selection process. Based on what is now known of the conditions existing at Yucca Mountain and the large number of safety, environmental and legal issues that have been raised in relation to the DOE Licence Application, it is suggested that it would be imprudent to include Yucca Mountain in a list of candidate sites for future evaluation in a consent-based process for site selection. Even if there were a desire at the local, tribal and state levels to act as hosts for such a repository, there would be enormous difficulties in attempting to develop an adequate post-closure safety case for such a facility, and in showing why this unsaturated environment should be preferred over other geological contexts that exist in the USA and that are more akin to those being studied and developed in other countries.

NMTC/JAM, Simulates High Energy Nuclear Reactions and Nuclear-Meson Transport Processes

International Nuclear Information System (INIS)

Furihata, Shiori

2002-01-01

1 - Description of program or function: NMTC/JAM is an upgraded version of the code system NMTC/JAERI97. NMTC/JAERI97 simulates high energy nuclear reactions and nucleon-meson transport processes. It implements an intra-nuclear cascade model taking account of the in-medium nuclear effects and the pre-equilibrium calculation model based on the exciton one. For treating the nucleon transport process, the nucleon-nucleus cross sections are revised to those derived by the systematics of Pearlstein. Moreover, the level density parameter derived by Ignatyuk is included as a new option for particle evaporation calculation. A geometry package based on the Combinatorial Geometry with multi-array system and the importance sampling technique is implemented in the code. Tally function is also employed for obtaining such physical quantities as neutron energy spectra, heat deposition and nuclide yield without editing a history file. The code can simulate both the primary spallation reaction and the secondary particle transport in the intermediate energy region from 20 MeV to 3.5 GeV by the use of the Monte Carlo technique. The code has been employed in combination with the neutron-photon transport codes available to the energy region below 20 MeV for neutronics calculation of accelerator-based subcritical reactors, analyses of thick target spallation experimented and so on. 2 - Methods: High energy nuclear reactions induced by incident high energy protons, neutrons and pions are simulated with the Monte Carlo Method by the intra-nuclear nucleon-nucleon reaction probabilities based on an intra-nuclear nucleon cascade model followed by the particle evaporation including high energy fission process. Jet-Aa Microscopic transport model (JAM) is employed to simulate high energy nuclear reactions in the energy range of GeV. All reaction channels are taken into account in the JAM calculation. An intra-nuclear cascade model (ISOBAR code) taking account of the in-medium nuclear effects

Saltstone processing startup at the Savannah River Plant

International Nuclear Information System (INIS)

Wilhite, E.L.; Langton, C.A.; Sturm, H.F.; Hooker, R.L.; Occhipinti, E.S.

1988-01-01

High-level nuclear wastes are stored in large underground tanks at the Savannah River Plant. Processing of this waste in preparation for ultimate disposal will begin in 1988. The waste will be processed to separate the high-level radioactive fraction from the low-level radioactive fraction. The separation will be made in existing waste tanks by a process combining precipitation, adsorption, and filtration. The high-level fraction will be vitrified into borosilicate glass in the Defense Waste Processing Facility (DWPF) for permanent disposal in a federal repository. The low-level fraction (decontaminated salt solution) will be mixed with a cementitious slag-flyash blend. The resulting wasteform, saltstone, will be disposed of onsite by emplacement in an engineered facility. Waste properties, disposal facility details, and wasteform characteristics are discussed. In particular, details of saltstone processing, focusing on experience obtained from facility startup, are presented

Thermodynamics of excited nuclei and nuclear level densities

International Nuclear Information System (INIS)

Ramamurthy, V.S.

1977-01-01

A review has been made of the different approaches that are being used for a theoretical calculation of nuclear level densities. It is pointed out that while the numerical calculations based on the partition function approach and shell model single particle level schemes have shed important insight into the influence of nuclear shell effects on level densities and its excitation energy dependence and have brought out the inadequacy of the conventional Bethe Formula, these calculations are yet to reach a level where they can be directly used for quantitative comparisons. Some of the important drawbacks of the numerical calculations are also discussed. In this context, a new semi-empirical level density formula is described which while retaining the simplicity of analytical formulae, takes into account nuclear shell effects in a more realistic manner. (author)

Geologic disposal as optimal solution of managing the spent nuclear fuel and high-level radioactive waste

International Nuclear Information System (INIS)

Ilie, P.; Didita, L.; Ionescu, A.; Deaconu, V.

2002-01-01

To date there exist three alternatives for the concept of geological disposal: 1. storing the high-level waste (HLW) and spent nuclear fuel (SNF) on ground repositories; 2. solutions implying advanced separation processes including partitioning and transmutation (P and T) and eventual disposal in outer space; 3. geological disposal in repositories excavated in rocks. Ground storing seems to be advantageous as it ensures a secure sustainable storing system over many centuries (about 300 years). On the other hand ground storing would be only a postponement in decision making and will be eventually followed by geological disposal. Research in the P and T field is expected to entail a significant reduction of the amount of long-lived radioactive waste although the long term geological disposal will be not eliminated. Having in view the high cost, as well as the diversity of conditions in the countries owning power reactors it appears as a reasonable regional solution of HLW disposal that of sharing a common geological disposal. In Romania legislation concerning of radioactive waste is based on the Law concerning Spent Nuclear Fuel and Radioactive Waste Management in View of Final Disposal. One admits at present that for Romania geological disposal is not yet a stressing issue and hence intermediate ground storing of SNF will allow time for finding a better final solution

A heated large block test for high level nuclear waste management

International Nuclear Information System (INIS)

Lin, W.; Wilder, D.G.; Blink, J.A.; Blair, S.C.; Buscheck, T.A.; Glassley, W.E.; Lee, K.; Owens, M.W.; Roberts, J.J.

1995-01-01

The radioactive decay heat from high-level nuclear waste may, depending on the thermal load, create coupled thermal-mechanical-hydrological-chemical (TMHC) processes in the host rock of a repository. A heated large block test (LBT) is designed to understand some of the TNMC processes. A block of Topopah Spring tuff of about 3 x 3 x 4.5 m was isolated at Fran Ridge, Nevada Test Site. Small blocks of the rock adjacent to the large block were collected for laboratory testing of some individual thermal-mechanical, thermal-hydrological, and thermal-chemical processes. The large block will be heated by heaters within so that a dryout zone and a condensate zone will exist simultaneously. Guard heaters on the block sides will be used to minimize horizontal heat losses. A constant load of about 4 MPa will be applied to the top and sides of the large block. The sides will be sealed with moisture and thermal barriers. Temperature, moisture content, pore pressure, chemical composition, stress, displacement, electrical resistivity, acoustic emissions, and acoustic velocities will be measured throughout the block during the heating and cool-down phases. The results from the experiments on small blocks and the tests on the large block will provide a better understanding of some concepts of the coupled TMHC processes. The progress of the project is presented in this paper

Westinghouse-DOE integration: Meeting the challenge

International Nuclear Information System (INIS)

Price, S.V.

1992-01-01

The Westinghouse Electric Corporation (WEC) is in a unique position to affect national environmental management policy approaching the 21st Century. Westinghouse companies are management and operating contractors (MOC,s) at several environmentally pivotal government-owned, contractor operated (GOCO) facilities within the Department of Energy's (DOE's) nuclear defense complex. One way the WEC brings its companies together is by activating teams to solve specific DOE site problems. For example, one challenging issue at DOE facilities involves the environmentally responsible, final disposal of transuranic and high-level nuclear wastes (TRUs and HLWS). To address these disposal issues, the DOE supports two Westinghouse-based task forces: The TRU Waste Acceptance Criteria Certification Committee (WACCC) and the HLW Vitrification Committee. The WACCC is developing methods to characterize an estimated 176,287 cubic meters of retrievably stored TRUs generated at DOE production sites. Once characterized, TRUs could be safely deposited in the WIPP repository. The Westinghouse HLW Vitrification Committee is dedicated to assess appropriate methods to process an estimated 380,702 cubic meters of HLWs currently stored in underground storage tanks (USTs). As planned, this processing will involve segregating, and appropriately treating, low level waste (LLW) and HLW tank constituents for eventual disposal. The first unit designed to process these nuclear wastes is the SRS Defense Waste Processing Facility (DWPF). Initiated in 1973, the DWPF project is scheduled to begin operations in 1991 or 1992. Westinghouse companies are also working together to achieve appropriate environmental site restoration at DOE sites via the GOCO Environmental Restoration Committee

Study of nuclear level density parameter and its temperature dependence

International Nuclear Information System (INIS)

Nasrabadi, M. N.; Behkami, A. N.

2000-01-01

The nuclear level density ρ is the basic ingredient required for theoretical studies of nuclear reaction and structure. It describes the statistical nuclear properties and is expressed as a function of various constants of motion such as number of particles, excitation energy and angular momentum. In this work the energy and spin dependence of nuclear level density will be presented and discussed. In addition the level density parameter α will be extracted from this level density information, and its temperature and mass dependence will be obtained

Development of high-level waste solidification technology 1

Energy Technology Data Exchange (ETDEWEB)

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

1999-02-01

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

Final Report - Engineering Study for DWPF Bubblers, VSL-10R1770-1, Rev. 0, dated 12/22/10

Energy Technology Data Exchange (ETDEWEB)

Kruger, Albert A.; Joseph, I.; Matlack, K. S.; Kot, W. K.; Diener, G. A.; Pegg, I. L.; Callow, R. A.

2013-11-13

The objective of this work was to perform an engineering assessment of the impact of implementation of bubblers to improve mixing of the glass pool, and thereby increase throughput, in the Defense Waste Processing Facility (DWPF) on the melter and off-gas system. Most of the data used for this evaluation were from extensive melter tests performed on non-SRS feeds. This information was supplemented by more recent results on SRS HLW simulants that were tested on a melter system at VSL under contracts from ORP and SRR. Per the work scope, the evaluation focused on the following areas: Glass production rate; Corrosion of melter components; Power requirements; Pouring stability; Off-gas characteristics; Safety and flammability.

The disposal of high level nuclear waste in the oceans

International Nuclear Information System (INIS)

Vilks, Gustavs

1976-01-01

A report is given on a meeting held at Woods Hole, Massachusetts to consider the feasibility of using the sea bed as a disposal site for highly radioactive nuclear waste. Some disadvantages are explained, the chief being that ocean water alone, regardless of depth, is a poor barrier. Some delegates discussed emplacement of vitrified waste on the ocean floor, and others its burial in rock or sediment below the floor. The most suitable sites are the mid-plate/mid-gyse localities of abyssal hills. Some engineering work on submarine burial has been done by Sandia Labs., in the U.S.A. Ocean disposal is particularly interesting to Britain and Japan. Data on biological transport rates are needed. (author)

Effect of pairing in nuclear level density at low temperatures

International Nuclear Information System (INIS)

Rhine Kumar, A.K.; Modi, Swati; Arumugam, P.

2013-01-01

The nuclear level density (NLD) has been an interesting topic for researchers, due its importance in many aspects of nuclear physics, nuclear astrophysics, nuclear medicine, and other applied areas. The calculation of NLD helps us to understand the energy distribution of the excited levels of nuclei, entropy, specific heat, reaction cross sections etc. In this work the effect of temperature and pairing on level-density of the nucleus 116 Sn has been studied

Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

Energy Technology Data Exchange (ETDEWEB)

Rechard, R.P. [ed.

1995-03-01

This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results

International Nuclear Information System (INIS)

Rechard, R.P.

1995-03-01

This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM

Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)

Energy Technology Data Exchange (ETDEWEB)

F. Hua; P. Pasupathi; N. Brown; K. Mon

2005-09-19

The safe disposal of radioactive waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels for plants, animals, and humans. All of the countries currently studying the options for disposing of high-level nuclear waste (HLW) have selected deep geologic formations to be the primary barrier for accomplishing this isolation. In U.S.A., the Nuclear Waste Policy Act of 1982 (as amended in 1987) designated Yucca Mountain in Nevada as the potential site to be characterized for high-level nuclear waste (HLW) disposal. Long-term containment of waste and subsequent slow release of radionuclides into the geosphere will rely on a system of natural and engineered barriers including a robust waste containment design. The waste package design consists of a highly corrosion resistant Ni-based Alloy 22 cylindrical barrier surrounding a Type 316 stainless steel inner structural vessel. The waste package is covered by a mailbox-shaped drip shield composed primarily of Ti Grade 7 with Ti Grade 24 structural support members. The U.S. Yucca Mountain Project has been studying and modeling the degradation issues of the relevant materials for some 20 years. This paper reviews the state-of-the-art understanding of the degradation processes based on the past 20 years studies on Yucca Mountain Project (YMP) materials degradation issues with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the 10,000 years regulatory period. This paper provides an overview of the current understanding of the likely degradation behavior of the waste package and drip shield in the repository after the permanent closure of the facility. The degradation scenario discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking and hydrogen induced

Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)

International Nuclear Information System (INIS)

Hua, F.; Pasupathi, P.; Brown, N.; Mon, K.

2005-01-01

The safe disposal of radioactive waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels for plants, animals, and humans. All of the countries currently studying the options for disposing of high-level nuclear waste (HLW) have selected deep geologic formations to be the primary barrier for accomplishing this isolation. In U.S.A., the Nuclear Waste Policy Act of 1982 (as amended in 1987) designated Yucca Mountain in Nevada as the potential site to be characterized for high-level nuclear waste (HLW) disposal. Long-term containment of waste and subsequent slow release of radionuclides into the geosphere will rely on a system of natural and engineered barriers including a robust waste containment design. The waste package design consists of a highly corrosion resistant Ni-based Alloy 22 cylindrical barrier surrounding a Type 316 stainless steel inner structural vessel. The waste package is covered by a mailbox-shaped drip shield composed primarily of Ti Grade 7 with Ti Grade 24 structural support members. The U.S. Yucca Mountain Project has been studying and modeling the degradation issues of the relevant materials for some 20 years. This paper reviews the state-of-the-art understanding of the degradation processes based on the past 20 years studies on Yucca Mountain Project (YMP) materials degradation issues with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the 10,000 years regulatory period. This paper provides an overview of the current understanding of the likely degradation behavior of the waste package and drip shield in the repository after the permanent closure of the facility. The degradation scenario discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking and hydrogen induced

Proposed classification scheme for high-level and other radioactive wastes

International Nuclear Information System (INIS)

Kocher, D.C.; Croff, A.G.

1986-01-01

The Nuclear Waste Policy Act (NWPA) of 1982 defines high-level (radioactive) waste (HLW) as (A) the highly radioactive material resulting from the reprocessing of spent nuclear fuel...that contains fission products in sufficient concentrations; and (B) other highly radioactive material that the Commission...determines...requires permanent isolation. This paper presents a generally applicable quantitative definition of HLW that addresses the description in paragraph B. The approach also results in definitions of other wastes classes, i.e., transuranic (TRU) and low-level waste (LLW). The basic waste classification scheme that results from the quantitative definitions of highly radioactive and requires permanent isolation is depicted. The concentrations of radionuclides that correspond to these two boundaries, and that may be used to classify radioactive wastes, are given

Selection criteria for container materials at the proposed Yucca Mountain high level nuclear waste repository

International Nuclear Information System (INIS)

Halsey, W.G.

1989-11-01

A geological repository has been proposed for the permanent disposal of the nation's high level nuclear waste at Yucca Mountain in the Nevada desert. The containers for this waste must remain intact for the unprecedented service lifetime of 1000 years. A combination of engineering, regulatory, and licensing requirements complicate the container material selection. In parallel to gathering information regarding the Yucca Mountain service environment and material performance data, a set of selection criteria have been established which compare candidate materials to the performance requirements, and allow a quantitative comparison of candidates. These criteria assign relative weighting to varied topic areas such as mechanical properties, corrosion resistance, fabricability, and cost. Considering the long service life of the waste containers, it is not surprising that the corrosion behavior of the material is a dominant factor. 7 refs

Recommendations for rheological testing and modelling of DWPF melter feed slurries

International Nuclear Information System (INIS)

Shadday, M.A. Jr.

1994-08-01

The melter feed in the DWPF process is a non-Newtonian slurry. In the melter feed system and the sampling system, this slurry is pumped at a wide range of flow rates through pipes of various diameters. Both laminar and turbulent flows are encountered. Good rheology models of the melter feed slurries are necessary for useful hydraulic models of the melter feed and sampling systems. A concentric cylinder viscometer is presently used to characterize the stress/strain rate behavior of the melter feed slurries, and provide the data for developing rheology models of the fluids. The slurries exhibit yield stresses, and they are therefore modelled as Bingham plastics. The ranges of strain rates covered by the viscometer tests fall far short of the entire laminar flow range, and therefore hydraulic modelling applications of the present rheology models frequently require considerable extrapolation beyond the range of the data base. Since the rheology models are empirical, this cannot be done with confidence in the validity of the results. Axial pressure drop versus flow rate measurements in a straight pipe can easily fill in the rest of the laminar flow range with stress/strain rate data. The two types of viscometer tests would be complementary, with the concentric cylinder viscometer providing accurate data at low strain rates, near the yield point if one exists, and pipe flow tests providing data at high strain rates up to and including the transition to turbulence. With data that covers the laminar flow range, useful rheological models can be developed. In the Bingham plastic model, linear behavior of the shear stress as a function of the strain rate is assumed once the yield stress is exceeded. Both shear thinning and shear thickening behavior have been observed in viscometer tests. Bingham plastic models cannot handle this non-linear behavior, but a slightly more complicated yield/power law model can

The importance of counting cows: Social and economic effects of a high-level nuclear waste repository in Texas

International Nuclear Information System (INIS)

Fleishman, J.; Brody, J.; Galavotti, C.

1987-01-01

Impact assessments that rely on existing records and extrapolation from broad geographic areas provide inadequate information about social and economic conditions important in siting a high-level nuclear waste repository. Texas has used an alternative approach, involving systematic surveys of representative samples of local residents, farm operators and businesses in the proposed site counties and comparison areas. Results show that this technique is useful in describing current economic conditions, including characteristics of key sectors of the economy, changes related to the siting process, and expectations that may influence investment. In addition, the surveys are useful in assessing the degree of consensus in local communities and in identifying possible differential effects of a repository on particular groups. They also provide a baseline for long-term monitoring of repository effects and contribute to their understanding of the underlying processes that shape public response to the nuclear waste program

Current status of high level radioactive waste disposal in Japan and foreign countries

International Nuclear Information System (INIS)

Tanaka, Satoru; Tanabe, Hiromi; Inagaki, Yusuke; Ishida, Hisahiro; Kato, Osamu; Kurata, Mitsuyuki; Yamachika, Hidehiko

2002-01-01

At a time point of 2002, there is no country actually disposing high level radioactive wastes into grounds, but in most of countries legislative preparation and practicing agents are carried out and site selection is promoted together with energetic advancement of its R and Ds. As disposal methods of the high level radioactive wastes, various methods such as space disposal, oceanic bottom disposal, ice bed disposal, ground disposal, and so on have been examined. And, a processing technology called partitioning and transmutation technology separating long-lived radionuclides from liquid high level radioactive waste and transmutation into short-lived or harmless radionuclides has also been studied. Here was introduced their wrestling conditions in Japan and main foreign countries, as a special issue of the Current status of high level radioactive waste disposal in Japan and foreign countries'. The high level radioactive wastes (glassification solids or spent nuclear fuels) are wastes always formed by nuclear power generation and establishment of technologies is an important subject for nuclear fuel cycle. (G.K.)

Hydraulic properties of buffer and backfill materials for high-level nuclear waste disposal

International Nuclear Information System (INIS)

Komine, Hideo; Ogata, Nobuhide

2001-01-01

The design and development of buffer and backfill materials, which fill up the disposal facility, are important for developing the technology of high-level nuclear wastes disposal. The compacted bentonite and sand-bentonite mixture are attracting greater attention as buffer and backfill materials because they have impermeable and swelling properties. This study investigated the hydraulic-conductivities at the different sand-bentonite mass ratio and dry density, which are the specifications of material, by the experimental works. This study also obtained the experimental data of hydraulic conductivities of the materials for 120 days at the farthest, and the permeability changes before and after swelling. Furthermore, this study proposed the evaluation method for hydraulic conductivity using the parameter 'Swelling volumetric strain of montmorillonite', which was proposed by the author. The evaluation method can obtain the hydraulic conductivity of buffer and backfill materials at various dry densities and bentonite contents. Therefore, the evaluation method can be used for designing the bentonite content and compaction density from the viewpoint of 'impermeability'. (author)

Mechanical properties of buffer materials for repositories of high-level nuclear waste, 2

International Nuclear Information System (INIS)

Komine, Hideo; Ogata, Nobuhide

1993-01-01

Compacted bentonites have attracted increasing attention as back filling (buffer) materials for repositories of high-level nuclear waste. However, since little has been known concerning the swelling characteristics of compacted bentonites, it is necessary to clarify the fundamental swelling characteristics and the quantitative evaluation on this characteristics is required. For this purpose, a theoretical model concerning the swelling characteristics (swelling deformation and swelling pressure) of compacted bentonites were developed. The following conclusions were drawn from this theoretical study; (1) The evaluation formula of the swelling characteristics of compacted bentonites based on the diffuse double layer theory has been proposed by combining the theoretical model and the theoretical equation to estimate the swelling characteristics of a crystal. (2) The applicability of the evaluation formula proposed in this study has been confirmed by the comparison of the experimental results with calculated results. The sensitivity of this evaluation formula has also been investigated to find that the swelling characteristics is strongly dependent on the ion concentration of pore water and on the montmorillonite content of bentonite. (author)

SUMO, System performance assessment for a high-level nuclear waste repository: Mathematical models

International Nuclear Information System (INIS)

Eslinger, P.W.; Miley, T.B.; Engel, D.W.; Chamberlain, P.J. II.

1992-09-01

Following completion of the preliminary risk assessment of the potential Yucca Mountain Site by Pacific Northwest Laboratory (PNL) in 1988, the Office of Civilian Radioactive Waste Management (OCRWM) of the US Department of Energy (DOE) requested the Performance Assessment Scientific Support (PASS) Program at PNL to develop an integrated system model and computer code that provides performance and risk assessment analysis capabilities for a potential high-level nuclear waste repository. The system model that has been developed addresses the cumulative radionuclide release criteria established by the US Environmental Protection Agency (EPA) and estimates population risks in terms of dose to humans. The system model embodied in the SUMO (System Unsaturated Model) code will also allow benchmarking of other models being developed for the Yucca Mountain Project. The system model has three natural divisions: (1) source term, (2) far-field transport, and (3) dose to humans. This document gives a detailed description of the mathematics of each of these three divisions. Each of the governing equations employed is based on modeling assumptions that are widely accepted within the scientific community

Ramifications of defining high-level waste

International Nuclear Information System (INIS)

Wood, D.E.; Campbell, M.H.; Shupe, M.W.

1987-01-01

The Nuclear Regulatory Commission (NRC) is considering rule making to provide a concentration-based definition of high-level waste (HLW) under authority derived from the Nuclear Waste Policy Act (NWPA) of 1982 and the Low Level Waste Policy Amendments Act of 1985. The Department of Energy (DOE), which has the responsibility to dispose of certain kinds of commercial waste, is supporting development of a risk-based classification system by the Oak Ridge National Laboratory to assist in developing and implementing the NRC rule. The system is two dimensional, with the axes based on the phrases highly radioactive and requires permanent isolation in the definition of HLW in the NWPA. Defining HLW will reduce the ambiguity in the present source-based definition by providing concentration limits to establish which materials are to be called HLW. The system allows the possibility of greater-confinement disposal for some wastes which do not require the degree of isolation provided by a repository. The definition of HLW will provide a firm basis for waste processing options which involve partitioning of waste into a high-activity stream for repository disposal, and a low-activity stream for disposal elsewhere. Several possible classification systems have been derived and the characteristics of each are discussed. The Defense High Level Waste Technology Lead Office at DOE - Richland Operations Office, supported by Rockwell Hanford Operations, has coordinated reviews of the ORNL work by a technical peer review group and other DOE offices. The reviews produced several recommendations and identified several issues to be addressed in the NRC rule making. 10 references, 3 figures

Process arrangement options for Defense waste immobilization

International Nuclear Information System (INIS)

1980-02-01

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

Structural changes in irreversibly densified fused silica: implications for the chemical resistance of high level nuclear waste glasses

International Nuclear Information System (INIS)

Susman, S.; Volin, K.J.; Liebermann, R.C.; Gwanmesia, G.D.; Yanbin Wang

1990-01-01

Energetic photons and energetic particles create changes in the structure of nuclear waste glasses. These can be observed as changes in the average bulk physical properties. For example, exposure of fused silica to high doses of neutron bombardment leads to a maximum average compaction of 3%. However, this does not reveal the true extent of the densification that takes place at a microscopic level. Recent advances in high pressure technology have yielded large samples of fused silica which have been permanently densified under pressure and whose bulk density has been increased by 20%. These specimens have an overall structure that replicates the microstructure of a radiation damaged glass. Measurements have been made for the first time of the structural changes in this pressure densified vitreous silica using neutron diffraction and infrared absorption spectrometry. Extensive alterations in intermediate range order have been observed with consequent anticipated changes in chemical reactivity. The resistance of high level waste glasses to leaching by groundwater must be considered in light of these experimental findings. (author)

Numerical investigation of high level nuclear waste disposal in deep anisotropic geologic repositories

KAUST Repository

Salama, Amgad

2015-11-01

One of the techniques that have been proposed to dispose high level nuclear waste (HLW) has been to bury them in deep geologic formations, which offer relatively enough space to accommodate the large volume of HLW accumulated over the years since the dawn of nuclear era. Albeit the relatively large number of research works that have been conducted to investigate temperature distribution surrounding waste canisters, they all abide to consider the host formations as homogeneous and isotropic. While this could be the case in some subsurface settings, in most cases, this is not true. In other words, subsurface formations are, in most cases, inherently anisotropic and heterogeneous. In this research, we show that even a slight difference in anisotropy of thermal conductivity of host rock with direction could have interesting effects on temperature fields. We investigate the effect of anisotropy angle (the angle the principal direction of anisotropy is making with the coordinate system) on the temperature field as well as on the maximum temperature attained in different barrier systems. This includes 0°, 30°, 45°, 60°, and 90°in addition to the isotropic case as a reference. We also consider the effect of anisotropy ratio (the ratio between the principal direction anisotropies) on the temperature fields and maximum temperature history. This includes ratios ranging between 1.5 and 4. Interesting patterns of temperature fields and profiles are obtained. It is found that the temperature contours are aligned more towards the principal direction of anisotropy. Furthermore the peak temperature in the buffer zone is found to be larger the smaller the anisotropy angle and vice versa. © 2015 Elsevier Ltd. All rights reserved.

High-level waste processing and disposal

International Nuclear Information System (INIS)

Crandall, J.L.; Krause, H.; Sombret, C.; Uematsu, K.

1984-01-01

The national high-level waste disposal plans for France, the Federal Republic of Germany, Japan, and the United States are covered. Three conclusions are reached. The first conclusion is that an excellent technology already exists for high-level waste disposal. With appropriate packaging, spent fuel seems to be an acceptable waste form. Borosilicate glass reprocessing waste forms are well understood, in production in France, and scheduled for production in the next few years in a number of other countries. For final disposal, a number of candidate geological repository sites have been identified and several demonstration sites opened. The second conclusion is that adequate financing and a legal basis for waste disposal are in place in most countries. Costs of high-level waste disposal will probably add about 5 to 10% to the costs of nuclear electric power. The third conclusion is less optimistic. Political problems remain formidable in highly conservative regulations, in qualifying a final disposal site, and in securing acceptable transport routes

Systematics of nuclear mass and level density formulas

Energy Technology Data Exchange (ETDEWEB)

Nakamura, Hisashi [Fuji Electric Co. Ltd., Kawasaki, Kanagawa (Japan)

1998-03-01

The phenomenological models of the nuclear mass and level density are close related to each other, the nuclear ground and excited state properties are described by using the parameter systematics on the mass and level density formulas. The main aim of this work is to provide in an analytical framework the improved energy dependent shell, pairing and deformation corrections generalized to the collective enhancement factors, which offer a systematic prescription over a great number of nuclear reaction cross sections. The new formulas are shown to be in close agreement with not only the empirical nuclear mass data but the measured slow neutron resonance spacings, and experimental systematics observed in the excitation energy dependent properties. (author)

Dissolution rates of DWPF glasses from long-term PCT

International Nuclear Information System (INIS)

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

1996-01-01

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

Nuclear moment of inertia and spin distribution of nuclear levels

International Nuclear Information System (INIS)

Alhassid, Y.; Fang, L.; Liu, S.; Bertsch, G.F.

2005-01-01

We introduce a simple model to calculate the nuclear moment of inertia at finite temperature. This moment of inertia describes the spin distribution of nuclear levels in the framework of the spin-cutoff model. Our model is based on a deformed single-particle Hamiltonian with pairing interaction and takes into account fluctuations in the pairing gap. We derive a formula for the moment of inertia at finite temperature that generalizes the Belyaev formula for zero temperature. We show that a number-parity projection explains the strong odd-even effects observed in shell model Monte Carlo studies of the nuclear moment of inertia in the iron region

Nuclear Level densities from drip line to drip line

International Nuclear Information System (INIS)

Hilaire, S.; Goriely, S.

2007-01-01

New energy-, spin-, parity-dependent level densities based on the microscopic combinatorial model are presented and compared with available experimental data as well as with other nuclear level densities usually employed in nuclear reaction codes. These microscopic level densities are made available in a table format for nearly 8500 nuclei

Devitrification of defense nuclear waste glasses: role of melt insolubles

International Nuclear Information System (INIS)

Bickford, D.F.; Jantzen, C.M.

1985-01-01

Time-temperature-transformation (TTT) curves have been determined for simulated nuclear waste glasses bounding the compositional range in the Defense Waste Processing Facility (DWPF). Formulations include all of the minor chemical elements such as ruthenium and chromium which have limited solubility in borosilicate glasses. Heterogeneous nucleation of spinel on ruthenium dioxide, and subsequent nucleation of acmite on spinel is the major devitrification path. Heterogeneous nucleation on melt insolubles causes more rapid growth of crystalline devitrification phases, than in glass free of melt insolubles. These studies point out the importance of simulating waste glass composition and processing as accurately as possible to obtain reliable estimates of glass performance. 11 refs., 8 figs., 1 tab

Update on ASME rules for spent nuclear fuel and high level radioactive material and waste storage containments

International Nuclear Information System (INIS)

Ralph S. Hill III; Foster, G.M.

2005-01-01

In 2004, a new Code Case, N-717, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) was published. The Code Case provides rules for construction of containments used for storage of spent nuclear fuel and high level radioactive material and waste. The Code Case has been incorporated into Section III of the Code as Division 3, Subsection WC, Class SC Storage Containments, and will be published in the 2005 Addenda. This paper provides an informative background and insight for these rules to provide Owners, regulators, designers, and fabricators with a more comprehensive understanding of the technical basis for these rules. (authors)

Dose estimation for nuclear power plant 4 accident in Taiwan at Fukushima nuclear meltdown emission level

International Nuclear Information System (INIS)

Tang, Mei-Ling; Tsuang, Ben-Jei; Kuo, Pei-Hsuan

2016-01-01

An advanced Gaussian trajectory dispersion model is used to evaluate the evacuation zone due to a nuclear meltdown at the Nuclear Power Plant 4 (NPP4) in Taiwan, with the same emission level as that occurred at Fukushima nuclear meltdown (FNM) in 2011. Our study demonstrates that a FNM emission level would pollute 9% of the island's land area with annual effective dose ≥50 mSv using the meteorological data on 11 March 2011 in Taiwan. This high dose area is also called permanent evacuation zone (denoted as PEZ). The PEZ as well as the emergency-planning zone (EPZ) are found to be sensitive to meteorological conditions on the event. In a sunny day under the dominated NE wind conditions, the EPZ can be as far as 100 km with the first 7-day dose ≥20 mSv. Three hundred sixty-five daily events using the meteorological data from 11 March 2011 to 9 March 2012 are evaluated. It is found that the mean land area of Taiwan in becoming the PEZ is 11%. Especially, the probabilities of the northern counties/cities (Keelung, New Taipei, Taipei, Taoyuan, Hsinchu City, Hsinchu County and Ilan County) to be PEZs are high, ranging from 15% in Ilan County to 51% in Keelung City. Note that the total population of the above cities/counties is as high as 10 million people. Moreover, the western valleys of the Central Mountain Range are also found to be probable being PEZs, where all of the reservoirs in western Taiwan are located. For example, the probability can be as high as 3% in the far southern-most tip of Taiwan Island in Pingtung County. This shows that the entire populations in western Taiwan can be at risk due to the shortage of clean water sources under an event at FNM emission level, especially during the NE monsoon period. - Highlights: • An advanced Gaussian-type trajectory model to evaluate the evacuation zone at Nuclear Power Plant 4 in Taiwan. • Mean land area of Taiwan in becoming the permanent evacuation zone is 11%. • The probabilities of the northern

Heat transfer effects in vertically emplaced high level nuclear waste container

International Nuclear Information System (INIS)

Moujaes, S.F.; Lei, Y.M.

1994-01-01

Modeling free convection heat transfer in a cylindrical annular enclosure is still an active area of research and an important problem to be addressed in the high level nuclear waste repository. For the vertically emplaced waste container, the air gap which is between the container shell and the rock borehole, have an important role of dissipating heat to surrounding rock. These waste containers are vertically emplaced in the borehole 300 meters just below ground, and in a horizontal grid of 30 x 8 meters apart. The borehole will be capped after the container emplacement. The expected initial heat generated is between 3-4.74 kW per container depending on the type of waste. The goal of this study is to use a computer simulation model to find the borehole wall, air-gap and the container outer wall temperature distributions. The borehole wall temperature history has been found in the previous study, and was estimated to reach a maximum temperature of about 218 degrees C after 18 years from the emplacement. The temperature history of the rock surface is then used for the air-gap simulation. The problem includes convection and radiation heat transfer in a vertical enclosure. This paper will present the results of the convection in the air-gap over one thousand years after the containers' emplacement. During this long simulation period it was also observed that a multi-cellular air flow pattern can be generated in the air gap

Socioeconomic impacts of large-scale developments: implications for high-level nuclear waste repositories

International Nuclear Information System (INIS)

Murdock, S.H.; Leistritz, F.L.; Hamm, R.R.

1983-01-01

High-level nuclear waste repositories will likely be located in sparsely settled rural areas in the US. These projects will significantly affect the economic, demographic, public service, fiscal, and social (the socioeconomic) dimensions of those rural areas. The impacts and means of mitigating them thus require careful analysis. This paper examines some of the potential socioeconomic impacts likely to occur in rural areas as a result of repository siting and development, and it describes some of the characteristics of mitigation programs that are likely to be necessary, if the impacts are to be addressed. Both (1) standard impacts, those resulting from the fact that, like many other large-scale developments, repositories will involve a substantial number of new workers and residents (relative to the size of existing communities), and (2) special impacts, those resulting from the fact that repositories store radioactive materials, are examined. The discussion indicates that economic, demographic, public service, fiscal, and social impacts (standard and special) of these repositories will be substantial and problematic in many cases. Unless the impacts are effectively addressed with carefully planned and well financed mitigation efforts that insure that high-quality planning information is provided to local residents, and that local residents are involved in impact planning and management throughout the siting and development process, repository siting is unlikely to be effectively and equitably achieved. 44 references

Study on high-level waste geological disposal metadata model

International Nuclear Information System (INIS)

Ding Xiaobin; Wang Changhong; Zhu Hehua; Li Xiaojun

2008-01-01

This paper expatiated the concept of metadata and its researches within china and abroad, then explain why start the study on the metadata model of high-level nuclear waste deep geological disposal project. As reference to GML, the author first set up DML under the framework of digital underground space engineering. Based on DML, a standardized metadata employed in high-level nuclear waste deep geological disposal project is presented. Then, a Metadata Model with the utilization of internet is put forward. With the standardized data and CSW services, this model may solve the problem in the data sharing and exchanging of different data form A metadata editor is build up in order to search and maintain metadata based on this model. (authors)

High-Level Waste Vitrification Facility Feasibility Study

Energy Technology Data Exchange (ETDEWEB)

D. A. Lopez