National Library of Energy BETA

Sample records for high level radioactive

  1. High-level radioactive wastes. Supplement 1

    SciTech Connect (OSTI)

    McLaren, L.H.

    1984-09-01

    This bibliography contains information on high-level radioactive wastes included in the Department of Energy's Energy Data Base from August 1982 through December 1983. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes, each preceded by a brief description, are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number. 1452 citations.

  2. Report on Separate Disposal of Defense High- Level Radioactive...

    Office of Environmental Management (EM)

    Radioactive Waste March 2015 This page left blank. i EXECUTIVE SUMMARY Purpose This report considers whether a separate repository for high-level radioactive waste (HLW) ...

  3. DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel

    Office of Environmental Management (EM)

    Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent ... level radioactive waste and spent nuclear fuel in a single repository or repositories. ...

  4. RETENTION OF SULFATE IN HIGH LEVEL RADIOACTIVE WASTE GLASS

    SciTech Connect (OSTI)

    Fox, K.

    2010-09-07

    High level radioactive wastes are being vitrified at the Savannah River Site for long term disposal. Many of the wastes contain sulfate at concentrations that can be difficult to retain in borosilicate glass. This study involves efforts to optimize the composition of a glass frit for combination with the waste to improve sulfate retention while meeting other process and product performance constraints. The fabrication and characterization of several series of simulated waste glasses are described. The experiments are detailed chronologically, to provide insight into part of the engineering studies used in developing frit compositions for an operating high level waste vitrification facility. The results lead to the recommendation of a specific frit composition and a concentration limit for sulfate in the glass for the next batch of sludge to be processed at Savannah River.

  5. Assessment of Disposal Options for DOE-Managed High-Level Radioactive...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear ...

  6. Long-term management of high-level radioactive waste (HLW) and...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) ...

  7. Spent Fuel and High-Level Radioactive Waste Transportation Report

    SciTech Connect (OSTI)

    Not Available

    1992-03-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by SSEB in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste Issues. In addition. this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  8. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect (OSTI)

    Not Available

    1989-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages sew be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  9. Spent fuel and high-level radioactive waste transportation report

    SciTech Connect (OSTI)

    Not Available

    1990-11-01

    This publication is intended to provide its readers with an introduction to the issues surrounding the subject of transportation of spent nuclear fuel and high-level radioactive waste, especially as those issues impact the southern region of the United States. It was originally issued by the Southern States Energy Board (SSEB) in July 1987 as the Spent Nuclear Fuel and High-Level Radioactive Waste Transportation Primer, a document patterned on work performed by the Western Interstate Energy Board and designed as a ``comprehensive overview of the issues.`` This work differs from that earlier effort in that it is designed for the educated layman with little or no background in nuclear waste issues. In addition, this document is not a comprehensive examination of nuclear waste issues but should instead serve as a general introduction to the subject. Owing to changes in the nuclear waste management system, program activities by the US Department of Energy and other federal agencies and developing technologies, much of this information is dated quickly. While this report uses the most recent data available, readers should keep in mind that some of the material is subject to rapid change. SSEB plans periodic updates in the future to account for changes in the program. Replacement pages will be supplied to all parties in receipt of this publication provided they remain on the SSEB mailing list.

  10. Report on Separate Disposal of Defense High-Level Radioactive Waste

    Broader source: Energy.gov [DOE]

    This is a report on the separate disposal of defense high-level radioactive waste and commercial nuclear waste.

  11. Deep borehole disposal of high-level radioactive waste.

    SciTech Connect (OSTI)

    Stein, Joshua S.; Freeze, Geoffrey A.; Brady, Patrick Vane; Swift, Peter N.; Rechard, Robert Paul; Arnold, Bill Walter; Kanney, Joseph F.; Bauer, Stephen J.

    2009-07-01

    Preliminary evaluation of deep borehole disposal of high-level radioactive waste and spent nuclear fuel indicates the potential for excellent long-term safety performance at costs competitive with mined repositories. Significant fluid flow through basement rock is prevented, in part, by low permeabilities, poorly connected transport pathways, and overburden self-sealing. Deep fluids also resist vertical movement because they are density stratified. Thermal hydrologic calculations estimate the thermal pulse from emplaced waste to be small (less than 20 C at 10 meters from the borehole, for less than a few hundred years), and to result in maximum total vertical fluid movement of {approx}100 m. Reducing conditions will sharply limit solubilities of most dose-critical radionuclides at depth, and high ionic strengths of deep fluids will prevent colloidal transport. For the bounding analysis of this report, waste is envisioned to be emplaced as fuel assemblies stacked inside drill casing that are lowered, and emplaced using off-the-shelf oilfield and geothermal drilling techniques, into the lower 1-2 km portion of a vertical borehole {approx}45 cm in diameter and 3-5 km deep, followed by borehole sealing. Deep borehole disposal of radioactive waste in the United States would require modifications to the Nuclear Waste Policy Act and to applicable regulatory standards for long-term performance set by the US Environmental Protection Agency (40 CFR part 191) and US Nuclear Regulatory Commission (10 CFR part 60). The performance analysis described here is based on the assumption that long-term standards for deep borehole disposal would be identical in the key regards to those prescribed for existing repositories (40 CFR part 197 and 10 CFR part 63).

  12. Control of high level radioactive waste-glass melters

    SciTech Connect (OSTI)

    Bickford, D.F.; Choi, A.S.

    1991-01-01

    Slurry Fed Melters (SFM) are being developed in the United States, Europe and Japan for the conversion of high-level radioactive waste to borosilicate glass for permanent disposal. The high transition metal, noble metal, nitrate, organic, and sulfate contents of these wastes lead to unique melter redox control requirements. Pilot waste-glass melter operations have indicated the possibility of nickel sulfide or noble-metal fission-product accumulation on melter floors, which can lead to distortion of electric heating patterns, and decrease melter life. Sulfide formation is prevented by control of the redox chemistry of the melter feed. The redox state of waste-glass melters is determined by balance between the reducing potential of organic compounds in the feed, and the oxidizing potential of gases above the melt, and nitrates and polyvalent elements in the waste. Semiquantitative models predicting limitations of organic content have been developed based on crucible testing. Computerized thermodynamic computations are being developed to predict the sequence and products of redox reactions and is assessing process variations. Continuous melter test results have been compared to improved computer staged-thermodynamic-models of redox behavior. Feed chemistry control to prevent sulfide and moderate noble metal accumulations are discussed. 17 refs., 3 figs.

  13. Long-term management of high-level radioactive waste (HLW) and spent

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    nuclear fuel (SNF) | Department of Energy Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) Long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor to generate nuclear energy but that has been removed from the reactor as no

  14. Synergistic Inhibitors for Dilute High-Level Radioactive Waste

    SciTech Connect (OSTI)

    Wiersma, B.J.; Zapp, P.E.

    1995-11-01

    Cyclic potentiodynamic polarization scans were conducted to determine the effectiveness of various combinations of anodic inhibitors in the prevention of pitting in carbon steel exposed to dilute radioactive waste. Chromate, molybdate, and phosphate were investigated as replacements for nitrite, whose effective concentrations are incompatible with the waste vitrification process. The polarization scans were performed in non-radioactive waste simulants. Their results showed that acceptable combinations of phosphate with chromate and phosphate with molybdate effectively prevented pitting corrosion. Chromate with molybdate could not replace nitrite.

  15. [Corrosion testing of high level radioactive waste. Final report

    SciTech Connect (OSTI)

    1996-06-01

    Alloys under consideration as candidates for the high level nuclear waste containers at Yucca Mountain were exposed to a range of corrosion conditions and their performance measured. The alloys tested were Incoloy 825, 70/30 Copper-Nickel, Monel 400, Hastelloy C- 22, and low carbon steel. The test conditions varied were: temperature, concentration, agitation, and crevice simulation. Only in the case of carbon steel was significant attack noted. This attack appeared to be transport limited.

  16. Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste

    Broader source: Energy.gov [DOE]

    The Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste is a framework for moving toward a sustainable program to deploy an integrated system capable of...

  17. Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste

    Broader source: Energy.gov [DOE]

    Issued on January 11, 2013, the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste is a framework for moving toward a sustainable program to deploy an...

  18. THERMAL ANALYSIS OF GEOLOGIC HIGH-LEVEL RADIOACTIVE WASTE PACKAGES

    SciTech Connect (OSTI)

    Hensel, S.; Lee, S.

    2010-04-20

    The engineering design of disposal of the high level waste (HLW) packages in a geologic repository requires a thermal analysis to provide the temperature history of the packages. Calculated temperatures are used to demonstrate compliance with criteria for waste acceptance into the geologic disposal gallery system and as input to assess the transient thermal characteristics of the vitrified HLW Package. The objective of the work was to evaluate the thermal performance of the supercontainer containing the vitrified HLW in a non-backfilled and unventilated underground disposal gallery. In order to achieve the objective, transient computational models for a geologic vitrified HLW package were developed by using a computational fluid dynamics method, and calculations for the HLW disposal gallery of the current Belgian geological repository reference design were performed. An initial two-dimensional model was used to conduct some parametric sensitivity studies to better understand the geologic system's thermal response. The effect of heat decay, number of co-disposed supercontainers, domain size, humidity, thermal conductivity and thermal emissivity were studied. Later, a more accurate three-dimensional model was developed by considering the conduction-convection cooling mechanism coupled with radiation, and the effect of the number of supercontainers (3, 4 and 8) was studied in more detail, as well as a bounding case with zero heat flux at both ends. The modeling methodology and results of the sensitivity studies will be presented.

  19. Reference design and operations for deep borehole disposal of high-level radioactive waste.

    SciTech Connect (OSTI)

    Herrick, Courtney Grant; Brady, Patrick Vane; Pye, Steven; Arnold, Bill Walter; Finger, John Travis; Bauer, Stephen J.

    2011-10-01

    A reference design and operational procedures for the disposal of high-level radioactive waste in deep boreholes have been developed and documented. The design and operations are feasible with currently available technology and meet existing safety and anticipated regulatory requirements. Objectives of the reference design include providing a baseline for more detailed technical analyses of system performance and serving as a basis for comparing design alternatives. Numerous factors suggest that deep borehole disposal of high-level radioactive waste is inherently safe. Several lines of evidence indicate that groundwater at depths of several kilometers in continental crystalline basement rocks has long residence times and low velocity. High salinity fluids have limited potential for vertical flow because of density stratification and prevent colloidal transport of radionuclides. Geochemically reducing conditions in the deep subsurface limit the solubility and enhance the retardation of key radionuclides. A non-technical advantage that the deep borehole concept may offer over a repository concept is that of facilitating incremental construction and loading at multiple perhaps regional locations. The disposal borehole would be drilled to a depth of 5,000 m using a telescoping design and would be logged and tested prior to waste emplacement. Waste canisters would be constructed of carbon steel, sealed by welds, and connected into canister strings with high-strength connections. Waste canister strings of about 200 m length would be emplaced in the lower 2,000 m of the fully cased borehole and be separated by bridge and cement plugs. Sealing of the upper part of the borehole would be done with a series of compacted bentonite seals, cement plugs, cement seals, cement plus crushed rock backfill, and bridge plugs. Elements of the reference design meet technical requirements defined in the study. Testing and operational safety assurance requirements are also defined. Overall, the results of the reference design development and the cost analysis support the technical feasibility of the deep borehole disposal concept for high-level radioactive waste.

  20. SPONTANEOUS CATALYTIC WET AIR OXIDATION DURING PRE-TREATMENT OF HIGH-LEVEL RADIOACTIVE WASTE SLUDGE

    SciTech Connect (OSTI)

    Koopman, D.; Herman, C.; Pareizs, J.; Bannochie, C.; Best, D.; Bibler, N.; Fellinger, T.

    2009-10-01

    Savannah River Remediation, LLC (SRR) operates the Defense Waste Processing Facility for the U.S. Department of Energy at the Savannah River Site. This facility immobilizes high-level radioactive waste through vitrification following chemical pretreatment. Catalytic destruction of formate and oxalate ions to carbon dioxide has been observed during qualification testing of non-radioactive analog systems. Carbon dioxide production greatly exceeded hydrogen production, indicating the occurrence of a process other than the catalytic decomposition of formic acid. Statistical modeling was used to relate the new reaction chemistry to partial catalytic wet air oxidation of both formate and oxalate ions driven by the low concentrations of palladium, rhodium, and/or ruthenium in the waste. Variations in process conditions led to increases or decreases in the total oxidative destruction, as well as partially shifting the preferred species undergoing destruction from oxalate ion to formate ion.

  1. Yucca Mountain, Nevada - A Proposed Geologic Repository for High-Level Radioactive Waste (Volume 1) Introduction

    SciTech Connect (OSTI)

    R.A. Levich; J.S. Stuckless

    2006-09-25

    Yucca Mountain in Nevada represents the proposed solution to what has been a lengthy national effort to dispose of high-level radioactive waste, waste which must be isolated from the biosphere for tens of thousands of years. This chapter reviews the background of that national effort and includes some discussion of international work in order to provide a more complete framework for the problem of waste disposal. Other chapters provide the regional geologic setting, the geology of the Yucca Mountain site, the tectonics, and climate (past, present, and future). These last two chapters are integral to prediction of long-term waste isolation.

  2. Collaboration, Automation, and Information Management at Hanford High Level Radioactive Waste (HLW) Tank Farms

    SciTech Connect (OSTI)

    Aurah, Mirwaise Y.; Roberts, Mark A.

    2013-12-12

    Washington River Protection Solutions (WRPS), operator of High Level Radioactive Waste (HLW) Tank Farms at the Hanford Site, is taking an over 20-year leap in technology, replacing systems that were monitored with clipboards and obsolete computer systems, as well as solving major operations and maintenance hurdles in the area of process automation and information management. While WRPS is fully compliant with procedures and regulations, the current systems are not integrated and do not share data efficiently, hampering how information is obtained and managed.

  3. Stress Corrosion Cracking Model for High Level Radioactive-Waste Packages

    SciTech Connect (OSTI)

    P. Andresen; G. Gordon; S. Lu

    2004-10-05

    A stress corrosion cracking (SCC) model has been adapted for performance prediction of high level radioactive-waste packages to be emplaced in the proposed Yucca Mountain repository. For waste packages of the proposed Yucca Mountain repository, the outer barrier material is the highly corrosion-resistant Alloy UNS-N06022 (Alloy 22), the environment is represented by aqueous brine films present on the surface of the waste package from dripping or deliquescence of soluble salts present in any surface deposits, and the tensile stress is principally from weld induced residual stress. SCC has historically been separated into ''initiation'' and ''propagation'' phases. Initiation of SCC will not occur on a smooth surface if the surface stress is below a threshold value defined as the threshold stress. Cracks can also initiate at and propagate from flaws (or defects) resulting from manufacturing processes (such as welding); or that develop from corrosion processes such as pitting or dissolution of inclusions. To account for crack propagation, the slip dissolution/film rupture (SDFR) model is adopted to provide mathematical formulae for prediction of the crack growth rate. Once the crack growth rate at an initiated SCC is determined, it can be used by the performance assessment to determine the time to through-wall penetration for the waste package. This paper presents the development of the SDFR crack growth rate model based on technical information in the literature as well as experimentally determined crack growth rates developed specifically for Alloy UNS-N06022 in environments relevant to high level radioactive-waste packages of the proposed Yucca Mountain radioactive-waste repository. In addition, a seismic damage related SCC crack opening area density model is briefly described.

  4. Considerations of the effects of high winds on a low-level radioactive interim storage pile

    SciTech Connect (OSTI)

    Smith, D.E. )

    1991-11-01

    On Wednesday, March 27, 1991, the St. Louis area experienced high winds that damaged a synthetic cover of a low-level radioactive waste storage pile at the US Department of Energy's (DOE's) Hazelwood Interim Storage Site (HISS) in Hazelwood, Missouri. Winds in the St. Louis area at the time of the incident were reported to be 35 mi/h with gusts up to 50 mi/h. Tornado warnings were in effect at the time. The purpose of this summary is to analyze the effects of uplift forces on a synthetic pile cover because of high winds. Consideration is given to anchoring the synthetic cover, type and placement of ballast on the pile, and the type of synthetic membranes best suited to this application. Discussion also includes the emergency procedures used in responding to the incident.

  5. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    SciTech Connect (OSTI)

    Gdowski, G.E.; Bullen, D.B. )

    1988-08-01

    Six alloys are being considered as possible materials for the fabrication of containers for the disposal of high-level radioactive waste. Three of these candidate materials are copper-based alloys: CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The other three are iron- to nickel-based austenitic materials: Types 304L and 316L stainless steels and Alloy 825. Radioactive waste will include spent-fuel assemblies from reactors as well as waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, the containers must be retrievable from the disposal site. Shortly after emplacement of the containers in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This radiation will promote the radiolytic decomposition of moist air to hydrogen. This volume surveys the available data on the effects of hydrogen on the six candidate alloys for fabrication of the containers. For copper, the mechanism of hydrogen embrittlement is discussed, and the effects of hydrogen on the mechanical properties of the copper-based alloys are reviewed. The solubilities and diffusivities of hydrogen are documented for these alloys. For the austenitic materials, the degradation of mechanical properties by hydrogen is documented. The diffusivity and solubility of hydrogen in these alloys are also presented. For the copper-based alloys, the ranking according to resistance to detrimental effects of hydrogen is: CDA 715 (best) > CDA 613 > CDA 102 (worst). For the austenitic alloys, the ranking is: Type 316L stainless steel {approx} Alloy 825 > Type 304L stainless steel (worst). 87 refs., 19 figs., 8 tabs.

  6. Control of high level radioactive waste-glass melters. Part 5, Modelling of complex redox effects

    SciTech Connect (OSTI)

    Bickford, D.F.; Choi, A.S.

    1991-12-31

    Slurry Fed Melters (SFM) are being developed in the United States, Europe and Japan for the conversion of high-level radioactive waste to borosilicate glass for permanent disposal. The high transition metal, noble metal, nitrate, organic, and sulfate contents of these wastes lead to unique melter redox control requirements. Pilot waste-glass melter operations have indicated the possibility of nickel sulfide or noble-metal fission-product accumulation on melter floors, which can lead to distortion of electric heating patterns, and decrease melter life. Sulfide formation is prevented by control of the redox chemistry of the melter feed. The redox state of waste-glass melters is determined by balance between the reducing potential of organic compounds in the feed, and the oxidizing potential of gases above the melt, and nitrates and polyvalent elements in the waste. Semiquantitative models predicting limitations of organic content have been developed based on crucible testing. Computerized thermodynamic computations are being developed to predict the sequence and products of redox reactions and is assessing process variations. Continuous melter test results have been compared to improved computer staged-thermodynamic-models of redox behavior. Feed chemistry control to prevent sulfide and moderate noble metal accumulations are discussed. 17 refs., 3 figs.

  7. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    SciTech Connect (OSTI)

    Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D. ); Bullen, D.B. )

    1988-04-01

    Three iron- to nickel-based austenitic alloys (Types 304L and 316L stainless steels and Alloy 825) are being considered as candidate materials for the fabrication of high-level radioactive-waste containers. Waste will include fuel assemblies from reactors as well as high-level waste in borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides in the repository will result in the generation of substantial heat and in fluences of gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including atmospheric oxidation; uniform aqueous phase corrosion; pitting; crevice corrosion; sensitization and intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This report is an analysis of data relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of the three austenitic candidate alloys. The candidates are compared in terms of their susceptibilities to these forms of corrosion. Although all three candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these types of localized corrosion (LC); such resistance is important because pits can penetrate the metal and serve as crack initiation sites. Both Types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented in Alloy 825 under comparable conditions. Gamma radiation has been found to enhance SCC in Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while the effects of microbiologically induced corrosion have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. 211 refs., 49 figs., 10 tabs.

  8. LONG-TERM CORROSION TESTING OF CANDIDATE MATERIALS FOR HIGH-LEVEL RADIOACTIVE WASTE CONTAINMENT

    SciTech Connect (OSTI)

    Estill, J. C.; Doughty, S.; Gdowski, G. E.; Gordon, S.; King, K.; McCright, R. D.; Wang, F.

    1997-10-01

    Preliminary results are presented from the long-term corrosion test program of candidate materials for the high-level radioactive waste packages that would be emplaced in the potential repository at Yucca Mountain, Nevada. The present waste package design is based on a multi-barrier concept having an inner container of a corrosion resistant material and an outer container of a corrosion allowance material. Test specimens have been exposed to simulated bounding environments that may credibly develop in the vicinity of the waste packages. Corrosion rates have been calculated for weight loss and crevice specimens, and U-bend specimens have been examined for evidence of stress corrosion cracking (SCC). Galvanic testing has been started recently and initial results are forthcoming. Pitting characterization of test specimens will be conducted in the coming year. This test program is expected to continue for a minimum of five years so that long-term corrosion data can be determined to support corrosion model development, performance assessment, and waste package design.

  9. Role of Congress in the High Level Radioactive Waste Odyssey: The Wisdom and Will of the Congress - 13096

    SciTech Connect (OSTI)

    Vieth, Donald L. [DOE/NVOO Project Manager for Yucca Mountain, 1982 thru 1987, 1154 Cheltenham Place, Maineville, OH 45039 (United States)] [DOE/NVOO Project Manager for Yucca Mountain, 1982 thru 1987, 1154 Cheltenham Place, Maineville, OH 45039 (United States); Voegele, Michael D. [Nye County Nuclear Waste Repository Project Office, 7404 Oak Grove Ave, Las Vegas, NV 89117 (United States)] [Nye County Nuclear Waste Repository Project Office, 7404 Oak Grove Ave, Las Vegas, NV 89117 (United States)

    2013-07-01

    Congress has had a dual role with regard to high level radioactive waste, being involved in both its creation and its disposal. A significant amount of time has passed between the creation of the nation's first high level radioactive waste and the present day. The pace of addressing its remediation has been highly irregular. Congress has had to consider the technical, regulatory, and political issues and all have had specific difficulties. It is a true odyssey framed by an imperative and accountability, by a sense of urgency, by an ability or inability to finish the job and by consequences. Congress had set a politically acceptable course by 1982. However, President Obama intervened in the process after he took office in January 2009. Through the efforts of his Administration, by the end of 2012, the US government has no program to dispose of high level radioactive waste and no reasonable prospect of a repository for high level radioactive waste. It is not obvious how the US government program will be reestablished or who will assume responsibility for leadership. The ultimate criteria for judging the consequences are 1) the outcome of the ongoing NRC's Nuclear Waste Confidence Rulemaking and 2) the concomitant permissibility of nuclear energy supplying electricity from operating reactors in the US. (authors)

  10. Evaluation of Options for Permanent Geologic Disposal of Spent NuclearFuel and High-Level Radioactive Waste

    Broader source: Energy.gov [DOE]

    [In Support of a Comprehensive National Nuclear Fuel Cycle Strategy, Volumes I and II (Appendices)] This study provides a technical basis for informing policy decisions regarding strategies for the management and permanent disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in the United States requiring geologic isolation.

  11. Granite disposal of U.S. high-level radioactive waste.

    SciTech Connect (OSTI)

    Freeze, Geoffrey A.; Mariner, Paul E.; Lee, Joon H.; Hardin, Ernest L.; Goldstein, Barry; Hansen, Francis D.; Price, Ronald H.; Lord, Anna Snider

    2011-08-01

    This report evaluates the feasibility of disposing U.S. high-level radioactive waste in granite several hundred meters below the surface of the earth. The U.S. has many granite formations with positive attributes for permanent disposal. Similar crystalline formations have been extensively studied by international programs, two of which, in Sweden and Finland, are the host rocks of submitted or imminent repository license applications. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in granite media. In this report we develop scoping performance analyses, based on the applicable features, events, and processes (FEPs) identified by international investigators, to support generic conclusions regarding post-closure safety. Unlike the safety analyses for disposal in salt, shale/clay, or deep boreholes, the safety analysis for a mined granite repository depends largely on waste package preservation. In crystalline rock, waste packages are preserved by the high mechanical stability of the excavations, the diffusive barrier of the buffer, and favorable chemical conditions. The buffer is preserved by low groundwater fluxes, favorable chemical conditions, backfill, and the rigid confines of the host rock. An added advantage of a mined granite repository is that waste packages would be fairly easy to retrieve, should retrievability be an important objective. The results of the safety analyses performed in this study are consistent with the results of comprehensive safety assessments performed for sites in Sweden, Finland, and Canada. They indicate that a granite repository would satisfy established safety criteria and suggest that a small number of FEPs would largely control the release and transport of radionuclides. In the event the U.S. decides to pursue a potential repository in granite, a detailed evaluation of these FEPs would be needed to inform site selection and safety assessment.

  12. A Low-Tech, Low-Budget Storage Solution for High Level Radioactive Sources

    SciTech Connect (OSTI)

    Brett Carlsen; Ted Reed; Todd Johnson; John Weathersby; Joe Alexander; Dave Griffith; Douglas Hamelin

    2014-07-01

    The need for safe, secure, and economical storage of radioactive material becomes increasingly important as beneficial uses of radioactive material expand (increases inventory), as political instability rises (increases threat), and as final disposal and treatment facilities are delayed (increases inventory and storage duration). Several vendor-produced storage casks are available for this purpose but are often costly due to the required design, analyses, and licensing costs. Thus the relatively high costs of currently accepted storage solutions may inhibit substantial improvements in safety and security that might otherwise be achieved. This is particularly true in areas of the world where the economic and/or the regulatory infrastructure may not provide the means and/or the justification for such an expense. This paper considers a relatively low-cost, low-technology radioactive material storage solution. The basic concept consists of a simple shielded storage container that can be fabricated locally using a steel pipe and a corrugated steel culvert as forms enclosing a concrete annulus. Benefits of such a system include 1) a low-tech solution that utilizes materials and skills available virtually anywhere in the world, 2) a readily scalable design that easily adapts to specific needs such as the geometry and radioactivity of the source term material), 3) flexible placement allows for free-standing above-ground or in-ground (i.e., below grade or bermed) installation, 4) the ability for future relocation without direct handling of sources, and 5) a long operational lifetime . Le mieux est lennemi du bien (translated: The best is the enemy of good) applies to the management of radioactive materials particularly where the economic and/or regulatory justification for additional investment is lacking. Development of a low-cost alternative that considerably enhances safety and security may lead to a greater overall risk reduction than insisting on solutions that remain economically and/or politically out of reach.

  13. Shale disposal of U.S. high-level radioactive waste.

    SciTech Connect (OSTI)

    Sassani, David Carl; Stone, Charles Michael; Hansen, Francis D.; Hardin, Ernest L.; Dewers, Thomas A.; Martinez, Mario J.; Rechard, Robert Paul; Sobolik, Steven Ronald; Freeze, Geoffrey A.; Cygan, Randall Timothy; Gaither, Katherine N.; Holland, John Francis; Brady, Patrick Vane

    2010-05-01

    This report evaluates the feasibility of high-level radioactive waste disposal in shale within the United States. The U.S. has many possible clay/shale/argillite basins with positive attributes for permanent disposal. Similar geologic formations have been extensively studied by international programs with largely positive results, over significant ranges of the most important material characteristics including permeability, rheology, and sorptive potential. This report is enabled by the advanced work of the international community to establish functional and operational requirements for disposal of a range of waste forms in shale media. We develop scoping performance analyses, based on the applicable features, events, and processes identified by international investigators, to support a generic conclusion regarding post-closure safety. Requisite assumptions for these analyses include waste characteristics, disposal concepts, and important properties of the geologic formation. We then apply lessons learned from Sandia experience on the Waste Isolation Pilot Project and the Yucca Mountain Project to develop a disposal strategy should a shale repository be considered as an alternative disposal pathway in the U.S. Disposal of high-level radioactive waste in suitable shale formations is attractive because the material is essentially impermeable and self-sealing, conditions are chemically reducing, and sorption tends to prevent radionuclide transport. Vertically and laterally extensive shale and clay formations exist in multiple locations in the contiguous 48 states. Thermal-hydrologic-mechanical calculations indicate that temperatures near emplaced waste packages can be maintained below boiling and will decay to within a few degrees of the ambient temperature within a few decades (or longer depending on the waste form). Construction effects, ventilation, and the thermal pulse will lead to clay dehydration and deformation, confined to an excavation disturbed zone within a few meters of the repository, that can be reasonably characterized. Within a few centuries after waste emplacement, overburden pressures will seal fractures, resaturate the dehydrated zones, and provide a repository setting that strongly limits radionuclide movement to diffusive transport. Coupled hydrogeochemical transport calculations indicate maximum extents of radionuclide transport on the order of tens to hundreds of meters, or less, in a million years. Under the conditions modeled, a shale repository could achieve total containment, with no releases to the environment in undisturbed scenarios. The performance analyses described here are based on the assumption that long-term standards for disposal in clay/shale would be identical in the key aspects, to those prescribed for existing repository programs such as Yucca Mountain. This generic repository evaluation for shale is the first developed in the United States. Previous repository considerations have emphasized salt formations and volcanic rock formations. Much of the experience gained from U.S. repository development, such as seal system design, coupled process simulation, and application of performance assessment methodology, is applied here to scoping analyses for a shale repository. A contemporary understanding of clay mineralogy and attendant chemical environments has allowed identification of the appropriate features, events, and processes to be incorporated into the analysis. Advanced multi-physics modeling provides key support for understanding the effects from coupled processes. The results of the assessment show that shale formations provide a technically advanced, scientifically sound disposal option for the U.S.

  14. Caustic leaching of high-level radioactive tank sludge: A critical literature review

    SciTech Connect (OSTI)

    McGinnis, C.P.; Welch, T.D.; Hunt, R.D.

    1998-08-01

    The Department of Energy (DOE) must treat and safely dispose of its radioactive tank contents, which can be separated into high-level waste (HLW) and low-level waste (LLW) fractions. Since the unit costs of treatment and disposal are much higher for HLW than for LLW, technologies to reduce the amount of HLW are being developed. A key process currently being studied to reduce the volume of HLW sludges is called enhanced sludge washing (ESW). This process removes, by water washes, soluble constituents such as sodium salts, and the washed sludge is then leached with 2--3 M NaOH at 60--100 C to remove nonradioactive metals such as aluminum. The remaining solids are considered to be HLW while the solutions are LLW after radionuclides such as {sup 137}Cs have been removed. Results of bench-scale tests have shown that the ESW will probably remove the required amounts of inert constituents. While both experimental and theoretical results have shown that leaching efficiency increases as the time and temperature of the leach are increased, increases in the caustic concentration above 2--3 M will only marginally improve the leach factors. However, these tests were not designed to validate the assumption that the caustic used in the ESW process will generate only a small increase (10 Mkg) in the amount of LLW; instead the test conditions were selected to maximize leaching in a short period and used more water and caustic than is planned during full-scale operations. Even though calculations indicate that the estimate for the amount of LLW generated by the ESW process appears to be reasonable, a detailed study of the amount of LLW from the ESW process is still required. If the LLW analysis indicates that sodium management is critical, then a more comprehensive evaluation of the clean salt process or caustic recycle would be needed. Finally, experimental and theoretical studies have clearly demonstrated the need for the control of solids formation during and after leaching.

  15. Caustic leaching of high-level radioactive tank sludge: A critical literature review

    SciTech Connect (OSTI)

    McGinnis, C.P.; Welch, T.D.; Hunt, R.D.

    1997-12-31

    The Department of Energy (DOE) must treat and safely dispose of its radioactive tank contents, which can be separated into high-level waste (HLW) and low-level waste (LLW) fractions. Since the unit costs of treatment and disposal are much higher for HLW than for LLW, technologies to reduce the amount of HLW are being developed. A key process currently being studied to reduce the volume of HLW sludges is called enhanced sludge washing (ESW). This process removes, by water washes, soluble constituents such as sodium salts, and the washed sludge is then leached with 2--3 M NaOH at 60--100 C to remove nonradioactive metals such as aluminum. The remaining solids are considered to be HLW while the solutions are LLW after radionuclides such as {sup 137}Cs have been removed. Results of bench-scale tests have shown that the ESW will probably remove the required amounts of inert constituents. While both experimental and theoretical results have shown that leaching efficiency increases as the time and temperature of the leach are increased, increases in the caustic concentration above 2--3 M will only marginally improve the leach factors. However, these tests were not designed to validate the assumption that the caustic used in the ESW process will generate only a small increase (10 Mkg) in the amount of LLW; instead, the test conditions were selected to maximize leaching in a short period and used more water and caustic than is planned during full-scale operations. Even though calculations indicate that the estimate for the amount of LLW generated by the ESW process appears to be reasonable, a detailed study of the amount of LLW from the ESW process is still required. If the LLW analysis indicates that sodium management is critical, then a more comprehensive evaluation of the clean salt process or caustic recycle would be needed. Finally, experimental and theoretical studies have clearly demonstrated the need for the control of solids formation during and after leaching.

  16. Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel

    Broader source: Energy.gov [DOE]

    The Assessment of Disposal Options for DOE-Managed High-Level Radioactive Waste and Spent Nuclear Fuel report assesses the technical options for the safe and permanent disposal of high-level radioactive waste (HLW) and spent nuclear fuel (SNF) managed by the Department of Energy. Specifically, it considers whether DOE-managed HLW and SNF should be disposed of with commercial SNF and HLW in one geologic repository or whether there are advantages to developing separate geologic disposal pathways for some DOE-managed HLW and SNF. The report recommends that the Department begin implementation of a phased, adaptive, and consent-based strategy with development of a separate mined repository for some DOE-managed HLW and cooler DOE-managed SNF.

  17. Geological Repository Layout for Radioactive High Level Long Lived Waste in Argilite

    SciTech Connect (OSTI)

    Gaussen, J.L.

    2006-07-01

    In the framework of the 1991 French radioactive waste act, ANDRA has studied the feasibility of a geological repository in the argillite layer of the Bure site for high-level long-lived waste. This presentation is focused on the underground facilities that constitute the specific component of this project. The preliminary underground layout, which has been elaborated, is based on four categories of data: - the waste characteristics and inventory; - the geological properties of the host argillite; - the long term performance objectives of the repository; - the specifications in term of operation and reversibility. The underground facilities consist of two types of works: the access works (shafts and drifts) and the disposal cells. The function of the access works is to permit the implementation of two concurrent activities: the nuclear operations (transfer and emplacement of the disposal packages into the disposal cells) and the construction of the next disposal cells. The design of the drifts network which matches up to this function is also influenced by two other specifications: the minimisation of the drift dimensions in order to limit their influence on the integrity of the geological formation and the necessity of a safe ventilation in case of fire. The resulting layout is a network of 4 parallel drifts (2 of them being dedicated to the operation, the other two being dedicated to the construction activities). The average diameter of these access drifts is 7 meters. 4 shafts ensure the link between the surface and the underground. The most important function of the disposal cells is to contribute to the long-term performance of the repository. In this regard, the thermal and geotechnical considerations play an important role. The B wastes (intermediate level wastes) are not (or not very) exothermic. Consequently, the design of their disposal cells result mainly from geotechnical considerations. The disposal packages (made of concrete) are piled up in big cavities the diameter of which is about 10 meters and the length of which is about 250 meters. On the other hand, the design of the C waste disposal cells (vitrified waste) is mainly derived from their thermal power (about 500 W after a 60 year period of interim storage). The disposal cell is a tunnel the diameter of which is about 0,70 m and the length of which is about 40 m. The number of the disposal packages (made of steel) per cell, the spacing between two adjacent canisters within a given cell and the spacing between two adjacent cells are adjusted to limit the peak of temperature in the host formation at 100 deg. C. The disposal cells are also characterized by favourable design factors that would facilitate the potential retrieval of the wastes. The whole underground layout would represent a surface area of several km{sup 2}. (authors)

  18. Disposing of High-Level Radioactive Waste in Germany - A Note from the Licensing Authority - 12530

    SciTech Connect (OSTI)

    Pick, Thomas Stefan; Bluth, Joachim; Lauenstein, Christof; Markhoefer, Joerg

    2012-07-01

    Following the national German consensus on the termination of utilisation of nuclear energy in the summer of 2011, the Federal and Laender Governments have declared their intention to work together on a national consensus on the disposal of radioactive waste as well. Projected in the early 1970's the Federal Government had started exploring the possibility to establish a repository for HLW at the Gorleben site in 1977. However, there is still no repository available in Germany today. The delay results mainly from the national conflict over the suitability of the designated Gorleben site, considerably disrupting German society along the crevice that runs between supporters and opponents of nuclear energy. The Gorleben salt dome is situated in Lower Saxony, the German state that also hosts the infamous Asse mine repository for LLW and ILW and the Konrad repository project designated to receive LLW and ILW as well. With the fourth German project, the Morsleben L/ILW repository only 20 km away across the state border, the state of Lower Saxony carries the main load for the disposal of radioactive waste in Germany. After more than 25 years of exploration and a 10 year moratorium the Gorleben project has now reached a cross-road. Current plans for setting up a new site selection procedure in Germany call for the selection and exploration of up to four alternative sites, depending only on suitable geology. In the meantime the discussion is still open on whether the Gorleben project should be terminated in order to pacify the societal conflict or being kept in the selection process on account of its promising geology. The Lower Saxony Ministry for Environment and Climate Protection proposes to follow a twelve-step-program for finding the appropriate site, including the Gorleben site in the process. With its long history of exploration the site is the benchmark that alternative sites will have to compare with. Following the national consensus of 2011 on the termination of nuclear energy utilisation, it is now the time to reach a national consensus on the disposal of radioactive waste as well. This is a task that the country and society, federal and state governments, political parties and the citizens will have to jointly master within the current generation and within German territory. The basis for the consensus will be a reset to the beginning of this process. It has to start with a new site selection procedure that will take into account and compare up to four alternative sites. This procedure will have to follow the principle of highest possible security. It should be based on a stepwise approach, strictly following scientific criteria. Public confidence in the process and trust can only be achieved by a transparent procedure allowing for the participation of the public and the stakeholders. It is therefore mandatory to consult, both on a national and regional level, all involved parties (public authority, scientist and citizen). The national consensus must also include a decision on the future of the Gorleben exploratory site. The site selection procedure must therefore take this site into account as well. Furthermore, the final decision on safe disposal of German radioactive wastes must be made by sovereign rule by Federal Parliament and Federal Council. (authors)

  19. Hazards and scenarios examined for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Hazards and scenarios examined for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste Rob P. Rechard a,n , Geoff A. Freeze b , Frank V. Perry c a Nuclear Waste Disposal Research & Analysis, Sandia National Laboratories, P.O. Box 5800, Albuquerque 87185-0747, NM, USA b Applied Systems Analysis & Research, Sandia National Laboratories, P.O. Box 5800, Albuquerque 87185-0747, NM, USA c Earth and Environmental Sciences Division, Los Alamos National

  20. Midwestern High-Level Radioactive Waste Transportation Project. Highway infrastructure report

    SciTech Connect (OSTI)

    Sattler, L.R.

    1992-02-01

    In addition to arranging for storage and disposal of radioactive waste, the US Department of Energy (DOE) must develop a safe and efficient transportation system in order to deliver the material that has accumulated at various sites throughout the country. The ability to transport radioactive waste safely has been demonstrated during the past 20 years: DOE has made over 2,000 shipments of spent fuel and other wastes without any fatalities or environmental damage related to the radioactive nature of the cargo. To guarantee the efficiency of the transportation system, DOE must determine the optimal combination of rail transport (which allows greater payloads but requires special facilities) and truck transport Utilizing trucks, in turn, calls for decisions as to when to use legal weight trucks or, if feasible, overweight trucks for fewer but larger shipments. As part of the transportation system, the Facility Interface Capability Assessment (FICA) study contributes to DOE`s development of transportation plans for specific facilities. This study evaluates the ability of different facilities to receive, load and ship the special casks in which radioactive materials will be housed during transport In addition, the DOE`s Near-Site Transportation Infrastructure (NSTI) study (forthcoming) will evaluate the rail, road and barge access to 76 reactor sites from which DOE is obligated to begin accepting spent fuel in 1998. The NSTI study will also assess the existing capabilities of each transportation mode and route, including the potential for upgrade.

  1. Environmental program overview for a high-level radioactive waste repository at Yucca Mountain

    SciTech Connect (OSTI)

    1988-12-01

    The United States plans to begin operating the first repository for the permanent disposal of high-level nuclear waste early in the next century. In February 1983, the US Department of Energy (DOE) identified Yucca Mountain, in Nevada, as one of nine potentially acceptable sites for a repository. To determine its suitability, the DOE evaluated the Yucca Mountain site, along with eight other potentially acceptable sites, in accordance with the DOE`s General Guidelines for the Recommendation of Sites for the Nuclear Waste Repositories. The purpose of the Environmental Program Overview (EPO) for the Yucca Mountain site is to provide an overview of the overall, comprehensive approach being used to satisfy the environmental requirements applicable to sitting a repository at Yucca Mountain. The EPO states how the DOE will address the following environmental areas: aesthetics, air quality, cultural resources (archaeological and Native American components), noise, radiological studies, soils, terrestrial ecosystems, and water resources. This EPO describes the environmental program being developed for the sitting of a repository at Yucca Mountain. 1 fig., 3 tabs.

  2. A guide for the ASME code for austenitic stainless steel containment vessels for high-level radioactive materials

    SciTech Connect (OSTI)

    Raske, D.T.

    1995-06-01

    The design and fabrication criteria recommended by the US Department of Energy (DOE) for high-level radioactive materials containment vessels used in packaging is found in Section III, Division 1, Subsection NB of the ASME Boiler and Pressure Vessel Code. This Code provides material, design, fabrication, examination, and testing specifications for nuclear power plant components. However, many of the requirements listed in the Code are not applicable to containment vessels made from austenitic stainless steel with austenitic or ferritic steel bolting. Most packaging designers, engineers, and fabricators are intimidated by the sheer volume of requirements contained in the Code; consequently, the Code is not always followed and many requirements that do apply are often overlooked during preparation of the Safety Analysis Report for Packaging (SARP) that constitutes the basis to evaluate the packaging for certification.

  3. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers

    SciTech Connect (OSTI)

    Vinson, D.W.; Nutt, W.M.; Bullen, D.B.

    1995-06-01

    Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27.

  4. National survey of crystalline rocks and recommendations of regions to be explored for high-level radioactive waste repository sites

    SciTech Connect (OSTI)

    Smedes, H.W.

    1983-04-01

    A reconnaissance of the geological literature on large regions of exposed crystalline rocks in the United States provides the basis for evaluating if any of those regions warrant further exploration toward identifying potential sites for development of a high-level radioactive waste repository. The reconnaissance does not serve as a detailed evaluation of regions or of any smaller subunits within the regions. Site performance criteria were selected and applied insofar as a national data base exists, and guidelines were adopted that relate the data to those criteria. The criteria include consideration of size, vertical movements, faulting, earthquakes, seismically induced ground motion, Quaternary volcanic rocks, mineral deposits, high-temperature convective ground-water systems, hydraulic gradients, and erosion. Brief summaries of each major region of exposed crystalline rock, and national maps of relevant data provided the means for applying the guidelines and for recommending regions for further study. It is concluded that there is a reasonable likelihood that geologically suitable repository sites exist in each of the major regions of crystalline rocks. The recommendation is made that further studies first be conducted of the Lake Superior, Northern Appalachian and Adirondack, and the Southern Appalachian Regions. It is believed that those regions could be explored more effectively and suitable sites probably could be found, characterized, verified, and licensed more readily there than in the other regions.

  5. EIS-0023: Long-Term Management of Defense High-Level Radioactive Wastes (Research and Development Program for Immobilization), Savannah River Plant, Aiken, South Carolina

    Broader source: Energy.gov [DOE]

    This environmental impact statement (EIS) analyzes the environmental implications of the proposed continuation of a large Federal research and development (R&D) program directed toward the immobilization of the high-level radioactive wastes resulting from chemical separations operations for defense radionuclides production at the DOE Savannah River Plant (SRP) near Aiken, South Carolina.

  6. EIS-0063: Waste Management Operations, Double-Shell Tanks for Defense High-Level Radioactive Waste Storage, Hanford Site, Richland, Washington

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy developed this statement to evaluate the existing tank design and consider additional specific design and safety feature alternatives for the thirteen tanks being constructed for storage of defense high-level radioactive liquid waste at the Hanford Site in Richland, Washington. This statement supplements ERDA-1538, "Final Environmental Statement on Waste Management Operation."

  7. Evaluation of melter system technologies for vitrification of high-sodium content low-level radioactive liquid wastes

    SciTech Connect (OSTI)

    Wilson, C.N.

    1994-03-21

    Westinghouse Hanford Company (WHC) is conducting a two-phased demonstration testing and evaluation of candidate melter system technologies for vitrification of Hanford Site low-level tank wastes. The testing is to be performed by melter equipment and vitrification technology commercial suppliers. This Statement of Work is for Phases 1 and 2 of the demonstration testing program. The primary objective of the demonstration testing is to identify the best available melter system technology for the Hanford Site LLW vitrification facility. Data obtained also will support various WHC engineering studies and conceptual design of the LLW vitrification facility. Multiple technologies will be selected for demonstration and evaluation. Testing will be conducted using non-radioactive LLW simulants in Seller-specified pilot/testing facilities.

  8. EIS-0074: Long-Term Management of Defense High-Level Radioactive Wastes Idaho Chemical Processing Plant, Idaho National Engineering Lab, Idaho

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy prepared this statement to analyze the environmental implications of the proposed selection of a strategy for long-term management of the high-level radioactive wastes generated as part of the national defense effort at the Department's Idaho Chemical Processing Plant at the Idaho National Engineering Laboratory. The project was cancelled after the Draft Environmental Impact Statement was produced.

  9. Progression of performance assessment modeling for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Progression of performance assessment modeling for the Yucca Mountain disposal system for spent nuclear fuel and high-level radioactive waste Rob P. Rechard a,n , Michael L. Wilson b , S. David Sevougian c a Nuclear Waste Disposal Research & Analysis, Sandia National Laboratories, Albuquerque, NM 87185-0747, USA b Systems Analysis/Operations Research, Sandia National Laboratories, Albuquerque, NM 87185-1138, USA c Applied Systems Analysis & Research, Sandia National Laboratories,

  10. High-Level Waste Requirements

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09

    The guide provides the criteria for determining which DOE radioactive wastes are to be managed as high-level waste in accordance with DOE M 435.1-1.

  11. Walk the Line: The Development of Route Selection Standards for Spent Nuclear Fuel and High-level Radioactive Waste in the United States - 13519

    SciTech Connect (OSTI)

    Dilger, Fred; Halstead, Robert J.; Ballard, James D.

    2013-07-01

    Although storage facilities for spent nuclear fuel (SNF) and high-level radioactive waste (HLRW) are widely dispersed throughout the United States, these materials are also relatively concentrated in terms of geographic area. That is, the impacts of storage occur in a very small geographic space. Once shipments begin to a national repository or centralized interim storage facility, the impacts of SNF and HLRW will become more geographically distributed, more publicly visible, and almost certainly more contentious. The selection of shipping routes will likely be a major source of controversy. This paper describes the development of procedures, regulations, and standards for the selection of routes used to ship spent nuclear fuel and high-level radioactive waste in the United States. The paper begins by reviewing the circumstances around the development of HM-164 routing guidelines. The paper discusses the significance of New York City versus the Department of Transportation and application of HM-164. The paper describes the methods used to implement those regulations. The paper will also describe the current HM-164 designated routes and will provide a summary data analysis of their characteristics. This analysis will reveal the relatively small spatial scale of the effects of HM 164. The paper will then describe subsequent developments that have affected route selection for these materials. These developments include the use of 'representative routes' found in the Department of Energy (DOE) 2008 Supplemental Environmental Impact Statement for the formerly proposed Yucca Mountain geologic repository. The paper will describe recommendations related to route selection found in the National Academy of Sciences 2006 report Going the Distance, as well as recommendations found in the 2012 Final Report of the Blue Ribbon Commission on America's Nuclear Future. The paper will examine recently promulgated federal regulations (HM-232) for selection of rail routes for hazardous materials transport. The paper concludes that while the HM 164 regime is sufficient for certain applications, it does not provide an adequate basis for a national plan to ship spent nuclear fuel and high-level radioactive waste to centralized storage and disposal facilities over a period of 30 to 50 years. (authors)

  12. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers. Final report

    SciTech Connect (OSTI)

    Vinson, D.W.; Bullen, D.B.

    1995-09-22

    One of the most significant factors impacting the performance of waste package container materials under repository relevant conditions is the thermal environment. This environment will be affected by the areal power density of the repository, which is dictated by facility design, and the dominant heat transfer mechanism at the site. The near-field environment will evolve as radioactive decay decreases the thermal output of each waste package. Recent calculations (Buscheck and Nitao, 1994) have addressed the importance of thermal loading conditions on waste package performance at the Yucca Mountain site. If a relatively low repository thermal loading design is employed, the temperature and relative humidity near the waste package may significantly affect the degradation of corrosion allowance barriers due to moist air oxidation and radiolytically enhanced corrosion. The purpose this report is to present a literature review of the potential degradation modes for moderately corrosion resistant nickel copper and nickel based candidate materials that may be applicable as alternate barriers for the ACD systems in the Yucca Mountain environment. This report presents a review of the corrosion of nickel-copper alloys, summaries of experimental evaluations of oxidation and atmospheric corrosion in nickel-copper alloys, views of experimental studies of aqueous corrosion in nickel copper alloys, a brief review of galvanic corrosion effects and a summary of stress corrosion cracking in these alloys.

  13. Milestones for Selection, Characterization, and Analysis of the Performance of a Repository for Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain.

    SciTech Connect (OSTI)

    Rechard, Robert P.

    2014-02-01

    This report presents a concise history in tabular form of events leading up to site identification in 1978, site selection in 1987, subsequent characterization, and ongoing analysis through 2008 of the performance of a repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain in southern Nevada. The tabulated events generally occurred in five periods: (1) commitment to mined geologic disposal and identification of sites; (2) site selection and analysis, based on regional geologic characterization through literature and analogous data; (3) feasibility analysis demonstrating calculation procedures and importance of system components, based on rough measures of performance using surface exploration, waste process knowledge, and general laboratory experiments; (4) suitability analysis demonstrating viability of disposal system, based on environment-specific laboratory experiments, in-situ experiments, and underground disposal system characterization; and (5) compliance analysis, based on completed site-specific characterization. Because the relationship is important to understanding the evolution of the Yucca Mountain Project, the tabulation also shows the interaction between four broad categories of political bodies and government agencies/institutions: (a) technical milestones of the implementing institutions, (b) development of the regulatory requirements and related federal policy in laws and court decisions, (c) Presidential and agency directives and decisions, and (d) critiques of the Yucca Mountain Project and pertinent national and world events related to nuclear energy and radioactive waste.

  14. Uncertainty and sensitivity analysis in the 2008 performance assessment for the proposed repository for high-level radioactive waste at Yucca Mountain, Nevada.

    SciTech Connect (OSTI)

    Helton, Jon Craig; Sallaberry, Cedric M.; Hansen, Clifford W.

    2010-05-01

    Extensive work has been carried out by the U.S. Department of Energy (DOE) in the development of a proposed geologic repository at Yucca Mountain (YM), Nevada, for the disposal of high-level radioactive waste. As part of this development, an extensive performance assessment (PA) for the YM repository was completed in 2008 [1] and supported a license application by the DOE to the U.S. Nuclear Regulatory Commission (NRC) for the construction of the YM repository [2]. This presentation provides an overview of the conceptual and computational structure of the indicated PA (hereafter referred to as the 2008 YM PA) and the roles that uncertainty analysis and sensitivity analysis play in this structure.

  15. CRAD, Low-Level Radioactive Waste Management - April 30, 2015...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Low-Level Radioactive Waste Management - April 30, 2015 (EA CRAD 31-11, Rev. 0) CRAD, Low-Level Radioactive Waste Management - April 30, 2015 (EA CRAD 31-11, Rev. 0) April 2015...

  16. Evaluation of plasma melter technology for verification of high-sodium content low-level radioactive liquid wastes: Demonstration test No. 4 preliminary test report

    SciTech Connect (OSTI)

    McLaughlin, D.F.; Gass, W.R.; Dighe, S.V.; D`Amico, N.; Swensrud, R.L.; Darr, M.F.

    1995-01-10

    This document provides a preliminary report of plasma arc vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System Low-Level Waste (LLW) Vitrification Program. Phase I test conduct included 26 hours (24 hours steady state) of melting of simulated high-sodium low-level radioactive liquid waste. Average processing rate was 4.9 kg/min (peak rate 6.2 kg/min), producing 7330 kg glass product. Free-flowing glass pour point was 1250 C, and power input averaged 1530 kW(e), for a total energy consumption of 19,800 kJ/kg glass. Restart capability was demonstrated following a 40-min outage involving the scrubber liquor heat exchanger, and glass production was continued for another 2 hours. Some volatility losses were apparent, probably in the form of sodium borates. Roughly 275 samples were collected and forwarded for analysis. Sufficient process data were collected for heat/material balances. Recommendations for future work include lower boron contents and improved tuyere design/operation.

  17. Illustration of sampling-based approaches to the calculation of expected dose in performance assessments for the proposed high level radioactive waste repository at Yucca Mountain, Nevada.

    SciTech Connect (OSTI)

    Helton, Jon Craig; Sallaberry, Cedric J. PhD.

    2007-04-01

    A deep geologic repository for high level radioactive waste is under development by the U.S. Department of Energy at Yucca Mountain (YM), Nevada. As mandated in the Energy Policy Act of 1992, the U.S. Environmental Protection Agency (EPA) has promulgated public health and safety standards (i.e., 40 CFR Part 197) for the YM repository, and the U.S. Nuclear Regulatory Commission has promulgated licensing standards (i.e., 10 CFR Parts 2, 19, 20, etc.) consistent with 40 CFR Part 197 that the DOE must establish are met in order for the YM repository to be licensed for operation. Important requirements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. relate to the determination of expected (i.e., mean) dose to a reasonably maximally exposed individual (RMEI) and the incorporation of uncertainty into this determination. This presentation describes and illustrates how general and typically nonquantitive statements in 40 CFR Part 197 and 10 CFR Parts 2, 19, 20, etc. can be given a formal mathematical structure that facilitates both the calculation of expected dose to the RMEI and the appropriate separation in this calculation of aleatory uncertainty (i.e., randomness in the properties of future occurrences such as igneous and seismic events) and epistemic uncertainty (i.e., lack of knowledge about quantities that are poorly known but assumed to have constant values in the calculation of expected dose to the RMEI).

  18. EIS-0250-S1: Final Supplemental Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada

    Broader source: Energy.gov [DOE]

    The Proposed Action defined in the Yucca Mountain FEIS is to construct, operate, monitor, and eventually close a geologic repository at Yucca Mountain to dispose of spent nuclear fuel and high-level radioactive waste. The Proposed Action includes transportation of these materials from commercial and DOE sites to the repository.

  19. Groundwater geochemical modeling and simulation of a breached high-level radioactive waste repository in the northern Tularosa Basin, New Mexico

    SciTech Connect (OSTI)

    Chappell, R.W.

    1989-01-01

    The northern Tularosa Basin in south-central New Mexico was ranked favorably as a potential location for a high-level radioactive waste repository by a US Geological Survey pilot screening study of the Basin and Range Province. The favorable ranking was based chiefly on hydrogeologic and descriptive geochemical evidence. A goal of this study was to develop a methodology for predicting the performance of this or any other basin as a potential repository site using geochemical methods. The approach involves first characterizing the groundwater geochemistry, both chemically and isotopically, and reconstructing the probable evolutionary history of, and controls on the ground water chemistry through modeling. In the second phase of the approach, a hypothetically breached repository is introduced into the system, and the mobility of the parent radionuclide, uranium, in the groundwater is predicted. Possible retardation of uranium transport in the downgradient flow direction from the repository by adsorption and mineral precipitation is then considered. The Permian Yeso Formation, the primary aquifer in the northern Tularosa Basin, was selected for study, development and testing of the methodology outlined above. The Yeso Formation contains abundant gypsum and related evaporite minerals, which impart a distinctive chemical signature to the ground water. Ground water data and solubility calculations indicate a conceptual model of irreversible gypsum and dolomite dissolution with concomitant calcite precipitation. Recharge areas are apparent from temperature, {delta}{sup 18}O and {delta}{sup 2} H, and {sup 3}H trends in the aquifer. Corrected {sup 14}C ages range between modern and 31,200 years, and suggest an average ground water velocity of 0.83 m/yr.

  20. Geology of the Yucca Mountain Region, Chapter in Stuckless, J.S., ED., Yucca Mountain, Nevada - A Proposed Geologic Repository for High-Level Radioactive Waste

    SciTech Connect (OSTI)

    J.S. Stuckless; D. O'Leary

    2006-09-25

    Yucca Mountain has been proposed as the site for the Nation's first geologic repository for high-level radioactive waste. This chapter provides the geologic framework for the Yucca Mountain region. The regional geologic units range in age from late Precambrian through Holocene, and these are described briefly. Yucca Mountain is composed dominantly of pyroclastic units that range in age from 11.4 to 15.2 Ma. The proposed repository would be constructed within the Topopah Spring Tuff, which is the lower of two major zoned and welded ash-flow tuffs within the Paintbrush Group. The two welded tuffs are separated by the partly to nonwelded Pah Canyon Tuff and Yucca Mountain Tuff, which together figure prominently in the hydrology of the unsaturated zone. The Quaternary deposits are primarily alluvial sediments with minor basaltic cinder cones and flows. Both have been studied extensively because of their importance in predicting the long-term performance of the proposed repository. Basaltic volcanism began about 10 Ma and continued as recently as about 80 ka with the eruption of cones and flows at Lathrop Wells, approximately 10 km south-southwest of Yucca Mountain. Geologic structure in the Yucca Mountain region is complex. During the latest Paleozoic and Mesozoic, strong compressional forces caused tight folding and thrust faulting. The present regional setting is one of extension, and normal faulting has been active from the Miocene through to the present. There are three major local tectonic domains: (1) Basin and Range, (2) Walker Lane, and (3) Inyo-Mono. Each domain has an effect on the stability of Yucca Mountain.

  1. Production of high intensity radioactive beams

    SciTech Connect (OSTI)

    Nitschke, J.M.

    1990-04-01

    The production of radioactive nuclear beams world-wide is reviewed. The projectile fragmentation and the ISOL approaches are discussed in detail, and the luminosity parameter is used throughout to compare different production methods. In the ISOL approach a thin and a thick target option are distinguished. The role of storage rings in radioactive beam research is evaluated. It is concluded that radioactive beams produced by the projectile fragmentation and the ISOL methods have complementary characteristics and can serve to answer different scientific questions. The decision which kind of facility to build has to depend on the significance and breadth of these questions. Finally a facility for producing a high intensity radioactive beams near the Coulomb barrier is proposed, with an expected luminosity of {approximately}10{sup 39} cm{sup {minus}2} s{sup {minus}1}, which would yield radioactive beams in excess of 10{sup 11} s{sup {minus}1}. 9 refs., 3 figs., 7 tabs.

  2. Directions in low-level radioactive waste management: A brief history of commercial low-level radioactive waste disposal

    SciTech Connect (OSTI)

    Not Available

    1990-10-01

    This report presents a history of commercial low-level radioactive waste management in the United States, with emphasis on the history of six commercially operated low-level radioactive waste disposal facilities. The report includes a brief description of important steps that have been taken during the 1980s to ensure the safe disposal of low-level waste in the 1990s and beyond. These steps include the issuance of Title 10 Code of Federal Regulations Part 61, Licensing Requirements for the Land Disposal of Radioactive Waste, the Low-Level Radioactive Waste Policy Act of 1980, the Low-Level Radioactive Waste Policy Amendments Act of 1985, and steps taken by states and regional compacts to establish additional disposal sites. 42 refs., 13 figs., 1 tab.

  3. System for chemically digesting low level radioactive, solid waste material

    DOE Patents [OSTI]

    Cowan, Richard G.; Blasewitz, Albert G.

    1982-01-01

    An improved method and system for chemically digesting low level radioactive, solid waste material having a high through-put. The solid waste material is added to an annular vessel (10) substantially filled with concentrated sulfuric acid. Concentrated nitric acid or nitrogen dioxide is added to the sulfuric acid within the annular vessel while the sulfuric acid is reacting with the solid waste. The solid waste is mixed within the sulfuric acid so that the solid waste is substantilly fully immersed during the reaction. The off gas from the reaction and the products slurry residue is removed from the vessel during the reaction.

  4. EIS-0250-S2: Supplemental EIS for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada- Nevada Rail Transportation Corridor

    Broader source: Energy.gov [DOE]

    This SEIS is to evaluate the potential environmental impacts of constructing and operating a railroad for shipments of spent nuclear fuel and high-level radioactive waste from an existing rail line in Nevada to a geologic repository at Yucca Mountain. The purpose of the evaluation is to assist the Department in deciding whether to construct and operate a railroad in Nevada, and if so, in which corridor and along which specific alignment within the selected corridor.

  5. Bibliographic Data on Low-Level Radioactive Waste Documents

    Energy Science and Technology Software Center (OSTI)

    1995-11-10

    The purpose of the system is to allow users (researchers, policy makers, etc) to identify existing documents on a range of subjects related to low-level radioactive waste management. The software is menu driven.

  6. EIS-0250: Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada

    Broader source: Energy.gov [DOE]

    This EIS analyzes DOE's proposed action to construct, operate, monitor, and eventually close a geologic repository at Yucca Mountain  for the disposal of spent nuclear fuel and high-level...

  7. EIS-0081: Long-Term Management of Liquid High-Level Radioactive Waste Stored at Western New York Nuclear Service Center, West Valley, New York

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy Office of Terminal Waste Disposal and Remedial Action prepared this environmental impact statement to analyze the environmental and socioeconomic impacts resulting from the Department’s proposed action to construct and operate facilities necessary to solidify the liquid high-level wastes currently stored in underground tanks at West Valley, New York.

  8. Low-level radioactive waste disposal facility closure

    SciTech Connect (OSTI)

    White, G.J.; Ferns, T.W.; Otis, M.D.; Marts, S.T.; DeHaan, M.S.; Schwaller, R.G.; White, G.J. )

    1990-11-01

    Part I of this report describes and evaluates potential impacts associated with changes in environmental conditions on a low-level radioactive waste disposal site over a long period of time. Ecological processes are discussed and baselines are established consistent with their potential for causing a significant impact to low-level radioactive waste facility. A variety of factors that might disrupt or act on long-term predictions are evaluated including biological, chemical, and physical phenomena of both natural and anthropogenic origin. These factors are then applied to six existing, yet very different, low-level radioactive waste sites. A summary and recommendations for future site characterization and monitoring activities is given for application to potential and existing sites. Part II of this report contains guidance on the design and implementation of a performance monitoring program for low-level radioactive waste disposal facilities. A monitoring programs is described that will assess whether engineered barriers surrounding the waste are effectively isolating the waste and will continue to isolate the waste by remaining structurally stable. Monitoring techniques and instruments are discussed relative to their ability to measure (a) parameters directly related to water movement though engineered barriers, (b) parameters directly related to the structural stability of engineered barriers, and (c) parameters that characterize external or internal conditions that may cause physical changes leading to enhanced water movement or compromises in stability. Data interpretation leading to decisions concerning facility closure is discussed. 120 refs., 12 figs., 17 tabs.

  9. Geologic and hydrologic characterization and evaluation of the Basin and Range Province relative to the disposal of high-level radioactive waste. Part I. Introduction and guidelines

    SciTech Connect (OSTI)

    Bedinger, M.S.; Sargent, K.A.; Reed, J.E.

    1984-12-31

    The US Geological Survey`s program for geologic and hydrologic evaluation of physiographic provinces to identify areas potentially suitable for locating repository sites for disposal of high-level nuclear wastes was announced to the Governors of the eight states in the Basin and Range Province on May 5, 1981. Representatives of Arizona, California, Idaho, New Mexico, Nevada, Oregon, Texas, and Utah, were invited to cooperate with the federal government in the evaluation process. Each governor was requested to nominate an earth scientist to represent the state in a province working group composed of state and US Geological Survey representatives. This report, Part I of a three-part report, provides the background, introduction and scope of the study. This part also includes a discussion of geologic and hydrologic guidelines that will be used in the evaluation process and illustrates geohydrologic environments and the effect of individual factors in providing multiple natural barriers to radionuclide migration. 27 refs., 6 figs., 1 tab.

  10. Process for solidifying high-level nuclear waste

    DOE Patents [OSTI]

    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.

  11. Operating Experience Level 3 2012-01: Control of Sealed Radioactive Sources

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    | Department of Energy Operating Experience Level 3 2012-01: Control of Sealed Radioactive Sources Operating Experience Level 3 2012-01: Control of Sealed Radioactive Sources February 1, 2012 OE-3 2012-01: Control of Sealed Radioactive Sources This Operating Experience Level 3 document provides information on a safety concern releated to use of a sealed radioactive source where hazard controls were not evaluated. PDF icon OE-3 2012-01: Control of Sealed Radioactive Sources More Documents

  12. Low-level radioactive waste technology: a selected, annotated bibliography

    SciTech Connect (OSTI)

    Fore, C.S.; Vaughan, N.D.; Hyder, L.K.

    1980-10-01

    This annotated bibliography of 447 references contains scientific, technical, economic, and regulatory information relevant to low-level radioactive waste technology. The bibliography focuses on environmental transport, disposal site, and waste treatment studies. The publication covers both domestic and foreign literature for the period 1952 to 1979. Major chapters selected are Chemical and Physical Aspects; Container Design and Performance; Disposal Site; Environmental Transport; General Studies and Reviews; Geology, Hydrology and Site Resources; Regulatory and Economic Aspects; Transportation Technology; Waste Production; and Waste Treatment. Specialized data fields have been incorporated into the data file to improve the ease and accuracy of locating pertinent references. Specific radionuclides for which data are presented are listed in the Measured Radionuclides field, and specific parameters which affect the migration of these radionuclides are presented in the Measured Parameters field. In addition, each document referenced in this bibliography has been assigned a relevance number to facilitate sorting the documents according to their pertinence to low-level radioactive waste technology. The documents are rated 1, 2, 3, or 4, with 1 indicating direct applicability to low-level radioactive waste technology and 4 indicating that a considerable amount of interpretation is required for the information presented to be applied. The references within each chapter are arranged alphabetically by leading author, corporate affiliation, or title of the document. Indexes are provide for (1) author(s), (2) keywords, (3) subject category, (4) title, (5) geographic location, (6) measured parameters, (7) measured radionuclides, and (8) publication description.

  13. Development of low-level radioactive waste disposal capacity in the United States - progress or stalemate?

    SciTech Connect (OSTI)

    Devgun, J.S. [Argonne National Lab., IL (United States); Larson, G.S. [Midwest Low-Level Radioactive Waste Commission, St. Paul, MN (United States)

    1995-12-31

    It has been fifteen years since responsibility for the disposal of commercially generated low-level radioactive waste (LLW) was shifted to the states by the United States Congress through the Low-Level Radioactive Waste Policy Act of 1980 (LLRWPA). In December 1985, Congress revisited the issue and enacted the Low-Level Radioactive Waste Policy Amendments Act of 1985 (LLRWPAA). No new disposal sites have opened yet, however, and it is now evident that disposal facility development is more complex, time-consuming, and controversial than originally anticipated. For a nation with a large nuclear power industry, the lack of availability of LLW disposal capacity coupled with a similar lack of high-level radioactive waste disposal capacity could adversely affect the future viability of the nuclear energy option. The U.S. nuclear power industry, with 109 operating reactors, generates about half of the LLW shipped to commercial disposal sites and faces dwindling access to waste disposal sites and escalating waste management costs. The other producers of LLW - industries, government (except the defense related research and production waste), academic institutions, and medical institutions that account for the remaining half of the commercial LLW - face the same storage and cost uncertainties. This paper will summarize the current status of U.S. low-level radioactive waste generation and the status of new disposal facility development efforts by the states. The paper will also examine the factors that have contributed to delays, the most frequently suggested alternatives, and the likelihood of change.

  14. EIS-0023: Long-Term Management of Defense High-Level Radioactive Wastes (Research and Development Program for Immobilization) Savannah River Plant, Aiken, South Carolina

    Broader source: Energy.gov [DOE]

    This EIS analyzes the potential environmental implications of the proposed continuation of a large Federal research and development (R&D) program directed toward the immobilization of the high...

  15. Test plan for glass melter system technologies for vitrification of high-sodium content low-level radioactive liquid waste, Project No. RDD-43288

    SciTech Connect (OSTI)

    Higley, B.A.

    1995-03-15

    This document provides a test plan for the conduct of combustion fired cyclone vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System, Low-Level Waste Vitrification Program. The vendor providing this test plan and conducting the work detailed within it is the Babcock & Wilcox Company Alliance Research Center in Alliance, Ohio. This vendor is one of seven selected for glass melter testing.

  16. Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) ...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW) A transuranic (TRU) waste shipment makes its way to the ...

  17. EIS-0375: Disposal of Greater-than-Class-C Low-Level Radioactive...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    5: Disposal of Greater-than-Class-C Low-Level Radioactive Waste and Department of Energy GTCC-like Waste EIS-0375: Disposal of Greater-than-Class-C Low-Level Radioactive Waste and ...

  18. Operating Experience Level 3 2012-01: Control of Sealed Radioactive...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Operating Experience Level 3 2012-01: Control of Sealed Radioactive Sources Operating Experience Level 3 2012-01: Control of Sealed Radioactive Sources February 1, 2012 OE-3 ...

  19. Test plan for evaluation of plasma melter technology for vitrification of high-sodium content low-level radioactive liquid wastes

    SciTech Connect (OSTI)

    McLaughlin, D.F.; Lahoda, E.J.; Gass, W.R.; D`Amico, N.

    1994-10-20

    This document provides a test plan for the conduct of plasma arc vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. The vendor providing this test plan and conducting the work detailed within it [one of seven selected for glass melter testing under Purchase Order MMI-SVV-384212] is the Westinghouse Science and Technology Center (WSTC) in Pittsburgh, PA. WSTC authors of the test plan are D. F. McLaughlin, E. J. Lahoda, W. R. Gass, and N. D`Amico. The WSTC Program Manager for this test is D. F. McLaughlin. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes melting of glass frit with Hanford LLW Double-Shell Slurry Feed waste simulant in a plasma arc fired furnace.

  20. Test Plan: Phase 1 demonstration of 3-phase electric arc melting furnace technology for vitrifying high-sodium content low-level radioactive liquid wastes

    SciTech Connect (OSTI)

    Eaton, W.C.

    1995-05-31

    This document provides a test plan for the conduct of electric arc vitrification testing by a vendor in support of the Hanford Tank Waste Remediation System (TWRS) Low-Level Waste (LLW) Vitrification Program. The vendor providing this test plan and conducting the work detailed within it [one of seven selected for glass melter testing under Purchase Order MMI-SVV-384216] is the US Bureau of Mines, Department of the Interior, Albany Research Center, Albany, Oregon. This test plan is for Phase I activities described in the above Purchase Order. Test conduct includes feed preparation activities and melting of glass with Hanford LLW Double-Shell Slurry Feed waste simulant in a 3-phase electric arc (carbon electrode) furnace.

  1. Commercial low-level radioactive waste transportation liability and radiological risk

    SciTech Connect (OSTI)

    Quinn, G.J.; Brown, O.F. II; Garcia, R.S.

    1992-08-01

    This report was prepared for States, compact regions, and other interested parties to address two subjects related to transporting low-level radioactive waste to disposal facilities. One is the potential liabilities associated with low-level radioactive waste transportation from the perspective of States as hosts to low-level radioactive waste disposal facilities. The other is the radiological risks of low-level radioactive waste transportation for drivers, the public, and disposal facility workers.

  2. Application of EPA regulations to low-level radioactive waste

    SciTech Connect (OSTI)

    Bowerman, B.S.; Piciulo, P.L.

    1985-01-01

    The survey reported here was conducted with the intent of identifying categories of low-level radioactive wastes which would be classified under EPA regulations 40 CFR Part 261 as hazardous due to the chemical properties of the waste. Three waste types are identified under these criteria as potential radioactive mixed wastes: wastes containing organic liquids; wastes containing lead metal; and wastes containing chromium. The survey also indicated that certain wastes, specific to particular generators, may also be radioactive mixed wastes. Ultimately, the responsibility for determining whether a facility's wastes are mixed wastes rest with the generator. However, the uncertainties as to which regulations are applicable, and the fact that no legal definition of mixed wastes exists, make such a determination difficult. In addition to identifying mixed wastes, appropriate methods for the management of mixed wastes must be defined. In an ongoing study, BNL is evaluating options for the management of mixed wastes. These options will include segregation, substitution, and treatments to reduce or eliminate chemical hazards associated with the wastes listed above. The impacts of the EPA regulations governing hazardous wastes on radioactive mixed waste cannot be assessed in detail until the applicability of these regulations is agreed upon. This issue is still being discussed by EPA and NRC and should be resolved in the near future. Areas of waste management which may affect generators of mixed wastes include: monitoring/tracking of wastes before shipment; chemical testing of wastes; permits for treatment of storage of wastes; and additional packaging requirements. 3 refs., 1 fig., 2 tabs.

  3. Hanford Site Shares Lessons Learned in Retrieving Highly Radioactive

    Energy Savers [EERE]

    Material | Department of Energy Shares Lessons Learned in Retrieving Highly Radioactive Material Hanford Site Shares Lessons Learned in Retrieving Highly Radioactive Material January 29, 2014 - 12:00pm Addthis A team from the Oak Ridge Office of Environmental Management observes equipment that will be used to retrieve highly radioactive sludge at the Hanford site. This pump was modified to fit the underwater environment where the sludge is stored. A team from the Oak Ridge Office of

  4. Low-level radioactive waste disposal technologies used outside the United States

    SciTech Connect (OSTI)

    Templeton, K.J.; Mitchell, S.J.; Molton, P.M.; Leigh, I.W.

    1994-01-01

    Low-level radioactive waste (LLW) disposal technologies are an integral part of the waste management process. In the United States, commercial LLW disposal is the responsibility of the State or groups of States (compact regions). The United States defines LLW as all radioactive waste that is not classified as spent nuclear fuel, high- level radioactive waste, transuranic waste, or by-product material as defined in Section II(e)(2) of the Atomic Energy Act. LLW may contain some long-lived components in very low concentrations. Countries outside the United States, however, may define LLW differently and may use different disposal technologies. This paper outlines the LLW disposal technologies that are planned or being used in Canada, China, Finland, France, Germany, Japan, Sweden, Taiwan, and the United Kingdom (UK).

  5. The siting record: An account of the programs of federal agencies and events that have led to the selection of a potential site for a geologic respository for high-level radioactive waste

    SciTech Connect (OSTI)

    Lomenick, T.F.

    1996-03-01

    This record of siting a geologic repository for high-level radioactive wastes (HLW) and spent fuel describes the many investigations that culminated on December 22, 1987 in the designation of Yucca Mountain (YM), as the site to undergo detailed geologic characterization. It recounts the important issues and events that have been instrumental in shaping the course of siting over the last three and one half decades. In this long task, which was initiated in 1954, more than 60 regions, areas, or sites involving nine different rock types have been investigated. This effort became sharply focused in 1983 with the identification of nine potentially suitable sites for the first repository. From these nine sites, five were subsequently nominated by the U.S. Department of Energy (DOE) as suitable for characterization and then, in 1986, as required by the Nuclear Waste Policy Act of 1982 (NWPA), three of these five were recommended to the President as candidates for site characterization. President Reagan approved the recommendation on May 28, 1986. DOE was preparing site characterization plans for the three candidate sites, namely Deaf Smith County, Texas; Hanford Site, Washington; and YM. As a consequence of the 1987 Amendment to the NWPA, only the latter was authorized to undergo detailed characterization. A final Site Characterization Plan for Yucca Mountain was published in 1988. Prior to 1954, there was no program for the siting of disposal facilities for high-level waste (HLW). In the 1940s and 1950s, the volume of waste, which was small and which resulted entirely from military weapons and research programs, was stored as a liquid in large steel tanks buried at geographically remote government installations principally in Washington and Tennessee.

  6. Low-level radioactive waste form qualification testing

    SciTech Connect (OSTI)

    Sohal, M.S.; Akers, D.W.

    1998-06-01

    This report summarizes activities that have already been completed as well as yet to be performed by the Idaho National Engineering and Environmental Laboratory (INEEL) to develop a plan to quantify the behavior of radioactive low-level waste forms. It briefly describes the status of various tasks, including DOE approval of the proposed work, several regulatory and environmental related documents, tests to qualify the waste form, preliminary schedule, and approximate cost. It is anticipated that INEEL and Brookhaven National Laboratory will perform the majority of the tests. For some tests, services of other testing organizations may be used. It should take approximately nine months to provide the final report on the results of tests on a waste form prepared for qualification. It is anticipated that the overall cost of the waste quantifying service is approximately $150,000. The following tests are planned: compression, thermal cycling, irradiation, biodegradation, leaching, immersion, free-standing liquid tests, and full-scale testing.

  7. High-Level Waste Inventory

    Office of Environmental Management (EM)

    Analysis of Alternatives for Disposition of the Idaho Calcined High-Level Waste Inventory ... of the Idaho Calcined High-Level Waste Inventory Volume 1- Summary Report April ...

  8. First of Hanford's Highly Radioactive Sludge Moved Away from River |

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Department of Energy of Hanford's Highly Radioactive Sludge Moved Away from River First of Hanford's Highly Radioactive Sludge Moved Away from River July 13, 2012 - 12:00pm Addthis Media Contacts Geoff Tyree, DOE Geoffrey.Tyree@rl.doe.gov 509-376-4171 Dee Millikin, CH2M HILL Dee_Millikin@rl.doe.gov 509-376-1297 RICHLAND, Wash. - Workers have started moving highly radioactive material, called sludge, away from the Columbia River, marking a significant milestone in the U. S. Department of

  9. High-Tech Collaboration Radioactive Learning

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Tech Collaboration Radioactive Learning NSTec, UNLV work together to design new chip. RSL gains attention for efforts during Super Bowl XLIX. Local students learn truths about radiation during visit. See page 4. See page 5. Deputy Administrator Creedon Visits the NNSS Madelyn Creedon, deputy administrator for the National Nuclear Security Administration (NNSA), made her first visit in her NNSA role to the Nevada National Security Site (NNSS) in early February. Creedon has a long history with the

  10. 1996 annual report on low-level radioactive waste management progress. Report to Congress

    SciTech Connect (OSTI)

    1997-11-01

    This report is prepared in response to the Low-Level Radioactive Waste Policy Act (the Act), Public Law 96-573, 1980, as amended by the Low-Level Radioactive Waste Policy Amendments Act of 1985, Public Law 99-240. The report summarizes the activities during calendar year 1996 related to the establishment of new disposal facilities for commercially-generated low-level radioactive waste. The report emphasizes significant issues and events that have affected progress in developing new disposal facilities, and also includes an introduction that provides background information and perspective on US policy for low-level radioactive waste disposal.

  11. Report to Congress: 1995 Annual report on low-level radioactive waste management progress

    SciTech Connect (OSTI)

    1996-06-01

    This report is prepared in response to the Low-Level Radioactive Waste Policy Act, Public Law 96-573, 1980, as amended by the Low-Level Radioactive Waste Policy Amendments Act of 1985, Public Law 99-240. The report summarizes the progress of states and compact regions during calendar year 1995 in establishing new disposal facilities for commercially-generated low-level radioactive waste. The report emphasizes significant issues and events that have affected progress, and also includes an introduction that provides background information and perspective on United States policy for low-level radioactive waste disposal.

  12. Selected radionuclides important to low-level radioactive waste management

    SciTech Connect (OSTI)

    1996-11-01

    The purpose of this document is to provide information to state representatives and developers of low level radioactive waste (LLW) management facilities about the radiological, chemical, and physical characteristics of selected radionuclides and their behavior in the environment. Extensive surveys of available literature provided information for this report. Certain radionuclides may contribute significantly to the dose estimated during a radiological performance assessment analysis of an LLW disposal facility. Among these are the radionuclides listed in Title 10 of the Code of Federal Regulations Part 61.55, Tables 1 and 2 (including alpha emitting transuranics with half-lives greater than 5 years). This report discusses these radionuclides and other radionuclides that may be significant during a radiological performance assessment analysis of an LLW disposal facility. This report not only includes essential information on each radionuclide, but also incorporates waste and disposal information on the radionuclide, and behavior of the radionuclide in the environment and in the human body. Radionuclides addressed in this document include technetium-99, carbon-14, iodine-129, tritium, cesium-137, strontium-90, nickel-59, plutonium-241, nickel-63, niobium-94, cobalt-60, curium -42, americium-241, uranium-238, and neptunium-237.

  13. Choosing solidification or vitrification for low-level radioactive and mixed waste treatment

    SciTech Connect (OSTI)

    Gimpel, R.F.

    1992-02-14

    Solidification (making concrete) and vitrification (making glass) are frequently the treatment methods recommended for treating inorganic or radioactive wastes. Solidification is generally perceived as the most economical treatment method. Whereas, vitrification is considered (by many) as the most effective of all treatment methods. Unfortunately, vitrification has acquired the stigma that it is too expensive to receive further consideration as an alternative to solidification in high volume treatment applications. Ironically, economic studies, as presented in this paper, show that vitrification may be more competitive in some high volume applications. Ex-situ solidification and vitrification are the competing methods for treating in excess of 450 000m{sup 3} of low-level radioactive and mixed waste at the Fernald Environmental Management Project (FEMP or simply, Fernald) located near Cincinnati, Ohio. This paper summarizes how Fernald is choosing between solidification and vitrification as the primary waste treatment method.

  14. Microbial degradation of low-level radioactive waste. Final report

    SciTech Connect (OSTI)

    Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W. Jr

    1996-06-01

    The Nuclear Regulatory Commission stipulates in 10 CFR 61 that disposed low-level radioactive waste (LLW) be stabilized. To provide guidance to disposal vendors and nuclear station waste generators for implementing those requirements, the NRC developed the Technical Position on Waste Form, Revision 1. That document details a specified set of recommended testing procedures and criteria, including several tests for determining the biodegradation properties of waste forms. Information has been presented by a number of researchers, which indicated that those tests may be inappropriate for examining microbial degradation of cement-solidified LLW. Cement has been widely used to solidify LLW; however, the resulting waste forms are sometimes susceptible to failure due to the actions of waste constituents, stress, and environment. The purpose of this research program was to develop modified microbial degradation test procedures that would be more appropriate than the existing procedures for evaluation of the effects of microbiologically influenced chemical attack on cement-solidified LLW. The procedures that have been developed in this work are presented and discussed. Groups of microorganisms indigenous to LLW disposal sites were employed that can metabolically convert organic and inorganic substrates into organic and mineral acids. Such acids aggressively react with cement and can ultimately lead to structural failure. Results on the application of mechanisms inherent in microbially influenced degradation of cement-based material are the focus of this final report. Data-validated evidence of the potential for microbially influenced deterioration of cement-solidified LLW and subsequent release of radionuclides developed during this study are presented.

  15. 18th U.S. Department of Energy Low-Level Radioactive Waste Management Conference. Program

    SciTech Connect (OSTI)

    1997-05-20

    This conference explored the latest developments in low-level radioactive waste management through presentations from professionals in both the public and the private sectors and special guests. The conference included two continuing education seminars, a workshop, exhibits, and a tour of Envirocare of Utah, Inc., one of America's three commercial low-level radioactive waste depositories.

  16. 1989 Annual report on low-level radioactive waste management progress

    SciTech Connect (OSTI)

    Not Available

    1990-10-01

    This report summarizes the progress during 1989 of states and compacts in establishing new low-level radioactive waste disposal facilities. It also provides summary information on the volume of low-level waste received for disposal in 1989 by commercially operated low-level waste disposal facilities. This report is in response to Section 7(b) of Title I of Public Law 99--240, the Low-Level Radioactive Waste Policy Amendments Act of 1985. 2 figs., 5 tabs.

  17. High Level Waste Management Division High. Level Waste System Plan

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    HLW -OVP-98-0037 High Level Waste Management Division High. Level Waste System Plan Revision 9 (U) April 1998 Westinghouse Savannah River Company Savannah River Site Aiken, SC 29808 Prepared for the U. S. Department of Energy under contract no. DE-AC09-96SR18500 HLW -OVP-98-0037 High Level Waste Management Division High Level Waste System Plan Revision 9 (U) Contributors: A. S. Choi P. Paul F. E. Wise Prepared by: ?1M.J II£) ~ N. R. Davis Approved by: HLW System Integration Manager ll\1-'-ft

  18. EIS-0303: Savannah River Site High-Level Waste Tank Closure

    Broader source: Energy.gov [DOE]

    This EIS evaluates alternatives for closing 49 high-level radioactive waste tanks and associated equipment such as evaporator systems, transfer pipelines, diversion boxes, and pump pits. DOE...

  19. Office of Enterprise Assessments Operational Awareness Record for the Observation of the Waste Treatment and Immobilization Plant High Level Waste Faciity Concentrate Receipt/Melter Feed/Glass Formers Reagent Hazard Analysis and Review of the Radioactive Liquid Disposal Hazards Analysis Event Tables - March 2015

    Energy Savers [EERE]

    Operational Awareness Record Report Number: EA-WTP-HLW-2014-10-20 Site: Hanford Site Subject: Observation of the Waste Treatment and Immobilization Plant High Level Waste Facility Concentrate Receipt/Melter Feed/ Glass Formers Reagent Hazards Analysis Activities and Review of the Radioactive Liquid Disposal Hazards Analysis Event Tables Dates of Activity: 10/20/14 - 11/06/14 Report Preparer: James O. Low Activity Description/Purpose: Bechtel National, Incorporated (BNI) is implementing a Safety

  20. Low-level radioactive waste management at the Nevada Test Site -- Current status

    SciTech Connect (OSTI)

    Becker, B.D.; Crowe, B.M.; Gertz, C.P.; Clayton, W.A.

    1999-04-01

    The performance objectives of the Department of Energy`s Low-Level Radioactive Waste (LLW) disposal facilities located at the Nevada Test Site transcend those of any other radioactive waste disposal site in the US. Situated at the southern end of the Great Basin, 800 feet above the water table, the Area 5 Radioactive Waste Management Site (RWMS) has utilized a combination of engineered shallow land disposal cells and deep augured shafts to dispose a variety of waste streams. These include high volume low-activity wastes, classified materials, and high-specific-activity special case wastes. Twenty miles north of Area 5 is the Area 3 RWMS. Here bulk LLW disposal takes place in subsidence craters formed from underground testing of nuclear weapons. Earliest records indicate that documented LLW disposal activities have occurred at the Area 5 and Area 3 RWMS`s since 1961 and 1968, respectively. However, these activities have only been managed under a formal program since 1978. This paper describes the technical attributes of the facilities, present and future capacities and capabilities, and provides a description of the process from waste approval to final disposition. The paper also summarizes the current status of the waste disposal operations.

  1. Greater-than-Class C low-level radioactive waste characterization. Appendix E-4: Packaging factors for greater-than-Class C low-level radioactive waste

    SciTech Connect (OSTI)

    Quinn, G.; Grant, P.; Winberg, M.; Williams, K.

    1994-09-01

    This report estimates packaging factors for several waste types that are potential greater-than-Class C (GTCC) low-level radioactive waste (LLW). The packaging factor is defined as the volume of a GTCC LLW disposal container divided by the as-generated or ``unpackaged`` volume of the waste loaded into the disposal container. Packaging factors reflect any processes that reduce or increase an original unpackaged volume of GTCC LLW, the volume inside a waste container not occupied by the waste, and the volume of the waste container itself. Three values are developed that represent (a) the base case or most likely value for a packaging factor, (b) a high case packaging factor that corresponds to the largest anticipated disposal volume of waste, and (c) a low case packaging factor for the smallest volume expected. GTCC LLW is placed in three categories for evaluation in this report: activated metals, sealed sources, and all other waste.

  2. Time of Compliance for Disposal of Low-Level Radioactive Waste | Department

    Energy Savers [EERE]

    of Energy Time of Compliance for Disposal of Low-Level Radioactive Waste Time of Compliance for Disposal of Low-Level Radioactive Waste Roger Seitz*, Savannah River National Laboratory ; Andrew Wallo, U.S. Department of Energy Abstract: The United States Department of Energy (DOE) has more than 25 years of experience conducting and overseeing performance assessments (PAs) for low-level waste (LLW) and mixed LLW from on-going operations, decommissioning and environmental restoration

  3. ECONOMIC ASSESSMENT ON VITRIFICATION FACILITY OF LOW-AND INTERMEDIATE-LEVEL RADIOACTIVE WASTES IN KOREA

    SciTech Connect (OSTI)

    Kim, Sung Il; Lee, Kun Jai; Ji, Pyung Kook; Park, Jong Kil; Ha, Jong Hyun; Song, Myung Jae

    2003-02-27

    The usefulness of vitrification technology for low-and intermediate-level radioactive wastes was demonstrated with high volume reduction capability and good mechanical and chemical stability of final waste forms, and commercial vitrification facility is expected to be constructed at Ulchin site of Korean Nuclear Power Plant Ulchin Unit 5 and 6. Hence, overall economic assessment was necessary to find out the economic advantage of the vitrification facility and to predict the construction and operation costs of the facility on the preliminary bases. Additionally, the generation characteristics of radioactive wastes were investigated. The results of the cost analysis showed that the disposal cost of radioactive wastes treated by vitrification facility reduced to 85 percent compared with that by current waste treatment system. And the present worth analysis was performed through the cost-benefit analysis method for the commercial vitrification facility. The results showed that the vitrification facility combining cold crucible melter (CCM) for treatment of combustible DAW, spent resin, and borated liquid waste concentrate and plasma torch melter (PTM) for non-combustible DAW and spent filter is more economical than current waste treatment system when the escalation rate of disposal cost of more than 10 percent per year was applied.

  4. West Valley low-level radioactive waste site revisited: Microbiological analysis of leachates

    SciTech Connect (OSTI)

    Gillow, J.B.; Francis, A.J.

    1990-10-01

    The abundance and types of microorganisms in leachate samples from the West Valley low-level radioactive waste disposal site were enumerated. This study was undertaken in support of the study conducted by Ecology and Environment, Inc., to assess the extent of radioactive gas emissions from the site. Total aerobic and anaerobic bacteria were enumerated as colony forming units (CFU) by dilution agar plate technique, and denitrifiers, sulfate-reducers and methanogens by the most probable number technique (MPN). Of the three trenches 3, 9, and 11 sampled, trench 11 contained the most number of organisms in the leachate. Concentrations of carbon-14 and tritium were highest in trench 11 leachate. Populations of aerobes and anaerobes in trench 9 leachate were one order of magnitude less than in trench 11 leachate while the methanogens were three orders of magnitude greater than in trench 11 leachate. The methane content from trench 9 was high due to the presence of a large number of methanogens; the gas in this trench also contained the most radioactivity. Trench 3 leachate contained the least number of microorganisms. Comparison of microbial populations in leachates sampled from trenches 3 and 9 during October 1978 and 1989 showed differences in the total number of microbial types. Variations in populations of the different types of organisms in the leachate reflect the changing nutrient conditions in the trenches. 14 refs., 3 figs., 4 tabs.

  5. Geologyy of the Yucca Mountain Site Area, Southwestern Nevada, Chapter in Stuckless, J.S., ED., Yucca Mountain, Nevada - A Proposed Geologic Repository for High-Level Radioactive Waste (Volume 1)

    SciTech Connect (OSTI)

    W.R. Keefer; J.W. Whitney; D.C. Buesch

    2006-09-25

    Yucca Mountain in southwestern Nevada is a prominent, irregularly shaped upland formed by a thick apron of Miocene pyroclastic-flow and fallout tephra deposits, with minor lava flows, that was segmented by through-going, large-displacement normal faults into a series of north-trending, eastwardly tilted structural blocks. The principal volcanic-rock units are the Tiva Canyon and Topopah Spring Tuffs of the Paintbrush Group, which consist of volumetrically large eruptive sequences derived from compositionally distinct magma bodies in the nearby southwestern Nevada volcanic field, and are classic examples of a magmatic zonation characterized by an upper crystal-rich (> 10% crystal fragments) member, a more voluminous lower crystal-poor (< 5% crystal fragments) member, and an intervening thin transition zone. Rocks within the crystal-poor member of the Topopah Spring Tuff, lying some 280 m below the crest of Yucca Mountain, constitute the proposed host rock to be excavated for the storage of high-level radioactive wastes. Separation of the tuffaceous rock formations into subunits that allow for detailed mapping and structural interpretations is based on macroscopic features, most importantly the relative abundance of lithophysae and the degree of welding. The latter feature, varying from nonwelded through partly and moderately welded to densely welded, exerts a strong control on matrix porosities and other rock properties that provide essential criteria for distinguishing hydrogeologic and thermal-mechanical units, which are of major interest in evaluating the suitability of Yucca Mountain to host a safe and permanent geologic repository for waste storage. A thick and varied sequence of surficial deposits mantle large parts of the Yucca Mountain site area. Mapping of these deposits and associated soils in exposures and in the walls of trenches excavated across buried faults provides evidence for multiple surface-rupturing events along all of the major faults during Pleistocene and Holocene times; these paleoseismic studies form the basis for evaluating the potential for future earthquakes and fault displacements. Thermoluminescence and U-series analyses were used to date the surficial materials involved in the Quaternary faulting events. The rate of erosional downcutting of bedrock on the ridge crests and hillslopes of Yucca Mountain, being of particular concern with respect to the potential for breaching of the proposed underground storage facility, was studied by using rock varnish cation-ratio and {sup 10}Be and {sup 36}Cl cosmogenic dating methods to determine the length of time bedrock outcrops and hillslope boulder deposits were exposed to cosmic rays, which then served as a basis for calculating long-term erosion rates. The results indicate rates ranging from 0.04 to 0.27 cm/k.y., which represent the maximum downcutting along the summit of Yucca Mountain under all climatic conditions that existed there during most of Quaternary time. Associated studies include the stratigraphy of surficial deposits in Fortymile Wash, the major drainage course in the area, which record a complex history of four to five cut-and-fill cycles within the channel during middle to late Quaternary time. The last 2 to 4 m of incision probably occurred during the last pluvial climatic period, 22 to 18 ka, followed by aggradation to the present time.

  6. Disposal of low-level and mixed low-level radioactive waste during 1990

    SciTech Connect (OSTI)

    Not Available

    1993-08-01

    Isotopic inventories and other data are presented for low-level radioactive waste (LLW) and mixed LLW disposed (and occasionally stored) during calendar year 1990 at commercial disposal facilities and Department of Energy (DOE) sites. Detailed isotopic information is presented for the three commercial disposal facilities located near Barnwell, SC, Richland, WA, and Beatty, NV. Less information is presented for the Envirocare disposal facility located near Clive, UT, and for LLW stored during 1990 at the West Valley site. DOE disposal information is included for the Savannah River Site (including the saltstone facility), Nevada Test Site, Los Alamos National Laboratory, Idaho National Engineering Laboratory, Hanford Site, Y-12 Site, and Oak Ridge National Laboratory. Summary information is presented about stored DOE LLW. Suggestions are made about improving LLW disposal data.

  7. South Carolina State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-08-01

    The South Carolina State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in South Carolina. The profile is the result of a survey of NRC licensees in South Carolina. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as definied by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in South Carolina.

  8. Pennsylvania State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-04-01

    The Pennsylvania State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Pennsylvania. The profile is the result of a survey of NRC licensees in Pennsylvania. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Pennsylvania.

  9. 1994 annual report on low-level radioactive waste management progress

    SciTech Connect (OSTI)

    1995-04-01

    This report for calendar year 1994 summarizes the progress that states and compact regions made during the year in establishing new low-level radioactive waste disposal facilities. Although events that have occurred in 1995 greatly alter the perspective in terms of storage versus disposal, the purpose of this report is to convey the concerns as evidenced during calendar year 1994. Significant developments occurring in 1995 are briefly outlined in the transmittal letter and will be detailed in the report for calendar year 1995. The report also provides summary information on the volume of low-level radioactive waste received for disposal in 1994 by commercially operated low-level radioactive waste disposal facilities, and is prepared is in response to Section 7(b) of Title I of Public Law 99-240, the Low-Level Radioactive Waste Policy Amendments Act of 1985.

  10. North Dakota State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    1981-10-01

    The North Dakota State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in North Dakota. The profile is the result of a survey of NRC licensees in North Dakota. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in North Dakota.

  11. Connecticut State Briefing Book for low-level radioactive-waste management

    SciTech Connect (OSTI)

    1981-06-01

    The Connecticut State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Connecticut. The profile is the result of a survey of Nuclear Regulatory Commission licensees in Connecticut. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may affect waste management practices in Connecticut.

  12. Greater-than-Class C Low-Level Radioactive Waste (GTCC LLW)

    Broader source: Energy.gov [DOE]

    In February 2016, DOE publicly issued the Final Environmental Impact Statement for the Disposal of Greater-Than-Class C (GTCC) Low-Level Radioactive Waste and GTCC-Like Waste (DOE/EIS-0375)(Final...

  13. Introduction to DOE Order 435.1 Low Level Radioactive Waste Disposal...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Introduction to DOE Order 435.1 Low Level Radioactive Waste Disposal Requirements Christine Gelles*, U.S. Department of Energy ; Edward Regnier, U.S. Department of Energy; Andrew ...

  14. Vermont State Briefing Book on low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-07-01

    The Vermont State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Vermont. The profile is the result of a survey of Nuclear Regulatory Commission licensees in Vermont. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may affect waste management practices in Vermont.

  15. Ohio State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-04-01

    The Ohio State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Ohio. The profile is the result of a survey of NRC licensees in Ohio. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Ohio.

  16. Tennessee State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-08-01

    The Tennessee State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Tennessee. The profile is the result of a survey of NRC licensees in Tennessee. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Tennessee.

  17. Massachusetts State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-03-12

    The Massachusetts State Briefing Book is one of a series of State briefing books on low-level radioactive waste management practices. It has been prepared to assist State and Federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Massachusetts. The profile is the result of a survey of NRC licensees in Massachusetts. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Massachusetts.

  18. Kentucky State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-08-01

    The Kentucky State Briefing Book is one of a series of State briefing books on low-level radioactive waste management practices. It has been prepared to assist State and Federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Kentucky. The profile is the result of a survey of NRC licensees in Kentucky. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Kentucky.

  19. North Carolina State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-08-01

    The North Carolina State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in North Carolina. The profile is the result of a survey of NRC licensees in North Carolina. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in North Carolina.

  20. Puerto Rico State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-10-01

    The Puerto Rico State Briefing Book is one of a series of state briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Puerto Rico. The profile is the result of a survey of NRC licensees in Puerto Rico. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Puerto Rico.

  1. New Jersey State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-04-01

    The New Jersey state Briefing Book is one of a series of State briefing books on low-level radioactive waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in New Jersey. The profile is the result of a survey of NRC licensees in New Jersey. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in New Jersey.

  2. Introduction to DOE Order 435.1 Low Level Radioactive Waste Disposal

    Energy Savers [EERE]

    Requirements | Department of Energy Introduction to DOE Order 435.1 Low Level Radioactive Waste Disposal Requirements Introduction to DOE Order 435.1 Low Level Radioactive Waste Disposal Requirements Christine Gelles*, U.S. Department of Energy ; Edward Regnier, U.S. Department of Energy; Andrew Wallo, U.S. Department of Energy Abstract: The Atomic Energy Act gives the U.S. Department of Energy (US DOE), the authority to regulate the management of radioactive waste generated by US DOE. This

  3. Working with SRNL - Our Facilities- High and Intermediate Level Shielded

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Cells High and Intermediate Level Shielded Cells Working with SRNL Our Facilities - High and Intermediate Level Shielded Cells The SRNL Shielded Cells Facility is a special containment facility that provide the shielding and confinement necessary for the safe examination, analysis and testing of highly radioactive materials. Skilled operators stand safely outside of each cell and use manipulator arms to control delicate procedures inside of each cell. The facility consists of sixteen 6x6

  4. Orex based {open_quotes}point of generation{close_quotes} low-level radioactive waste reduction program

    SciTech Connect (OSTI)

    Haynes, B.

    1995-11-01

    Nuclear power facilities, both commercial and government operated, generate material called Dry Active Waste (DAW). DAW is a by-product of maintenance and operation of the power systems which contain radioactive materials. DAW can be any material contaminated with radioactive particles as long as it is not a fluid, typically: paper, cardboard, wood, plastics, cloth, and any other solid which is contaminated and determined to be dry. DAW is generated when any material is exposed to loose radioactive particles and subsequently becomes contaminated. In the United States, once a material is contaminated it must be treated as radioactive waste and disposed of in accordance with the requirements of Title 10 of the Code of Federal Regulations. Problems facing all commercial and non-commercial nuclear facilities are escalating costs of processing DAW and volumetric reduction of the DAW generated. Currently, approximately 85% of all DAW generated at a typical facility is comprised of anti-contamination clothing and protective barrier materials. A typical 800 megawatt power station will generate between 6,000 to 10,000 cubic feet of DAW annually. Facilities that generate low-level radioactive waste need to dramatically reduce their waste volumes. This curtailment is required for several reasons: (1). The number of radioactive waste repositories now accepting new waste is limited. (2). The current cost of burial at an operating dump site is significant. Costs can be as high as $4,000 for a single 55 gallon drum. (3). The cost of burial is constantly increasing. (4). Onsite storage of low-level radioactive waste is costly and results in a burial fee at plant decommissioning. In order to address this issue, the industry must look to the application of {open_quotes}point of generation{close_quotes} technologies.

  5. Low-Level Radioactive Waste Management at the Nevada Test Site - Current Status

    SciTech Connect (OSTI)

    Bruce D. Becker, Bechtel Nevada; Bruce M. Crowe, Los Alamos National Laboratory; Carl P. Gertz, DOE Nevada Operations Office; Wendy A. Clayton, DOE Nevada Operations Office

    1999-02-01

    The performance objective of the Department of Energy's Low-Level Radioactive Waste disposal facility at the Nevada Test Site transcends those of any other radioactive waste disposal site in the United States. This paper describes the technical attributes of the facility, present and future capacities and capabilities, and provides a description of the process from waste approval to final disposition. The paper also summarizes the current status of the waste disposal operations.

  6. Low-Level Radioactive Waste Management at the Nevada Test Site - Year 2000 Current Status

    SciTech Connect (OSTI)

    Bruce D. Becker, Bechtel Nevada; Bruce M. Crowe, Los Alamos National Laboratory; Carl P. Gertz, DOE Nevada; Wendy A. Clayton, DOE Nevada

    1999-08-06

    The performance objectives of the Department of Energy's Low-level radioactive waste disposal facilities at the Nevada Test Site transcend those of any other radioactive waste disposal site in the United States. The expanded paper will describe the technical attributes of the facilities, the present and the future disposal capacities and capabilities, and includes a description of the process from waste approval to final disposition. The paper also summarizes the current status of the waste disposal operations.

  7. High pressure liquid level monitor

    DOE Patents [OSTI]

    Bean, Vern E.; Long, Frederick G.

    1984-01-01

    A liquid level monitor for tracking the level of a coal slurry in a high-pressure vessel including a toroidal-shaped float with magnetically permeable bands thereon disposed within the vessel, two pairs of magnetic field generators and detectors disposed outside the vessel adjacent the top and bottom thereof and magnetically coupled to the magnetically permeable bands on the float, and signal processing circuitry for combining signals from the top and bottom detectors for generating a monotonically increasing analog control signal which is a function of liquid level. The control signal may be utilized to operate high-pressure control valves associated with processes in which the high-pressure vessel is used.

  8. Mixed Low-Level Radioactive Waste (MLLW) Primer

    SciTech Connect (OSTI)

    W. E. Schwinkendorf

    1999-04-01

    This document presents a general overview of mixed low-level waste, including the regulatory definitions and drivers, the manner in which the various kinds of mixed waste are regulated, and a discussion of the waste treatment options.

  9. Texas State Briefing Book for low-level radioactive waste management

    SciTech Connect (OSTI)

    Not Available

    1981-08-01

    The Texas State Briefing Book is one of a series of state briefing books on low-level radioactivee waste management practices. It has been prepared to assist state and federal agency officials in planning for safe low-level radioactive waste disposal. The report contains a profile of low-level radioactive waste generators in Texas. The profile is the result of a survey of NRC licensees in Texas. The briefing book also contains a comprehensive assessment of low-level radioactive waste management issues and concerns as defined by all major interested parties including industry, government, the media, and interest groups. The assessment was developed through personal communications with representatives of interested parties, and through a review of media sources. Lastly, the briefing book provides demographic and socioeconomic data and a discussion of relevant government agencies and activities, all of which may impact waste management practices in Texas.

  10. Operational Strategies for Low-Level Radioactive Waste Disposal Site in Egypt - 13513

    SciTech Connect (OSTI)

    Mohamed, Yasser T.

    2013-07-01

    The ultimate aims of treatment and conditioning is to prepare waste for disposal by ensuring that the waste will meet the waste acceptance criteria of a disposal facility. Hence the purpose of low-level waste disposal is to isolate the waste from both people and the environment. The radioactive particles in low-level waste emit the same types of radiation that everyone receives from nature. Most low-level waste fades away to natural background levels of radioactivity in months or years. Virtually all of it diminishes to natural levels in less than 300 years. In Egypt, The Hot Laboratories and Waste Management Center has been established since 1983, as a waste management facility for LLW and ILW and the disposal site licensed for preoperational in 2005. The site accepts the low level waste generated on site and off site and unwanted radioactive sealed sources with half-life less than 30 years for disposal and all types of sources for interim storage prior to the final disposal. Operational requirements at the low-level (LLRW) disposal site are listed in the National Center for Nuclear Safety and Radiation Control NCNSRC guidelines. Additional procedures are listed in the Low-Level Radioactive Waste Disposal Facility Standards Manual. The following describes the current operations at the LLRW disposal site. (authors)

  11. Design and operational considerations of United States commercial near-surface low-level radioactive waste disposal facilities

    SciTech Connect (OSTI)

    Birk, S.M.

    1997-10-01

    In accordance with the Low-Level Radioactive Waste Policy Amendments Act of 1985, states are responsible for providing for disposal of commercially generated low-level radioactive waste (LLW) within their borders. LLW in the US is defined as all radioactive waste that is not classified as spent nuclear fuel, high-level radioactive waste, transuranic waste, or by-product material resulting from the extraction of uranium from ore. Commercial waste includes LLW generated by hospitals, universities, industry, pharmaceutical companies, and power utilities. LLW generated by the country`s defense operations is the responsibility of the Federal government and its agency, the Department of Energy. The commercial LLRW disposal sites discussed in this report are located near: Sheffield, Illinois (closed); Maxey Flats, Kentucky (closed); Beatty, Nevada (closed); West Valley, New York (closed); Barnwell, South Carolina (operating); Richland, Washington (operating); Ward Valley, California, (proposed); Sierra Blanca, Texas (proposed); Wake County, North Carolina (proposed); and Boyd County, Nebraska (proposed). While some comparisons between the sites described in this report are appropriate, this must be done with caution. In addition to differences in climate and geology between sites, LLW facilities in the past were not designed and operated to today`s standards. This report summarizes each site`s design and operational considerations for near-surface disposal of low-level radioactive waste. The report includes: a description of waste characteristics; design and operational features; post closure measures and plans; cost and duration of site characterization, construction, and operation; recent related R and D activities for LLW treatment and disposal; and the status of the LLW system in the US.

  12. FLUIDIZED BED STEAM REFORMING ENABLING ORGANIC HIGH LEVEL WASTE DISPOSAL

    SciTech Connect (OSTI)

    Williams, M

    2008-05-09

    Waste streams planned for generation by the Global Nuclear Energy Partnership (GNEP) and existing radioactive High Level Waste (HLW) streams containing organic compounds such as the Tank 48H waste stream at Savannah River Site have completed simulant and radioactive testing, respectfully, by Savannah River National Laboratory (SRNL). GNEP waste streams will include up to 53 wt% organic compounds and nitrates up to 56 wt%. Decomposition of high nitrate streams requires reducing conditions, e.g. provided by organic additives such as sugar or coal, to reduce NOX in the off-gas to N2 to meet Clean Air Act (CAA) standards during processing. Thus, organics will be present during the waste form stabilization process regardless of the GNEP processes utilized and exists in some of the high level radioactive waste tanks at Savannah River Site and Hanford Tank Farms, e.g. organics in the feed or organics used for nitrate destruction. Waste streams containing high organic concentrations cannot be stabilized with the existing HLW Best Developed Available Technology (BDAT) which is HLW vitrification (HLVIT) unless the organics are removed by pretreatment. The alternative waste stabilization pretreatment process of Fluidized Bed Steam Reforming (FBSR) operates at moderate temperatures (650-750 C) compared to vitrification (1150-1300 C). The FBSR process has been demonstrated on GNEP simulated waste and radioactive waste containing high organics from Tank 48H to convert organics to CAA compliant gases, create no secondary liquid waste streams and create a stable mineral waste form.

  13. Low-Level Radioactive Waste Management in the United States: What Have We Wrought? The Richard S. Hodes, M.D. Honor Lecture Award - 12222

    SciTech Connect (OSTI)

    Jacobi, Lawrence R.

    2012-07-01

    In 1979, radioactive waste disposal was an important national issue. State governors were closing the gates on the existing low-level radioactive waste disposal sites and the ultimate disposition of spent fuel was undecided. A few years later, the United States Congress thought they had solved both problems by passing the Low-Level Radioactive Waste Policy Act of 1981, which established a network of regional compacts for low-level radioactive waste disposal, and by passing the Nuclear Waste Policy Act of 1982 to set out how a final resting place for high-level waste would be determined. Upon passage of the acts, State, Regional and Federal officials went to work. Here we are some 30 years later with little to show for our combined effort. The envisioned national repository for high-level radioactive waste has not materialized. Efforts to develop the Yucca Mountain high-level radioactive waste disposal facility were abandoned after spending $13 billion on the failed project. Recently, the Blue Ribbon Commission on America's Nuclear Future issued its draft report that correctly concludes the existing policy toward high-level nuclear waste is 'all but completely broken down'. A couple of new low-level waste disposal facilities have opened since 1981, but neither were the result of efforts under the act. What the Act has done is interject a system of interstate compacts with a byzantine interstate import and export system to complicate the handling of low-level radioactive waste, with attendant costs. As this paper is being written in the fourth-quarter of 2011, after 30 years of political and bureaucratic turmoil, a new comprehensive low-level waste disposal facility at Andrews Texas is approaching its initial operating date. The Yucca Mountain project might be completed or it might not. The US Nuclear Regulatory Commission is commencing a review of their 1981 volume reduction policy statement. The Department of Energy after 26 years has yet to figure out how to implement its obligations under the 1985 amendments to the Low-Level Radioactive Waste Policy Act. But, the last three decades have not been a total loss. A great deal has been learned about radioactive waste disposal since 1979 and the efforts of the public and private sector have shaped and focused the work to be done in the future. So, this lecturer asks the question: 'What have we wrought?' to which he provides his perspective and his recommendations for radioactive waste management policy for the next 30 years. (author)

  14. Gas generation from low-level radioactive waste: Concerns for disposal

    SciTech Connect (OSTI)

    Siskind, B.

    1992-01-01

    The Advisory Committee on Nuclear Waste (ACNW) has urged the Nuclear Regulatory Commission (NRC) to reexamine the topic of hydrogen gas generation from low-level radioactive waste (LLW) in closed spaces to ensure that the slow buildup of hydrogen from water-bearing wastes in sealed containers does not become a problem for long-term safe disposal. Brookhaven National Laboratory (BNL) has prepared a report, summarized in this paper, for the NRC to respond to these concerns. The paper discusses the range of values for G(H{sub 2}) reported for materials of relevance to LLW disposal; most of these values are in the range of 0.1 to 0.6. Most studies of radiolytic hydrogen generation indicate a leveling off of pressurization, probably because of chemical kinetics involving, in many cases, the radiolysis of water within the waste. Even if no leveling off occurs, realistic gas leakage rates (indicating poor closure by gaskets on drums and liners) will result in adequate relief of pressure for radiolytic gas generation from the majority of commercial sector LLW packages. Biodegradative gas generation, however, could pose a pressurization hazard even at realistic gas leakage rates. Recommendations include passive vents on LLW containers (as already specified for high integrity containers) and upper limits to the G values and/or the specific activity of the LLW.

  15. Gas generation from low-level radioactive waste: Concerns for disposal

    SciTech Connect (OSTI)

    Siskind, B.

    1992-04-01

    The Advisory Committee on Nuclear Waste (ACNW) has urged the Nuclear Regulatory Commission (NRC) to reexamine the topic of hydrogen gas generation from low-level radioactive waste (LLW) in closed spaces to ensure that the slow buildup of hydrogen from water-bearing wastes in sealed containers does not become a problem for long-term safe disposal. Brookhaven National Laboratory (BNL) has prepared a report, summarized in this paper, for the NRC to respond to these concerns. The paper discusses the range of values for G(H{sub 2}) reported for materials of relevance to LLW disposal; most of these values are in the range of 0.1 to 0.6. Most studies of radiolytic hydrogen generation indicate a leveling off of pressurization, probably because of chemical kinetics involving, in many cases, the radiolysis of water within the waste. Even if no leveling off occurs, realistic gas leakage rates (indicating poor closure by gaskets on drums and liners) will result in adequate relief of pressure for radiolytic gas generation from the majority of commercial sector LLW packages. Biodegradative gas generation, however, could pose a pressurization hazard even at realistic gas leakage rates. Recommendations include passive vents on LLW containers (as already specified for high integrity containers) and upper limits to the G values and/or the specific activity of the LLW.

  16. Potential for Subsidence at the Low-Level Radioactive Waste Disposal Area

    SciTech Connect (OSTI)

    Keck, K.A.; Seitz, R.R.

    2002-09-26

    U.S. Department of Energy (DOE) Order 435.1, Radioactive Waste Management requires that DOE low-level radioactive waste (LLW) disposal facilities receive a Disposal Authorization Statement (DAS) from DOE-Headquarters. The DAS for the LLW disposal facility at the Radioactive Waste Management Complex (RWMC) at the Idaho National Engineering and Environmental Laboratory (INEEL) was granted in April 2000 and included a number of conditions that must be addressed. A maintenance plan (Schuman 2000) was prepared that identifies the tasks to be completed to address the conditions in the DAS as well as a schedule for their completion. The need for a subsidence analysis was one of the conditions identified for the DAS, and thus, a task to prepare a subsidence analysis was included in the maintenance plan. This document provides the information necessary to satisfy that requirement.

  17. Summary report. Low-level radioactive waste management activities in the states and compacts. Volume 4, No. 2

    SciTech Connect (OSTI)

    NONE

    1996-08-01

    `Low-Level Radioactive Waste Management Activities in the States and Compacts` is a supplement to `LLW Notes` and is distributed periodically by Afton Associates, Inc. to state, compact and federal officials that receive `LLW Notes`. The Low-Level Radioactive Waste Forum (LLW Forum) is an association of state and compact representatives, appointed by governors and compact commissions, established to facilitate state and compact implementation of the Low- Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The LLW Forum provides an opportunity for state and compact officials to share information with one another and to exchange views with officials of federal agencies and other interested parties.

  18. Summary report, low-level radioactive waste management activities in the states and compacts. Vol. 4. No. 1

    SciTech Connect (OSTI)

    NONE

    1996-01-01

    `Low-Level Radioactive Waste Management Activities in the States and Compacts` is a supplement to `LLW Notes` and is distributed periodically by Afton Associates, Inc. to state, compact and federal officials that receive `LLW Notes`. The Low-Level Radioactive Waste Forum (LLW Forum) is an association of state and compact representatives, appointed by governors and compact commissions, established to facilitate state and compact implementation of the Low- Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive Waste Policy Amendments Act of 1985 and to promote the objectives of low-level radioactive waste regional compacts. The LLW Forum provides an opportunity for state and compact officials to share information with one another and to exchange views with officials of federal agencies and other interested parties.

  19. Environmental assessment for the treatment of Class A low-level radioactive waste and mixed low-level waste generated by the West Valley Demonstration Project

    SciTech Connect (OSTI)

    NONE

    1995-11-01

    The U.S. Department of Energy (DOE) is currently evaluating low-level radioactive waste management alternatives at the West Valley Demonstration Project (WVDP) located on the Western New York Nuclear Service Center (WNYNSC) near West Valley, New York. The WVDP`s mission is to vitrify high-level radioactive waste resulting from commercial fuel reprocessing operations that took place at the WNYNSC from 1966 to 1972. During the process of high-level waste vitrification, low-level radioactive waste (LLW) and mixed low-level waste (MILLW) will result and must be properly managed. It is estimated that the WVDP`s LLW storage facilities will be filled to capacity in 1996. In order to provide sufficient safe storage of LLW until disposal options become available and partially fulfill requirements under the Federal Facilities Compliance Act (FFCA), the DOE is proposing to use U.S. Nuclear Regulatory Commission-licensed and permitted commercial facilities in Oak Ridge, Tennessee; Clive, Utah; and Houston, Texas to treat (volume-reduce) a limited amount of Class A LLW and MLLW generated from the WVDP. Alternatives for ultimate disposal of the West Valley LLW are currently being evaluated in an environmental impact statement. This proposed action is for a limited quantity of waste, over a limited period of time, and for treatment only; this proposal does not include disposal. The proposed action consists of sorting, repacking, and loading waste at the WVDP; transporting the waste for commercial treatment; and returning the residual waste to the WVDP for interim storage. For the purposes of this assessment, environmental impacts were quantified for a five-year operating period (1996 - 2001). Alternatives to the proposed action include no action, construction of additional on-site storage facilities, construction of a treatment facility at the WVDP comparable to commercial treatment, and off-site disposal at a commercial or DOE facility.

  20. Low Level Radioactive Wastes Conditioning during Decommissioning of Salaspils Research Reactor

    SciTech Connect (OSTI)

    Abramenkova, G.; Klavins, M.; Abramenkovs, A.

    2008-01-15

    The decommissioning of Salaspils research reactor is connected with the treatment of 2200 tons different materials. The largest part of all materials ({approx}60 % of all dismantled materials) is connected with low level radioactive wastes conditioning activities. Dismantled radioactive materials were cemented in concrete containers using water-cement mortar. According to elaborated technology, the tritiated water (150 tons of liquid wastes from special canalization tanks) was used for preparation of water-cement mortar. Such approach excludes the emissions of tritiated water into environment and increases the efficiency of radioactive wastes management system for decommissioning of Salaspils research reactor. The Environmental Impact Assessment studies for Salaspils research reactor decommissioning (2004) and for upgrade of repository 'Radons' for decommissioning purposes (2005) induced the investigations of radionuclides release parameters from cemented radioactive waste packages. These data were necessary for implementation of quality assurance demands during conditioning of radioactive wastes and for safety assessment modeling for institutional control period during 300 years. Experimental studies indicated, that during solidification of water- cement samples proceeds the increase of temperature up to 81 deg. C. It is unpleasant phenomena since it can result in damage of concrete container due to expansion differences for mortar and concrete walls. Another unpleasant factor is connected with the formation of bubbles and cavities in the mortar structure which can reduce the mechanical stability of samples and increase the release of radionuclides from solidified cement matrix. The several additives, fly ash and PENETRON were used for decrease of solidification temperature. It was found, that addition of fly ash to the cement-water mortar can reduce the solidification temperature up to 62 deg. C. Addition of PENETRON results in increasing of solidification temperature up to 83 deg. C. Experimental data shows, that water/cement ratio significantly influences on water-cement mortar's viscosity and solidified samples mechanical stability. Increasing of water ratio from 0.45 up to 0.65 decreases water-cement mortar's viscosity from 1100 mPas up to 90 mPas. Significant reduction of viscosity is an important factor, which facilitates the fulfillment all gaps and cavities with the mortar during conditioning of solid radioactive wastes in containers. On the other hand, increase water ratio from 0.45 up to 0.65 decreases mechanical stability of water-cement samples from 23 N/mm{sup 2} to the 12 N/mm{sup 2}. It means that water-cement bulk stability significantly decreases with increasing of water content. Technologically is important to increase the tritiated water content in container with cemented radioactive wastes. It gives a possibility to increase the fulfillment of container with radioactive materials. On the other hand, additional water significantly reduces bulk stability of containers with cemented radioactive wastes, which can result in disintegration of radioactive wastes packages in repository during 300 years. Taking into account the experimental results, it is not recommended to exceed the water/cement ratio more than 0.60. Tritium and Cs{sup 137} leakage tests show, that radionuclides release curves has a complicate structure. Experimental results indicated that addition of fly ash result in facilitation of tritium and cesium release in water phase. This is unpleasant factor, which significantly decreases the safety of disposed radioactive wastes. Despite the positive impact on solidification temperature drop, the addition of fly ash to the cement-water mortar is not recommended in case of cementation of radionuclides in concrete containers. In conclusion: The cementation processes of solid radioactive wastes in concrete containers were investigated. The influence of additives on cementation processes was studied. It was shown, that the increasing of water ratio from 0.45 up to 0.65 decreases water-cement mortar

  1. High level white noise generator

    DOE Patents [OSTI]

    Borkowski, Casimer J.; Blalock, Theron V.

    1979-01-01

    A wide band, stable, random noise source with a high and well-defined output power spectral density is provided which may be used for accurate calibration of Johnson Noise Power Thermometers (JNPT) and other applications requiring a stable, wide band, well-defined noise power spectral density. The noise source is based on the fact that the open-circuit thermal noise voltage of a feedback resistor, connecting the output to the input of a special inverting amplifier, is available at the amplifier output from an equivalent low output impedance caused by the feedback mechanism. The noise power spectral density level at the noise source output is equivalent to the density of the open-circuit thermal noise or a 100 ohm resistor at a temperature of approximately 64,000 Kelvins. The noise source has an output power spectral density that is flat to within 0.1% (0.0043 db) in the frequency range of from 1 KHz to 100 KHz which brackets typical passbands of the signal-processing channels of JNPT's. Two embodiments, one of higher accuracy that is suitable for use as a standards instrument and another that is particularly adapted for ambient temperature operation, are illustrated in this application.

  2. Extended storage of low-level radioactive waste: potential problem areas

    SciTech Connect (OSTI)

    Siskind, B.; Dougherty, D.R.; MacKenzie, D.R.

    1985-01-01

    If a state or state compact does not have adequate disposal capacity for low-level radioactive waste (LLRW) by 1986 as required by the Low-Level Waste Policy Act, then extended storage of certain LLRW may be necessary. The issue of extended storage of LLRW is addressed in order to determine for the Nuclear Regulatory Commission the areas of concern and the actions recommended to resolve these concerns. The focus is on the properties and behavior of the waste form and waste container. Storage alternatives are considered in order to characterize the likely storage environments for these wastes. The areas of concern about extended storage of LLRW are grouped into two categories: 1. Behavior of the waste form and/or container during storage, e.g., radiolytic gas generation, radiation-enhanced degradation of polymeric materials, and corrosion. 2. Effects of extended storage on the properties of the waste form and/or container that are important after storage (e.g., radiation-induced oxidative embrittlement of high-density polyethylene and the weakening of steel containers resulting from corrosion by the waste). The additional information and actions required to address these concerns are discussed and, in particular, it is concluded that further information is needed on the rates of corrosion of container material by Class A wastes and on the apparent dose-rate dependence of radiolytic processes in Class B and C waste packages. Modifications to the guidance for solidified wastes and high-integrity containers in NRC's Technical Position on Waste Form are recommended. 27 references.

  3. Taiwan industrial cooperation program technology transfer for low-level radioactive waste final disposal - phase I.

    SciTech Connect (OSTI)

    Knowlton, Robert G.; Cochran, John Russell; Arnold, Bill Walter; Jow, Hong-Nian; Mattie, Patrick D.; Schelling, Frank Joseph Jr.

    2007-01-01

    Sandia National Laboratories and the Institute of Nuclear Energy Research, Taiwan have collaborated in a technology transfer program related to low-level radioactive waste (LLW) disposal in Taiwan. Phase I of this program included regulatory analysis of LLW final disposal, development of LLW disposal performance assessment capabilities, and preliminary performance assessments of two potential disposal sites. Performance objectives were based on regulations in Taiwan and comparisons to those in the United States. Probabilistic performance assessment models were constructed based on limited site data using software including GoldSim, BLT-MS, FEHM, and HELP. These software codes provided the probabilistic framework, container degradation, waste-form leaching, groundwater flow, radionuclide transport, and cover infiltration simulation capabilities in the performance assessment. Preliminary performance assessment analyses were conducted for a near-surface disposal system and a mined cavern disposal system at two representative sites in Taiwan. Results of example calculations indicate peak simulated concentrations to a receptor within a few hundred years of LLW disposal, primarily from highly soluble, non-sorbing radionuclides.

  4. LOW-LEVEL RADIOACTIVE & MIXED WASTE SHIPMENTS TO THE NEVADA NATIONAL SECURITY SITE

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    NV NY NY NV OH TN TN TN, WA, CA TN TN TN TX y ct 2 Plant Perma-Fix LOW-LEVEL RADIOACTIVE & MIXED WASTE SHIPMENTS TO THE NEVADA NATIONAL SECURITY SITE FIRST QUARTER REPORT, FY 2011 (OCTOBER, NOVEMBER, DECEMBER 2010) Livermore National Laboratory ergy Alliance onal Laboratory Mixed Waste Treatment Project gonne National Laboratory ational Laboratory Gaseous Diffusion Plant Proving Ground terra, LLC en National Laboratory y Environmental Services ecurity Technologies, Inc. h Gaseous Diffusion

  5. Immobilization of sodium-bearing high-level radioactive waste in synroc containing (Na{sub 0.5}Nd{sub 0.5})TiO{sub 3}-type perovskite

    SciTech Connect (OSTI)

    Li, L.; Luo, S.; Tang, B.; Wang, D.

    1997-01-01

    This study on the immobilization of high-sodium-bearing HLW in synroc indicates that (Na{sub 0.5}Nd{sub 0.5})TiO{sub 3}-type perovskite can be used to incorporate a high content of sodium in synroc. Synroc samples containing 13.0 wt% waste oxide and 5.7 wt% Na{sub 2}O show very well chemical durability and physical properties. The standard Synroc-C formulation can incorporate only 2 wt% Na{sub 2}O, so this study greatly improved the immobilization ability of sodium in Synroc-related material.

  6. Identification of radioactive mixed wastes in commercial low-level wastes

    SciTech Connect (OSTI)

    Bowerman, B.S.; Kempf, C.R.; MacKenzie, D.R.; Siskind, B.; Piciulo, P.L.

    1985-01-01

    A literature review and survey were conducted on behalf of the US NRC Division of Waste Management to determine whether any commercial low-level radioactive wastes (LLW) could be considered hazardous as defined by EPA under 40 CFR Part 261. The purpose of the study was to identify broad categories of LLW which may require special management as radioactive mixed waste, and to help address uncertainties regarding the regulation of such wastes. Of 239 questionnaires sent out to reactor and non-reactor LLW generators, there were 91 responses representing 29% by volume of all low-level wastes disposed of at commercial disposal sites in 1984. The analysis of the survey results indicated that three waste streams generic to commercial LLW may be potential radioactive mixed wastes. These are as follows: (1) wastes containing organic liquids, disposed of by all types of generators and representing approx. =2.3% by volume of all wastes reported; (2) wastes containing lead metal, i.e., discarded shielding and lead containers, representing <0.1% by volume of all wastes reported; and (3) wastes containing chromium, i.e., process wastes from nuclear power plants which use chromates as corrosion inhibitors; these represent 0.6% of the total volume reported in the survey. Certain wastes, specific to particular generators, were identified as potential mixed wastes as well. 4 refs., 5 tabs.

  7. Identification of radioactive mixed wastes in commercial low-level wastes

    SciTech Connect (OSTI)

    Bowerman, B.S.; Kempf, C.R.; MacKenzie, D.R.; Siskind, B.; Piciulo, P.L.

    1986-01-01

    A literature review and survey were conducted on behalf of the US NRC Division of Waste Management to determine whether any commercial low-level radioactive wastes (LLW) could be considered hazardous as defined by EPA under 40 CFR Part 261. The purpose of the study was to identify broad categories of LLW which may require special management as radioactive mixed waste, and to help address uncertainties regarding the regulation of such wastes. Of 239 questionnaires sent out to reactor and non-reactor LLW generators, there were 91 responses representing 29% by volume of all low-level wastes disposed of at commercial disposal sites in 1984. The analysis of the survey results indicated that the following waste types generic to commercial LLW may be potential radioactive mixed wastes: Wastes containing oil, disposed of by reactors and industrial facilities, and representing 4.2% of the total LLW volume reported in the survey. Wastes containing organic liquids, disposed of by all types of generators, and representing 2.3% by volume of all wastes reported. Wastes containing lead metal, i.e., discarded shielding and lead containers, representing <0.1% by volume of all wastes reported. Wastes containing chromium, i.e., process wastes from nuclear power plants which use chromates as corrosion inhibitors; these represent 0.6% of the total volume reported in the survey. Certain wastes, specific to particular generators, were identified as potential mixed wastes as well.

  8. New York State Low-Level Radioactive Waste Status Report for 1992

    SciTech Connect (OSTI)

    Attridge, T.; Rapaport, S.; Yang, Qian

    1993-06-01

    This report summarizes data on low-level radioactive waste (LLRW) generation in New York State for calendar year 1992. It is based on reports from generators that must be filed annually with the New York State Energy Research and Development Authority (Energy Authority) and on data from the US Department of Energy. The New York State Low-Level Radioactive Waste Management Act (State Act) requires LLRW generators in the State to submit annual reports detailing the classes and quantities of waste generated. This is the seventh year generators have been required to submit reports on their waste to the Energy Authority. The data are summarized in a series of tables and figures. There are three sections in the report. Section 1 covers volume, radioactivity and other characteristics of waste generated in 1992. Section 2 shows historical LLRW generation over the years and includes generators` projections for the next five years. Section 3 provides a list of all facilities for which 1992 LLRW reports were received.

  9. New York State low-level radioactive waste status report for 1998

    SciTech Connect (OSTI)

    Voelk, H.

    1999-06-01

    This report summarizes data on low-level radioactive waste (LLRW) generated in New York State: it is based on reports from generators that must be filed annually with the New York State Energy Research and Development Authority (NYSERDA) and on data from the US Department of Energy (US DOE). The New York State Low-Level Radioactive Waste Management Act (State Act) requires LLRW generators in the State to submit annual reports detailing the classes and quantities of waste generated. This is the 13th year generators have been required to submit these reports to NYSERDA. The data are summarized in a series of tables and figures. There are four sections in the report. Section 1 covers volume, activity, and other characteristics of waste shipped for disposal in 1998. Activity is the measure of a material`s radioactivity, or the number of radiation-emitting events occurring each second. Section 2 summarizes volume, activity, and other characteristics of waste held for storage as of December 31, 1998. Section 3 shows historical LLRW generation and includes generators` projections for the next five years. Section 4 provides a list, by county, of all facilities from which 1998 LLRW reports were received. 2 figs., 23 tabs.

  10. Improvement to low-level radioactive-waste vitrification processes. Master's thesis

    SciTech Connect (OSTI)

    Horton, W.S.

    1986-05-01

    Low-level radioactive waste vitrification (LLWV) is a technically feasible and cost-competitive alternative to the traditional immobilization options, i.e., cementation or bituminization. This thesis analyzes cementation, bituminization and vitrification, reviews the impact of the low-level Waste-stream composition on the vitrification process, then proposes and discusses several techniques to control the volatile radionuclides in a Process Improved LLWV system (PILLWV). The techniques that control the volatile radionuclides include chemical precipitation, electrodialysis, and ion exchange. Ion exchange is preferred. A comparison of the technical specifications, of the regulatory compliance, and of the cost considerations shows the PILLWV to be the superior LLW immobilization option.

  11. Letter report: Minor component study for low-level radioactive waste glasses

    SciTech Connect (OSTI)

    Li, H.

    1996-03-01

    During the waste vitrification process, troublesome minor components in low-level radioactive waste streams could adversely affect either waste vitrification rate or melter life-time. Knowing the solubility limits for these minor components is important to determine pretreatment options for waste streams and glass formulation to prevent or to minimize these problems during the waste vitrification. A joint study between Pacific Northwest Laboratory and Rensselaer Polytechnic Institute has been conducted to determine minor component impacts in low-level nuclear waste glass.

  12. Radioactive Waste Management Complex low-level waste radiological performance assessment

    SciTech Connect (OSTI)

    Maheras, S.J.; Rood, A.S.; Magnuson, S.O.; Sussman, M.E.; Bhatt, R.N.

    1994-04-01

    This report documents the projected radiological dose impacts associated with the disposal of radioactive low-level waste at the Radioactive Waste Management Complex at the Idaho National Engineering Laboratory. This radiological performance assessment was conducted to evaluate compliance with applicable radiological criteria of the US Department of Energy and the US Environmental Protection Agency for protection of the public and the environment. The calculations involved modeling the transport of radionuclides from buried waste, to surface soil and subsurface media, and eventually to members of the public via air, groundwater, and food chain pathways. Projections of doses were made for both offsite receptors and individuals inadvertently intruding onto the site after closure. In addition, uncertainty and sensitivity analyses were performed. The results of the analyses indicate compliance with established radiological criteria and provide reasonable assurance that public health and safety will be protected.

  13. New York State low-level radioactive waste status report for 1997

    SciTech Connect (OSTI)

    1998-06-01

    This report summarizes data on low-level radioactive waste (LLRW) generated in New York State. It is based on reports from generators that must be filed annually with the New York State Energy Research and Development Authority (NYSERDA) and on data from the US Department of Energy (US DOE). The data are summarized in a series of tables and figures. There are four sections in this report. Section 1 covers volume, activity, and other characteristics of waste shipped for disposal in 1997. (Activity is the measure of a material`s radioactivity, or the number of radiation-emitting events occurring each second.) Section 2 summarizes volume, activity, and other characteristics of waste held for storage as of December 31, 1997. Section 3 shows historical LLRW generation and includes generators` projections for the next five years. Section 4 provides a list, by county, of all facilities from which 1997 LLRW reports were received.

  14. Commercial disposal options for Idaho National Engineering Laboratory low-level radioactive waste

    SciTech Connect (OSTI)

    Porter, C.L.; Widmayer, D.A.

    1995-09-01

    The Idaho National Engineering Laboratory (INEL) is a Department of Energy (DOE)-owned, contractor-operated site. Significant quantities of low-level radioactive waste (LLW) have been generated and disposed of onsite at the Radioactive Waste Management Complex (RWMC). The INEL expects to continue generating LLW while performing its mission and as aging facilities are decommissioned. An on-going Performance Assessment process for the RWMC underscores the potential for reduced or limited LLW disposal capacity at the existing onsite facility. In order to properly manage the anticipated amount of LLW, the INEL is investigating various disposal options. These options include building a new facility, disposing the LLW at other DOE sites, using commercial disposal facilities, or seeking a combination of options. This evaluation reports on the feasibility of using commercial disposal facilities.

  15. High temperature liquid level sensor

    DOE Patents [OSTI]

    Tokarz, Richard D. (West Richland, WA)

    1983-01-01

    A length of metal sheathed metal oxide cable is perforated to permit liquid access to the insulation about a pair of conductors spaced close to one another. Changes in resistance across the conductors will be a function of liquid level, since the wetted insulation will have greater electrical conductivity than that of the dry insulation above the liquid elevation.

  16. Low-level radioactive-waste compacts. Status report as of July 1982

    SciTech Connect (OSTI)

    Not Available

    1982-07-01

    The Low-Level Radioactive Waste Policy Act (P.L. 96-573), enacted in December 1980, established as federal policy that states take responsibility for providing disposal capacity for low-level radioactive waste (LLW) generated within their borders, except for defense waste and Federal R and D. At the request of Senator James A. McClure, Chairman of the Senate Committee on Energy and Natural Resources, DOE has documented the progress of states individually and collectively in fulfilling their responsibilities under the Public Law. Regionalization through formation of low-level waste compacts has been the primary vehicle by which many states are assuming this responsibility. To date seven low-level waste compacts have been drafted and six have been enacted by state legislatures or ratified by a governor. As indicated by national progress to date, DOE considers the task of compacting achievable by the January 1, 1986, exclusionary date set in law, although several states and NRC questioned this.

  17. High Level Waste System Plan Revision 9

    SciTech Connect (OSTI)

    Davis, N.R.; Wells, M.N.; Choi, A.S.; Paul, P.; Wise, F.E.

    1998-04-01

    Revision 9 of the High Level Waste System Plan documents the current operating strategy of the HLW System at SRS to receive, store, treat, and dispose of high-level waste.

  18. 1992 annual report on low-level radioactive waste management progress; Report to Congress in response to Public Law 99-240

    SciTech Connect (OSTI)

    1993-11-01

    This report summarizes the progress States and compact regions made during 1992 in establishing new low-level radioactive waste disposal facilities. It also provides summary information on the volume of low-level radioactive waste received for disposal in 1992 by commercially operated low-level radioactive waste disposal facilities. This report is in response to section 7 (b) of the Low-Level Radioactive Waste Policy Act.

  19. High-level waste tank farm set point document

    SciTech Connect (OSTI)

    Anthony, J.A. III

    1995-01-15

    Setpoints for nuclear safety-related instrumentation are required for actions determined by the design authorization basis. Minimum requirements need to be established for assuring that setpoints are established and held within specified limits. This document establishes the controlling methodology for changing setpoints of all classifications. The instrumentation under consideration involve the transfer, storage, and volume reduction of radioactive liquid waste in the F- and H-Area High-Level Radioactive Waste Tank Farms. The setpoint document will encompass the PROCESS AREA listed in the Safety Analysis Report (SAR) (DPSTSA-200-10 Sup 18) which includes the diversion box HDB-8 facility. In addition to the PROCESS AREAS listed in the SAR, Building 299-H and the Effluent Transfer Facility (ETF) are also included in the scope.

  20. New York State`s regulations for low-level radioactive waste disposal facilities

    SciTech Connect (OSTI)

    Youngberg, B.; Merges, P.; Owen, K.

    1994-12-31

    The New York State Department of Environmental Conservation`s (NYSDEC) regulations for low-level radioactive waste (LLRW) disposal facilities set primarily performance-based criteria for LLRW disposal facilities. The regulations (Part 383 of Title 6 of the New York State Codes of Rules and Regulations) set requirements for design, construction, operation, monitoring, site safety planning, financial assurance, closure, post closure monitoring and maintenance, and institutional control. The regulations are unique in their detail and in presenting specific requirements for below ground disposal units, above ground disposal units, and underground mined repositories.

  1. Identifying industrial best practices for the waste minimization of low-level radioactive materials

    SciTech Connect (OSTI)

    Levin, V.

    1996-04-01

    In US DOE, changing circumstances are affecting the management and disposal of solid, low-level radioactive waste (LLW). From 1977 to 1991, the nuclear power industry achieved major reductions in solid waste disposal, and DOE is interested in applying those practices to reduce solid waste at DOE facilities. Project focus was to identify and document commercial nuclear industry best practices for radiological control programs supporting routine operations, outages, and decontamination and decommissioning activities. The project team (DOE facility and nuclear power industry representatives) defined a Work Control Process Model, collected nuclear power industry Best Practices, and made recommendations to minimize LLW at DOE facilities.

  2. Northeast High-Level Radioactive Waste Transportation Task Force Agenda |

    Office of Environmental Management (EM)

    Synchrophasor Technology and Renewables Integration | Department of Energy Synchrophasor Technology and Renewables Integration North American SynchroPhasor Initiative (NASPI) Technical Report - Synchrophasor Technology and Renewables Integration This technical report was developed in June 2012 by the North American SynchroPhasor Initiative, a collaboration between the North American electric industry (utilities, grid operators, vendors and consultants), the North American Electric

  3. Northeast High-Level Radioactive Waste Transportation Task Force Agenda

    Office of Environmental Management (EM)

    Department of Energy Normal Conditions of Transport Truck Test of a Surrogate Fuel Assembly Normal Conditions of Transport Truck Test of a Surrogate Fuel Assembly This report describes a test of an instrumented surrogate PWR fuel assembly on a truck trailer conducted to simulate normal conditions of truck transport. The purpose of the test was to measure strains and accelerations on a Zircaloy-4 fuel rod during the transport of the assembly on the truck. This test complements tests conducted

  4. Stakeholder Engagement on the Environmental Impact Statement for the Disposal of Greater-Than-Class C Low-Level Radioactive Waste -12565

    SciTech Connect (OSTI)

    Gelles, Christine; Joyce, James; Edelman, Arnold

    2012-07-01

    The Department of Energy's (DOE) Office of Disposal Operations is responsible for developing a permanent disposal capability for a small volume, but highly radioactive, class of commercial low-level radioactive waste, known as Greater-Than-Class C (GTCC) low-level radioactive waste. DOE has issued a draft environmental impact statement (EIS) and will be completing a final EIS under the National Environmental Policy Act (NEPA) that evaluates a range of disposal alternatives. Like other classes of radioactive waste, proposing and evaluating disposal options for GTCC waste is highly controversial, presents local and national impacts, and generates passionate views from stakeholders. Recent national and international events, such as the cancellation of the Yucca Mountain project and the Fukushima Daiichi nuclear accident, have heighten stakeholder awareness of everything nuclear, including disposal of radioactive waste. With these challenges, the Office of Disposal Operations recognizes that informed decision-making that will result from stakeholder engagement and participation is critical to the success of the GTCC EIS project. This paper discusses the approach used by the Office of Disposal Operations to engage stakeholders on the GTCC EIS project, provides advice based on our experiences, and proffers some ideas for future engagements in today's open, always connected cyber environment. (authors)

  5. Melton Valley liquid low-level radioactive waste storage tanks evaluation

    SciTech Connect (OSTI)

    1995-06-01

    The Melton Valley Liquid Low-Level Radioactive Waste Storage Tanks (MVSTs) store the evaporator concentrates from the Liquid Low-Level Radioactive Waste (LLLW) System at the Oak Ridge National Laboratory (ORNL). The eight stainless steel tanks contain approximately 375,000 gallons of liquid and sludge waste. These are some of the newer, better-designed tanks in the LLLW System. They have been evaluated and found by the US Environmental Protection Agency (EPA) and the Tennessee Department of Environment and Conservation to comply with all Federal Facility Agreement requirements for double containment. The operations and maintenance aspects of the tanks were also reviewed by the Defense Nuclear Facilities Safety Board (DNFSB) in September 1994. This document also contains an assessment of the risk to the public and ORNL workers from a leak in one of the MVSTs. Two primary scenarios were investigated: (1) exposure of the public to radiation from drinking Clinch River water contaminated by leaked LLLW, and (2) exposure of on-site workers to radiation by inhaling air contaminated by leaked LLLW. The estimated frequency of a leak from one of the MVSTs is about 8 {times} 10{sup {minus}4} events per year, or about once in 1200 years (with a 95% confidence level). If a leak were to occur, the dose to a worker from inhalation would be about 2.3 {times} 10{sup {minus}1} mrem (with a 95% confidence level). The dose to a member of the public through the drinking water pathway is estimated to be about 7 {times} 10{sup {minus}1} mrem (with a 95% confidence level). By comparison with EPA Safe Drinking Water regulations, the allowable lifetime radiation dose is about 300 mrem. Thus, a postulated LLLW leak from the MVSTs would not add appreciably to an individual`s lifetime radiation dose.

  6. Assessment of microbial processes on gas production at radioactive low-level waste disposal sites

    SciTech Connect (OSTI)

    Weiss, A.J.; Tate, R.L. III; Colombo, P.

    1982-05-01

    Factors controlling gaseous emanations from low level radioactive waste disposal sites are assessed. Importance of gaseous fluxes of methane, carbon dioxide, and possible hydrogen from the site, stems from the inclusion of tritium and/or carbon-14 into the elemental composition of these compounds. In that the primary source of these gases is the biodegradation of organic components of the waste material, primary emphasis of the study involved an examination of the biochemical pathways producing methane, carbon dioxide, and hydrogen, and the environmental parameters controlling the activity of the microbial community involved. Initial examination of the data indicates that the ecosystem is anaerobic. As the result of the complexity of the pathway leading to methane production, factors such as substrate availability, which limit the initial reaction in the sequence, greatly affect the overall rate of methane evolution. Biochemical transformations of methane, hydrogen and carbon dioxide as they pass through the soil profile above the trench are discussed. Results of gas studies performed at three commercial low level radioactive waste disposal sites are reviewed. Methods used to obtain trench and soil gas samples are discussed. Estimates of rates of gas production and amounts released into the atmosphere (by the GASFLOW model) are evaluated. Tritium and carbon-14 gaseous compounds have been measured in these studies; tritiated methane is the major radionuclide species in all disposal trenches studied. The concentration of methane in a typical trench increases with the age of the trench, whereas the concentration of carbon dioxide is similar in all trenches.

  7. Estimating costs of low-level radioactive waste disposal alternatives for the Commonwealth of Massachusetts

    SciTech Connect (OSTI)

    Not Available

    1994-02-01

    This report was prepared for the Commonwealth of Massachusetts by the Idaho National Engineering Laboratory, National Low-Level Waste Management Program. It presents planning life-cycle cost (PLCC) estimates for four sizes of in-state low-level radioactive waste (LLRW) disposal facilities. These PLCC estimates include preoperational and operational expenditures, all support facilities, materials, labor, closure costs, and long-term institutional care and monitoring costs. It is intended that this report bc used as a broad decision making tool for evaluating one of the several complex factors that must be examined when deciding between various LLRW management options -- relative costs. Because the underlying assumptions of these analyses will change as the Board decides how it will manage Massachusett`s waste and the specific characteristics any disposal facility will have, the results of this study are not absolute and should only be used to compare the relative costs of the options presented. The disposal technology selected for this analysis is aboveground earth-mounded vaults. These vaults are reinforced concrete structures where low-level waste is emplaced and later covered with a multi-layered earthen cap. The ``base case`` PLCC estimate was derived from a preliminary feasibility design developed for the Illinois Low-Level Radioactive Waste Disposal Facility. This PLCC report describes facility operations and details the procedure used to develop the base case PLCC estimate for each facility component and size. Sensitivity analyses were performed on the base case PLCC estimate by varying several factors to determine their influences upon the unit disposal costs. The report presents the results of the sensitivity analyses for the five most significant cost factors.

  8. High Level Waste ManagemenfDivision ..

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    High Level Waste ManagemenfDivision .. . . HLWSystem Plan Revision 2(U) Westinghouse Savannah River Company . Aiken; South Carolina Jam,lary 14,1994 HIGH LEVEL WASTE SYSTEM PLAN REVISION 2 _--JANUARY 14, 1994 APPROVAL SHEET Deputy General Manager High Level Waste Management Westinghouse Savannah River Company fO ..... R. E. Erickson Director,- Vitrification Projects Division U. S. Department of Energy, Headquarters Date I Date Date " " HLW System Plan - Revision 2 (U) Table of Contents

  9. Vitrification of low-level radioactive waste in a slagging combustor

    SciTech Connect (OSTI)

    Holmes, M.J.; Downs, W.; Higley, B.A. [and others

    1995-07-01

    The suitability of a Babcock & Wilcox cyclone furnace to vitrify a low-level radioactive liquid waste was evaluated. The feed stream contained a mixture of simulated radioactive liquid waste and glass formers. The U.S. Department of Energy is testing technologies to vitrify over 60,000,000 gallons of this waste at the Hanford site. The tests reported here demonstrated the technical feasibility of Babcock & Wilcox`s cyclone vitrification technology to produce a glass for near surface disposal. Glass was produced over a period of 24-hours at a rate of 100 to 150 lb/hr. Based on glass analyses performed by an independent laboratory, all of the glass samples had leachabilities at least as low as those of the laboratory glass that the recipe was based upon. This paper presents the results of this demonstration, and includes descriptions of feed preparation, glass properties, system operation, and flue gas composition. The paper also provides discussions on key technical issues required to match cyclone furnace vitrification technology to this U.S. Department of Energy Hanford site application.

  10. Greater-than-Class C low-level radioactive waste characterization. Appendix A-2: Timing of greater-than-Class C low-level radioactive waste from nuclear power plants

    SciTech Connect (OSTI)

    Steinke, W.F.

    1994-09-01

    Planning for the storage or disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of that waste. Timing, or the date the waste will require storage or disposal, is an integral aspect of that planning. The majority of GTCC LLW is generated by nuclear power plants, and the length of time a reactor remains operational directly affects the amount of GTCC waste expected from that reactor. This report uses data from existing literature to develop high, base, and low case estimates for the number of plants expected to experience (a) early shutdown, (b) 40-year operation, or (c) life extension to 60-year operation. The discussion includes possible effects of advanced light water reactor technology on future GTCC LLW generation. However, the main focus of this study is timing for shutdown of current technology reactors that are under construction or operating.

  11. A comparison of solidification media for the stabilization of low- level radioactive wastes

    SciTech Connect (OSTI)

    Cowgill, M.G.

    1991-10-01

    When requirements exist to stabilize low-level radioactive waste (LLW) prior to disposal, efforts to achieve this stability often center on the mixing of the waste with a solidification medium. Although historically the medium of choice has been based on the use of portland cement as the binder material, several other options have been developed and subsequently implemented. These include thermoplastic polymers, thermosetting polymers and gypsum. No one medium has thus far been successful in providing stability to all forms of LLW. The characteristics and attributes of these different binder materials are reviewed and compared. The aspects examined include availability of information, limitations to use, sensitivity to process or waste chemistry changes, radionuclide retention ability, modeling of radionuclide release processes, ease and safety of use, and relative costs.

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

    DOE Patents [OSTI]

    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

  13. Measurement of Radioactive Contamination in the High-Resistivity Silicon CCDs of the DAMIC Experiment

    SciTech Connect (OSTI)

    Aguilar-Arevalo, A.

    2015-08-25

    We present measurements of radioactive contamination in the high-resistivity silicon charge-coupled devices (CCDs) used by the DAMIC experiment to search for dark matter particles. Novel analysis methods, which exploit the unique spatial resolution of CCDs, were developed to identify α and β particles. Uranium and thorium contamination in the CCD bulk was measured through α spectroscopy, with an upper limit on the 238U (232Th) decay rate of 5 (15) kg-1 d-1 at 95% CL. We also searched for pairs of spatially correlated electron tracks separated in time by up to tens of days, as expected from 32Si –32P or 210Pb –210Bi sequences of b decays. The decay rate of 32Si was found to be 80+110-65 (95% CI). An upper limit of ~35 kg -1 d-1 (95% CL) on the 210Pb decay rate was obtained independently by α spectroscopy and the β decay sequence search. Furthermore, these levels of radioactive contamination are sufficiently low for the successful operation of CCDs in the forthcoming 100 g DAMIC detector.

  14. Measurement of Radioactive Contamination in the High-Resistivity Silicon CCDs of the DAMIC Experiment

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Aguilar-Arevalo, A.

    2015-08-25

    We present measurements of radioactive contamination in the high-resistivity silicon charge-coupled devices (CCDs) used by the DAMIC experiment to search for dark matter particles. Novel analysis methods, which exploit the unique spatial resolution of CCDs, were developed to identify ? and ? particles. Uranium and thorium contamination in the CCD bulk was measured through ? spectroscopy, with an upper limit on the 238U (232Th) decay rate of 5 (15) kg-1 d-1 at 95% CL. We also searched for pairs of spatially correlated electron tracks separated in time by up to tens of days, as expected from 32Si 32P or 210Pbmore210Bi sequences of b decays. The decay rate of 32Si was found to be 80+110-65 (95% CI). An upper limit of ~35 kg -1 d-1 (95% CL) on the 210Pb decay rate was obtained independently by ? spectroscopy and the ? decay sequence search. Furthermore, these levels of radioactive contamination are sufficiently low for the successful operation of CCDs in the forthcoming 100 g DAMIC detector.less

  15. A data base for low-level radioactive waste disposal sites

    SciTech Connect (OSTI)

    Daum, M.L.; Moskowitz, P.D.

    1989-07-01

    A computerized database was developed to assist the US Environmental Protection Agency (EPA) in evaluating methods and data for characterizing health hazards associated with land and ocean disposal options for low-level radioactive wastes. The data cover 1984 to 1987. The types of sites considered include Nuclear Regulatory Commission (NRC) licensed commercial disposal sites, EPA National Priority List (NPL) sites, US Department of Energy (DOE) Formerly Utilized Sites Remedial Action Project (FUSRAP) and DOE Surplus Facilities Management Program (SFMP) sites, inactive US ocean disposal sites, and DOE/Department of Defense facilities. Sources of information include reports from EPA, the US Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC), as well as direct communication with individuals associated with specific programs. The data include site descriptions, waste volumes and activity levels, and physical and radiological characterization of low-level wastes. Additional information on mixed waste, packaging forms, and disposal methods were compiled, but are not yet included in the database. 55 refs., 4 figs., 2 tabs.

  16. Protecting Lake Ontario - Treating Wastewater from the Remediated Low-Level Radioactive Waste Management Facility - 13227

    SciTech Connect (OSTI)

    Freihammer, Till; Chaput, Barb; Vandergaast, Gary; Arey, Jimi

    2013-07-01

    The Port Granby Project is part of the larger Port Hope Area Initiative, a community-based program for the development and implementation of a safe, local, long-term management solution for historic low level radioactive waste (LLRW) and marginally contaminated soils (MCS). The Port Granby Project involves the relocation and remediation of up to 0.45 million cubic metres of such waste from the current Port Granby Waste Management Facility located in the Municipality of Clarington, Ontario, adjacent to the shoreline of Lake Ontario. The waste material will be transferred to a new suitably engineered Long-Term Waste Management Facility (LTWMF) to be located inland approximately 700 m from the existing site. The development of the LTWMF will include construction and commissioning of a new Wastewater Treatment Plant (WWTP) designed to treat wastewater consisting of contaminated surface run off and leachate generated during the site remediation process at the Port Granby Waste Management Facility as well as long-term leachate generated at the new LTWMF. Numerous factors will influence the variable wastewater flow rates and influent loads to the new WWTP during remediation. The treatment processes will be comprised of equalization to minimize impacts from hydraulic peaks, fine screening, membrane bioreactor technology, and reverse osmosis. The residuals treatment will comprise of lime precipitation, thickening, dewatering, evaporation and drying. The distribution of the concentration of uranium and radium - 226 over the various process streams in the WWTP was estimated. This information was used to assess potential worker exposure to radioactivity in the various process areas. A mass balance approach was used to assess the distribution of uranium and radium - 226, by applying individual contaminant removal rates for each process element of the WTP, based on pilot scale results and experience-based assumptions. The mass balance calculations were repeated for various flow and load scenarios. (authors)

  17. Fifteenth annual U.S. Department of Energy low-level radioactive waste management conference: Agenda and abstracts

    SciTech Connect (OSTI)

    1993-12-31

    The goal of the conference was to give the opportunity to identify and discuss low-level radioactive waste management issues, share lessons learned, and hear about some of the latest advances in technology. Abstracts of the presentations are arranged into the following topical sections: (1) Performance Management Track: Performance assessment perspectives; Site characterization; Modeling and performance assessment; and Remediation; (2) Technical Track: Strategic planning; Tools and options; Characterization and validation; Treatment updates; Technology development; and Storage; (3) Institutional Track: Orders and regulatory issues; Waste management options; Legal, economic, and social issues; Public involvement; Siting process; and Low-level radioactive waste policy amendment acts.

  18. Management of Low-Level Radioactive Waste from Research, Hospitals and Nuclear Medical Centers in Egypt - 13469

    SciTech Connect (OSTI)

    Hasan, M.A.; Selim, Y.T.; Lasheen, Y.F.

    2013-07-01

    The application of radioisotopes and radiation sources in medical diagnosis and therapy is an important issue. Physicians can use radioisotopes to diagnose and treat diseases. Methods of treatment, conditioning and management of low level radioactive wastes from the use of radiation sources and radioisotopes in hospitals and nuclear medicine application, are described. Solid Radioactive waste with low-level activity after accumulation, minimization, segregation and measurement, are burned or compressed in a compactor according to the international standards. Conditioned drums are transported to the interim storage site at the Egyptian Atomic Energy Authority (EAEA) represented in Hot Labs and Waste Management Center (HLWMC) for storage and monitoring. (authors)

  19. National profile on commercially generated low-level radioactive mixed waste

    SciTech Connect (OSTI)

    Klein, J.A.; Mrochek, J.E.; Jolley, R.L.; Osborne-Lee, I.W.; Francis, A.A.; Wright, T.

    1992-12-01

    This report details the findings and conclusions drawn from a survey undertaken as part of a joint US Nuclear Regulatory Commission and US Environmental Protection Agency-sponsored project entitled ``National Profile on Commercially Generated Low-Level Radioactive Mixed Waste.`` The overall objective of the work was to compile a national profile on the volumes, characteristics, and treatability of commercially generated low-level mixed waste for 1990 by five major facility categories-academic, industrial, medical, and NRC-/Agreement State-licensed goverment facilities and nuclear utilities. Included in this report are descriptions of the methodology used to collect and collate the data, the procedures used to estimate the mixed waste generation rate for commercial facilities in the United States in 1990, and the identification of available treatment technologies to meet applicable EPA treatment standards (40 CFR Part 268) and, if possible, to render the hazardous component of specific mixed waste streams nonhazardous. The report also contains information on existing and potential commercial waste treatment facilities that may provide treatment for specific waste streams identified in the national survey. The report does not include any aspect of the Department of Energy`s (DOES) management of mixed waste and generally does not address wastes from remedial action activities.

  20. US Army facility for the consolidation of low-level radioactive waste

    SciTech Connect (OSTI)

    Stein, S.L.; Tanner, J.E.; Murphy, B.L.; Gillings, J.C.; Hadley, R.T.; Lyso, O.M.; Gilchrist, R.L.; Murphy, D.W.

    1983-12-01

    A preliminary study of a waste consolidation facility for the Department of the Army's low-level radioactive waste was carried out to determine a possible site and perform a cost-benefit analysis. Four sites were assessed as possible locations for such a facility, using predetermined site selection criteria. To assist in the selection of a site, an evaluation of environmental issues was included as part of each site review. In addition, a preliminary design for a waste consolidation facility was developed, and facilities at each site were reviewed for their availability and suitability for this purpose. Currently available processes for volume reduction, as well as processes still under development, were then investigated, and the support and handling equipment and the staff needed for the safe operation of a waste consolidation facility were studied. Using current costs for the transportation and burial of low-level waste, a cost comparison was then made between waste disposal with and without the utilization of volume reduction. Finally, regulations that could affect the operation of a waste consolidation facility were identified and their impact was assessed. 11 references, 5 figures, 16 tables.

  1. Radioactive Waste Management Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09

    This Manual further describes the requirements and establishes specific responsibilities for implementing DOE O 435.1, Radioactive Waste Management, for the management of DOE high-level waste, transuranic waste, low-level waste, and the radioactive component of mixed waste. Change 1 dated 6/19/01 removes the requirement that Headquarters is to be notified and the Office of Environment, Safety and Health consulted for exemptions for use of non-DOE treatment facilities. Certified 1-9-07.

  2. Geochemical information for sites contaminated with low-level radioactive wastes: I. Niagara Falls Storage Site

    SciTech Connect (OSTI)

    Seeley, F.G.; Kelmers, A.D.

    1984-11-01

    The Niagara Falls Storage Site (NFSS) became radioactively contaminated as a result of wastes that were being stored from operations carried out to recover uranium from pitchblende ore in the 1940s and 1950s. The US Department of Energy (DOE) is considering various remedial action options for the NFSS. This report describes the results of geochemical investigations performed to help provide a quantitative evaluation of the effects of various options. NFSS soil and groundwater samples were characterized; and uranium and radium sorption ratios, as well as apparent concentration limit values, were measured in site soil/groundwater systems by employing batch contact methodology. The results suggest that any uranium which is in solution in the groundwater at the NFSS may be poorly retarded due to the low uranium sorption ratio values and high solubility measured. Further, appreciable concentrations of uranium in groundwater could be attained from soluble wastes. Release of uranium via groundwater migration could be a significant release pathway. Solubilized radium would be expected to be effectively retarded by soil at the NFSS as a result of the very high radium sorption ratios observed. The addition of iron oxyhydroxide to NFSS soils resulted in much higher uranium sorption ratios. Additional field testing of this potential remedial action additive could be desirable. 10 references.

  3. Source team evaluation for radioactive low-level waste disposal performance assessment

    SciTech Connect (OSTI)

    Cowgill, M.G.; Sullivan, T.M.

    1993-01-01

    Information compiled on the low-level radioactive waste disposed at the three currently operating commercial disposal sites during the period 1987--1989 have been reviewed and processed in order to determine the total activity distribution in terms of waste stream, waste classification and waste form. The review identified deficiencies in the information currently being recorded on shipping manifests and the development of a uniform manifest is recommended (the NRC is currently developing a rule to establish a uniform manifest). The data from waste disposed during 1989 at one of the sites (Richland, WA) were more detailed than the data available during other years and at other sites, and thus were amenable to a more in-depth treatment. This included determination of the distribution of activity for each radionuclide by waste form, and thus enabled these data to be evaluated in terms of the specific needs for improved modeling of releases from waste packages. From the results, preliminary lists have been prepared of the isotopes which might be the most significant from the aspect of the development of a source term model.

  4. Equity of commercial low-level radioactive waste disposal fees. Report to Congress

    SciTech Connect (OSTI)

    1998-02-01

    In the Report accompanying the Fiscal Year 1997 Senate Energy and Water Development Appropriations Bill, the Senate Appropriations Committee directed the Department of Energy (DOE) to prepare a study of the costs of operating a low-level radioactive waste (LLW) disposal facility such as the one at Barnwell, South Carolina, and to determine whether LLW generators are paying equitable disposal fees. The disposal costs of four facilities are reviewed in this report, two operating facilities and two planned facilities. The operating facilities are located at Barnwell, South Carolina, and Richland, Washington. They are operated by Chem-Nuclear, LLC, (Chem-Nuclear), and US Ecology, Inc., (US Ecology), respectively. The planned facilities are expected to be built at Ward Valley, California, and Sierra Blanca, Texas. They will be operated by US Ecology and the State of Texas, respectively. This report found that disposal fees vary significantly among facilities for a variety of reasons. However, the information suggests that at each disposal facility, LLW generators pay equitable disposal fees.

  5. Potential problem areas: extended storage of low-level radioactive waste

    SciTech Connect (OSTI)

    Siskind, B.

    1985-01-01

    If a state or regional compact does not have adequate disposal capacity for low-level radioactive waste (LLRW), then extended storage of certain LLRW may be necessary. The Nuclear Regulatory Commission (NRC) has contracted with Brookhaven National Laboratory to address the technical issues of extended storage. The dual objectives of this study are (1) to provide practical technical assessments for NRC to consider in evaluating specific proposals for extended storage and (2) to help ensure adequate consideration by NRC, Agreement States, and licensees of potential problems that may arise from existing or proposed extended storage practices. Storage alternatives are considered in order to characterize the likely storage environments for these wastes. In particular, the range of storage alternatives considered and being implemented by the nuclear power plant utilities is described. The properties of the waste forms and waste containers are discussed. An overview is given of the performance of the waste package and its contents during storage (e.g., radiolytic gas generation, corrosion) and of the effects of extended storage on the performance of the waste package after storage (e.g., radiation-induced embrittlement of polyethylene, the weakening of steel containers by corrosion). Additional information and actions required to address these concerns, including possible mitigative measures, are discussed. 26 refs., 1 tab.

  6. Methods for verifying compliance with low-level radioactive waste acceptance criteria

    SciTech Connect (OSTI)

    NONE

    1993-09-01

    This report summarizes the methods that are currently employed and those that can be used to verify compliance with low-level radioactive waste (LLW) disposal facility waste acceptance criteria (WAC). This report presents the applicable regulations representing the Federal, State, and site-specific criteria for accepting LLW. Typical LLW generators are summarized, along with descriptions of their waste streams and final waste forms. General procedures and methods used by the LLW generators to verify compliance with the disposal facility WAC are presented. The report was written to provide an understanding of how a regulator could verify compliance with a LLW disposal facility`s WAC. A comprehensive study of the methodology used to verify waste generator compliance with the disposal facility WAC is presented in this report. The study involved compiling the relevant regulations to define the WAC, reviewing regulatory agency inspection programs, and summarizing waste verification technology and equipment. The results of the study indicate that waste generators conduct verification programs that include packaging, classification, characterization, and stabilization elements. The current LLW disposal facilities perform waste verification steps on incoming shipments. A model inspection and verification program, which includes an emphasis on the generator`s waste application documentation of their waste verification program, is recommended. The disposal facility verification procedures primarily involve the use of portable radiological survey instrumentation. The actual verification of generator compliance to the LLW disposal facility WAC is performed through a combination of incoming shipment checks and generator site audits.

  7. Radioactive Waste Management Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09

    This Manual further describes the requirements and establishes specific responsibilities for implementing DOE O 435.1, Radioactive Waste Management, for the management of DOE high-level waste, transuranic waste, low-level waste, and the radioactive component of mixed waste. The purpose of the Manual is to catalog those procedural requirements and existing practices that ensure that all DOE elements and contractors continue to manage DOE's radioactive waste in a manner that is protective of worker and public health and safety, and the environment. Does not cancel other directives.

  8. Comparison of selected DOE and non-DOE requirements, standards, and practices for Low-Level Radioactive Waste Disposal

    SciTech Connect (OSTI)

    Cole, L.; Kudera, D.; Newberry, W.

    1995-12-01

    This document results from the Secretary of Energy`s response to Defense Nuclear Facilities Safety Board Recommendation 94--2. The Secretary stated that the US Department of Energy (DOE) would ``address such issues as...the need for additional requirements, standards, and guidance on low-level radioactive waste management. `` The authors gathered information and compared DOE requirements and standards for the safety aspects Of low-level disposal with similar requirements and standards of non-DOE entities.

  9. Greater-than-Class C low-level radioactive waste shipping package/container identification and requirements study. National Low-Level Waste Management Program

    SciTech Connect (OSTI)

    Tyacke, M.

    1993-08-01

    This report identifies a variety of shipping packages (also referred to as casks) and waste containers currently available or being developed that could be used for greater-than-Class C (GTCC) low-level waste (LLW). Since GTCC LLW varies greatly in size, shape, and activity levels, the casks and waste containers that could be used range in size from small, to accommodate a single sealed radiation source, to very large-capacity casks/canisters used to transport or dry-store highly radioactive spent fuel. In some cases, the waste containers may serve directly as shipping packages, while in other cases, the containers would need to be placed in a transport cask. For the purpose of this report, it is assumed that the generator is responsible for transporting the waste to a Department of Energy (DOE) storage, treatment, or disposal facility. Unless DOE establishes specific acceptance criteria, the receiving facility would need the capability to accept any of the casks and waste containers identified in this report. In identifying potential casks and waste containers, no consideration was given to their adequacy relative to handling, storage, treatment, and disposal. Those considerations must be addressed separately as the capabilities of the receiving facility and the handling requirements and operations are better understood.

  10. PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)

    SciTech Connect (OSTI)

    CERTA, P.J.

    2006-02-22

    The primary objective of this study is to demonstrate a mission scenario that uses pairwise and incidental blending of high level waste (HLW) to reduce the total mass of HLW glass. Secondary objectives include understanding how recent refinements to the tank waste inventory and solubility assumptions affect the mass of HLW glass and how logistical constraints may affect the efficacy of HLW blending.

  11. Geochemical information for sites contaminated with low-level radioactive wastes. III. Weldon Spring Storage Site

    SciTech Connect (OSTI)

    Seeley, F.G.; Kelmers, A.D.

    1985-02-01

    The Weldon Spring Storage Site (WSSS), which includes both the chemical site and the quarry, became radioactively contaminated as the result of wastes that were being stored from operations to recover uranium from pitchblende ores in the 1940s and 1950s. The US Department of Energy (DOE) is considering various remedial action options for the WSSS. This report describes the results of geochemical investigations carried out at Oak Ridge National Laboratory (ORNL) to support these activities and to help quantify various remedial action options. Soil and groundwater samples were characterized, and uranium and radium sorption ratios were measured in site soil/groundwater systems by batch contact methodology. Soil samples from various locations around the raffinate pits were found to contain major amounts of silica, along with illite as the primary clay constituent. Particle sizes of the five soil samples were variable (50% distribution point ranging from 12 to 81 ..mu..m); the surface areas varied from 13 to 62 m/sup 2//g. Elemental analysis of the samples showed them to be typical of sandy clay and silty clay soils. Groundwater samples included solution from Pit 3 and well water from Well D. Anion analyses showed significant concentrations of sulfate and nitrate (>350 and >7000 mg/L, respectively) in the solution from Pit 3. These anions were also present in the well water, but in lower concentrations. Uranium sorption ratios for four of the soil samples contacted with the solution from Pit 3 were moderate to high (approx. 300 to approx. 1000 mL/g). The fifth sample had a ratio of only 12 mL/g. Radium sorption ratios for the five samples were moderate to high (approx. 600 to approx. 1000 mL/g). These values indicate that soil at the WSSS may show favorable retardation of uranium and radium in the groundwater. 13 references, 13 figures, 10 tables.

  12. High-level waste management technology program plan

    SciTech Connect (OSTI)

    Harmon, H.D.

    1995-01-01

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

  13. Field testing an OREX{reg_sign} based {open_quotes}point of generation{close_quotes} low-level radioactive waste reduction program at FP&L`s St. Lucie Plant

    SciTech Connect (OSTI)

    Payne, K.; Haynes, B.

    1996-10-01

    Nuclear power facilities, both commercial and government operated, generate material called Dry Active Waste (DAW). DAW is a by-product of maintenance and operation of the power systems which contain radioactive materials. DAW can be any material contaminated with radioactive particles as long as it is not a fluid, typically: paper, cardboard, wood, plastics, cloth, and any other solid which is contaminated and determined to be dry. DAW is generated when any material is exposed to loose radioactive particles and subsequently becomes contaminated. In the United States, once a material is contaminated it must be treated as radioactive waste and disposed of in accordance with the requirements of Title 10 of the Code of Federal Regulations. Problems facing all commercial and non-commercial nuclear facilities are escalating costs of processing DAW and volumetric reduction of the DAW generated. Currently, approximately 85% of all DAW generated at a typical facility is comprised of anti-contamination clothing and protective barrier materials. Facilities that generate low-level radioactive waste need to dramatically reduce their waste volumes. This curtailment is required for several reasons: the number of radioactive waste repositories now accepting new waste is limited; the current cost of burial at an operating dump site is significant. Costs can be as high as $4,000 for a single 55 gallon drum; the cost of burial is constantly increasing; onsite storage of low-level radioactive waste is costly and results in a burial fee at plant decommissioning.

  14. High-Level Waste Melter Study Report

    SciTech Connect (OSTI)

    Perez Jr, Joseph M; Bickford, Dennis F; Day, Delbert E; Kim, Dong-Sang; Lambert, Steven L; Marra, Sharon L; Peeler, David K; Strachan, Denis M; Triplett, Mark B; Vienna, John D; Wittman, Richard S

    2001-07-13

    At the Hanford Site in Richland, Washington, the path to site cleanup involves vitrification of the majority of the wastes that currently reside in large underground tanks. A Joule-heated glass melter is the equipment of choice for vitrifying the high-level fraction of these wastes. Even though this technology has general national and international acceptance, opportunities may exist to improve or change the technology to reduce the enormous cost of accomplishing the mission of site cleanup. Consequently, the U.S. Department of Energy requested the staff of the Tanks Focus Area to review immobilization technologies, waste forms, and modifications to requirements for solidification of the high-level waste fraction at Hanford to determine what aspects could affect cost reductions with reasonable long-term risk. The results of this study are summarized in this report.

  15. Characterization of radionuclide-chelating agent complexes found in low-level radioactive decontamination waste. Literature review

    SciTech Connect (OSTI)

    Serne, R.J.; Felmy, A.R.; Cantrell, K.J.; Krupka, K.M.; Campbell, J.A.; Bolton, H. Jr.; Fredrickson, J.K.

    1996-03-01

    The US Nuclear Regulatory Commission is responsible for regulating the safe land disposal of low-level radioactive wastes that may contain organic chelating agents. Such agents include ethylenediaminetetraacetic acid (EDTA), diethylenetriaminepentaacetic acid (DTPA), picolinic acid, oxalic acid, and citric acid, and can form radionuclide-chelate complexes that may enhance the migration of radionuclides from disposal sites. Data from the available literature indicate that chelates can leach from solidified decontamination wastes in moderate concentration (1--100 ppm) and can potentially complex certain radionuclides in the leachates. In general it appears that both EDTA and DTPA have the potential to mobilize radionuclides from waste disposal sites because such chelates can leach in moderate concentration, form strong radionuclide-chelate complexes, and can be recalcitrant to biodegradation. It also appears that oxalic acid and citric acid will not greatly enhance the mobility of radionuclides from waste disposal sites because these chelates do not appear to leach in high concentration, tend to form relatively weak radionuclide-chelate complexes, and can be readily biodegraded. In the case of picolinic acid, insufficient data are available on adsorption, complexation of key radionuclides (such as the actinides), and biodegradation to make definitive predictions, although the available data indicate that picolinic acid can chelate certain radionuclides in the leachates.

  16. Department of Energy treatment capabilities for greater-than-Class C low-level radioactive waste

    SciTech Connect (OSTI)

    Morrell, D.K.; Fischer, D.K.

    1995-01-01

    This report provides brief profiles for 26 low-level and high-level waste treatment capabilities available at the Idaho National Engineering Laboratory (INEL), Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Oak Ridge National Laboratory (ORNL), Pacific Northwest Laboratory (PNL), Rocky Flats Plant (RFP), Savannah River Site (SRS), and West Valley Demonstration Plant (WVDP). Six of the treatments have potential use for greater-than-Class C low-level waste (GTCC LLW). They include: (a) the glass ceramic process and (b) the Waste Experimental Reduction Facility incinerator at INEL; (c) the Super Compaction and Repackaging Facility and (d) microwave melting solidification at RFP; (e) the vitrification plant at SRS; and (f) the vitrification plant at WVDP. No individual treatment has the capability to treat all GTCC LLW streams. It is recommended that complete physical and chemical characterizations be performed for each GTCC waste stream, to permit using multiple treatments for GTCC LLW.

  17. High Level Waste Management Division High-Level Waste System Plan

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    6 (U) December 20, 1995 Westinghouse Savannah River Company Savannah River Site Aiken, SC 29808 HLW-OVP-95-0102 Westinghouse Savannah River Company 2 0 GtC 1995 Mr. A. L. Watkins, Assistant Manager High Level Waste U. S. Department of Energy Savannah River Operations Office P. O. Box A Aiken, SC 29802 Dear Mr. Watkins: A. B. Scott, Jr. Vice President and General Manager High Level Waste Management Division P. O. Box 616 Aiken, SC 29802 HLW-OVP-95-0102 HIGH LEVEL WASTE SYSTEM PLAN. REVISION 6 (U)

  18. Decomposition of tetraphenylborate precipitates used to isolate Cs-137 from Savannah River Site high-level waste

    SciTech Connect (OSTI)

    Ferrara, D.M.; Bibler, N.E.; Ha, B.C.

    1993-03-01

    This paper presents results of the radioactive demonstration of the Precipitate Hydrolysis Process (PHP) that will be performed in the Defense Waste Processing Facility (DWPF) at the Savannah River Site. The PHP destroys the tetraphenylborate precipitate that is used at SRS to isolate Cs-137 from caustic High-Level Waste (HLW) supernates. This process is necessary to decrease the amount of organic compounds going to the melter in the DWPF. Actual radioactive precipitate containing Cs-137 was used for this demonstration.

  19. ISO standardization of scaling factor method for low and intermediate level radioactive wastes generated at nuclear power plants

    SciTech Connect (OSTI)

    Kashiwagi, Makoto; Masui, Hideki; Denda, Yasutaka; James, David; Lantes, Bertrand; Mueller, Wolfgang; Garamszeghy, Mike; Leganes, Jose Luis; Maxeiner, Harald; Van Velzen, Leo

    2007-07-01

    Low- and intermediate-level radioactive wastes (L-ILW ) generated at nuclear power plants are disposed of in various countries. In the disposal of such wastes, it is required that the radioactivity concentrations of waste packages should be declared with respect to difficult-to-measure nuclides (DTM nuclides), such as C-14, Ni-63 and a-emitting nuclides, which are often limited to maximum values in disposal licenses, safety cases and/or regulations for maximum radioactive concentrations. To fulfill this requirement, the Scaling Factor method (SF method) has been applied in various countries as a principal method for determining the concentrations of DTM nuclides. In the SF method, the concentrations of DTM nuclides are determined by multiplying the concentrations of certain key nuclides by SF values (the determined ratios of radioactive concentration between DTM nuclides and those key nuclides). The SF values used as conversion factors are determined from the correlation between DTM nuclides and key nuclides such as Co-60. The concentrations of key nuclides are determined by {gamma} ray measurements which can be made comparatively easily from outside the waste package. The SF values are calculated based on the data obtained from the radiochemical analysis of waste samples. The use of SFs, which are empirically based on analytical data, has become established as a widely recognized 'de facto standard'. A number of countries have independently collected nuclide data by analysis over many years and each has developed its own SF method, but all the SF methods that have been adopted are similar. The project team for standardization had been organized for establishing this SF method as a 'de jure standard' in the international standardization system of the International Organization for Standardization (ISO). The project team for standardization has advanced the standardization through technical studies, based upon each country's study results and analysis data. The conclusions reached by the project team was published as ISO International Standard 21238:2007 'The Scaling Factor method to determine the radioactivity of low- and intermediate-level radioactive waste packages generated at nuclear power plants'. This paper gives an introduction to the international standardization process for the SF method and the contents of the recently published International Standard. (authors)

  20. Greater-than-Class C low-level radioactive waste characterization. Appendix A-3: Basis for greater-than-Class C low-level radioactive waste light water reactor projections

    SciTech Connect (OSTI)

    Mancini, A.; Tuite, P.; Tuite, K.; Woodberry, S.

    1994-09-01

    This study characterizes low-level radioactive waste types that may exceed Class C limits at light water reactors, estimates the amounts of waste generated, and estimates radionuclide content and distribution within the waste. Waste types that may exceed Class C limits include metal components that become activated during operations, process wastes such as cartridge filters and decontamination resins, and activated metals from decommissioning activities. Operating parameters and current management practices at operating plants are reviewed and used to estimate the amounts of low-level waste exceeding Class C limits that is generated per fuel cycle, including amounts of routinely generated activated metal components and process waste. Radionuclide content is calculated for specific activated metals components. Empirical data from actual low-level radioactive waste are used to estimate radionuclide content for process wastes. Volumes and activities are also estimated for decommissioning activated metals that exceed Class C limits. To estimate activation levels of decommissioning waste, six typical light water reactors are modeled and analyzed. This study does not consider concentration averaging.

  1. High accuracy electronic material level sensor

    DOE Patents [OSTI]

    McEwan, Thomas E.

    1997-01-01

    The High Accuracy Electronic Material Level Sensor (electronic dipstick) is a sensor based on time domain reflectometry (TDR) of very short electrical pulses. Pulses are propagated along a transmission line or guide wire that is partially immersed in the material being measured; a launcher plate is positioned at the beginning of the guide wire. Reflected pulses are produced at the material interface due to the change in dielectric constant. The time difference of the reflections at the launcher plate and at the material interface are used to determine the material level. Improved performance is obtained by the incorporation of: 1) a high accuracy time base that is referenced to a quartz crystal, 2) an ultrawideband directional sampler to allow operation without an interconnect cable between the electronics module and the guide wire, 3) constant fraction discriminators (CFDs) that allow accurate measurements regardless of material dielectric constants, and reduce or eliminate errors induced by triple-transit or "ghost" reflections on the interconnect cable. These improvements make the dipstick accurate to better than 0.1%.

  2. High accuracy electronic material level sensor

    DOE Patents [OSTI]

    McEwan, T.E.

    1997-03-11

    The High Accuracy Electronic Material Level Sensor (electronic dipstick) is a sensor based on time domain reflectometry (TDR) of very short electrical pulses. Pulses are propagated along a transmission line or guide wire that is partially immersed in the material being measured; a launcher plate is positioned at the beginning of the guide wire. Reflected pulses are produced at the material interface due to the change in dielectric constant. The time difference of the reflections at the launcher plate and at the material interface are used to determine the material level. Improved performance is obtained by the incorporation of: (1) a high accuracy time base that is referenced to a quartz crystal, (2) an ultrawideband directional sampler to allow operation without an interconnect cable between the electronics module and the guide wire, (3) constant fraction discriminators (CFDs) that allow accurate measurements regardless of material dielectric constants, and reduce or eliminate errors induced by triple-transit or ``ghost`` reflections on the interconnect cable. These improvements make the dipstick accurate to better than 0.1%. 4 figs.

  3. Radiolytic gas generation from cement-based waste hosts for DOE low-level radioactive wastes

    SciTech Connect (OSTI)

    Dole, L.R.; Friedman, H.A.

    1986-01-01

    Using cement-based immobilization binders with simulated radioactive waste containing sulfate, nitrate, nitrite, phosphate, and fluoride anions, the gamma- and alpha-radiolytic gas generation factors (G/sub t/, molecules/100 eV) and gas compositions were measured on specimens of cured grouts. These tests studied the effects of; (1) waste composition; (2) the sample surface-to-volume ratio; (3) the waste slurry particle size; and (4) the water content of the waste host formula. The radiolysis test vessels were designed to minimize the ''dead'' volume and to simulate the configuration of waste packages.

  4. Characterization of Class A low-level radioactive waste 1986--1990. Volume 6: Appendices G--J

    SciTech Connect (OSTI)

    Dehmel, J.C.; Loomis, D.; Mauro, J.; Kaplan, M.

    1994-01-01

    Under contract to the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, the firms of S. Cohen & Associates, Inc. (SC&A) and Eastern Research Group (ERG) have compiled a report that describes the physical, chemical, and radiological properties of Class-A low-level radioactive waste. The report also presents information characterizing various methods and facilities used to treat and dispose non-radioactive waste. A database management program was developed for use in accessing, sorting, analyzing, and displaying the electronic data provided by EG&G. The program was used to present and aggregate data characterizing the radiological, physical, and chemical properties of the waste from descriptions contained in shipping manifests. The data thus retrieved are summarized in tables, histograms, and cumulative distribution curves presenting radionuclide concentration distributions in Class-A waste as a function of waste streams, by category of waste generators, and regions of the United States. The report also provides information characterizing methods and facilities used to treat and dispose non-radioactive waste, including industrial, municipal, and hazardous waste regulated under Subparts C and D of the Resource Conservation and Recovery Act (RCRA). The information includes a list of disposal options, the geographical locations of the processing and disposal facilities, and a description of the characteristics of such processing and disposal facilities. Volume 1 contains the Executive Summary, Volume 2 presents the Class-A waste database, Volume 3 presents the information characterizing non-radioactive waste management practices and facilities, and Volumes 4 through 7 contain Appendices A through P with supporting information.

  5. Characterization of Class A low-level radioactive waste 1986--1990. Volume 2: Main report -- Part A

    SciTech Connect (OSTI)

    Dehmel, J.C.; Loomis, D.; Mauro, J.; Kaplan, M.

    1994-01-01

    Under contract to the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, the firms of S. Cohen & Associates, Inc. (SC&A) and Eastern Research Group (ERG) have compiled a report that describes the physical, chemical, and radiological properties of Class-A low-level radioactive waste. The report also presents information characterizing various methods and facilities used to treat and dispose non-radioactive waste. A database management program was developed for use in accessing, sorting, analyzing, and displaying the electronic data provided by EG&G. The program was used to present and aggregate data characterizing the radiological, physical, and chemical properties of the, waste from descriptions contained in shipping manifests. The data thus retrieved are summarized in tables, histograms, and cumulative distribution curves presenting radionuclide concentration distributions in Class-A waste as a function of waste streams, by category of waste generators, and regions of the United States. The report also provides information characterizing methods and facilities used to treat and dispose non-radioactive waste, including industrial, municipal, and hazardous waste regulated under Subparts C and D of the Resource Conservation and Recovery Act (RCRA). The information includes a list of disposal options, the geographical locations of the processing and disposal facilities, and a description of the characteristics of such processing and disposal facilities. Volume 1 contains the Executive Summary, Volume 2 presents the Class-A waste database, Volume 3 presents the information characterizing non-radioactive waste management practices and facilities, and Volumes 4 through 7 contain Appendices A through P with supporting information.

  6. Characterization of Class A low-level radioactive waste 1986--1990. Volume 3: Main report -- Part B

    SciTech Connect (OSTI)

    Dehmel, J.C.; Loomis, D.; Mauro, J.; Kaplan, M.

    1994-01-01

    Under contract to the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, the firms of S. Cohen & Associates, Inc. (SC&A) and Eastern Research Group (ERG) have compiled a report that describes the physical, chemical, and radiological properties of Class-A low-level radioactive waste. The report also presents information characterizing various methods and facilities used to treat and dispose non-radioactive waste. A database management program was developed for use in accessing, sorting, analyzing, and displaying the electronic data provided by EG&G. The program was used to present and aggregate data characterizing the radiological, physical, and chemical properties of the waste from descriptions contained in shipping manifests. The data thus retrieved are summarized in tables, histograms, and cumulative distribution curves presenting radionuclide concentration distributions in Class-A waste as a function of waste streams, by category of waste generators, and regions of the United States. The report also provides information characterizing methods and facilities used to treat and dispose non-radioactive waste, including industrial, municipal, and hazardous waste regulated under Subparts C and D of the Resource Conservation and Recovery Act (RCRA). The information includes a list of disposal options, the geographical locations of the processing and disposal facilities, and a description of the characteristics of such processing and disposal facilities. Volume 1 contains the Executive Summary, Volume 2 presents the Class-A waste database, Volume 3 presents the information characterizing non-radioactive waste management practices and facilities, and Volumes 4 to 7 contain Appendices A to P with supporting information.

  7. Characterization of Class A low-level radioactive waste 1986--1990. Volume 7: Appendices K--P

    SciTech Connect (OSTI)

    Dehmel, J.C.; Loomis, D.; Mauro, J.; Kaplan, M.

    1994-01-01

    Under contract to the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, the firms of S. Cohen & Associates, Inc. (SC&A) and Eastern Research Group (ERG) have compiled a report that describes the physical, chemical, and radiological properties of Class-A low-level radioactive waste. The report also presents information characterizing various methods and facilities used to treat and dispose non-radioactive waste. A database management program was developed for use in accessing, sorting, analyzing, and displaying the electronic data provided by EG&G. The program was used to present and aggregate data characterizing the radiological, physical, and chemical properties of the waste from descriptions contained in shipping manifests. The data thus retrieved are summarized in tables, histograms, and cumulative distribution curves presenting radionuclide concentration distributions in Class-A waste as a function of waste streams, by category of waste generators, and regions of the United States. The report also provides information characterizing methods and facilities used to treat and dispose non-radioactive waste, including industrial, municipal, and hazardous waste regulated under Subparts C and D of the Resource Conservation and Recovery Act (RCRA). The information includes a list of disposal options, the geographical locations of the processing and disposal facilities, and a description of the characteristics of such processing and disposal facilities. Volume 1 contains the Executive Summary, Volume 2 presents the Class-A waste database, Volume 3 presents the information characterizing non-radioactive waste management practices and facilities, and Volumes 4 through 7 contain Appendices A through P with supporting information.

  8. Progress of the High Level Waste Program at the Defense Waste Processing Facility - 13178

    SciTech Connect (OSTI)

    Bricker, Jonathan M.; Fellinger, Terri L.; Staub, Aaron V.; Ray, Jeff W.; Iaukea, John F. [Savannah River Remediation, Aiken, South Carolina, 29808 (United States)] [Savannah River Remediation, Aiken, South Carolina, 29808 (United States)

    2013-07-01

    The Defense Waste Processing Facility at the Savannah River Site treats and immobilizes High Level Waste into a durable borosilicate glass for safe, permanent storage. The High Level Waste program significantly reduces environmental risks associated with the storage of radioactive waste from legacy efforts to separate fissionable nuclear material from irradiated targets and fuels. In an effort to support the disposition of radioactive waste and accelerate tank closure at the Savannah River Site, the Defense Waste Processing Facility recently implemented facility and flowsheet modifications to improve production by 25%. These improvements, while low in cost, translated to record facility production in fiscal years 2011 and 2012. In addition, significant progress has been accomplished on longer term projects aimed at simplifying and expanding the flexibility of the existing flowsheet in order to accommodate future processing needs and goals. (authors)

  9. EA-1061: The Off-site Volume Reduction of Low-level Radioactive Waste From the Savannah River Site, Aiken, South Carolina

    Broader source: Energy.gov [DOE]

    This EA evaluates the environmental impacts of the proposal for off-site volume reduction of low-level radioactive wastes generated at the U.S. Department of Energy's Savannah River Site located...

  10. Technology and System Level Demonstration of Highly Efficient...

    Broader source: Energy.gov (indexed) [DOE]

    Technology and System Level Demonstration of Highly Efficient and Clean, Diesel Powered Class 8 Trucks Technology and System Level Demonstration of Highly Efficient and Clean, ...

  11. NERC Presentation: Accommodating High Levels of Variable Generation...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    NERC Presentation: Accommodating High Levels of Variable Generation, October 29, 2010 NERC Presentation: Accommodating High Levels of Variable Generation, October 29, 2010 North ...

  12. Residual radioactive contamination from decommissioning: Technical basis for translating contamination levels to annual dose

    SciTech Connect (OSTI)

    Kennedy, W.E. Jr.; Peloquin, R.A. )

    1990-01-01

    This document describes the generic modeling of the total effective dose equivalent (TEDE) to an individual in a population from a unit concentration of residual radioactive contamination. Radioactive contamination inside buildings and soil contamination are considered. Unit concentration TEDE factors by radionuclide, exposure pathway, and exposure scenario are calculated. Reference radiation exposure scenarios are used to derive unit concentration TEDE factors for about 200 individual radionuclides and parent-daughter mixtures. For buildings, these unit concentration factors list the annual TEDE for volume and surface contamination situations. For soil, annual TEDE factors are presented for unit concentrations of radionuclides in soil during residential use of contaminated land and the TEDE per unit total inventory for potential use of drinking water from a ground-water source. Because of the generic treatment of potentially complex ground-water systems, the annual TEDE factors for drinking water for a given inventory may only indicate when additional site data or modeling sophistication are warranted. Descriptions are provided of the models, exposure pathways, exposure scenarios, parameter values, and assumptions used. An analysis of the potential annual TEDE resulting from reference mixtures of residual radionuclides is provided to demonstrate application of the TEDE factors. 62 refs., 5 figs., 66 tabs.

  13. Hydrologic and geologic aspects of low-level radioactive-waste site management. [Shallow land burial at Oak Ridge

    SciTech Connect (OSTI)

    Cutshall, N.H.; Vaughan, N.D.; Haase, C.S.; Olsen, C.R.; Huff, D.D.

    1982-01-01

    Hydrologic and geologic site characterization is a critical phase in development of shallow land-burial sites for low-level radioactive-waste disposal, especially in humid environments. Structural features such as folds, faults, and bedding and textural features such as formation permeability, porosity, and mineralogy all affect the water balance and water movement and, in turn, radionuclide migration. Where these features vary over short distance scales, detailed mapping is required in order to enable accurate model predictions of site performance and to provide the basis for proper design and planning of site-disposal operations.

  14. Low-level radioactive waste (LLW) management at the Nevada Test Site (NTS)

    SciTech Connect (OSTI)

    Becker, B.D.; Gertz, C.P.; Clayton, W.A.; Crowe, B.M.

    1998-12-31

    In 1978, the Department of Energy, Nevada Operations Office (DOE/NV), established a managed LLW disposal project at the Nevada Test Site (NTS). Two, sites which were already accepting limited amounts of on-site generated waste for disposal and off-site generated Transuranic Waste for interim storage, were selected to house the disposal facilities. In those early days, these sites, located about 15 miles apart, afforded the DOE/NV the opportunity to use at least two technologies to manage its waste cost effectively. The Area 5 Radioactive Waste Management Site (RWMS) uses engineered shallow-land burial cells to dispose packaged waste while the Area 3 RWMS uses subsidence craters formed from underground testing of nuclear weapons for the disposal of packaged and unpackaged bulk waste. The paper describes the technical attributes of both Area 5 and Area 3 facilities, the acceptance process, the disposal processes, and present and future capacities of both sites.

  15. Greater-than-Class C low-level radioactive waste characterization. Appendix E-2: Mixed GTCC LLW assessment

    SciTech Connect (OSTI)

    Kirner, N.P. [Ebasco Environmental, Idaho Falls, ID (United States)

    1994-09-01

    Mixed greater-than-Class C low-level radioactive waste (mixed GTCC LLW) is waste that combines two characteristics: it is radioactive, and it is hazardous. This report uses information compiled from Greater-Than-Class C Low-Level Radioactive Waste Characterization: Estimated Volumes, Radionuclide Activities, and Other Characteristics (DOE/LLW 1 14, Revision 1), and applies it to the question of how much and what types of mixed GTCC LLW are generated and are likely to require disposal in facilities jointly regulated by the DOE and the NRC. The report describes how to classify a RCRA hazardous waste, and then applies that classification process to the 41 GTCC LLW waste types identified in the DOE/LLW-114 (Revision 1). Of the 41 GTCC LLW categories identified, only six were identified in this study as potentially requiring regulation as hazardous waste under RCRA. These wastes can be combined into the following three groups: fuel-in decontamination resins, organic liquids, and process waste consisting of lead scrap/shielding from a sealed source manufacturer. For the base case, no mixed GTCC LLW is expected from nuclear utilities or sealed source licensees, whereas only 177 ml of mixed GTCC LLW are expected to be produced by other generators through the year 2035. This relatively small volume represents approximately 40% of the base case estimate for GTCC wastes from other generators. For these other generators, volume estimates for mixed GTCC LLW ranged from less than 1 m{sup 3} to 187 m{sup 3}, depending on assumptions and treatments applied to the wastes.

  16. Radioactive Waste Management Manual

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1999-07-09

    This Manual further describes the requirements and establishes specific responsibilities for implementing DOE O 435.1, Radioactive Waste Management, for the management of DOE high-level waste, transuranic waste, low-level waste, and the radioactive component of mixed waste. Change 1 dated 6/19/01 removes the requirement that Headquarters is to be notified and the Office of Environment, Safety and Health consulted for exemptions for use of non-DOE treatment facilities. Certified 1-9-07. Admin Chg 2, dated 6-8-11, supersedes DOE M 435.1-1 Chg 1.

  17. Summary of expenditures of rebates from the DOE low-level radioactive waste surcharge escrow account for calendar year 1986

    SciTech Connect (OSTI)

    Not Available

    1987-06-01

    The Low-Level Radioactive Waste Policy Amendments Act of 1985, Public Law 99-240, requires the Department of Energy (DOE) to manage an escrow account creatd by collection of 25% of the non-penalty surcharge fees paid by the generators in non-sited regions and nonmember states to sited states for disposal of low-level radioactive waste. For the milestone period ending June 30, 1986, a total of $921,807.84, representing surcharge fees collected and interest earned, was in escrow during 1986 for rebate to the nonmember states, non-sited compact regions, and sited states. As of December 31, 1986, $802,194.54 had been rebated from the Escrow Account with an additional $118,517.62 scheduled for rebate in early 1987. The remaining rebate to be disbursed under this milestone is $1,095.68 for the state of Delaware. At the request of the state of Delaware, this rebate amount is being held in the Escrow Account until the state provides specific instructions for its disbursement. Individual rebate expenditure reports were submitted to DOE by all the non-sited compact regions and nonmember states that received rebates in 1986. Only $14.00 of these rebates were expended in 1986. DOE reviewed all of these reports and concluded that the single expenditure complies with the expenditure limitations stated in the Act.

  18. DOE/EIS-0287 Idaho High-Level Waste & Facilities Disposition...

    Office of Environmental Management (EM)

    ... Prevention of Significant Deterioration regulations require that agencies evaluate new projects to see if they increase air pollution levels. These regulations apply to radioactive ...

  19. In-tank pretreatment of high-level tank wastes: The SIPS system

    SciTech Connect (OSTI)

    Reich, M.; Powell, J.; Barletta, R.

    1996-03-01

    A new approach, termed SIPS (Small In-Tank Processing System), that enables the in-tank processing and separation of high-level tank wastes into high-level waste (HLW) and low-level waste (LLW) streams that are suitable for vitrification, is described. Presently proposed pretreatment systems, such as enhanced sludge washing (ESW) and TRUEX, require that the high-level tank wastes be retrieved and pumped to a large, centralized processing facility, where the various waste components are separated into a relatively small, radioactively concentrated stream (HLW), and a relatively large, predominantly non-radioactive stream (LLW). In SIPS, a small process module, typically on the order of 1 meter in diameter and 4 meters in length, is inserted into a tank. During a period of approximately six months, it processes the solid/liquid materials in the tank, separating them into liquid HLW and liquid LLW output streams that are pumped away in two small diameter (typically 3 cm o.d.) pipes. The SIPS concept appears attractive for pretreating high level wastes, since it would: (1) process waste in-situ in the tanks, (2) be cheaper and more reliable than a larger centralized facility, (3) be quickly demonstrable at full scale, (4) have less technical risk, (5) avoid having to transfer unstable slurries for long distances, and (6) be simple to decommission and dispose of. Further investigation of the SIPS concept appears desirable, including experimental testing and development of subscale demonstration units.

  20. Determination of low-level radioactivity in environmental samples by gamma spectroscopy at Argonne National Laboratory

    SciTech Connect (OSTI)

    Streets, W.E.; Heinrich, R.R.; Lamoureux, L.L.

    1988-01-01

    We currently have six Ge/Ge(Li) detectors that are being used for gamma counting of environmental samples (three with horizontal geometry, three with vertical geometry). The detectors were calibrated for close-geometry efficiency by counting standards in a set configuration (a height of 6.4 cm in a 4-oz. wide-mouthed Naglene bottle) immediately adjacent to the cryostat face. This configuration was chosen so that the standards would be symmetrical to the centers of both the horizontal and vertical detectors. Two solid standards were prepared by adding a known amount of Standard Reference Material NBL No. 6-A (Pitchblende) and NBL No. 7-A (Monazite Sand) to a blanked soil and mixing. Homogeneity of the standards was checked by counting each standard at each quadrant (on a horizontal detector). The final mixtures showed less than 1% deviation between the high and low quadrant counts. Two liquid secondary standards were prepared from stock solutions of /sup 137/Cs, /sup 131/I, and /sup 110m/Ag, which has been characterized as point sources using several detector efficiency curves. These efficiencies were determined using point source standards from the National Bureau of Standards (NBS) and International Atomic Energy Agency (IAEA). All standards had activity levels that allowed less than 1% counting statistics to be obtained on the major peaks (i.e., those with stronger branching ratio) within two hours. Analysis of the resulting data yield smooth efficiency curves for each of the six detectors. Although the standard compositions varied, solid and liquid, the densities were all approx. =1.0 g/cm/sup 3/. 3 refs.

  1. Assessing Potential Exposure from Truck Transport of Low-level Radioactive Waste to the Nevada Test Site

    SciTech Connect (OSTI)

    J. Miller; D. Shafer; K. Gray; B. Church; S. Campbell; B. Holz

    2005-08-01

    Since 1980, over 651,558 m{sup 3} (23,000,000 ft{sup 3}) of low-level radioactive waste (LLW) have been disposed of at the Nevada Test Site (NTS) by shallow land burial. Since 1988, the majority of this waste has been generated at other United States (U.S.) Department of Energy (DOE) and Department of Defense (DoD) sites and facilities in the U.S. Between fiscal year (FY) 2002 and the publication date, the volumes of LLW being shipped by truck to the NTS increased sharply with the accelerated closure of DOE Environmental Management (EM) Program sites (DOE, 2002). The NTS is located 105 km (65 mi) northwest of Las Vegas, Nevada, in the U.S. There continue to be public concerns over the safety of LLW shipments to the NTS. They can be broadly divided into two categories: (1) the risk of accidents involving trucks traveling on public highways; and (2) whether residents along transportation routes receive cumulative exposure from individual LLW shipments that pose a long-term health risk. The DOE and U.S. Department of Transportation (DOT) regulations ensure that radiation exposure from truck shipments to members of the public is negligible. Nevertheless, particularly in rural communities along transportation routes in Utah and Nevada, there is a perceived risk from members of the public about cumulative exposure, particularly when ''Main Street'' and the routes being used by LLW trucks are one in the same. To provide an objective assessment of gamma radiation exposure to members of the public from LLW transport by truck, the Desert Research Institute (DRI) and the DOE, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) established a stationary and automated array of four pressurized ion chambers (PICs) in a vehicle pullout for LLW trucks to pass through just outside the entrance to the NTS. The PICs were positioned at a distance of 1.0 m (3.3 ft) from the sides of the truck trailer and at a height of 1.5 m (5.0 ft) to simulate conditions that a member of the public (Turner, 1995) might experience if a truck were to pass while the person was on the side of the road, or if a truck were to come to a stop at a stoplight in one of the smaller towns along the transportation routes. The 1.0-m (3.3-ft) distance also allowed for comparison with gamma readings of trucks taken with portable, hand-held instruments at the two LLW disposal sites at the NTS: the Area 5 Radioactive Waste Management Complex (RWMC) and the Area 3 Radioactive Waste Management Site (RWMS). The purpose in automating the system was to provide the most objective and consistent measurement and calculation of radiation exposure from the trucks possible. The array was set up in November 2002 and equipment was tested and calibrated over the next two months. Data collection on trucks began on February 13, 2003, and continued to the end of December 2003. In all, external gamma readings were collected from 1,012 of the 2,260 trucks that delivered LLW to the NTS during this period. Because DOE could not contractually require waste generators to participate in the study, the database is biased toward voluntary participants; however, data were collected from the 10 generators that represented 92 percent of the LLW shipments to the NTS during the study period, with another eight generators accounting for the balance of the shipments. Because of the voluntary nature of the participation, the identity of the waste generators is not used in the report. Previous studies on potential exposure to the public from transporting LLW to the NTS either relied on calculated exposures (Davis et al., 2002) or was based on a small population of trucks (e.g., 88) where a relatively high-background value of 50 microRoentgens per hour (R/h) (background value measured at the LLW disposal sites) were subtracted from the gross reading of the truck trailer as measured by portable, handheld instruments (Gertz, 2001). The dataset that resulted from the DRI study is the largest collection of measurements of LLW trucks in transit of which the authors are aware.

  2. Assessing Potential Exposure from Truck Transport of Low-level Radioactive Waste to the Nevada Test Site

    SciTech Connect (OSTI)

    Miller, J; Shafer, D; Gray, K; Church, B; Campbell, S; Holtz, B.

    2005-08-15

    Since 1980, over 651,558 m{sup 3} (23,000,000 ft{sup 3}) of low-level radioactive waste (LLW) have been disposed of at the Nevada Test Site (NTS) by shallow land burial. Since 1988, the majority of this waste has been generated at other United States (U.S.) Department of Energy (DOE) and Department of Defense (DoD) sites and facilities in the U.S. Between fiscal year (FY) 2002 and the publication date, the volumes of LLW being shipped by truck to the NTS increased sharply with the accelerated closure of DOE Environmental Management (EM) Program sites (DOE, 2002). The NTS is located 105 km (65 mi) northwest of Las Vegas, Nevada, in the U.S. There continue to be public concerns over the safety of LLW shipments to the NTS. They can be broadly divided into two categories: (1) the risk of accidents involving trucks traveling on public highways; and (2) whether residents along transportation routes receive cumulative exposure from individual LLW shipments that pose a long-term health risk. The DOE and U.S. Department of Transportation (DOT) regulations ensure that radiation exposure from truck shipments to members of the public is negligible. Nevertheless, particularly in rural communities along transportation routes in Utah and Nevada, there is a perceived risk from members of the public about cumulative exposure, particularly when ''Main Street'' and the routes being used by LLW trucks are one in the same. To provide an objective assessment of gamma radiation exposure to members of the public from LLW transport by truck, the Desert Research Institute (DRI) and the DOE, National Nuclear Security Administration Nevada Site Office (NNSA/NSO) established a stationary and automated array of four pressurized ion chambers (PICs) in a vehicle pullout for LLW trucks to pass through just outside the entrance to the NTS. The PICs were positioned at a distance of 1.0 m (3.3 ft) from the sides of the truck trailer and at a height of 1.5 m (5.0 ft) to simulate conditions that a member of the public (Turner, 1995) might experience if a truck were to pass while the person was on the side of the road, or if a truck were to come to a stop at a stoplight in one of the smaller towns along the transportation routes. The 1.0-m (3.3-ft) distance also allowed for comparison with gamma readings of trucks taken with portable, hand-held instruments at the two LLW disposal sites at the NTS: the Area 5 Radioactive Waste Management Complex (RWMC) and the Area 3 Radioactive Waste Management Site (RWMS). The purpose in automating the system was to provide the most objective and consistent measurement and calculation of radiation exposure from the trucks possible. The array was set up in November 2002 and equipment was tested and calibrated over the next two months. Data collection on trucks began on February 13, 2003, and continued to the end of December 2003. In all, external gamma readings were collected from 1,012 of the 2,260 trucks that delivered LLW to the NTS during this period. Because DOE could not contractually require waste generators to participate in the study, the database is biased toward voluntary participants; however, data were collected from the 10 generators that represented 92 percent of the LLW shipments to the NTS during the study period, with another eight generators accounting for the balance of the shipments. Because of the voluntary nature of the participation, the identity of the waste generators is not used in the report. Previous studies on potential exposure to the public from transporting LLW to the NTS either relied on calculated exposures (Davis et al., 2002) or was based on a small population of trucks (e.g., 88) where a relatively high-background value of 50 microRoentgens per hour ({micro}R/h) (background value measured at the LLW disposal sites) were subtracted from the gross reading of the truck trailer as measured by portable, handheld instruments (Gertz, 2001). The dataset that resulted from the DRI study is the largest collection of measurements of LLW trucks in transit of which the authors are aware.

  3. High temperature materials for radioactive waste incineration and vitrification. Revision 1

    SciTech Connect (OSTI)

    Bickford, D F; Ondrejcin, R S; Salley, L

    1986-01-01

    Incineration or vitrification of radioactive waste subjects equipment to alkaline or acidic fluxing, oxidation, sulfidation, carburization, and thermal shock. It is necessary to select appropriate materials of construction and control operating conditions to avoid rapid equipment failure. Nickel- and cobalt-based alloys with high chromium or aluminum content and aluminum oxide/chromium oxide refractories with high chromium oxide content have provided the best service in pilot-scale melter tests. Inconel 690 and Monofrax K-3 are being used for waste vitrification. Haynes 188 and high alumina refractory are undergoing pilot scale tests for incineration equipment. Laboratory tests indicate that alloys and refractories containing still higher concentrations of chromium or chromium oxide, such as Inconel 671 and Monofrax E, may provide superior resistance to attack in glass melter environments.

  4. PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION

    SciTech Connect (OSTI)

    D.C. Richardson

    2003-03-19

    In accordance with the Nuclear Waste Policy Amendments Act of 1987, Yucca Mountain was designated as the site to be investigated as a potential repository for the disposal of high-level radioactive waste. The Yucca Mountain site is an undeveloped area located on the southwestern edge of the Nevada Test Site (NTS), about 100 miles northwest of Las Vegas. The site currently lacks rail service or an existing right-of-way. If the Yucca Mountain site is found suitable for the repository, rail service is desirable to the Office of Civilian Waste Management (OCRWM) Program because of the potential of rail transportation to reduce costs and to reduce the number of shipments relative to highway transportation. A Preliminary Rail Access Study evaluated 13 potential rail spur options. Alternative routes within the major options were also developed. Each of these options was then evaluated for potential land use conflicts and access to regional rail carriers. Three potential routes having few land use conflicts and having access to regional carriers were recommended for further investigation. Figure 1-1 shows these three routes. The Jean route is estimated to be about 120 miles long, the Carlin route to be about 365 miles long, and Caliente route to be about 365 miles long. The remaining ten routes continue to be monitored and should any of the present conflicts change, a re-evaluation of that route will be made. Complete details of the evaluation of the 13 routes can be found in the previous study. The DOE has not identified any preferred route and recognizes that the transportation issues need a full and open treatment under the National Environmental Policy Act. The issue of transportation will be included in public hearings to support development of the Environmental Impact Statement (EIS) proceedings for either the Monitored Retrievable Storage Facility or the Yucca Mountain Project or both.

  5. The consequences of disposal of low-level radioactive waste from the Fernald Environmental Management Project: Report of the DOE/Nevada Independent Panel

    SciTech Connect (OSTI)

    Crowe, B.; Hansen, W.; Waters, R.; Sully, M.; Levitt, D.

    1998-04-01

    The Department of Energy (DOE) convened a panel of independent scientists to assess the performance impact of shallow burial of low-level radioactive waste from the Fernald Environmental Management Project, in light of a transportation incident in December 1997 involving this waste stream. The Fernald waste has been transported to the Nevada Test Site and disposed in the Area 5 Radioactive Waste Management Site (RWMS) since 1993. A separate DOE investigation of the incident established that the waste has been buried in stress-fractured metal boxes, and some of the waste contained excess moisture (high-volumetric water contents). The Independent Panel was charged with determining whether disposition of this waste in the Area 5 RWMS has impacted the conclusions of a previously completed performance assessment in which the site was judged to meet required performance objectives. To assess the performance impact on Area 5, the panel members developed a series of questions. The three areas addressed in these questions were (1) reduced container integrity, (2) the impact of reduced container integrity on subsidence of waste in the disposal pits and (3) excess moisture in the waste. The panel has concluded that there is no performance impact from reduced container integrity--no performance is allocated to the container in the conservative assumptions used in performance assessment. Similarly, the process controlling post-closure subsidence results primarily from void space within and between containers, and the container is assumed to degrade and collapse within 100 years.

  6. Low-level radioactive waste management: transitioning to off-site disposal at Los Alamos National Laboratory

    SciTech Connect (OSTI)

    Dorries, Alison M

    2010-11-09

    Facing the closure of nearly all on-site management and disposal capability for low-level radioactive waste (LLW), Los Alamos National Laboratory (LANL) is making ready to ship the majority of LLW off-site. In order to ship off-site, waste must meet the Treatment, Storage, and Disposal Facility's (TSDF) Waste Acceptance Criteria (WAC). In preparation, LANL's waste management organization must ensure LANL waste generators characterize and package waste compliantly and waste characterization documentation is complete and accurate. Key challenges that must be addressed to successfully make the shift to off-site disposal of LLW include improving the detail, accuracy, and quality of process knowledge (PK) and acceptable knowledge (AK) documentation, training waste generators and waste management staff on the higher standard of data quality and expectations, improved WAC compliance for off-site facilities, and enhanced quality assurance throughout the process. Certification of LANL generators will allow direct off-site shipping of LLW from their facilities.

  7. Environmental monitoring report for commercial low-level radioactive waste disposal sites (1960`s through 1990`s)

    SciTech Connect (OSTI)

    1996-11-01

    During the time period covered in this report (1960`s through early 1990`s), six commercial low-level radioactive waste (LLRW) disposal facilities have been operated in the US. This report provides environmental monitoring data collected at each site. The report summarizes: (1) each site`s general design, (2) each site`s inventory, (3) the environmental monitoring program for each site and the data obtained as the program has evolved, and (4) what the program has indicated about releases to off-site areas, if any, including a statement of the actual health and safety significance of any release. A summary with conclusions is provided at the end of each site`s chapter. The six commercial LLRW disposal sites discussed are located near: Sheffield, Illinois; Maxey Flats, Kentucky; Beatty, Nevada; West Valley, New York; Barnwell, South Carolina; Richland, Washington.

  8. Radiological Survey of Contaminated Installations of Research Reactor before Dismantling in High Dose Conditions with Complex for Remote Measurements of Radioactivity - 12069

    SciTech Connect (OSTI)

    Danilovich, Alexey; Ivanov, Oleg; Lemus, Alexey; Smirnov, Sergey; Stepanov, Vyacheslav; Volkovich, Anatoly

    2012-07-01

    Decontamination and decommissioning of the research reactors MR (Testing Reactor) and RFT (Reactor of Physics and Technology) has recently been initiated in the National Research Center (NRC) 'Kurchatov institute', Moscow. These research reactors have a long history and many installations - nine loop facilities for experiments with different kinds of fuel. When decommissioning nuclear facilities it is necessary to measure the distribution of radioactive contamination in the rooms and at the equipment at high levels of background radiation. At 'Kurchatov Institute' some special remote control measuring systems were developed and they are applied during dismantling of the reactors MR and RFT. For a survey of high-level objects a radiometric system mounted on the robotic Brokk vehicle is used. This system has two (4? and collimated) dose meters and a high resolution video camera. Maximum measured dose rate for this system is ?8.5 Sv/h. To determine the composition of contaminants, a portable spectrometric system is used. It is a remotely controlled, collimated detector for scanning the distribution of radioactive contamination. To obtain a detailed distribution of contamination a remote-controlled gamma camera is applied. For work at highly contaminated premises with non-uniform background radiation, another camera is equipped with rotating coded mask (coded aperture imaging). As a result, a new system of instruments for remote radioactivity measurements with wide range of sensitivity and angular resolution was developed. The experience and results of measurements in different areas of the reactor and at its loop installations, with emphasis on the radioactive survey of highly-contaminated samples, are presented. These activities are conducted under the Federal Program for Nuclear and Radiation Safety of Russia. Adaptation of complex remote measurements of radioactivity and survey of contaminated installations of research reactor before dismantling in high dose conditions has proven successful. The radioactivity measuring devices for operation at high, non-uniform dose background were tested in the field and a new data of measurement of contamination distribution in the premises and installations were obtained. (authors)

  9. Summary of expenditures of rebates from the low-level radioactive waste surcharge escrow account for calendar year 1995

    SciTech Connect (OSTI)

    1996-06-01

    This report is submitted in response to Title 1 of the 1980 Low-Level Radioactive Waste Policy Act, as amended, (the Act). The report summarizes expenditures made by compact regions and unaffiliated states during calendar year 1995 of surcharge rebates from the July 1, 1986, January 1, 1988, and January 1, 1990, milestones, and the January 1, 1993, deadline. Section 5(d)(2)(A) of the Act requires the Department of Energy (DOE) to administer a surcharge escrow account. This account consists of a portion of the surcharge fees paid by generators of low-level radioactive waste in nonsited compact regions (compact regions currently without disposal sites) and nonmember states (states without disposal sites that are not members of compact regions) to the three sited states (states with operating disposal facilities--Nevada, South Carolina, and Washington) for the use of facilities in sited states through the end of 1992. In administering the surcharge escrow account, the Act requires DOE to: (1) Invest the funds in interest-bearing United States Government securities with the highest available yield; (2) Determine eligibility for rebates of the funds by evaluating compact region and state progress toward developing new disposal sites against the milestone requirements set forth in the Act; (3) Disburse the collected rebates and accrued interest to eligible compact regions, states, or generators; (4) Assess compliance of rebate expenditures in accordance with the conditions and limitations prescribed in the Act; and (5) Submit a report annually to Congress summarizing rebate expenditures by state and compact region and assessing the compliance of each such state or compact region with the requirement for expenditure of the rebates as provided in section 5(d)(2)(E) of the Act.

  10. Remote Sensing Analysis of the Sierra Blanca (Faskin Ranch) Low-Level Radioactive Waste Disposal Site, Hudspeth County, Texas

    SciTech Connect (OSTI)

    LeMone, D. V.; Dodge, R.; Xie, H.; Langford, R. P.; Keller, G. R.

    2002-02-26

    Remote sensing images provide useful physical information, revealing such features as geological structure, vegetation, drainage patterns, and variations in consolidated and unconsolidated lithologies. That technology has been applied to the failed Sierra Blanca (Faskin Ranch) shallow burial low-level radioactive waste disposal site selected by the Texas Low-Level Radioactive Waste Disposal Authority. It has been re-examined using data from LANDSAT satellite series. The comparison of the earlier LANDSAT V (5/20/86) (30-m resolution) with the later new, higher resolution ETM imagery (10/23/99) LANDSAT VII data (15-m resolution) clearly shows the superiority of the LANDSAT VII data. The search for surficial indications of evidence of fatal flaws at the Sierra Blanca site utilizing was not successful, as it had been in the case of the earlier remote sensing analysis of the failed Fort Hancock site utilizing LANDSAT V data. The authors conclude that the tectonic activity at the Sierra Blanca site is much less recent and active than in the previously studied Fort Hancock site. The Sierra Blanca site failed primarily on the further needed documentation concerning a subsurface fault underneath the site and environmental justice issues. The presence of this fault was not revealed using the newer LANDSAT VII data. Despite this fact, it must be remembered that remote sensing provides baseline documentation for determining future physical and financial remediation responsibilities. On the basis of the two sites examined by LANDSAT remote sensing imaging, it is concluded that it is an essential, cost-effective tool that should be utilized not only in site examination but also in all nuclear-related facilities.

  11. A strategy for resolving high-priority Hanford Site radioactive waste storage tank safety issues

    SciTech Connect (OSTI)

    Babad, H.; DeFigh-Price, C.; Fulton, J.C.

    1993-02-01

    High-activity radioactive waste has been stored in large underground storage tanks at the US Department of Energy`s (DOE) Hanford Site in Eastern Washington State since 1944. Since then, more than 227,000 m{sup 3} (60 Mgal) of waste have been accumulated in 177 tanks. These caustic wastes consist of many different chemicals. The waste forms include liquids, slurries, salt cakes, and sludges. A number of safety issues have been raised about these wastes, and resolution of these issues is a top priority of DOE. A Waste Tank Safety Program has been established to resolve these high-priority safety issues. This paper will deal with three of these issues. The issues described are the release of flammable vapors from single- and double-shell tanks, the existence of organic chemicals, and/or ferrocyanide ion-containing fuel-rich mixtures of nitrate and nitrite salts in single-shell tanks.

  12. Radionuclide Concentration in Soils and Vegetation at Low-Level Radioactive Waste Disposal Area G during 2005

    SciTech Connect (OSTI)

    P.R. Fresquez; M.W. McNaughton; M.J. Winch

    2005-10-01

    Soil samples were collected at 15 locations and unwashed overstory and understory vegetation samples were collected from up to nine locations within and around the perimeter of Area G, the primary disposal facility for low-level radioactive solid waste at Los Alamos National Laboratory (LANL). Soil and plant samples were also collected from the proposed expansion area west of Area G for the purpose of gaining preoperational baseline data. Soil and plant samples were analyzed for radionuclides that have shown a history of detection in past years; these included {sup 3}H, {sup 238}Pu, {sup 239,240}Pu, {sup 241}Am, {sup 234}U, {sup 235}U, and {sup 238}U for soils and {sup 3}H, {sup 238}Pu, and {sup 239,240}Pu for plants. As in previous years, the highest levels of {sup 3}H in soils and vegetation were detected at the south portion of Area G near the {sup 3}H shafts; whereas, the highest concentrations of the Pu isotopes were detected in the northern and northeastern portions near the pads for transuranic waste. All concentrations of radionuclides in soils and vegetation, however, were still very low (pCi range) and far below LANL screening levels and regulatory standards.

  13. Management of radioactive mixed wastes in commercial low-level wastes

    SciTech Connect (OSTI)

    Kempf, C.R.; MacKenzie, D.R.; Piciulo, P.L.; Bowerman, B.S.; Siskind, B.

    1986-01-01

    Potential mixed wastes in commercial low-level wastes have been identified and management options applicable to these wastes have been evaluated. Both the identification and management evaluation have necessarily been based on review of NRC and EPA regulations and recommendations. The underlying intent of both agencies is protection of man and/or environment, but differences may occur in the means by which intent is achieved. Apparent discrepancies, data gaps and unresolved issues that have surfaced during the course of this work are discussed.

  14. Greater-than-Class C low-level radioactive waste characterization: Estimated volumes, radionuclide activities, and other characteristics. Revision 1

    SciTech Connect (OSTI)

    Not Available

    1994-09-01

    The Department of Energy`s (DOE`s) planning for the disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of the waste. This report estimates volumes, radionuclide activities, and waste forms of GTCC LLW to the year 2035. It groups the waste into four categories, representative of the type of generator or holder of the waste: Nuclear Utilities, Sealed Sources, DOE-Held, and Other Generator. GTCC LLW includes activated metals (activation hardware from reactor operation and decommissioning), process wastes (i.e., resins, filters, etc.), sealed sources, and other wastes routinely generated by users of radioactive material. Estimates reflect the possible effect that packaging and concentration averaging may have on the total volume of GTCC LLW. Possible GTCC mixed LLW is also addressed. Nuclear utilities will probably generate the largest future volume of GTCC LLW with 65--83% of the total volume. The other generators will generate 17--23% of the waste volume, while GTCC sealed sources are expected to contribute 1--12%. A legal review of DOE`s obligations indicates that the current DOE-Held wastes described in this report will not require management as GTCC LLW because of the contractual circumstances under which they were accepted for storage. This report concludes that the volume of GTCC LLW should not pose a significant management problem from a scientific or technical standpoint. The projected volume is small enough to indicate that a dedicated GTCC LLW disposal facility may not be justified. Instead, co-disposal with other waste types is being considered as an option.

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

    SciTech Connect (OSTI)

    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.

  16. High Level Overview of DOE Biomass Logistics II Project Activities...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Level Overview of DOE Biomass Logistics II Project Activities High Level Overview of DOE Biomass Logistics II Project Activities Breakout Session 1B-Integration of Supply Chains I: ...

  17. EIS-0375: Disposal of Greater-than-Class-C Low-Level Radioactive Waste and Department of Energy GTCC-like Waste

    Broader source: Energy.gov [DOE]

    This EIS evaluates the reasonably foreseeable environmental impacts associated with the proposed development, operation, and long-term management of a disposal facility or facilities for Greater-Than-Class C (GTCC) low-level radioactive waste and GTCC-like waste. The Environmental Protection Agency is a cooperating agency in the preparation of this EIS.

  18. DOE issues Finding of No Significant Impact on Environmental Assessment for Replacement Capability for Disposal of Remote-Handled Low Level Radioactive Waste Generated at Idaho Site

    Broader source: Energy.gov [DOE]

    Idaho Falls, ID – After completing a careful assessment, the U.S. Department of Energy has determined that building a new facility at its Idaho National Laboratory site for continued disposal of remote-handled low level radioactive waste generated by operations at the site will not have a significant impact on the environment.

  19. Draft Environmental Impact Statement for the Disposal of Greater-Than-Class C Low-Level Radioactive Waste and GTCC-Like Waste

    Broader source: Energy.gov [DOE]

    WASHINGTON – The Department of Energy (DOE) has issued a Draft Environmental Impact Statement for the Disposal of Greater-Than-Class C (GTCC) Low-Level Radioactive Waste (LLRW) and GTCC-Like Waste (DOE/EIS-0375D, Draft EIS) as required under the National Environmental Policy Act for public review and comment.

  20. Greater-than-Class C low-level radioactive waste characterization: Estimated volumes, radionuclide activities, and other characteristics

    SciTech Connect (OSTI)

    Hulse, R.A.

    1991-08-01

    Planning for storage or disposal of greater-than-Class C low-level radioactive waste (GTCC LLW) requires characterization of that waste to estimate volumes, radionuclide activities, and waste forms. Data from existing literature, disposal records, and original research were used to estimate the characteristics and project volumes and radionuclide activities to the year 2035. GTCC LLW is categorized as: nuclear utilities waste, sealed sources waste, DOE-held potential GTCC LLW; and, other generator waste. It has been determined that the largest volume of those wastes, approximately 57%, is generated by nuclear power plants. The Other Generator waste category contributes approximately 10% of the total GTCC LLW volume projected to the year 2035. Waste held by the Department of Energy, which is potential GTCC LLW, accounts for nearly 33% of all waste projected to the year 2035; however, no disposal determination has been made for that waste. Sealed sources are less than 0.2% of the total projected volume of GTCC LLW.

  1. Hydrologic evaluation methodology for estimating water movement through the unsaturated zone at commercial low-level radioactive waste disposal sites

    SciTech Connect (OSTI)

    Meyer, P.D.; Rockhold, M.L.; Nichols, W.E.; Gee, G.W. [Pacific Northwest Lab., Richland, WA (United States)

    1996-01-01

    This report identifies key technical issues related to hydrologic assessment of water flow in the unsaturated zone at low-level radioactive waste (LLW) disposal facilities. In addition, a methodology for incorporating these issues in the performance assessment of proposed LLW disposal facilities is identified and evaluated. The issues discussed fall into four areas: estimating the water balance at a site (i.e., infiltration, runoff, water storage, evapotranspiration, and recharge); analyzing the hydrologic performance of engineered components of a facility; evaluating the application of models to the prediction of facility performance; and estimating the uncertainty in predicted facility performance. To illustrate the application of the methodology, two examples are presented. The first example is of a below ground vault located in a humid environment. The second example looks at a shallow land burial facility located in an arid environment. The examples utilize actual site-specific data and realistic facility designs. The two examples illustrate the issues unique to humid and arid sites as well as the issues common to all LLW sites. Strategies for addressing the analytical difficulties arising in any complex hydrologic evaluation of the unsaturated zone are demonstrated.

  2. Analysis of the suitability of DOE facilities for treatment of commercial low-level radioactive mixed waste

    SciTech Connect (OSTI)

    1996-02-01

    This report evaluates the capabilities of the United States Department of Energy`s (DOE`s) existing and proposed facilities to treat 52 commercially generated low-level radioactive mixed (LLMW) waste streams that were previously identified as being difficult-to-treat using commercial treatment capabilities. The evaluation was performed by comparing the waste matrix and hazardous waste codes for the commercial LLMW streams with the waste acceptance criteria of the treatment facilities, as identified in the following DOE databases: Mixed Waste Inventory Report, Site Treatment Plan, and Waste Stream and Technology Data System. DOE facility personnel also reviewed the list of 52 commercially generated LLMW streams and provided their opinion on whether the wastes were technically acceptable at their facilities, setting aside possible administrative barriers. The evaluation tentatively concludes that the DOE is likely to have at least one treatment facility (either existing or planned) that is technically compatible for most of these difficult-to-treat commercially generated LLMW streams. This conclusion is tempered, however, by the limited amount of data available on the commercially generated LLMW streams, by the preliminary stage of planning for some of the proposed DOE treatment facilities, and by the need to comply with environmental statutes such as the Clean Air Act.

  3. Alternative Chemical Cleaning Methods for High Level Waste Tanks: Simulant Studies

    SciTech Connect (OSTI)

    Rudisill, T.; King, W.; Hay, M.; Jones, D.

    2015-11-19

    Solubility testing with simulated High Level Waste tank heel solids has been conducted in order to evaluate two alternative chemical cleaning technologies for the dissolution of sludge residuals remaining in the tanks after the exhaustion of mechanical cleaning and sludge washing efforts. Tests were conducted with non-radioactive pure phase metal reagents, binary mixtures of reagents, and a Savannah River Site PUREX heel simulant to determine the effectiveness of an optimized, dilute oxalic/nitric acid cleaning reagent and pure, dilute nitric acid toward dissolving the bulk non-radioactive waste components. A focus of this testing was on minimization of oxalic acid additions during tank cleaning. For comparison purposes, separate samples were also contacted with pure, concentrated oxalic acid which is the current baseline chemical cleaning reagent. In a separate study, solubility tests were conducted with radioactive tank heel simulants using acidic and caustic permanganate-based methods focused on the “targeted” dissolution of actinide species known to be drivers for Savannah River Site tank closure Performance Assessments. Permanganate-based cleaning methods were evaluated prior to and after oxalic acid contact.

  4. Technology and System Level Demonstration of Highly Efficient...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Evaluation Meeting arravt081vssnewhouse2012o.pdf More Documents & Publications Technology and System Level Demonstration of Highly Efficient and Clean, Diesel Powered Class 8...

  5. Characterization of High Level Waste from a Hybrid LIFE Engine...

    Office of Scientific and Technical Information (OSTI)

    Title: Characterization of High Level Waste from a Hybrid LIFE Engine for Enhanced Repository Performance Authors: Beckett, E ; Fratoni, M Publication Date: 2010-08-25 OSTI ...

  6. High Level Computational Chemistry Approaches to the Prediction...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Energetic Properties of Chemical Hydrogen Storage Systems High Level Computational Chemistry Approaches to the Prediction of Energetic Properties of Chemical Hydrogen Storage ...

  7. Idaho High-Level Waste & Facilities Disposition, Final Environmental...

    Office of Environmental Management (EM)

    must prepare an Environmental Impact Statement (EIS). Copies of the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement are available at the...

  8. High Level Waste Corporate Board Newsletter - 06/03/08

    Office of Environmental Management (EM)

    2008 UPCOMING EVENTS: Next High-Level Waste Corporate Board meeting will be held at ... Needs Collection Prioritization * Waste Acceptance Product Specification This ...

  9. Qualification of Innovative High Level Waste Pipeline Unplugging Technologies

    SciTech Connect (OSTI)

    McDaniel, D.; Gokaltun, S.; Varona, J.; Awwad, A.; Roelant, D.; Srivastava, R.

    2008-07-01

    In the past, some of the pipelines have plugged during high level waste (HLW) transfers resulting in schedule delays and increased costs. Furthermore, pipeline plugging has been cited by the 'best and brightest' technical review as one of the major issues that can result in unplanned outages at the Waste Treatment Plant causing inconsistent operation. As the DOE moves toward a more active high level waste retrieval, the site engineers will be faced with increasing cross-site pipeline waste slurry transfers that will result in increased probability of a pipeline getting plugged. Hence, availability of a pipeline unplugging tool/technology is crucial to ensure smooth operation of the waste transfers and in ensuring tank farm cleanup milestones are met. FIU had earlier tested and evaluated various unplugging technologies through an industry call. Based on mockup testing, two technologies were identified that could withstand the rigors of operation in a radioactive environment and with the ability to handle sharp 90 elbows. We present results of the second phase of detailed testing and evaluation of pipeline unplugging technologies and the objective is to qualify these pipeline unplugging technologies for subsequent deployment at a DOE facility. The current phase of testing and qualification comprises of a heavily instrumented 3-inch diameter (full-scale) pipeline facilitating extensive data acquisition for design optimization and performance evaluation, as it applies to three types of plugs atypical of the DOE HLW waste. Furthermore, the data from testing at three different lengths of pipe in conjunction with the physics of the process will assist in modeling the unplugging phenomenon that will then be used to scale-up process parameters and system variables for longer and site typical pipe lengths, which can extend as much as up to 19,000 ft. Detailed information resulting from the testing will provide the DOE end-user with sufficient data and understanding of the technology, and its limitations to aid in the benefit-cost analysis for management decision whether to deploy the technology or to abandon the pipeline as has been done in the past. In conclusion: The ultimate objective of this study is to qualify NuVision's unplugging technology for use at Hanford. Experimental testing has been conducted using three pipeline lengths and three types of blockages. Erosion rates have been obtained and pressure data is being analyzed. An amplification of the inlet pressure has been observed along the pipeline and is the key to determining up to what pipe lengths the technology can be used without surpassing the site pressure limit. In addition, we will attempt to establish what the expected unplugging rates will be at the longer pipe lengths for each of the three blockages tested. Detailed information resulting from the testing will provide the DOE end-user with sufficient data and understanding of the technology, and its limitations so that management decisions can be made whether the technology has a reasonable chance to successfully unplug a pipeline, such as a cross site transfer line or process transfer pipeline at the Waste Treatment Plant. (authors)

  10. Cold Crucible Induction Melting Technology for Vitrification of High Level Waste: Development and Status in India

    SciTech Connect (OSTI)

    Sugilal, G.; Sengar, P.B.S. [Nuclear Recycle Group, Bhabha Atomic Research Centre, Trombay, Mumbai (India)

    2008-07-01

    Cold crucible induction melting is globally emerging as an alternative technology for the vitrification of high level radioactive waste. The new technology offers several advantages such as high temperature availability with long melter life, high waste loading, high specific capacity etc. Based on the laboratory and bench scale studies, an engineering scale cold crucible induction melter was locally developed in India. The melter was operated continuously to assess its performance. The electrical and thermal efficiencies were found to be in the range of 70-80 % and 10-20 % respectively. Glass melting capacities up to 200 kg m{sup -2} hr{sup -1} were accomplished using the ESCCIM. Industrially adaptable melter operating procedures for start-up, melting and pouring operations were established (author)

  11. Spray Calciner/In-Can Melter high-level waste solidification technical manual

    SciTech Connect (OSTI)

    Larson, D.E.

    1980-09-01

    This technical manual summarizes process and equipment technology developed at Pacific Northwest Laboratory over the last 20 years for vitrification of high-level liquid waste by the Spray Calciner/In-Can Melter process. Pacific Northwest Laboratory experience includes process development and demonstration in laboratory-, pilot-, and full-scale equipment using nonradioactive synthetic wastes. Also, laboratory- and pilot-scale process demonstrations have been conducted using actual high-level radioactive wastes. In the course of process development, more than 26 tonnes of borosilicate glass have been produced in 75 canisters. Four of these canisters contained radioactive waste glass. The associated process and glass chemistry is discussed. Technology areas described include calciner feed treatment and techniques, calcination, vitrification, off-gas treatment, glass containment (the canister), and waste glass chemistry. Areas of optimization and site-specific development that would be needed to adapt this base technology for specific plant application are indicated. A conceptual Spray Calciner/In-Can Melter system design and analyses are provided in the manual to assist prospective users in evaluating the process for plant application, to provide equipment design information, and to supply information for safety analyses and environmental reports. The base (generic) technology for the Spray Calciner/In-Can Melter process has been developed to a point at which it is ready for plant application.

  12. ASSESSING EXPOSURE TO THE PUBLIC FROM LOW LEVEL RADIOACTIVE WASTE (LLW) TRANSPORTATION TO THE NEVADA TEST SITE.

    SciTech Connect (OSTI)

    Miller, J.J.; Campbell, S.; Church, B.W.; Shafer, D. S.; Gillespie, D.; Sedano, S.; Cebe, J.J.

    2003-02-27

    The United States (U.S.) Department of Energy (DOE) Nevada Test Site (NTS) is one of two regional sites where low-level radioactive waste (LLW) from approved DOE and U.S. DOD generators across the United States is disposed. In federal fiscal year (FY) 2002, over 57,000 cubic meters of waste was transported to and disposed at the NTS. DOE and U.S. Department of Transportation (DOT) regulations ensure that radiation exposure from truck shipments to members of the public is negligible. Nevertheless, particularly in rural communities along transportation routes in Utah and Nevada, there is perceived risk from members of the public about incremental exposure from LLW trucks, especially when ''Main Street'' and the LLW transportation route are the same. To better quantify the exposure to gamma radiation, a stationary monitoring array of four pressurized ion chambers (PICs) have been set up in a pullout just before LLW trucks reach the entrance to the NTS. The PICs are positioned at a distance of one meter from the sides of the truck trailer and at a height appropriate for the design of the trucks that will be used in FY2003 to haul LLW to the NTS. The use of four PICs (two on each side of the truck) is to minimize and to correct for non-uniformity where radiation levels from waste packages vary from side to side, and from front to back in the truck trailer. The PIC array is being calibrated by collecting readings from each PIC exposed to a known 137Cs source that was positioned at different locations on a flatbed stationed in the PIC array, along with taking secondary readings from other known sources. Continuous data collection using the PICs, with and without a truck in the array, is being used to develop background readings. In addition, acoustic sensors are positioned on each side of the PIC array to record when a large object (presumably a truck) enters the array. In FY2003, PIC surveys from as many incoming LLW trucks as possible will be made and survey data recorded automatically by dataloggers that will be periodically downloaded. Solar panels provide power for the batteries to run both the dataloggers and PICs. Truck drivers have been asked to park their truck within the PIC array for only the time it takes to complete an information log before moving on to one of two Radioactive Waste Management Sites (RWMS) on the NTS. On the log, the truck drivers record their shipment identification number, the time of day, where the waste originated, and information on the route they used to reach the NTS. This data will facilitate comparison of PIC readings with waste manifests and other waste disposal operations data collected at the RWMSs. Gamma radiation measurements collected from the PICs will be analyzed using standard health physics and statistical methods for comparison to DOT standards, but with the added benefit of obtaining an improved understanding of the variability of readings that can occur in the near vicinity of a LLW truck. The data collected will be combined with measurements of street width and other information about transportation routes through towns to develop realistic dose scenarios for citizens in Nevada and Utah towns.

  13. Geochemical information for sites contaminated with low-level radioactive wastes: II. St. Louis Airport Storage Site

    SciTech Connect (OSTI)

    Seeley, F.G.; Kelmers, A.D.

    1985-01-01

    The St. Louis Airport Storage Site (SLASS) became radioactively contaminated as a result of wastes that were being stored from operations to recover uranium from pitchblende ores in the 1940s and 1950s. The US Department of Energy is considering various remedial action options for the SLASS under the Formerly Utilized Site Remedial Action Program (FUSRAP). This report describes the results of geochemical investigations, carried out to support the FUSRAP activities and to aid in quantifying various remedial action options. Soil and groundwater samples from the site were characterized, and sorption ratios for uranium and radium and apparent concentration limit values for uranium were measured in soil/groundwater systems by batch contact methodology. The uranium and radium concentrations in soil samples were significantly above background near the old contaminated surface horizon (now at the 0.3/sup -/ to 0.9/sup -/m depth); the maximum values were 1566 ..mu..g/g and 101 pCi/g, respectively. Below about the 6/sup -/m depth, the concentrations appeared to be typical of those naturally present in soils of this area (3.8 +- 1.2 ..mu..g/g and 3.1 +- 0.6 pCi/g). Uranium sorption ratios showed stratigraphic trends but were generally moderate to high (100 to 1000 L/kg). The sorption isotherm suggested an apparent uranium concentration limit of about 200 mg/L. This relatively high solubility can probably be correlated with the carbonate content of the soil/groundwater systems. The lower sorption ratio values obtained from the sorption isotherm may have resulted from changes in the experimental procedure or the groundwater used. The SLASS appears to exhibit generally favorable behavior for the retardation of uranium solubilized from waste in the site. Parametric tests were conducted to estimate the sensitivity of uranium sorption and solubility to the pH and carbonate content of the system.

  14. Low-level radioactive waste from nuclear power generating stations: Characterization, classification and assessment of activated metals and waste streams

    SciTech Connect (OSTI)

    Thomas, V.W.; Robertson, D.E.; Thomas, C.W.

    1993-02-01

    Since the enactment of 10 CFR Part 61, additional difficult-to-measure long-lived radionuclides, not specified in Tables 1 2 of Part 61, have been identified (e.g., {sup 108m}Ag, {sup 93}Mo, {sup 36}Cl, {sup 10}Be, {sup 113m}Cd, {sup 121m}Sn, {sup 126}Sn, {sup 93m}Nb) that may be of concern in certain types of waste. These nuclides are primarily associated with activated metal and perhaps other nuclear power low-level waste (LLW) being sent to disposal facilities. The concentration of a radionuclide in waste materials is normally determined by direct measurement or by indirect calculational methods, such as using a scaling factor to relate inferred concentration of a difficult-to-measure radionuclide to another that is easily measured. The total disposal site inventory of certain difficult-to-measure radionuclides (e.g., {sup 14}C, {sup 129}I, and {sup 99}Tc) often control the total quantities of radioactive waste permitted in LLW burial facilities. Overly conservative scaling factors based on lower limits of detection (LLD), often used in the nuclear power industry to estimate these controlling nuclides, could lead to premature closure of a disposal facility. Samples of LLW (Class B and C activated metals [AM] and other waste streams) are being collected from operating nuclear power stations and analyzed for radionuclides covered in 10 CFR Part 61 and the additional difficult-to-measure radionuclides. This analysis will enhance the NRC`s understanding of the distribution and projected quantities of radionuclides within AM and LLW streams from commercial nuclear power stations. This research will also provide radiological characterization of AM specimens for others to use in leach-rate and lysimeter experiments to determine nuclide releases and subsequent movement in natural soil environments.

  15. Low-level radioactive waste from nuclear power generating stations: Characterization, classification and assessment of activated metals and waste streams

    SciTech Connect (OSTI)

    Thomas, V.W.; Robertson, D.E.; Thomas, C.W.

    1993-02-01

    Since the enactment of 10 CFR Part 61, additional difficult-to-measure long-lived radionuclides, not specified in Tables 1 2 of Part 61, have been identified (e.g., [sup 108m]Ag, [sup 93]Mo, [sup 36]Cl, [sup 10]Be, [sup 113m]Cd, [sup 121m]Sn, [sup 126]Sn, [sup 93m]Nb) that may be of concern in certain types of waste. These nuclides are primarily associated with activated metal and perhaps other nuclear power low-level waste (LLW) being sent to disposal facilities. The concentration of a radionuclide in waste materials is normally determined by direct measurement or by indirect calculational methods, such as using a scaling factor to relate inferred concentration of a difficult-to-measure radionuclide to another that is easily measured. The total disposal site inventory of certain difficult-to-measure radionuclides (e.g., [sup 14]C, [sup 129]I, and [sup 99]Tc) often control the total quantities of radioactive waste permitted in LLW burial facilities. Overly conservative scaling factors based on lower limits of detection (LLD), often used in the nuclear power industry to estimate these controlling nuclides, could lead to premature closure of a disposal facility. Samples of LLW (Class B and C activated metals [AM] and other waste streams) are being collected from operating nuclear power stations and analyzed for radionuclides covered in 10 CFR Part 61 and the additional difficult-to-measure radionuclides. This analysis will enhance the NRC's understanding of the distribution and projected quantities of radionuclides within AM and LLW streams from commercial nuclear power stations. This research will also provide radiological characterization of AM specimens for others to use in leach-rate and lysimeter experiments to determine nuclide releases and subsequent movement in natural soil environments.

  16. HLW-OVP-97-0068 High Level Waste Management Division High-Level...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ... kiln or secondary combustion chamber for treatment ... Influents - F-Canyon Low Heat Waste (LHW) and High Heat Waste ... IV tanks, which lack internal structures, thereby Page ...

  17. Moisture measurement for high-level-waste tanks using copper activation probe in cone penetrometer

    SciTech Connect (OSTI)

    Reeder, P.L.; Stromswold, D.C.; Brodzinski, R.L.; Reeves, J.H.; Wilson, W.E.

    1995-10-01

    Laboratory tests have established the feasibility of using neutron activation of copper as a means for measuring the moisture in Hanford`s high-level radioactive waste tanks. The performance of the neutron activation technique to measure moisture is equivalent to the neutron moisture gauges or neutron logs commonly used in commercial well-logging. The principle difference is that the activation of {sup 64}Cu (t{sub 1/2} = 12.7 h) replaces the neutron counters used in moisture gauges or neutron logs. For application to highly radioactive waste tanks, the Cu activation technique has the advantage that it is insensitive to very strong gamma radiation fields or high temperatures. In addition, this technique can be deployed through tortuous paths or in confined spaces such as within the bore of a cone penetrometer. However, the results are not available in ``real-time``. The copper probe`s sensitivity to moisture was measured using simulated tank waste of known moisture content. This report describes the preparation of the simulated waste mixtures and the experiments performed to demonstrate the capabilities of the neutron activation technique. These experiments included determination of the calibration curve of count rate versus moisture content using a single copper probe, measurement of the calibration curve based on ``near-field `` to ``far-field`` counting ratios using a multiple probe technique, and profiling the activity of the copper probe as a function of the vertical height within a simulated waste barrel.

  18. Reference commercial high-level waste glass and canister definition.

    SciTech Connect (OSTI)

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

    1981-09-01

    This report presents technical data and performance characteristics of a high-level waste glass and canister intended for use in the design of a complete waste encapsulation package suitable for disposal in a geologic repository. The borosilicate glass contained in the stainless steel canister represents the probable type of high-level waste product that will be produced in a commercial nuclear-fuel reprocessing plant. Development history is summarized for high-level liquid waste compositions, waste glass composition and characteristics, and canister design. The decay histories of the fission products and actinides (plus daughters) calculated by the ORIGEN-II code are presented.

  19. Technical Exchange on Improved Design and Performance of High Level Waste Melters - Final Report

    SciTech Connect (OSTI)

    SK Sundaram; ML Elliott; D Bickford

    1999-11-19

    SIA Radon is responsible for management of low- and intermediate-level radioactive waste (LILW) produced in Central Russia. In cooperation with Minatom organizations Radon carries out R and D programs on treatment of simulated high level waste (HLW) as well. Radon scientists deal with a study of materials for LILW, HLW, and Nuclear Power Plants (NPP) wastes immobilization, and development and testing of processes and technologies for waste treatment and disposal. Radon is mostly experienced in LILW vitrification. This experience can be carried over to HLW vitrification especially in field of melting systems. The melter chosen as a basic unit for the vitrification plant is a cold crucible. Later on Radon experience in LILW vitrification as well as our results on simulated HLW vitrification are briefly described.

  20. Advanced waste form and melter development for treatment of troublesome high-level wastes

    SciTech Connect (OSTI)

    Marra, James; Kim, Dong -Sang; Maio, Vincent

    2015-09-02

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these "troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approached to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.

  1. An informal expert judgment assessment of subsidence mitigation options for low-level radioactive waste management sites on the Nevada Test Site

    SciTech Connect (OSTI)

    Crowe, B.M. |; Leary, K.; Jacobson, R.; Bensinger, H.; Dolenc, M.

    1999-03-01

    An assessment of options to mitigate the effects of subsidence at low-level radioactive waste disposal sites on the Nevada Test Site was conducted using an informal method of expert judgment. Mitigation options for existing waste cells and future waste cells were identified by a committee composed of knowledgeable personnel from the DOE and DOE-contractors. Eight ranking factors were developed to assess the mitigation options and these factors were scored through elicitation of consensus views from the committee. Different subsets of the factors were applied respectively, to existing waste cells and future waste cells, and the resulting scores were ranked using weighted and unweighted scores. These scores show that there is a large number of viable mitigation options and considerable flexibility in assessing the subsidence issue with a greater range of options for future waste cells compared to existing waste cells. A highly ranked option for both existing and future waste cells is covering the waste cells with a thick closure cap of native alluvium.

  2. High Level Overview of DOE Biomass Logistics II Project Activities

    Broader source: Energy.gov [DOE]

    Breakout Session 1B—Integration of Supply Chains I: Breaking Down Barriers High Level Overview of DOE Biomass Logistics II Project Activities Kevin Comer, Associate Principal, Antares Group Inc.

  3. A TWP-ICE High-Level Cloud Case Study

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    A TWP-ICE High-Level Cloud Case Study Mace, Gerald University of Utah Category: Field Campaigns The Tropical Warm Pool International Cloud Experiment (TWP ICE) was conducted near...

  4. Neptunium estimation in dissolver and high-level-waste solutions

    SciTech Connect (OSTI)

    Pathak, P.N.; Prabhu, D.R.; Kanekar, A.S.; Manchanda, V.K.

    2008-07-01

    This papers deals with the optimization of the experimental conditions for the estimation of {sup 237}Np in spent-fuel dissolver/high-level waste solutions using thenoyltrifluoroacetone as the extractant. (authors)

  5. EIS-0287: Idaho High-Level Waste & Facilities Disposition

    Broader source: Energy.gov [DOE]

    This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liquid...

  6. High Level Waste Corporate Board Newsletter - 09/11/08

    Office of Environmental Management (EM)

    Waste Federal Review Group (LFRG) in Washington, DC on 16-18 September 2008. Contact Maureen O'Dell for details (MAUREEN.O'DELL@hq.doe.gov) Next High-Level Waste Corporate ...

  7. Avoided level crossings in very highly charged ions (Journal...

    Office of Scientific and Technical Information (OSTI)

    Title: Avoided level crossings in very highly charged ions Authors: Beiersdorfer, P. ; Scofield, J. H. ; Brown, G. V. ; Chen, M. H. ; Hell, N. ; Osterheld, A. L. ; Vogel, D. A. ; ...

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

    DOE Patents [OSTI]

    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.

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

    SciTech Connect (OSTI)

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

    1983-02-25

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

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

    SciTech Connect (OSTI)

    Dees, L.A.

    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.

  11. Method for making a low density polyethylene waste form for safe disposal of low level radioactive material

    DOE Patents [OSTI]

    Colombo, P.; Kalb, P.D.

    1984-06-05

    In the method of the invention low density polyethylene pellets are mixed in a predetermined ratio with radioactive particulate material, then the mixture is fed through a screw-type extruder that melts the low density polyethylene under a predetermined pressure and temperature to form a homogeneous matrix that is extruded and separated into solid monolithic waste forms. The solid waste forms are adapted to be safely handled, stored for a short time, and safely disposed of in approved depositories.

  12. Maintenance Free Fluidic Transfer and Mixing Devices for Highly Radioactive Applications - Design, Development, Deployment and Operational Experience

    SciTech Connect (OSTI)

    Phillips, C.; Richardson, J. E.; Fallows, P.

    2006-07-01

    Power Fluidics is the generic name for a range of maintenance-free fluid transfer and mixing devices, capable of handling a wide range of highly radioactive fluids, jointly developed by British Nuclear Group, its US-based subsidiary BNG America, and AEA Technology. Power Fluidic devices include Reverse Flow Diverters (RFDs), Vacuum Operated Slug Lifts (VOSLs), and Air Lifts, all of which have an excellent proven record for pumping radioactive liquids and sludges. Variants of the RFD, termed Pulse Jet Mixers (PJMs) are used to agitate and mix tank contents, where maintenance-free equipment is desirable, and where a high degree of homogenization is necessary. The equipment is designed around the common principle of using compressed air to provide the motive force to transfer liquids and sludges. These devices have no moving parts in contact with the radioactive medium and therefore require no maintenance in radioactive areas of processing plants. Once commissioned, Power Fluidic equipment has been demonstrated to operate for the life of the facility. Over 800 fluidic devices continue to operate safely and reliably in British Nuclear Group's nuclear facilities at the Sellafield site in the United Kingdom, and some of these have done so for almost 40 years. More than 400 devices are being supplied by AEA Technology and BNG America for the Waste Treatment Plant (WTP) at the Hanford Site in southeastern Washington State, USA. This paper discusses: - Principles of operation of fluidic pumps and mixers. - Selection criteria and design of fluidic pumps and mixers. - Operational experience of fluidic pumps and mixers in the United Kingdom. - Applications of fluidic pumps and mixers at the U.S. Department of Energy nuclear sites. (authors)

  13. Comparison of selected foreign plans and practices for spent fuel and high-level waste management

    SciTech Connect (OSTI)

    Schneider, K.J.; Mitchell, S.J.; Lakey, L.T.; Johnson, A.B. Jr.; Hazelton, R.F.; Bradley, D.J.

    1990-04-01

    This report describes the major parameters for management of spent nuclear fuel and high-level radioactive wastes in selected foreign countries as of December 1989 and compares them with those in the United States. The foreign countries included in this study are Belgium, Canada, France, the Federal Republic of Germany, Japan, Sweden, Switzerland, and the United Kingdom. All the countries are planning for disposal of spent fuel and/or high-level wastes in deep geologic repositories. Most countries (except Canada and Sweden) plan to reprocess their spent fuel and vitrify the resultant high-level liquid wastes; in comparison, the US plans direct disposal of spent fuel. The US is planning to use a container for spent fuel as the primary engineered barrier. The US has the most developed repository concept and has one of the earliest scheduled repository startup dates. The repository environment presently being considered in the US is unique, being located in tuff above the water table. The US also has the most prescriptive regulations and performance requirements for the repository system and its components. 135 refs., 8 tabs.

  14. Hanford High-Level Waste Vitrification Program at the Pacific Northwest National Laboratory: technology development - annotated bibliography

    SciTech Connect (OSTI)

    Larson, D.E.

    1996-09-01

    This report provides a collection of annotated bibliographies for documents prepared under the Hanford High-Level Waste Vitrification (Plant) Program. The bibliographies are for documents from Fiscal Year 1983 through Fiscal Year 1995, and include work conducted at or under the direction of the Pacific Northwest National Laboratory. The bibliographies included focus on the technology developed over the specified time period for vitrifying Hanford pretreated high-level waste. The following subject areas are included: General Documentation; Program Documentation; High-Level Waste Characterization; Glass Formulation and Characterization; Feed Preparation; Radioactive Feed Preparation and Glass Properties Testing; Full-Scale Feed Preparation Testing; Equipment Materials Testing; Melter Performance Assessment and Evaluations; Liquid-Fed Ceramic Melter; Cold Crucible Melter; Stirred Melter; High-Temperature Melter; Melter Off-Gas Treatment; Vitrification Waste Treatment; Process, Product Control and Modeling; Analytical; and Canister Closure, Decontamination, and Handling

  15. Operating experience during high-level waste vitrification at the West Valley Demonstration Project

    SciTech Connect (OSTI)

    Valenti, P.J.; Elliott, D.I.

    1999-01-01

    This report provides a summary of operational experiences, component and system performance, and lessons learned associated with the operation of the Vitrification Facility (VF) at the West Valley Demonstration Project (WVDP). The VF was designed to convert stored high-level radioactive waste (HLW) into a stable waste form (borosilicate glass) suitable for disposal in a federal repository. Following successful completion on nonradioactive test, HLW processing began in July 1995. Completion of Phase 1 of HLW processing was reached on 10 June 1998 and represented the processing of 9.32 million curies of cesium-137 (Cs-137) and strontium-90 (Sr-90) to fill 211 canisters with over 436,000 kilograms of glass. With approximately 85% of the total estimated curie content removed from underground waste storage tanks during Phase 1, subsequent operations will focus on removal of tank heel wastes.

  16. Locations of Spent Nuclear Fuel and High-Level Radioactive Waste |

    Energy Savers [EERE]

    Industrial Assessment Centers (IACs) » Locations of Industrial Assessment Centers Locations of Industrial Assessment Centers To apply for an assessment, contact one of the 24 schools across the country that currently participate in the IAC Program. Click on a university name below for contact information for each location. Map of participating schools West Oregon State University San Diego State University San Francisco State University Boise State University Midwest Bradley University Indiana

  17. Locations of spent nuclear fuel and high-level radioactive waste ultimately destined for geologic disposal

    SciTech Connect (OSTI)

    Not Available

    1994-09-01

    Since the late 1950s, Americans have come to rely more and more on energy generated from nuclear reactors. Today, 109 commercial nuclear reactors supply over one-fifth of the electricity used to run our homes, schools, factories, and farms. When the nuclear fuel can no longer sustain a fission reaction in these reactors it becomes `spent` or `used` and is removed from the reactors and stored onsite. Most of our Nation`s spent nuclear fuel is currently being stored in specially designed deep pools of water at reactor sites; some is being stored aboveground in heavy thick-walled metal or concrete structures. Sites currently using aboveground dry storage systems include Virginia Power`s Surry Plant, Carolina Power and Light`s H.B. Robinson Plant, Duke Power`s Oconee Nuclear Station, Colorado Public Service Company`s shutdown reactor at Fort St. Vrain, Baltimore Gas and Electric`s Calvert Cliffs Plant, and Michigan`s Consumer Power Palisades Plant.

  18. Report on interim storage of spent nuclear fuel. Midwestern high-level radioactive waste transportation project

    SciTech Connect (OSTI)

    Not Available

    1993-04-01

    The report on interim storage of spent nuclear fuel discusses the technical, regulatory, and economic aspects of spent-fuel storage at nuclear reactors. The report is intended to provide legislators state officials and citizens in the Midwest with information on spent-fuel inventories, current and projected additional storage requirements, licensing, storage technologies, and actions taken by various utilities in the Midwest to augment their capacity to store spent nuclear fuel on site.

  19. Silicon-Polymer Encapsulation of High-Level Calcine Waste for Transportation or Disposal

    SciTech Connect (OSTI)

    G. G. Loomis; C. M. Miller; J. A. Giansiracusa; R. Kimmel; S. V. Prewett

    2000-01-01

    This report presents the results of an experimental study investigating the potential uses for silicon-polymer encapsulation of High Level Calcine Waste currently stored within the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). The study investigated two different applications of silicon polymer encapsulation. One application uses silicon polymer to produce a waste form suitable for disposal at a High Level Radioactive Waste Disposal Facility directly, and the other application encapsulates the calcine material for transportation to an offsite melter for further processing. A simulated waste material from INTEC, called pilot scale calcine, which contained hazardous materials but no radioactive isotopes was used for the study, which was performed at the University of Akron under special arrangement with Orbit Technologies, the originators of the silicon polymer process called Polymer Encapsulation Technology (PET). This document first discusses the PET process, followed by a presentation of past studies involving PET applications to waste problems. Next, the results of an experimental study are presented on encapsulation of the INTEC calcine waste as it applies to transportation or disposal of calcine waste. Results relating to long-term disposal include: (1) a characterization of the pilot calcine waste; (2) Toxicity Characteristic Leaching Procedure (TCLP) testing of an optimum mixture of pilot calcine, polysiloxane and special additives; and, (3) Material Characterization Center testing MCC-1P evaluation of the optimum waste form. Results relating to transportation of the calcine material for a mixture of maximum waste loading include: compressive strength testing, 10-m drop test, melt testing, and a Department of Transportation (DOT) oxidizer test.

  20. Polysiloxane Encapsulation of High Level Calcine Waste for Transportation or Disposal

    SciTech Connect (OSTI)

    Loomis, Guy George

    2000-03-01

    This report presents the results of an experimental study investigating the potential uses for silicon-polymer encapsulation of High Level Calcine Waste currently stored within the Idaho Nuclear Technology and Engineering Center (INTEC) at the Idaho National Engineering and Environmental Laboratory (INEEL). The study investigated two different applications of silicon polymer encapsulation. One application uses silicon polymer to produce a waste form suitable for disposal at a High Level Radioactive Waste Disposal Facility directly, and the other application encapsulates the calcine material for transportation to an offsite melter for further processing. A simulated waste material from INTEC, called pilot scale calcine, which contained hazardous materials but no radioactive isotopes was used for the study, which was performed at the University of Akron under special arrangement with Orbit Technologies, the originators of the silicon polymer process called Polymer Encapsulation Technology (PET). This document first discusses the PET process, followed by a presentation of past studies involving PET applications to waste problems. Next, the results of an experimental study are presented on encapsulation of the INTEC calcine waste as it applies to transportation or disposal of calcine waste. Results relating to long-term disposal include: 1) a characterization of the pilot calcine waste; 2) Toxicity Characteristic Leaching Procedure (TCLP) testing of an optimum mixture of pilot calcine, polysiloxane and special additives; and, 3) Material Characterization Center testing MCC-1P evaluation of the optimum waste form. Results relating to transportation of the calcine material for a mixture of maximum waste loading include: compressive strength testing, 10-m drop test, melt testing, and a Department of Transportation (DOT) oxidizer test.

  1. Methods of calculating the post-closure performance of high-level waste repositories

    SciTech Connect (OSTI)

    Ross, B.

    1989-02-01

    This report is intended as an overview of post-closure performance assessment methods for high-level radioactive waste repositories and is designed to give the reader a broad sense of the state of the art of this technology. As described here, ''the state of the art'' includes only what has been reported in report, journal, and conference proceedings literature through August 1987. There is a very large literature on the performance of high-level waste repositories. In order to make a review of this breadth manageable, its scope must be carefully defined. The essential principle followed is that only methods of calculating the long-term performance of waste repositories are described. The report is organized to reflect, in a generalized way, the logical order to steps that would be taken in a typical performance assessment. Chapter 2 describes ways of identifying scenarios and estimating their probabilities. Chapter 3 presents models used to determine the physical and chemical environment of a repository, including models of heat transfer, radiation, geochemistry, rock mechanics, brine migration, radiation effects on chemistry, and coupled processes. The next two chapters address the performance of specific barriers to release of radioactivity. Chapter 4 treats engineered barriers, including containers, waste forms, backfills around waste packages, shaft and borehole seals, and repository design features. Chapter 5 discusses natural barriers, including ground water systems and stability of salt formations. The final chapters address optics of general applicability to performance assessment models. Methods of sensitivity and uncertainty analysis are described in Chapter 6, and natural analogues of repositories are treated in Chapter 7. 473 refs., 19 figs., 2 tabs.

  2. Vitrification of hazardous and radioactive wastes

    SciTech Connect (OSTI)

    Bickford, D.F.; Schumacher, R.

    1995-12-31

    Vitrification offers many attractive waste stabilization options. Versatility of waste compositions, as well as the inherent durability of a glass waste form, have made vitrification the treatment of choice for high-level radioactive wastes. Adapting the technology to other hazardous and radioactive waste streams will provide an environmentally acceptable solution to many of the waste challenges that face the public today. This document reviews various types and technologies involved in vitrification.

  3. Electrosorption on carbon aerogel electrodes as a means of treating low-level radioactive wastes and remediating contaminated ground water

    SciTech Connect (OSTI)

    Tran, Tri Duc; Farmer, Joseph C.; DePruneda, Jean H.; Richardson, Jeffery H.

    1997-07-01

    A novel separation process based upon carbon aerogel electrodes has been recently developed for the efficient removal of ionic impurities from aqueous streams. This process can be used as an electrical y- regenerated alternative to ion exchange, thereby reducing-the need for large quantities of chemical regenerants. Once spent (contaminated), these regenerants contribute to the waste that must be disposed of in landfills. The elimination of such wastes is especially beneficial in situations involving radioactive contaminants, and pump and treat processing of massive volumes of ground water. A review and analysis of potential applications will be presented.

  4. Operational considerations for high level blast furnace fuel injection

    SciTech Connect (OSTI)

    Poveromo, J.J.

    1996-12-31

    Injection levels of over 400 lbs/NTHM for coal, over 250 lbs/NTHM for natural gas and over 200 lbs/NTHM for oil have been achieved. Such high levels of fuel injection has a major impact on many aspects of blast furnace operation. In this paper the author begins by reviewing the fundamentals of fuel injection with emphasis on raceway thermochemical phenomena. The operational impacts which are generic to high level injection of any injectant are then outlined. The author will then focus on the particular characteristics of each injectant, with major emphasis on coal and natural gas. Operational considerations for coping with these changes and methods of maximizing the benefits of fuel injection will be reviewed.

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

    SciTech Connect (OSTI)

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

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

    SciTech Connect (OSTI)

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

    2013-01-09

    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.

  7. High-Octane Mid-Level Ethanol Blend Market Assessment

    Alternative Fuels and Advanced Vehicles Data Center [Office of Energy Efficiency and Renewable Energy (EERE)]

    High-Octane Mid-Level Ethanol Blend Market Assessment Caley Johnson, Emily Newes, Aaron Brooker, and Robert McCormick National Renewable Energy Laboratory Steve Peterson Lexidyne, LLC Paul Leiby, Rocio Uria Martinez, and Gbadebo Oladosu Oak Ridge National Laboratory Maxwell L. Brown Colorado School of Mines Technical Report NREL/TP-5400-63698 December 2015 NREL is a national laboratory of the U.S. Department of Energy Office of Energy Efficiency & Renewable Energy Operated by the Alliance

  8. World first in high level waste vitrification - A review of French vitrification industrial achievements

    SciTech Connect (OSTI)

    Brueziere, J.; Chauvin, E. [AREVA, 1 place Jean Millier, 92084 Paris La Defense (France); Piroux, J.C. [Joint Vitrification Laboratory - LCV, Marcoule, BP171, 30207 Bagnols sur Ceze (France)

    2013-07-01

    AREVA has more than 30 years experience in operating industrial HLW (High Level radioactive Waste) vitrification facilities (AVM - Marcoule Vitrification Facility, R7 and T7 facilities). This vitrification technology was based on borosilicate glasses and induction-heating. AVM was the world's first industrial HLW vitrification facility to operate in-line with a reprocessing plant. The glass formulation was adapted to commercial Light Water Reactor fission products solutions, including alkaline liquid waste concentrates as well as platinoid-rich clarification fines. The R7 and T7 facilities were designed on the basis of the industrial experience acquired in the AVM facility. The AVM vitrification process was implemented at a larger scale in order to operate the R7 and T7 facilities in-line with the UP2 and UP3 reprocessing plants. After more than 30 years of operation, outstanding record of operation has been established by the R7 and T7 facilities. The industrial startup of the CCIM (Cold Crucible Induction Melter) technology with enhanced glass formulation was possible thanks to the close cooperation between CEA and AREVA. CCIM is a water-cooled induction melter in which the glass frit and the waste are melted by direct high frequency induction. This technology allows the handling of highly corrosive solutions and high operating temperatures which permits new glass compositions and a higher glass production capacity. The CCIM technology has been implemented successfully at La Hague plant.

  9. Towards increased waste loading in high level waste glasses: Developing a better understanding of crystallization behavior

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Marra, James C.; Kim, Dong -Sang

    2014-12-18

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JCHM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these ''troublesome'' waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Thus, recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized.more » Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating. The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (with higher Al2O3). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group.« less

  10. Towards increased waste loading in high level waste glasses: developing a better understanding of crystallization behavior

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Marra, James C.; Kim, Dong-Sang

    2014-12-18

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these troublesome" waste species cause crystallization in the glass that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glasses and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulationsmorehave been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating. The Hanford site AZ-101 composition represents a waste group that is waste loading limited primarily due to high concentration of Fe2O3. Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste group.less

  11. Towards increased waste loading in high level waste glasses: Developing a better understanding of crystallization behavior

    SciTech Connect (OSTI)

    Marra, James C.; Kim, Dong -Sang

    2014-12-18

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JCHM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these ''troublesome'' waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Thus, recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating. The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (with higher Al2O3). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group.

  12. Investigation of Flammable Gas Releases from High Level Waste Tanks during Periodic Mixing

    SciTech Connect (OSTI)

    Swingle, R.F.

    1999-01-07

    The Savannah River Site processes high-level radioactive waste through precipitation by the addition of sodium tetraphenylborate in a large (approximately 1.3 million gallon) High Level Waste Tank. Radiolysis of water produces a significant amount of hydrogen gas in this slurry. During quiescent periods the tetraphenylborate slurry retains large amounts of hydrogen as dissolved gas and small bubbles. When mixing pumps start, large amounts of hydrogen release due to agitation of the slurry. Flammability concerns necessitate an understanding of the hydrogen retention mechanism in the slurry and a model of how the hydrogen releases from the slurry during pump operation. Hydrogen concentration data collected from the slurry tank confirmed this behavior in the full-scale system. These measurements also provide mass transfer results for the hydrogen release during operation. The authors compared these data to an existing literature model for mass transfer in small, agitated reactors and developed factors to scale this existing model to the 1.3 million gallon tanks in use at the Savannah River Site. The information provides guidance for facility operations.

  13. Towards increased waste loading in high level waste glasses: developing a better understanding of crystallization behavior

    SciTech Connect (OSTI)

    Marra, James C.; Kim, Dong-Sang

    2014-12-18

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these troublesome" waste species cause crystallization in the glass that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glasses and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating. The Hanford site AZ-101 composition represents a waste group that is waste loading limited primarily due to high concentration of Fe2O3. Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste group.

  14. Lessons learned by southern states in transportation of radioactive materials

    SciTech Connect (OSTI)

    Not Available

    1992-03-01

    This report has been prepared under a cooperative agreement with DOE`s Office of Civilian Radioactive Waste Management (OCRWM) and is a summary of the lessons learned by southern states regarding the transportation of radioactive materials including High-Level Radioactive Wastes (HLRW) and Spent Nuclear Fuel (SNF). Sources used in this publication include interviews of state radiological health and public safety officials that are members of the Southern States Energy Board (SSEB) Advisory Committee on Radioactive Materials Transportation, as well as the Board`s Transuranic (TRU) Waste Transportation Working Group. Other sources include letters written by the above mentioned committees concerning various aspects of DOE shipment campaigns.

  15. A user's guide to the GoldSim/BLT-MS integrated software package:a low-level radioactive waste disposal performance assessment model.

    SciTech Connect (OSTI)

    Knowlton, Robert G.; Arnold, Bill Walter; Mattie, Patrick D.

    2007-03-01

    Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in the assessment of radioactive waste disposal and at the time of this publication is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. In countries with small radioactive waste programs, international technology transfer program efforts are often hampered by small budgets, schedule constraints, and a lack of experienced personnel. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available software codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission (NRC) and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, revitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a credible and solid computational platform for constructing probabilistic safety assessment models. This document is a reference users guide for the GoldSim/BLT-MS integrated modeling software package developed as part of a cooperative technology transfer project between Sandia National Laboratories and the Institute of Nuclear Energy Research (INER) in Taiwan for the preliminary assessment of several candidate low-level waste repository sites. Breach, Leach, and Transport-Multiple Species (BLT-MS) is a U.S. NRC sponsored code which simulates release and transport of contaminants from a subsurface low-level waste disposal facility. GoldSim is commercially available probabilistic software package that has radionuclide transport capabilities. The following report guides a user through the steps necessary to use the integrated model and presents a successful application of the paradigm of renewing legacy codes for contemporary application.

  16. 3-D MAPPING TECHNOLOGIES FOR HIGH LEVEL WASTE TANKS

    SciTech Connect (OSTI)

    Marzolf, A.; Folsom, M.

    2010-08-31

    This research investigated four techniques that could be applicable for mapping of solids remaining in radioactive waste tanks at the Savannah River Site: stereo vision, LIDAR, flash LIDAR, and Structure from Motion (SfM). Stereo vision is the least appropriate technique for the solids mapping application. Although the equipment cost is low and repackaging would be fairly simple, the algorithms to create a 3D image from stereo vision would require significant further development and may not even be applicable since stereo vision works by finding disparity in feature point locations from the images taken by the cameras. When minimal variation in visual texture exists for an area of interest, it becomes difficult for the software to detect correspondences for that object. SfM appears to be appropriate for solids mapping in waste tanks. However, equipment development would be required for positioning and movement of the camera in the tank space to enable capturing a sequence of images of the scene. Since SfM requires the identification of distinctive features and associates those features to their corresponding instantiations in the other image frames, mockup testing would be required to determine the applicability of SfM technology for mapping of waste in tanks. There may be too few features to track between image frame sequences to employ the SfM technology since uniform appearance may exist when viewing the remaining solids in the interior of the waste tanks. Although scanning LIDAR appears to be an adequate solution, the expense of the equipment ($80,000-$120,000) and the need for further development to allow tank deployment may prohibit utilizing this technology. The development would include repackaging of equipment to permit deployment through the 4-inch access ports and to keep the equipment relatively uncontaminated to allow use in additional tanks. 3D flash LIDAR has a number of advantages over stereo vision, scanning LIDAR, and SfM, including full frame time-of-flight data (3D image) collected with a single laser pulse, high frame rates, direct calculation of range, blur-free images without motion distortion, no need for precision scanning mechanisms, ability to combine 3D flash LIDAR with 2D cameras for 2D texture over 3D depth, and no moving parts. The major disadvantage of the 3D flash LIDAR camera is the cost of approximately $150,000, not including the software development time and repackaging of the camera for deployment in the waste tanks.

  17. 324 Building radiochemical engineering cells, high-level vault, low-level vault, and associated areas closure plan

    SciTech Connect (OSTI)

    Barnett, J.M.

    1998-03-25

    The Hanford Site, located adjacent to and north of Richland, Washington, is operated by the US Department of Energy, Richland Operations Office (RL). The 324 Building is located in the 300 Area of the Hanford Site. The 324 Building was constructed in the 1960s to support materials and chemical process research and development activities ranging from laboratory/bench-scale studies to full engineering-scale pilot plant demonstrations. In the mid-1990s, it was determined that dangerous waste and waste residues were being stored for greater than 90 days in the 324 Building Radiochemical Engineering Cells (REC) and in the High-Level Vault/Low-Level Vault (HLV/LLV) tanks. [These areas are not Resource Conservation and Recovery Act of 1976 (RCRA) permitted portions of the 324 Building.] Through the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-89, agreement was reached to close the nonpermitted RCRA unit in the 324 Building. This closure plan, managed under TPA Milestone M-20-55, addresses the identified building areas targeted by the Tri-Party Agreement and provides commitments to achieve the highest degree of compliance practicable, given the special technical difficulties of managing mixed waste that contains high-activity radioactive materials, and the physical limitations of working remotely in the areas within the subject closure unit. This closure plan is divided into nine chapters. Chapter 1.0 provides the introduction, historical perspective, 324 Building history and current mission, and the regulatory basis and strategy for managing the closure unit. Chapters 2.0, 3.0, 4.0, and 5.0 discuss the detailed facility description, process information, waste characteristics, and groundwater monitoring respectively. Chapter 6.0 deals with the closure strategy and performance standard, including the closure activities for the B-Cell, D-Cell, HLV, LLV; piping and miscellaneous associated building areas. Chapter 7.0 addresses the closure activities identified in Chapter 6.0, and also adds information on closure activities for the soil directly beneath the unit, regulated material removed during closure, and the schedule for closure. Chapter 8.0 provides Surveillance, monitoring and post-closure information and Chapter 9.0 provides a list of references used throughout the document.

  18. Implementation plan for liquid low-level radioactive waste systems under the FFA for fiscal years 1996 and 1997 at Oak Ridge National Laboratory, Oak Ridge, Tennessee

    SciTech Connect (OSTI)

    1996-06-01

    This document is the fourth annual revision of the plans and schedules for implementing the Federal Facility Agreement (FFA) compliance program, originally submitted in 1992 as ES/ER-17&D1, Federal Facility Agreement Plans and Schedules for Liquid Low-Level Radioactive Waste Tank Systems at Oak Ridge National Laboratory, Oak Ridge, Tennessee. This document summarizes the progress that has been made to date implementing the plans and schedules for meeting the FFA commitments for the Liquid Low-Level Waste (LLLW) System at Oak Ridge National Laboratory (ORNL). In addition, this document lists FFA activities planned for FY 1997. Information presented in this document provides a comprehensive summary to facilitate understanding of the FFA compliance program for LLLW tank systems and to present plans and schedules associated with remediation, through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) process, of LLLW tank systems that have been removed from service.

  19. Permitting plan for the high-level waste interim storage

    SciTech Connect (OSTI)

    Deffenbaugh, M.L.

    1997-04-23

    This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist.

  20. ASTATINE-211 RADIOCHEMISTRY: THE DEVELOPMENT OF METHODOLOGIES FOR HIGH ACTIVITY LEVEL RADIOSYNTHESIS

    SciTech Connect (OSTI)

    MICHAEL R. ZALUTSKY

    2012-08-08

    Targeted radionuclide therapy is emerging as a viable approach for cancer treatment because of its potential for delivering curative doses of radiation to malignant cell populations while sparing normal tissues. Alpha particles such as those emitted by 211At are particularly attractive for this purpose because of their short path length in tissue and high energy, making them highly effective in killing cancer cells. The current impact of targeted radiotherapy in the clinical domain remains limited despite the fact that in many cases, potentially useful molecular targets and labeled compounds have already been identified. Unfortunately, putting these concepts into practice has been impeded by limitations in radiochemistry methodologies. A critical problem is that the synthesis of therapeutic radiopharmaceuticals provides additional challenges in comparison to diagnostic reagents because of the need to perform radio-synthesis at high levels of radioactivity. This is particularly important for {alpha}-particle emitters such as 211At because they deposit large amounts of energy in a highly focal manner. The overall objective of this project is to develop convenient and reproducible radiochemical methodologies for the radiohalogenation of molecules with the {alpha}-particle emitter 211At at the radioactivity levels needed for clinical studies. Our goal is to address two problems in astatine radiochemistry: First, a well known characteristic of 211At chemistry is that yields for electrophilic astatination reactions decline as the time interval after radionuclide isolation from the cyclotron target increases. This is a critical problem that must be addressed if cyclotrons are to be able to efficiently supply 211At to remote users. And second, when the preparation of high levels of 211At-labeled compounds is attempted, the radiochemical yields can be considerably lower than those encountered at tracer dose. For these reasons, clinical evaluation of promising 211At-labeled targeted radiotherapeutics currently is a daunting task. Our central hypothesis is that improvements in 211At radiochemistry are critically dependent on gaining an understanding of and compensating for the effects of radiolysis induced by 211At {alpha}-particles. Because of the widespread interest in labeling antibodies, antibody fragments and peptides with 211At, our proposed work plan will initially focus on reagents that we have developed for this purpose. Part of our strategy is the use of synthetic precursors immobilized on polymeric resins or perfluorous and triarylphosphonium supports. Their use could eliminate the need for a purification step to separate unreacted tin precursor from labeled product and hopefully provide a simple kit technology that could be utilized at other institutions. The specific aims of this project are: (1) To optimze methods for 211At production and isolation of 211At from cyclotron targets; (2) To develop convenient and reproducible methodologies for high activity level and high specific activity radiohalogenation of biomolecules with 211At; (3) to develop a procedure for extending the shelf-life of 211At beyond a few hours so that this radionuclide can be utilized at centers remote from its site of production; and (4) to work out high activity level synthesis methods for utilizing support immobilized tin precursors for 211At labeling. If we are successful in achieving our goals, the radiochemical methodologies that are developed could greatly facilitate the use of 211At-labeled targeted cancer therapeutics in patients, even at institutions that are distant from the few sites currently available for 211At production.

  1. Adsorption of Ruthenium, Rhodium and Palladium from Simulated High-Level Liquid Waste by Highly Functional Xerogel - 13286

    SciTech Connect (OSTI)

    Onishi, Takashi [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan)] [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan); Koyama, Shin-ichi [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan)] [Fukushima Fuels and Materials Department O-arai Research and Development Center Japan Atomic Energy Agency, Narita-cho 4002, O-arai-machi, Ibaraki, 311-1393 (Japan); Mimura, Hitoshi [Dept. of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2,Aoba-ku, Sendai-shi, Miyagi-ken, 980-8579 (Japan)] [Dept. of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University Aramaki-Aza-Aoba 6-6-01-2,Aoba-ku, Sendai-shi, Miyagi-ken, 980-8579 (Japan)

    2013-07-01

    Fission products are generated by fission reactions in nuclear fuel. Platinum group (Pt-G) elements, such as palladium (Pd), rhodium (Rh) and ruthenium (Ru), are also produced. Generally, Pt-G elements play important roles in chemical and electrical industries. Highly functional xerogels have been developed for recovery of these useful Pt-G elements from high - level radioactive liquid waste (HLLW). An adsorption experiment from simulated HLLW was done by the column method to study the selective adsorption of Pt-G elements, and it was found that not only Pd, Rh and Ru, but also nickel, zirconium and tellurium were adsorbed. All other elements were not adsorbed. Adsorbed Pd was recovered by washing the xerogel-packed column with thiourea solution and thiourea - nitric acid mixed solution in an elution experiment. Thiourea can be a poison for automotive exhaust emission system catalysts, so it is necessary to consider its removal. Thermal decomposition and an acid digestion treatment were conducted to remove sulfur in the recovered Pd fraction. The relative content of sulfur to Pd was decreased from 858 to 0.02 after the treatment. These results will contribute to design of the Pt-G element separation system. (authors)

  2. CEMENTITIOUS GROUT FOR CLOSING SRS HIGH LEVEL WASTE TANKS - #12315

    SciTech Connect (OSTI)

    Langton, C.; Burns, H.; Stefanko, D.

    2012-01-10

    In 1997, the first two United States Department of Energy (US DOE) high level waste tanks (Tanks 17-F and 20-F: Type IV, single shell tanks) were taken out of service (permanently closed) at the Savannah River Site (SRS). In 2012, the DOE plans to remove from service two additional Savannah River Site (SRS) Type IV high-level waste tanks, Tanks 18-F and 19-F. These tanks were constructed in the late 1950's and received low-heat waste and do not contain cooling coils. Operational closure of Tanks 18-F and 19-F is intended to be consistent with the applicable requirements of the Resource Conservation and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) and will be performed in accordance with South Carolina Department of Health and Environmental Control (SCDHEC). The closure will physically stabilize two 4.92E+04 cubic meter (1.3 E+06 gallon) carbon steel tanks and isolate and stabilize any residual contaminants left in the tanks. The closure will also fill, physically stabilize and isolate ancillary equipment abandoned in the tanks. A Performance Assessment (PA) has been developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closure of the F-Area Tank Farm (FTF) waste tanks. Next generation flowable, zero-bleed cementitious grouts were designed, tested, and specified for closing Tanks 18-F and 19-F and for filling the abandoned equipment. Fill requirements were developed for both the tank and equipment grouts. All grout formulations were required to be alkaline with a pH of 12.4 and chemically reduction potential (Eh) of -200 to -400 to stabilize selected potential contaminants of concern. This was achieved by including Portland cement and Grade 100 slag in the mixes, respectively. Ingredients and proportions of cementitious reagents were selected and adjusted, respectively, to support the mass placement strategy developed by closure operations. Subsequent down selection was based on compressive strength and saturated hydraulic conductivity results. Fresh slurry property results were used as the first level of screening. A high range water reducing admixture and a viscosity modifying admixture were used to adjust slurry properties to achieve flowable grouts. Adiabatic calorimeter results were used as the second level screening. The third level of screening was used to design mixes that were consistent with the fill material parameters used in the F-Tank Farm Performance Assessment which was developed to assess the long-term fate and transport of residual contamination in the environment resulting from the operational closures.

  3. Stagnation Region Heat Transfer Augmentation at Very High Turbulence Levels

    SciTech Connect (OSTI)

    Ames, Forrest; Kingery, Joseph E.

    2015-06-17

    A database for stagnation region heat transfer has been extended to include heat transfer measurements acquired downstream from a new high intensity turbulence generator. This work was motivated by gas turbine industry heat transfer designers who deal with heat transfer environments with increasing Reynolds numbers and very high turbulence levels. The new mock aero-combustor turbulence generator produces turbulence levels which average 17.4%, which is 37% higher than the older turbulence generator. The increased level of turbulence is caused by the reduced contraction ratio from the liner to the exit. Heat transfer measurements were acquired on two large cylindrical leading edge test surfaces having a four to one range in leading edge diameter (40.64 cm and 10.16 cm). Gandvarapu and Ames [1] previously acquired heat transfer measurements for six turbulence conditions including three grid conditions, two lower turbulence aero-combustor conditions, and a low turbulence condition. The data are documented and tabulated for an eight to one range in Reynolds numbers for each test surface with Reynolds numbers ranging from 62,500 to 500,000 for the large leading edge and 15,625 to 125,000 for the smaller leading edge. The data show augmentation levels of up to 136% in the stagnation region for the large leading edge. This heat transfer rate is an increase over the previous aero-combustor turbulence generator which had augmentation levels up to 110%. Note, the rate of increase in heat transfer augmentation decreases for the large cylindrical leading edge inferring only a limited level of turbulence intensification in the stagnation region. The smaller cylindrical leading edge shows more consistency with earlier stagnation region heat transfer results correlated on the TRL (Turbulence, Reynolds number, Length scale) parameter. The downstream regions of both test surfaces continue to accelerate the flow but at a much lower rate than the leading edge. Bypass transition occurs in these regions providing a useful set of data to ground the prediction of transition onset and length over a wide range of Reynolds numbers and turbulence intensity and scales.

  4. Mixing Processes in High-Level Waste Tanks

    Office of Scientific and Technical Information (OSTI)

    C. Berkeley Proposal to DOE/OER (1) Project ID: 54656 Mixing Processes in High-Level Waste Tanks Per F. Peterson Professor Department of Nuclear Engineering University of California, Berkeley Berkeley, CA 94720-1730 DOE/EM Contract: FDDE-FG07-96ER14731-09/99 Annual Progress Report For the Period: Sept. 15, 1998 - Sept. 14, 1999 to: Mark Gilbertson, EM-52 U.S. Department of Energy Office of Environmental Management Office of Science and Risk Policy 1000 Independence Avenue SW Washington, DC 20585

  5. Market Designs for High Levels of Variable Generation: Preprint

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Market Designs for High Levels of Variable Generation Preprint M. Milligan, 1 H. Holttinen, 2 J. Kiviluoma, 2 A. Orths, 3 M. Lynch, 4 and L. Söder 5 1 National Renewable Energy Laboratory 2 VTT Technical Research Centre of Finland 3 Energinet 4 University College Dublin 5 KTH Presented at the IEEE Power and Energy Society General Meeting National Harbor, Maryland July 27-31, 2014 Conference Paper NREL/CP-5D00-62280 October 2014 NOTICE The submitted manuscript has been offered by an employee of

  6. Market Designs for High Levels of Variable Generation: Preprint

    SciTech Connect (OSTI)

    Milligan, M.; Holttinen, H.; Kiviluoma, J.; Orths, A.; Lynch, M.; Soder, L.

    2014-10-01

    Variable renewable generation is increasing in penetration in modern power systems, leading to higher variability in the supply and price of electricity as well as lower average spot prices. This raises new challenges, particularly in ensuring sufficient capacity and flexibility from conventional technologies. Because the fixed costs and lifetimes of electricity generation investments are significant, designing markets and regulations that ensure the efficient integration of renewable generation is a significant challenge. This papers reviews the state of play of market designs for high levels of variable generation in the United States and Europe and considers new developments in both regions.

  7. Letter report: Pre-conceptual design study for a pilot-scale Non-Radioactive Low-Level Waste Vitrification Facility

    SciTech Connect (OSTI)

    Thompson, R.A.; Morrissey, M.F.

    1996-03-01

    This report presents a pre-conceptual design study for a Non-Radioactive Low-Level Waste, Pilot-Scale Vitrification System. This pilot plant would support the development of a full-scale LLW Vitrification Facility and would ensure that the full-scale facility can meet its programmatic objectives. Use of the pilot facility will allow verification of process flowsheets, provide data for ensuring product quality, assist in scaling to full scale, and support full-scale start-up. The facility will vitrify simulated non-radioactive LLW in a manner functionally prototypic to the full-scale facility. This pre-conceptual design study does not fully define the LLW Pilot-Scale Vitrification System; rather, it estimates the funding required to build such a facility. This study includes identifying all equipment necessary. to prepare feed, deliver it into the melter, convert the feed to glass, prepare emissions for atmospheric release, and discharge and handle the glass. The conceived pilot facility includes support services and a structure to contain process equipment.

  8. Application of Probabilistic Performance Assessment Modeling for Optimization of Maintenance Studies for Low-Level Radioactive Waste Disposal Sites at the Nevada Test Site

    SciTech Connect (OSTI)

    Crowe, B.; Yucel, V.; Rawlinson, S.; Black, P.; Carilli, J.; DiSanza, F.

    2002-02-25

    The U.S. Department of Energy (DOE), National Nuclear Security Administration of the Nevada Operations Office (NNSA/NV) operates and maintains two active facilities on the Nevada Test Site (NTS) that dispose defense-generated low-level radioactive waste (LLW), mixed radioactive waste, and ''classified waste'' in shallow trenches and pits. The operation and maintenance of the LLW disposal sites are self-regulated by the DOE under DOE Order 435.1. This Order requires formal review of a performance assessment (PA) and composite analysis (CA; assessment of all interacting radiological sources) for each LLW disposal system followed by an active maintenance program that extends through and beyond the site closure program. The Nevada disposal facilities continue to receive NTS-generated LLW and defense-generated LLW from across the DOE complex. The PA/CAs for the sites have been conditionally approved and the facilities are now under a formal maintenance program that requires testing of conceptual models, quantifying and attempting to reduce uncertainty, and implementing confirmatory and long-term background monitoring, all leading to eventual closure of the disposal sites. To streamline and reduce the cost of the maintenance program, the NNSA/NV is converting the deterministic PA/CAs to probabilistic models using GoldSim, a probabilistic simulation computer code. The output of probabilistic models will provide expanded information supporting long-term decision objectives of the NTS disposal sites.

  9. Modeling the corrosion of high-level waste containers CAM-CRM interface

    SciTech Connect (OSTI)

    Farmer, J.C.; McCright, M.

    1997-12-09

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design,the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 and C-22, while the outer barrier is made of a corrosion allowance material (CAM) such as carbon steel or Monel 400. Initially, the containers will be hot and dry due to the heat generated by radioactive decay. However, the temperature will eventually drop to levels where both humid air and aqueous phase corrosion will be possible. As the outer barrier is penetrated, uniform corrosion of the CRM will be possible of exfoliated areas. The possibility of crevice formation between the CAM and CRM will also exist. In the case of either Alloy 625 or C-22, a crevice will have to form before significant penetration of the CRM can occur. Crevice corrosion of the CRMs has been well documented.

  10. DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER

    SciTech Connect (OSTI)

    G. Radulesscu; J.S. Tang

    2000-06-07

    The purpose of ''Design Analysis for the Defense High-Level Waste Disposal Container'' analysis is to technically define the defense high-level waste (DHLW) disposal container/waste package using the Waste Package Department's (WPD) design methods, as documented in ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000a). The DHLW disposal container is intended for disposal of commercial high-level waste (HLW) and DHLW (including immobilized plutonium waste forms), placed within disposable canisters. The U.S. Department of Energy (DOE)-managed spent nuclear fuel (SNF) in disposable canisters may also be placed in a DHLW disposal container along with HLW forms. The objective of this analysis is to demonstrate that the DHLW disposal container/waste package satisfies the project requirements, as embodied in Defense High Level Waste Disposal Container System Description Document (SDD) (CRWMS M&O 1999a), and additional criteria, as identified in Waste Package Design Sensitivity Report (CRWMS M&Q 2000b, Table 4). The analysis briefly describes the analytical methods appropriate for the design of the DHLW disposal contained waste package, and summarizes the results of the calculations that illustrate the analytical methods. However, the analysis is limited to the calculations selected for the DHLW disposal container in support of the Site Recommendation (SR) (CRWMS M&O 2000b, Section 7). The scope of this analysis is restricted to the design of the codisposal waste package of the Savannah River Site (SRS) DHLW glass canisters and the Training, Research, Isotopes General Atomics (TRIGA) SNF loaded in a short 18-in.-outer diameter (OD) DOE standardized SNF canister. This waste package is representative of the waste packages that consist of the DHLW disposal container, the DHLW/HLW glass canisters, and the DOE-managed SNF in disposable canisters. The intended use of this analysis is to support Site Recommendation reports and to assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the Development Plan ''Design Analysis for the Defense High-Level Waste Disposal Container'' (CRWMS M&O 2000c) with no deviations from the plan.

  11. High Level Waste Remote Handling Equipment in the Melter Cave Support Handling System at the Hanford Waste Treatment Plant

    SciTech Connect (OSTI)

    Bardal, M.A. [PaR Systems, Inc., Shoreview, MN (United States); Darwen, N.J. [Bechtel National, Inc., Richland, WA (United States)

    2008-07-01

    Cold war plutonium production led to extensive amounts of radioactive waste stored in tanks at the Department of Energy's (DOE) Hanford site. Bechtel National, Inc. is building the largest nuclear Waste Treatment Plant in the world located at the Department of Energy's Hanford site to immobilize the millions of gallons of radioactive waste. The site comprises five main facilities; Pretreatment, High Level Waste vitrification, Low Active Waste vitrification, an Analytical Lab and the Balance of Facilities. The pretreatment facilities will separate the high and low level waste. The high level waste will then proceed to the HLW facility for vitrification. Vitrification is a process of utilizing a melter to mix molten glass with radioactive waste to form a stable product for storage. The melter cave is designated as the High Level Waste Melter Cave Support Handling System (HSH). There are several key processes that occur in the HSH cell that are necessary for vitrification and include: feed preparation, mixing, pouring, cooling and all maintenance and repair of the process equipment. Due to the cell's high level radiation, remote handling equipment provided by PaR Systems, Inc. is required to install and remove all equipment in the HSH cell. The remote handling crane is composed of a bridge and trolley. The trolley supports a telescoping tube set that rigidly deploys a TR 4350 manipulator arm with seven degrees of freedom. A rotating, extending, and retracting slewing hoist is mounted to the bottom of the trolley and is centered about the telescoping tube set. Both the manipulator and slewer are unique to this cell. The slewer can reach into corners and the manipulator's cross pivoting wrist provides better operational dexterity and camera viewing angles at the end of the arm. Since the crane functions will be operated remotely, the entire cell and crane have been modeled with 3-D software. Model simulations have been used to confirm operational and maintenance functional and timing studies throughout the design process. Since no humans can go in or out of the cell, there are several recovery options that have been designed into the system including jack-down wheels for the bridge and trolley, recovery drums for the manipulator hoist, and a wire rope cable cutter for the slewer jib hoist. If the entire crane fails in cell, the large diameter cable reel that provides power, signal, and control to the crane can be used to retrieve the crane from the cell into the crane maintenance area. (authors)

  12. Standard Review Plan for the review of a license application for a low-level radioactive waste disposal facility. Revision 3

    SciTech Connect (OSTI)

    Not Available

    1994-04-01

    The Standard Review Plan (SRP) (NUREG-1200) provides guidance to staff reviewers in the Office of Nuclear Material Safety and Safeguards who perform safety reviews of applications to construct and operate low-level radioactive waste disposal facilities. The SRP ensures the quality and uniformity of the staff reviews and presents a well-defined base from which to evaluate proposed changes in the scope and requirements of the staff reviews. The SRP makes information about the regulatory licensing process widely available and serves to improve the understanding of the staff`s review process by interested members of the public and the industry. Each individual SRP addresses the responsibilities of persons performing the review, the matters that are reviewed, the Commission`s regulations and acceptance criteria necessary for the review, how the review is accomplished, the conclusions that are appropriate, and the implementation requirements.

  13. Life-Cycle Cost and Risk Analysis of Alternative Configurations for Shipping Low-Level Radioactive Waste to the Nevada Test Site

    SciTech Connect (OSTI)

    PM Daling; SB Ross; BM Biwer

    1999-12-17

    The Nevada Test Site (NTS) is a major receiver of low-level radioactive waste (LLW) for disposal. Currently, all LLW received at NTS is shipped by truck. The trucks use highway routes to NTS that pass through the Las Vegas Valley and over Hoover Dam, which is a concern of local stakeholder groups in the State of Nevada. Rail service offers the opportunity to reduce transportation risks and costs, according to the Waste Management Programmatic Environmental Impact Statement (WM-PEIS). However, NTS and some DOE LLW generator sites are not served with direct rail service so intermodal transport is under consideration. Intermodal transport involves transport via two modes, in this case truck and rail, from the generator sites to NTS. LLW shipping containers would be transferred between trucks and railcars at intermodal transfer points near the LLW generator sites, NTS, or both. An Environmental Assessment (EA)for Intermodal Transportation of Low-Level Radioactive Waste to the Nevada Test Site (referred to as the NTSIntermodal -M) has been prepared to determine whether there are environmental impacts to alterations to the current truck routing or use of intermodal facilities within the State of Nevada. However, an analysis of the potential impacts outside the State of Nevada are not addressed in the NTS Intermodal EA. This study examines the rest of the transportation network between LLW generator sites and the NTS and evaluates the costs, risks, and feasibility of integrating intermodal shipments into the LLW transportation system. This study evaluates alternative transportation system configurations for NTS approved and potential generators based on complex-wide LLW load information. Technical judgments relative to the availability of DOE LLW generators to ship from their sites by rail were developed. Public and worker risk and life-cycle cost components are quantified. The study identifies and evaluates alternative scenarios that increase the use of rail (intermodal where needed) to transport LLW from generator sites to NTS.

  14. Calculates Neutron Production in Canisters of High-level Waste

    Energy Science and Technology Software Center (OSTI)

    1993-01-15

    ALPHN calculates the (alpha,n) neutron production rate of a canister of vitrified high-level waste. The user supplies the chemical composition of the glass or glass-ceramic and the curies of the alpha-emitting actinides present. The output of the program gives the (alpha,n) neutron production of each actinide in neutrons per second and the total for the canister. The (alpha,n) neutron production rates are source terms only; that is, they are production rates within the glass andmore » do not take into account the shielding effect of the glass. For a given glass composition, the user can calculate up to eight cases simultaneously; these cases are based on the same glass composition but contain different quantities of actinides per canister.« less

  15. High Level Waste System Impacts from Acid Dissolution of Sludge

    SciTech Connect (OSTI)

    KETUSKY, EDWARD

    2006-04-20

    This research evaluates the ability of OLI{copyright} equilibrium based software to forecast Savannah River Site High Level Waste system impacts from oxalic acid dissolution of Tank 1-15 sludge heels. Without further laboratory and field testing, only the use of oxalic acid can be considered plausible to support sludge heel dissolution on multiple tanks. Using OLI{copyright} and available test results, a dissolution model is constructed and validated. Material and energy balances, coupled with the model, identify potential safety concerns. Overpressurization and overheating are shown to be unlikely. Corrosion induced hydrogen could, however, overwhelm the tank ventilation. While pH adjustment can restore the minimal hydrogen generation, resultant precipitates will notably increase the sludge volume. OLI{copyright} is used to develop a flowsheet such that additional sludge vitrification canisters and other negative system impacts are minimized. Sensitivity analyses are used to assess the processability impacts from variations in the sludge/quantities of acids.

  16. Comments of Santiago Grijalva: High-Level Response to DOE RFI...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Santiago Grijalva: High-Level Response to DOE RFI on Smart Grid Policy Comments of Santiago Grijalva: High-Level Response to DOE RFI on Smart Grid Policy High-Level Response to DOE ...

  17. EIS-0110: Central Waste Disposal Facility for Low-Level Radioactive Waste, Oak Ridge Reservation, Oak Ridge, Tennessee

    Broader source: Energy.gov [DOE]

    This EIS assessed the environmental impacts of alternatives for the disposal of low-level waste and by-product materials generated by the three major plants on the Oak Ridge Reservation (ORR). In addition to the no-action alternative, two classes of alternatives were evaluated: facility design alternatives and siting alternatives. This project was cancelled after the Draft Environmental Impact Statement was issued.

  18. Defense High Level Waste Disposal Container System Description

    SciTech Connect (OSTI)

    2000-10-12

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials will be selected for the disposal container inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lids will be a barrier made of high-nickel alloy. The defense HLW disposal container interfaces with the emplacement drift environment and the internal waste by transferring heat from the canisters to the external environment and by protecting the canisters and their contents from damage/degradation by the external environment. The disposal container also interfaces with the canisters by limiting access of moderator and oxidizing agents to the waste. A loaded and sealed disposal container (waste package) interfaces with the Emplacement Drift System's emplacement drift waste package supports upon which the waste packages are placed. The disposal container interfaces with the Canister Transfer System, Waste Emplacement /Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement, and retrieval for the disposal container/waste package.

  19. Canister arrangement for storing radioactive waste

    DOE Patents [OSTI]

    Lorenzo, Donald K.; Van Cleve, Jr., John E.

    1982-01-01

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

  20. Canister arrangement for storing radioactive waste

    DOE Patents [OSTI]

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

    1980-04-23

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

  1. Review of high-level waste form properties. [146 bibliographies

    SciTech Connect (OSTI)

    Rusin, J.M.

    1980-12-01

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

  2. A COMPLETE HISTORY OF THE HIGH-LEVEL WASTE PLANT AT THE WEST VALLEY DEMONSTRATION PROJECT

    SciTech Connect (OSTI)

    Petkus, Lawrence L.; Paul, James; Valenti, Paul J.; Houston, Helene; May, Joseph

    2003-02-27

    The West Valley Demonstration Project (WVDP) vitrification melter was shut down in September 2002 after being used to vitrify High Level Waste (HLW) and process system residuals for six years. Processing of the HLW occurred from June 1996 through November 2001, followed by a program to flush the remaining HLW through to the melter. Glass removal and shutdown followed. The facility and process equipment is currently in a standby mode awaiting deactivation. During HLW processing operations, nearly 24 million curies of radioactive material were vitrified into 275 canisters of HLW glass. At least 99.7% of the curies in the HLW tanks at the WVDP were vitrified using the melter. Each canister of HLW holds approximately 2000 kilograms of glass with an average contact dose rate of over 2600 rem per hour. After vitrification processing ended, two more cans were filled using the Evacuated Canister Process to empty the melter at shutdown. This history briefly summarizes the initial stages of process development and earlier WVDP experience in the design and operation of the vitrification systems, followed by a more detailed discussion of equipment availability and failure rates during six years of operation. Lessons learned operating a system that continued to function beyond design expectations also are highlighted.

  3. Technical considerations for evaluating substantially complete containment of high-level waste within the waste package

    SciTech Connect (OSTI)

    Manaktala, H.K. (Southwest Research Inst., San Antonio, TX (USA). Center for Nuclear Waste Regulatory Analyses); Interrante, C.G. (Nuclear Regulatory Commission, Washington, DC (USA). Div. of High-Level Waste Management)

    1990-12-01

    This report deals with technical information that is considered essential for demonstrating the ability of the high-level radioactive waste package to provide substantially complete containment'' of its contents (vitrified waste form or spent light-water reactor fuel) for a period of 300 to 1000 years in a geological repository environment. The discussion is centered around technical considerations of the repository environment, materials and fabrication processes for the waste package components, various degradation modes of the materials of construction of the waste packages, and inspection and monitoring of the waste package during the preclosure and retrievability period, which could begin up to 50 years after initiation of waste emplacement. The emphasis in this report is on metallic materials. However, brief references have been made to other materials such as ceramics, graphite, bonded ceramic-metal systems, and other types of composites. The content of this report was presented to an external peer review panel of nine members at a workshop held at the Center for Nuclear Waste Regulatory Analyses (CNWRA), Southwest Research Institute, San Antonio, Texas, April 2--4, 1990. The recommendations of the peer review panel have been incorporated in this report. There are two companion reports; the second report in the series provides state-of-the-art techniques for uncertainty evaluations. 97 refs., 1 fig.

  4. High-level waste borosilicate glass a compendium of corrosion characteristics. Volume 1

    SciTech Connect (OSTI)

    Cunnane, J.C.; Bates, J.K.; Bradley, C.R.

    1994-03-01

    Current plans call for the United States Department of Energy (DOE) to start up facilities for vitrification of high-level radioactive waste (HLW) stored in tanks at the Savannah River Site, Aiken, South Carolina, in 1995; West Valley Demonstration Project, West Valley, New York, in 1996; and at the Hanford Site, Richland, Washington, after the year 2000. The product from these facilities will be canistered HLW borosilicate glass, which will be stored, transported, and eventually disposed of in a geologic repository. The behavior of this glass waste product, under the range of likely service conditions, is the subject of considerable scientific and public interest. Over the past few decades, a large body of scientific information on borosilicate waste glass has been generated worldwide. The intent of this document is to consolidate information pertaining to our current understanding of waste glass corrosion behavior and radionuclide release. The objective, scope, and organization of the document are discussed in Section 1.1, and an overview of borosilicate glass corrosion is provided in Section 1.2. The history of glass as a waste form and the international experience with waste glass are summarized in Sections 1.3 and 1.4, respectively.

  5. SUMO, System performance assessment for a high-level nuclear waste repository: Mathematical models

    SciTech Connect (OSTI)

    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.

  6. Evaluation of a performance assessment methodology for low-level radioactive waste disposal facilities: Validation needs. Volume 2

    SciTech Connect (OSTI)

    Kozak, M.W.; Olague, N.E.

    1995-02-01

    In this report, concepts on how validation fits into the scheme of developing confidence in performance assessments are introduced. A general framework for validation and confidence building in regulatory decision making is provided. It is found that traditional validation studies have a very limited role in developing site-specific confidence in performance assessments. Indeed, validation studies are shown to have a role only in the context that their results can narrow the scope of initial investigations that should be considered in a performance assessment. In addition, validation needs for performance assessment of low-level waste disposal facilities are discussed, and potential approaches to address those needs are suggested. These areas of topical research are ranked in order of importance based on relevance to a performance assessment and likelihood of success.

  7. Survey of statistical and sampling needs for environmental monitoring of commercial low-level radioactive waste disposal facilities

    SciTech Connect (OSTI)

    Eberhardt, L.L.; Thomas, J.M.

    1986-07-01

    This project was designed to develop guidance for implementing 10 CFR Part 61 and to determine the overall needs for sampling and statistical work in characterizing, surveying, monitoring, and closing commercial low-level waste sites. When cost-effectiveness and statistical reliability are of prime importance, then double sampling, compositing, and stratification (with optimal allocation) are identified as key issues. If the principal concern is avoiding questionable statistical practice, then the applicability of kriging (for assessing spatial pattern), methods for routine monitoring, and use of standard textbook formulae in reporting monitoring results should be reevaluated. Other important issues identified include sampling for estimating model parameters and the use of data from left-censored (less than detectable limits) distributions.

  8. Defense High-Level Waste Leaching Mechanisms Program. Final report

    SciTech Connect (OSTI)

    Mendel, J.E.

    1984-08-01

    The Defense High-Level Waste Leaching Mechanisms Program brought six major US laboratories together for three years of cooperative research. The participants reached a consensus that solubility of the leached glass species, particularly solubility in the altered surface layer, is the dominant factor controlling the leaching behavior of defense waste glass in a system in which the flow of leachant is constrained, as it will be in a deep geologic repository. Also, once the surface of waste glass is contacted by ground water, the kinetics of establishing solubility control are relatively rapid. The concentrations of leached species reach saturation, or steady-state concentrations, within a few months to a year at 70 to 90/sup 0/C. Thus, reaction kinetics, which were the main subject of earlier leaching mechanisms studies, are now shown to assume much less importance. The dominance of solubility means that the leach rate is, in fact, directly proportional to ground water flow rate. Doubling the flow rate doubles the effective leach rate. This relationship is expected to obtain in most, if not all, repository situations.

  9. Advanced waste form and Melter development for treatment of troublesome high-level wastes

    SciTech Connect (OSTI)

    Marra, James; Kim, Dong -Sang; Maio, Vincent

    2015-10-01

    A number of waste components in US defense high level radioactive wastes (HLW) have proven challenging for current Joule heated ceramic melter (JHCM) operations and have limited the ability to increase waste loadings beyond already realized levels. Many of these “troublesome" waste species cause crystallization in the glass melt that can negatively impact product quality or have a deleterious effect on melter processing. Recent efforts at US Department of Energy laboratories have focused on understanding crystallization behavior within HLW glass melts and investigating approaches to mitigate the impacts of crystallization so that increases in waste loading can be realized. Advanced glass formulations have been developed to highlight the unique benefits of next-generation melter technologies such as the Cold Crucible Induction Melter (CCIM). Crystal-tolerant HLW glasses have been investigated to allow sparingly soluble components such as chromium to crystallize in the melter but pass out of the melter before accumulating.The Hanford site AZ-101 tank waste composition represents a waste group that is waste loading limited primarily due to high concentrations of Fe2O3 (also with high Al2O3 concentrations). Systematic glass formulation development utilizing slightly higher process temperatures and higher tolerance to spinel crystals demonstrated that an increase in waste loading of more than 20% could be achieved for this waste composition, and by extension higher loadings for wastes in the same group. An extended duration CCIM melter test was conducted on an AZ-101 waste simulant using the CCIM platform at the Idaho National Laboratory (INL). The melter was continually operated for approximately 80 hours demonstrating that the AZ-101 high waste loading glass composition could be readily processed using the CCIM technology. The resulting glass was close to the targeted composition and exhibited excellent durability in both the as poured state and after being slowly cooled according to the canister centerline cooling (CCC) profile. Glass formulation development was also completed on other Hanford tank wastes that were identified to further challenge waste loading due to the presence of appreciable quantities (>750 g) of plutonium in the waste tanks. In addition to containing appreciable Pu quantities, the C-102 waste tank and the 244-TX waste tank contain high concentrations of aluminum and iron, respectively that will further challenge vitrification processing. Glass formulation testing also demonstrated that high waste loadings could be achieved with these tank compositions using the attributes afforded by the CCIM technology.

  10. Milestones for Selection Characterization and Analysis of the Performance of a Repository for Spent Nuclear Fuel and HIh-Level Radioactive Waste at Yucca Mountain.

    SciTech Connect (OSTI)

    Rechard, Robert P.

    2015-02-01

    This report presents a concise history in tabular form of events leading up to site identification in 1978, site selection in 1987, subsequent characterization, and ongoing analysis throu gh 2009 of the performance of a repository for spent nuclear fuel and high - level radi oactive waste at Yucca Mou ntain in southern Nevada. The tabulated events generally occurred in five periods: (1) commitment to mined geologic disposal and identification of sites; (2) site selection and analysis, based on regional geologic characterization through literature and an alogous data; (3) feasibility analysis demonstrating calculation procedures and importance of system components, based on rough measures of performance using surface exploration, waste process knowledge, and general laboratory experiments; (4) suitability analysis demonstrating viability of disposal system, based on environment - specific laboratory experiments, in - situ experiments, and underground disposal system characterization; and (5) compliance analysis, based on completed site - specific characterization . The current sixth period beyond 2010 represents a new effort to set waste management policy in the United States. Because the relationship is important to understanding the evolution of the Yucca Mountain Project , the tabulation also shows the interaction between the policy realm and technical realm using four broad categories of events : (a ) R egulatory requirements and related federal policy in laws and court decisions, (c ) Presidential and agency directives, (c) technical milestones of implemen ting institutions, and (d ) critiques of the Yucca Mountai n P roject and pertinent national and world events related to nuclear energy and radioactive waste. Preface The historical progression of technical milestones for the Yucca Mountain Project was originally developed for 10 journal articles in a special issue of Reliability Engineering System Safe ty on the performance assessment for the Yucca Mountain license application [ 1 - 10 ] . The listing of mile stones , a distinct feature of those articles, has been collected and tabulated here. A lthough a brief description is presented here (based on the summaries in the 10 journal articles) , t he emphasis remains on the tab ulation because of its usefulness in pro viding a comprehensive but concise history of the Yucca Mountain Project . T he tabulation presented here is more elaborate than originally presented in that many of the interactions that occurred between the technical realm and policy realm can be depicted in separate columns . The usefulness of the milestones table is due in part to L.A. Connolly, for editorial and reference support, and S.K. Best, Raytheon, and L. May s, Sandia National Laboratories ( SNL ) , for illustration support. Reviewers P.N. Swift, SNL , and K. Gupta, University of Oklahoma, helped improve the discussion. The historical perspective presented is that of the author and is not necessarily held by reviewers, Sandia National Laboratories , and the US Department of Energy. As a historic perspect ive, the author is reporting on the work of others; however, any interpretative erro r s of the documentation are those of the author alone. The characterization and modeling of the Yucca Mountain disposal system required numerous participants with expertise in many areas of science a nd technology, as evident from the extensive reference list. Their d iligent efforts are generally acknowledged here and through the many references to their impressive work , but the 10 journal articles acknowledge by name many of the numerous participants that contributed to the Yucca Mountain Project .

  11. Next Generation Extractants for Cesium Separation from High-Level Waste

    SciTech Connect (OSTI)

    Moyer, Bruce A; Bazelaire, Eve; Bonnesen, Peter V; Custelcean, Radu; Delmau, Laetitia Helene; Ditto, Mary E; Engle, Nancy L; Gorbunova, Maryna; Haverlock, Tamara; Levitskaia, Tatiana G.; Bartsch, Richard A.; Surowiec, Malgorzata A.; Marquez, Manuel; Zhou, Hui

    2006-01-01

    This project seeks a fundamental understanding and major improvement in cesium separation from high-level waste by cesium-selective calixcrown extractants. Systems of particular interest involve novel solvent-extraction systems containing specific members of the calix[4]arene-crown-6 family, alcohol solvating agents, and alkylamines. Questions being addressed bear upon cesium binding strength, extraction selectivity, cesium stripping, and extractant solubility. Enhanced properties in this regard will specifically benefit applied projects funded by the USDOE Office of Environmental Management to clean up sites such as the Savannah River Site (SRS), Hanford, and the Idaho National Environmental and Engineering Laboratory. The most direct beneficiary will be the SRS Salt Processing Project, which has recently identified the Caustic-Side Solvent Extraction (CSSX) process employing a calixcrown as its preferred technology for cesium removal from SRS high-level tank waste. Disposal of high-level waste is horrendously expensive, in large part because the actual radioactive matter in underground waste tanks at various USDOE sites has been diluted over 1000-fold by ordinary inorganic chemicals. To vitrify the entire mass of the high-level waste would be prohibitively expensive. Accordingly, an urgent need has arisen for technologies to remove radionuclides such as {sup 137}Cs from the high-level waste so that the bulk of it may be diverted to cheaper low-level waste forms and cheaper storage. To address this need in part, chemical research at Oak Ridge National Laboratory (ORNL) has focused on calixcrown extractants, molecules that combine a crown ether with a calixarene. This hybrid possesses a cavity that is highly complementary for the Cs{sup +} ion vs. the Na+ ion, making it possible to cleanly separate cesium from wastes that contain 10,000- to 1,000,000-fold higher concentrations of sodium. Previous EMSP results in Project 55087 elucidated the underlying extraction equilibria in cesium nitrate extraction by the calixcrown used in the CSSX process, calix[4]arene-bis(t-octylbenzo-crown-6), designated here as BOBCalixC6 (see structure). This understanding led to key improvements in the development of the CSSX process under the EM Efficient Separations and Crosscutting Program, entailing a method to back-extract or 'strip' cesium from the calixcrown subsequent to cesium extraction from waste. Having this stripping method allowed the cesium to be concentrated in a relatively pure aqueous stream and the extractant to be regenerated for recycle. Closing the cycle then made possible the design of a process flowsheet and successful demonstration through collaboration with Argonne National Laboratory and Savannah River Technology Center under funding from the USDOE Office of Project Completion and Tanks Focus Area. Despite these successes, the CSSX process represents young technology that can benefit substantially from further fundamental inquiry. First, reversibility of the process (stripping efficiency) still presents the greatest potential for problems and the greatest potential for improvement. Second, although the calixcrown extractants for cesium are two orders of magnitude stronger than the next best simple crown ether, a minor fraction of the extractant capacity is utilized. Third, potassium competes significantly with cesium for the calixcrown binding site, an important issue in dealing with Hanford wastes having potassium concentrations as high as 1 M. Fourth, the calixcrown solubility needs to be improved. And finally, the mechanism of extraction must be understood in detail to provide the base of knowledge from which further development of the technology can be rationally made.

  12. Potential co-disposal of greater-than-class C low-level radioactive waste with Department of Energy special case waste - greater-than-class C low-level waste management program

    SciTech Connect (OSTI)

    Allred, W.E.

    1994-09-01

    This document evaluates the feasibility of co-disposing of greater-than-Class C low-level radioactive waste (GTCC LLW) with U.S. Department of Energy (DOE) special case waste (SCW). This document: (1) Discusses and evaluates key issues concerning co-disposal of GTCC LLW with SCW. This includes examining these issues in terms of regulatory concerns, technical feasibility, and economics; (2) Examines advantages and disadvantages of such co-disposal; and (3) Makes recommendations. Research and analysis of the issues presented in this report indicate that it would be technically and economically feasible to co-dispose of GTCC LLW with DOE SCW. However, a dilemma will likely arise in the current division of regulatory responsibilities between the U.S. Nuclear Regulatory Commission and DOE (i.e., current requirement for disposal of GTCC LLW in a facility licensed by the Nuclear Regulatory Commission). DOE SCW is currently not subject to this licensing requirement.

  13. Radioactive tank waste remediation focus area

    SciTech Connect (OSTI)

    1996-08-01

    EM`s Office of Science and Technology has established the Tank Focus Area (TFA) to manage and carry out an integrated national program of technology development for tank waste remediation. The TFA is responsible for the development, testing, evaluation, and deployment of remediation technologies within a system architecture to characterize, retrieve, treat, concentrate, and dispose of radioactive waste stored in the underground stabilize and close the tanks. The goal is to provide safe and cost-effective solutions that are acceptable to both the public and regulators. Within the DOE complex, 335 underground storage tanks have been used to process and store radioactive and chemical mixed waste generated from weapon materials production and manufacturing. Collectively, thes tanks hold over 90 million gallons of high-level and low-level radioactive liquid waste in sludge, saltcake, and as supernate and vapor. Very little has been treated and/or disposed or in final form.

  14. A pilot test of partitioning for the simulated highly saline high level waste

    SciTech Connect (OSTI)

    Chen, Jing; Wang, Jianchen; Jing, Shan

    2007-07-01

    It is a problem how to treat the highly saline high level waste (HLW). A partitioning process for HLW was developed at INET. The partitioning process includes the removal of actinides by TRPO extraction, the removal of Sr by crown ether extraction, and the removal of Cs by ion exchange. A 72-hour test was carried out in a pilot facility using the simulated HLW. Nd and Zr were used to simulate Am and Pu, respectively. The decontamination factors are >3000, >500, >1000, {approx}150 and {approx}94 for U, Nd, Zr, Sr and Cs, respectively. The results meet the requirement to change the highly saline HLW into a non-{alpha} and intermediate level waste. (authors)

  15. SIPS: A small modular process unit for the in-tank pretreatment of high-level wastes

    SciTech Connect (OSTI)

    Reich, M.; Powell, J.; Barletta, R. [Brookhaven National Lab., Upton, NY (United States)

    1996-12-31

    As a result of the U.S. weapons production program, there are now hundreds of large tanks containing highly radioactive wastes. Safe disposal of these wastes requires their processing and separations into a small volume of highly radioactive waste (HLW) and a much larger volume of low-level waste (LLW). The HLW waste would then be vitrified and transported to a geologic repository. To date, the principal approach proposed for the separation envisions a large, centralized process facility. The small in-tank processing system (SIPS) is a proposed new, small modular concept for the in-tank processing and separation of wastes into HLW and LLW output streams suitable for vitrification. Instead of pumping the retrieved tank wastes as a solid/liquid slurry over long distances to a centralized process facility, SIPS would employ a small process module, typically {approximately}1 m in diameter and 4 m long, which would be inserted into the tank. Over a period of {approx} 6 months, the module would process the solid/liquid materials in the tank, producing separated liquid HLW and liquid LLW output streams that are pumped away in two small-diameter ({approx}3-cm outside diameter) pipes. The SIPS module would be serviced by five auxiliary small pipes - a water feed pipe, a water feed pipe containing micron-size ferromagnetic particles, a nitric acid ({approx}3 M) feed pipe, and input/out pipes to hydraulically load/unload ion exchange beads.

  16. HIGH LEVEL WASTE SLUDGE BATCH 4 VARIABILITY STUDY

    SciTech Connect (OSTI)

    Fox, K; Tommy Edwards, T; David Peeler, D; David Best, D; Irene Reamer, I; Phyllis Workman, P

    2006-10-02

    The Defense Waste Processing Facility (DWPF) is preparing for vitrification of High Level Waste (HLW) Sludge Batch 4 (SB4) in early FY2007. To support this process, the Savannah River National Laboratory (SRNL) has provided a recommendation to utilize Frit 503 for vitrifying this sludge batch, based on the composition projection provided by the Liquid Waste Organization on June 22, 2006. Frit 418 was also recommended for possible use during the transition from SB3 to SB4. A critical step in the SB4 qualification process is to demonstrate the applicability of the durability models, which are used as part of the DWPF's process control strategy, to the glass system of interest via a variability study. A variability study is an experimentally-driven assessment of the predictability and acceptability of the quality of the vitrified waste product that is anticipated from the processing of a sludge batch. At the DWPF, the durability of the vitrified waste product is not directly measured. Instead, the durability is predicted using a set of models that relate the Product Consistency Test (PCT) response of a glass to the chemical composition of that glass. In addition, a glass sample is taken during the processing of that sludge batch, the sample is transmitted to SRNL, and the durability is measured to confirm acceptance. The objective of a variability study is to demonstrate that these models are applicable to the glass composition region anticipated during the processing of the sludge batch - in this case the Frit 503 - SB4 compositional region. The success of this demonstration allows the DWPF to confidently rely on the predictions of the durability/composition models as they are used in the control of the DWPF process.

  17. Level 1 Tornado PRA for the High Flux Beam Reactor

    SciTech Connect (OSTI)

    Bozoki, G.E.; Conrad, C.S.

    1994-05-01

    This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data, were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.

  18. Testimony of Mark Whitney Principal Deputy Assistant Secretary for Environmental Management Before the Subcommittee on Environment and the Economy House Energy and Commerce Committee Disposal of Low-Level Radioactive Waste and Greater-Than-Class C Waste

    Broader source: Energy.gov [DOE]

    Testimony of Mark Whitney Principal Deputy Assistant Secretary for Environmental Management Before the Subcommittee on Environment and the Economy House Energy and Commerce Committee Disposal of Low-Level Radioactive Waste and Greater-Than-Class C Waste October 28, 2015

  19. Public Invited to Comment on Draft Environmental Assessment for Replacement Capability for Disposal of Remote-Handled Low Level Radioactive Waste Generated at the U.S. Department of Energy’s Idaho Site

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy invites the public to read and comment on a draft environmental assessment it has prepared, for a proposal to provide a replacement capability for continued disposal of remote-handled low-level radioactive waste that is generated at the Idaho National Laboratory site.

  20. West Valley Demonstration Project High-Level Waste Management

    Office of Environmental Management (EM)

    of low-level salt solution into 19,877 drums of cemented LLW that were placed in storage in the Drum Cell 2006 - 07 LLW drums safely removed and successfully shipped to...

  1. Improved Alumina Loading in High-Level Waste Glasses

    SciTech Connect (OSTI)

    Kim, D.; Vienna, J.D. [Pacific Northwest National Laboratory, Richland, WA (United States); Peeler, D.K.; Fox, K.M. [Savannah River National Laboratory, Aiken, SC (United States); Aloy, A.; Trofimenko, A.V. [V.G. Khlopin Radium Institute, St. Petersburg (Russian Federation); Gerdes, K.D. [EM-21, Office of Waste Processing, U.S. Department of Energy, Washington, DC (United States)

    2008-07-01

    Recent tank retrieval, blending, and treatment strategies at both the Savannah River Site (SRS) and Hanford have identified increased amounts of high-Al{sub 2}O{sub 3} waste streams that are scheduled to be processed through their respective high-level waste (HLW) vitrification facilities. It is well known that the addition of small amounts of Al{sub 2}O{sub 3} to borosilicate glasses generally enhances the durability of the waste glasses. However, at higher Al{sub 2}O{sub 3} concentrations nepheline (NaAlSiO{sub 4}) formation can result in a severe deterioration of the chemical durability of the slowly cooled glass near the center of the canister. Additionally, higher concentrations of Al{sub 2}O{sub 3} generally increase the liquidus temperature of the melt and decrease the processing rate. Pacific Northwest National Laboratory (PNNL), Savannah River National Laboratory (SRNL), and Khlopin Radium Institute (KRI) are jointly performing laboratory and scaled-melter tests, through US Department of Energy, EM-21 Office of Waste Processing program, to develop glass formulations with increased Al{sub 2}O{sub 3} concentrations. These glasses are formulated for specific DOE waste compositions at Hanford and Savannah River Site. The objectives are to avoid nepheline formation while maintaining or meeting waste loading and/or waste throughput expectations as well as satisfying critical process and product performance related constraints such as viscosity, liquidus temperature, and glass durability. This paper summarizes the results of recent tests of simulated Hanford HLW glasses containing up to 26 wt% Al{sub 2}O{sub 3} in glass. In summary: Glasses with Al{sub 2}O{sub 3} loading ranging from 25 to 27 wt% were formulated and tested at a crucible scale. Successful glass formulations with up to 26 wt% Al{sub 2}O{sub 3} that do not precipitate nepheline during CCC treatment and had spinel crystals 1 vol% or less after 24 hr heat treatment at 950 deg. C were obtained. The selected glass, HAL-17 with 26 wt% Al{sub 2}O{sub 3}, had viscosity and electrical conductivity within the boundaries for adequate processing in the Joule heated melters operated at 1150 deg. C. This HAL-17 glass was successfully processed using small-scale (SMK) and larger scale (EP-5) melters. There was no indication of spinel settling during processing. The product glass samples from these melter tests contained 1 to 4 vol% spinel crystals that are likely formed during cooling. The PCT tests on the product glasses are underway. The present study demonstrated that it is possible to formulate the glasses with up to 26 wt% Al{sub 2}O{sub 3} that satisfy the property requirements and is processable with Joule-heated melters operated at 1150 deg. C. The 'nepheline discriminator' for HAL-17 glass is 0.45, which supports that claim that the current rule ('nepheline discriminator' < 0.62) is too restrictive. Considering that the cost of HLW treatment is highly dependent on loading of waste in glass, this result provides a potential for significant cost saving for Hanford. The maximum Al{sub 2}O{sub 3} loading that can be achieved will also depend on concentrations of other components in wastes. For example, the loading of waste used in this study was also limited by the spinel crystallization after 950 deg. C 24 hr heat treatment, which suggests that the concentrations of spinel-forming components such as Fe{sub 2}O{sub 3}, Cr{sub 2}O{sub 3}, NiO, ZnO, and MnO would be critical in addition to Al{sub 2}O{sub 3} for the maximum Al{sub 2}O{sub 3} loading achievable. The observed glass production rate per unit melter surface area of 0.75 MT/(d.m{sup 2}) for SMK test is comparable to the design capacity of WTP HLW melters at 0.8 MT/(d.m{sup 2}). However, the test with EP-5 melter achieved 0.38 MT/(d.m{sup 2}), which is roughly a half of the WTP design capacity. This result may imply that the glass with 26 wt% Al{sub 2}O{sub 3} may not achieve the WTP design production rate. However, this hypothesis is not conclusive because of unknown effects of melter size and operation

  2. Technology and System Level Demonstration of Highly Efficient...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    of Highly Efficient and Clean, Diesel Powered Class 8 Trucks Vehicle Technologies Office Merit Review 2015: Volvo SuperTruck - Powertrain Technologies for Efficiency Improvement...

  3. Application of Synergistic Technologies to Achieve High Levels...

    Broader source: Energy.gov (indexed) [DOE]

    Discussed technologies applied in highly downsized efficient gasoline engine concept such as multiple injection, advanced boosting, cooled exhaust gas recirculation, and electrical ...

  4. CENTRAL STORAGE FACILITY PROJECT IN COLOMBIA TO PROVIDE THE SAFE STORAGE AND PROTECTION OF HIGH-ACTIVITY RADIOACTIVE SOURCES

    SciTech Connect (OSTI)

    Greenberg, Raymond; Wright, Kyle A.; McCaw, Erica E.; Vallejo, Jorge

    2009-10-07

    The Global Threat Reduction Initiative (GTRI) reduces and protects vulnerable nuclear and radiological material located at civilian sites worldwide. Internationally, over 40 countries are cooperating with GTRI to enhance the security of these materials. The GTRI program has worked successfully with foreign countries to remove and protect nuclear and radioactive materials, including orphaned and disused high-activity sources. GTRI began cooperation with the Republic of Colombia in April 2004. This cooperation has been a resounding success by securing forty high-risk sites, consolidating disused/orphan sources at an interim secure national storage facility, and developing a comprehensive approach to security, training, and sustainability. In 2005 the Colombian Ministry of Mines and Energy requested the Department of Energys support in the construction of a new Central Storage Facility (CSF). In December 2005, the Ministry selected to construct this facility at the Institute of Geology and Mining (Ingeominas) site in Bogota. This site already served as Colombias national repository, where disused sources were housed in various buildings around the complex. The CSF project was placed under contract in May 2006, but environmental issues and public protests, which led to a class action lawsuit against the Colombian Government, forced the Ministry to quickly suspend activities, thereby placing the project in jeopardy. Despite these challenges, however, the Ministry of Mines and Energy worked closely with public and environmental authorities to resolve these issues, and continued to be a strong advocate of the GTRI program. In June 2008, the Ministry of Mines and Energy was granted the construction and environmental licenses. As a result, construction immediately resumed and the CSF was completed by December 2008. A commissioning ceremony was held for the new facility in January 2009, which was attended by representatives from the Department of Energy, U.S. Embassy, and the Ministry of Mines and Energy, including the Minister and Vice Minister.

  5. APPLICATION OF A THIN FILM EVAPORATOR SYSTEM FOR MANAGEMENT OF LIQUID HIGH-LEVEL WASTES AT HANFORD

    SciTech Connect (OSTI)

    TEDESCHI AR; WILSON RA

    2010-01-14

    A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORP/DOE), through Columbia Energy & Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper discusses results of pre-project pilot-scale testing by Columbia Energy and ongoing technology maturation development scope through fiscal year 2012, including planned additional pilot-scale and full-scale simulant testing and operation with actual radioactive tank waste.

  6. Advanced High-Level Waste Glass Research and Development Plan

    SciTech Connect (OSTI)

    Peeler, David K.; Vienna, John D.; Schweiger, Michael J.; Fox, Kevin M.

    2015-07-01

    The U.S. Department of Energy Office of River Protection (ORP) has implemented an integrated program to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. The integrated ORP program is focused on providing a technical, science-based foundation from which key decisions can be made regarding the successful operation of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) facilities. The fundamental data stemming from this program will support development of advanced glass formulations, key process control models, and tactical processing strategies to ensure safe and successful operations for both the low-activity waste (LAW) and high-level waste (HLW) vitrification facilities with an appreciation toward reducing overall mission life. The purpose of this advanced HLW glass research and development plan is to identify the near-, mid-, and longer-term research and development activities required to develop and validate advanced HLW glasses and their associated models to support facility operations at WTP, including both direct feed and full pretreatment flowsheets. This plan also integrates technical support of facility operations and waste qualification activities to show the interdependence of these activities with the advanced waste glass (AWG) program to support the full WTP mission. Figure ES-1 shows these key ORP programmatic activities and their interfaces with both WTP facility operations and qualification needs. The plan is a living document that will be updated to reflect key advancements and mission strategy changes. The research outlined here is motivated by the potential for substantial economic benefits (e.g., significant increases in waste throughput and reductions in glass volumes) that will be realized when advancements in glass formulation continue and models supporting facility operations are implemented. Developing and applying advanced glass formulations will reduce the cost of Hanford tank waste management by reducing the schedule for tank waste treatment and reducing the amount of HLW glass for storage, transportation, and disposal. Additional benefits will be realized if advanced glasses are developed that demonstrate more tolerance for key components in the waste (such as Al2O3, Cr2O3, SO3 and Na2O) above the currently defined WTP constraints. Tolerating these higher concentrations of key waste loading limiters may reduce the burden on (or even eliminate the need for) leaching to remove Cr and Al and washing to remove excess S and Na from the HLW fraction. Advanced glass formulations may also make direct vitrification of the HLW fraction without significant pretreatment more cost effective. Finally, the advanced glass formulation efforts seek not only to increase waste loading in glass, but also to increase glass production rate. When coupled with higher waste loading, ensuring that all of the advanced glass formulations are processable at or above the current contract processing rate leads to significant improvements in waste throughput (the amount of waste being processed per unit time),which could significantly reduce the overall WTP mission life. The integration of increased waste loading, reduced leaching/washing requirements, and improved melting rates provides a system-wide approach to improve the effectiveness of the WTP process.

  7. Next Generation Extractants for Cesium Separation from High-Level Waste: From Fundamental Concepts to Site Implementation

    SciTech Connect (OSTI)

    Moyer, Bruce A.; Bazelaire, Eve; Bonnesen, Peter V.; Bryan, Jeffrey C.; Delmau, Latitia H.; Engle, Nancy L.; Gorbunova, Maryna G.; Keever, Tamara J.; Levitskaia, Tatiana G.; Sachleben, Richard A.; Tomkins, Bruce A.

    2004-06-30

    General project objectives. This project seeks a fundamental understanding and major improvement in cesium separation from high-level waste by cesium-selective calixcrown extractants. Systems of particular interest involve novel solvent-extraction systems containing specific members of the calix[4]arene-crown-6 family, alcohol solvating agents, and alkylamines. Questions being addressed pertain to cesium binding strength, extraction selectivity, cesium stripping, and extractant solubility. Enhanced properties in this regard will specifically benefit cleanup projects funded by the USDOE Office of Environmental Management to treat and dispose of high-level radioactive wastes currently stored in underground tanks at the Savannah River Site (SRS), the Hanford site, and the Idaho National Environmental and Engineering Laboratory.1 The most direct beneficiary will be the SRS Salt Processing Project, which has recently identified the Caustic-Side Solvent Extraction (CSSX) process employing a calixcrown as its preferred technology for cesium removal from SRS high level tank waste.2 This technology owes its development in part to fundamental results obtained in this program.

  8. Next Generation Extractants for Cesium Separation from High-Level Waste: From Fundamental Concepts to Site Implementation

    SciTech Connect (OSTI)

    Moyer, Bruce A; Bazelaire, Eve; Bonnesen, Peter V.; Bryan, Jeffrey C.; Delmau, Laetitia H.; Engle, Nancy L.; Gorbunova, Maryna G.; Keever, Tamara J.; Levitskaia, Tatiana G.; Sachleben, Richard A.; Tomkins, Bruce A.; Bartsch, Richard A.; Talanov, Vladimir S.; Gibson, Harry W.; Jones, Jason W.; Hay, Benjamin P.

    2003-09-01

    This project seeks a fundamental understanding and major improvement in cesium separation from high-level waste by cesium-selective calixcrown extractants. Systems of particular interest involve novel solvent-extraction systems containing specific members of the calix[4]arene-crown-6 family, alcohol solvating agents, and alkylamines. Questions being addressed pertain to cesium binding strength, extraction selectivity, cesium stripping, and extractant solubility. Enhanced properties in this regard will specifically benefit cleanup projects funded by the USDOE Office of Environmental Management to treat and dispose of high-level radioactive wastes currently stored in underground tanks at the Savannah River Site (SRS), the Hanford site, and the Idaho National Environmental and Engineering Laboratory.1 The most direct beneficiary will be the SRS Salt Processing Project, which has recently identified the Caustic-Side Solvent Extraction (CSSX) process employing a calixcrown as its preferred technology for cesium removal from SRS high-level tank waste.2 This technology owes its development in part to fundamental results obtained in this program.

  9. LIFE ESTIMATION OF HIGH LEVEL WASTE TANK STEEL FOR F-TANK FARM CLOSURE PERFORMANCE ASSESSMENT - 9310

    SciTech Connect (OSTI)

    Subramanian, K; Bruce Wiersma, B; Stephen Harris, S

    2009-01-12

    High level radioactive waste (HLW) is stored in underground carbon steel storage tanks at the Savannah River Site. The underground tanks will be closed by removing the bulk of the waste, chemical cleaning, heel removal, stabilizing remaining residuals with tailored grout formulations, and severing/sealing external penetrations. The life of the carbon steel materials of construction in support of the performance assessment has been completed. The estimation considered general and localized corrosion mechanisms of the tank steel exposed to grouted conditions. A stochastic approach was followed to estimate the distributions of failures based upon mechanisms of corrosion accounting for variances in each of the independent variables. The methodology and results used for one-type of tank is presented.

  10. Unsaturated flow and transport through fractured rock related to high-level waste repositories; Final report, Phase 3

    SciTech Connect (OSTI)

    Evans, D.D.; Rasmussen, T.C. [Arizona Univ., Tucson, AZ (USA). Dept. of Hydrology and Water Resources

    1991-01-01

    Research results are summarized for a US Nuclear Regulatory Commission contract with the University of Arizona focusing on field and laboratory methods for characterizing unsaturated fluid flow and solute transport related to high-level radioactive waste repositories. Characterization activities are presented for the Apache Leap Tuff field site. The field site is located in unsaturated, fractured tuff in central Arizona. Hydraulic, pneumatic, and thermal characteristics of the tuff are summarized, along with methodologies employed to monitor and sample hydrologic and geochemical processes at the field site. Thermohydrologic experiments are reported which provide laboratory and field data related to the effects conditions and flow and transport in unsaturated, fractured rock. 29 refs., 17 figs., 21 tabs.

  11. Preliminary total-system analysis of a potential high-level nuclear waste repository at Yucca Mountain

    SciTech Connect (OSTI)

    Eslinger, P.W.; Doremus, L.A.; Engel, D.W.; Miley, T.B.; Murphy, M.T.; Nichols, W.E.; White, M.D. [Pacific Northwest Lab., Richland, WA (United States); Langford, D.W.; Ouderkirk, S.J. [Westinghouse Hanford Co., Richland, WA (United States)

    1993-01-01

    The placement of high-level radioactive wastes in mined repositories deep underground is considered a disposal method that would effectively isolate these wastes from the environment for long periods of time. This report describes modeling performed at PNL for Yucca Mountain between May and November 1991 addressing the performance of the entire repository system related to regulatory criteria established by the EPA in 40 CFR Part 191. The geologic stratigraphy and material properties used in this study were chosen in cooperation with performance assessment modelers at Sandia National Laboratories (SNL). Sandia modeled a similar problem using different computer codes and a different modeling philosophy. Pacific Northwest Laboratory performed a few model runs with very complex models, and SNL performed many runs with much simpler (abstracted) models.

  12. Tracer-level radioactive pilot-scale test of in situ vitrification for the stabilization of contaminated soil sites at ORNL

    SciTech Connect (OSTI)

    Spalding, B.P.; Jacobs, G.K.; Naney, M.T. ); Dunbar, N.W. ); Tixier, J.S.; Powell, T.D. )

    1992-11-01

    A field demonstration of in situ vitrification (ISV) was completed in May 1991, and produced approximately 12 Mg of melted earthen materials containing 12.7 mCi of radioactivity within 500 g of sludge in amodel of an old seepage trench waste disposal unit. Past waste disposal operations at Oak Ridge National Laboratory have left several contaminated seepage sites. In planning for remediation of such sites, ISV technology has been identified as a leading candidate because of the high risks associated with any retrieval option and because of the usual high quality of vitreous waste form. Major isotopes placed in the test trench were [sup 137]Cs and [sup 90]Sr, with lesser amounts of [sup 6O]Co, [sup 241]Am, and [sup 239,240]Pu. A total of 29 MWh of electrical power was delivered to the ground over a 5-day period producing a melt depth of 8.5 ft. During melting, 2.4% of the [sup 137]Cs volatilized from the melt into an off-gas containment hood and was captured quantitatively on a high efficiency particulate air filter. No volatilization of [sup 90]Sr, [sup 241]Am, or [sup 239,240]Pu was detected and > 99.993% retention of these isotopes in the melt was estimated. The use of added rare earth tracers (Ce, La, and Nd), as surrogates for transuranic isotopes, led to estimated melt retentions of >99.9995% during the test. The molten material, composed of the native soil and dolomitic limestone used for filling the test trench, reached a processing temperature of 1500[degrees]C. Standardized leaching procedures using Product Consistency Testing indicated that the ISV product has excellent characteristics relative to other vitreous nuclear waste forms.

  13. Tracer-level radioactive pilot-scale test of in situ vitrification for the stabilization of contaminated soil sites at ORNL

    SciTech Connect (OSTI)

    Spalding, B.P.; Jacobs, G.K.; Naney, M.T.; Dunbar, N.W.; Tixier, J.S.; Powell, T.D.

    1992-11-01

    A field demonstration of in situ vitrification (ISV) was completed in May 1991, and produced approximately 12 Mg of melted earthen materials containing 12.7 mCi of radioactivity within 500 g of sludge in amodel of an old seepage trench waste disposal unit. Past waste disposal operations at Oak Ridge National Laboratory have left several contaminated seepage sites. In planning for remediation of such sites, ISV technology has been identified as a leading candidate because of the high risks associated with any retrieval option and because of the usual high quality of vitreous waste form. Major isotopes placed in the test trench were {sup 137}Cs and {sup 90}Sr, with lesser amounts of {sup 6O}Co, {sup 241}Am, and {sup 239,240}Pu. A total of 29 MWh of electrical power was delivered to the ground over a 5-day period producing a melt depth of 8.5 ft. During melting, 2.4% of the {sup 137}Cs volatilized from the melt into an off-gas containment hood and was captured quantitatively on a high efficiency particulate air filter. No volatilization of {sup 90}Sr, {sup 241}Am, or {sup 239,240}Pu was detected and > 99.993% retention of these isotopes in the melt was estimated. The use of added rare earth tracers (Ce, La, and Nd), as surrogates for transuranic isotopes, led to estimated melt retentions of >99.9995% during the test. The molten material, composed of the native soil and dolomitic limestone used for filling the test trench, reached a processing temperature of 1500{degrees}C. Standardized leaching procedures using Product Consistency Testing indicated that the ISV product has excellent characteristics relative to other vitreous nuclear waste forms.

  14. The demonstration of continuous stirred tank reactor operations with high level waste

    SciTech Connect (OSTI)

    Peterson, R.A.

    2000-07-19

    This report contains the results of testing performed at the request of High Level Waste Engineering. These tests involved the operation of two continuous stirred tank reactors with high level waste.

  15. Road Map for Development of Crystal-Tolerant High Level Waste...

    Office of Scientific and Technical Information (OSTI)

    Road Map for Development of Crystal-Tolerant High Level Waste Glasses Citation Details In-Document Search Title: Road Map for Development of Crystal-Tolerant High Level Waste ...

  16. Integration of US Department of Energy contractor installations for the purpose of optimizing treatment, storage, and disposal of low-level radioactive waste (LLW)

    SciTech Connect (OSTI)

    Lucas, M.; Gnoose, J.; Coony, M.; Martin, E.; Piscitella, R.

    1998-02-01

    The US Department of Energy (DOE) manages a multibillion dollar environmental management (EM) program. In June 1996, the Assistant Secretary of Energy for EM issued a memorandum with guidance and a vision for a ten year planning process for the EM Program. The purpose of this process, which became known as the Accelerated Cleanup: Focus on 2006, is to make step changes within the DOE complex regarding the approach for making meaningful environmental cleanup progress. To augment the process, Assistant Secretary requested the site contractors to engage in an effort to identify and evaluate integration alternatives for EM waste stream treatment, storage, and disposal (TSD) that would parallel the 2006 Plan. In October 1996, ten DOE contractor installations began the task of identifying alternative opportunities for low level radioactive waste (LLW). Cost effective, efficient solutions were necessary to meet all requirements associated with storing, characterizing, treating, packaging, transporting, and disposing of LLW while protecting the workers` health and safety, and minimizing impacts to the environment. To develop these solutions, a systems engineering approach was used to establish the baseline requirements, to develop alternatives, and to evaluate the alternatives. Key assumptions were that unique disposal capabilities exist within the DOE that must be maintained; private sector disposal capability for some LLW may not continue to exist into the foreseeable future; and decisions made by the LLW Team must be made on a system or complex wide basis to fully realize the potential cost and schedule benefits. This integration effort promoted more accurate waste volume estimates and forecasts; enhanced recognition of existing treatment, storage, and disposal capabilities and capacities; and improved identification of cost savings across the complex.

  17. Application of Synergistic Technologies to Achieve High Levels of Gasoline

    Broader source: Energy.gov (indexed) [DOE]

    Engine Downsizing | Department of Energy Discussed technologies applied in highly downsized efficient gasoline engine concept such as multiple injection, advanced boosting, cooled exhaust gas recirculation, and electrical supercharger PDF icon deer11_boggs.pdf More Documents & Publications Meeting the CO2 Challenge DEER 2002 DEER 2007 Presentation Bosch Powertrain Technologies

  18. Radionuclides, Heavy Metals, and Polychlorinated Biphenyls in Soils Collected Around the Perimeter of Low-Level Radioactive Waste Disposal Area G during 2006

    SciTech Connect (OSTI)

    P. R. Fresquez

    2007-02-28

    Twenty-one soil surface samples were collected in March around the perimeter of Area G, the primary disposal facility for low-level radioactive solid waste at Los Alamos National Laboratory (LANL). Three more samples were collected in October around the northwest corner after elevated tritium levels were detected on an AIRNET station located north of pit 38 in May. Also, four soil samples were collected along a transect at various distances (48, 154, 244, and 282 m) from Area G, starting from the northeast corner and extending to the Pueblo de San Ildefonso fence line in a northeasterly direction (this is the main wind direction). Most samples were analyzed for radionuclides ({sup 3}H, {sup 238}Pu, {sup 239,240}Pu, {sup 241}Am, {sup 234}U, {sup 235}U, and {sup 238}U), inorganic elements (Al, Ba, Be, Ca, Cr, Co, Cu, Fe, Mg, Mn, Ni, K, Na, V, Hg, Zn, Sb, As, Cd, Pb, Se, Ag, and Tl) and polychlorinated biphenyl (PCB) concentrations. As in previous years, the highest levels of {sup 3}H in soils (690 pCi/mL) were detected along the south portion of Area G near the {sup 3}H shafts; whereas, the highest concentrations of {sup 241}Am (1.2 pCi/g dry) and the Pu isotopes (1.9 pCi/g dry for {sup 238}Pu and 5 pCi/g dry for {sup 239,240}Pu) were detected along the northeastern portions near the transuranic waste pads. Concentrations of {sup 3}H in three soil samples and {sup 241}Am and Pu isotopes in one soil sample collected around the northwest corner in October increased over concentrations found in soils collected at the same locations earlier in the year. Almost all of the heavy metals, with the exception of Zn and Sb in one sample each, in soils around the perimeter of Area G were below regional statistical reference levels (mean plus three standard deviations) (RSRLs). Similarly, only one soil sample collected on the west side contained PCB concentrations--67 {micro}g/kg dry of aroclor-1254 and 94 {micro}g/kg dry of aroclor-1260. Radionuclide and inorganic element concentrations in soils collected along a transect from Area G to the Pueblo de San Ildefonso fence line show that most contained concentrations of {sup 241}Am, {sup 238}Pu, and {sup 239,240}Pu above the RSRLs. Overall, all concentrations of radionuclides, heavy metals, and PCBs that were detected above background levels in soils collected around the perimeter of Area G and towards the Pueblo de San Ildefonso boundary were still very low and far below LANL screening levels and regulatory standards.

  19. Pump Jet Mixing and Pipeline Transfer Assessment for High-Activity Radioactive Wastes in Hanford Tank 241-AZ-102

    SciTech Connect (OSTI)

    Onishi, Yasuo; Recknagle, Kurtis P.; Wells, Beric E.

    2000-08-09

    This report evaluates how two 300-hp mixer pumps would mix solid and liquid radioactive wastes stored in Hanford double-shell Tank 241-AZ-102. It also assesses and confirms the adequacy of a 3-inch pipeline to transfer the resulting mixed waste slurry to the AP Tank Farm and ultimately to a planned waste treatment/vitrification plant on the Hanford Site.

  20. RADIOACTIVE BATTERY

    DOE Patents [OSTI]

    Birden, J.H.; Jordan, K.C.

    1959-11-17

    A radioactive battery which includes a capsule containing the active material and a thermopile associated therewith is presented. The capsule is both a shield to stop the radiations and thereby make the battery safe to use, and an energy conventer. The intense radioactive decay taking place inside is converted to useful heat at the capsule surface. The heat is conducted to the hot thermojunctions of a thermopile. The cold junctions of the thermopile are thermally insulated from the heat source, so that a temperature difference occurs between the hot and cold junctions, causing an electrical current of a constant magnitude to flow.

  1. ROAD MAP FOR DEVELOPMENT OF CRYSTAL-TOLERANT HIGH LEVEL WASTE GLASSES

    SciTech Connect (OSTI)

    Fox, K.; Peeler, D.; Herman, C.

    2014-05-15

    The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. This road map guides the research and development for formulation and processing of crystaltolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford waste glasses without jeopardizing melter operation by crystal accumulation in the melter or melter discharge riser. The potential applicability to the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) will also be addressed in this road map. The planned research described in this road map is motivated by the potential for substantial economic benefits (significant reductions in glass volumes) that will be realized if the current constraints (T1% for WTP and TL for DWPF) are approached in an appropriate and technically defensible manner for defense waste and current melter designs. 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 high waste loadings while still meeting process related limits and melter lifetime expectancies. The modeling effort will be an iterative process, where model form and a broader range of conditions, e.g., glass composition and temperature, will evolve as additional data on crystal accumulation are gathered. Model validation steps will be included to guide the development process and ensure the value of the effort (i.e., increased waste loading and waste throughput). A summary of the stages of the road map for developing the crystal-tolerant glass approach, their estimated durations, and deliverables is provided.

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

    SciTech Connect (OSTI)

    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. The incorporation of 1 wt % Pu in the glass did not adversely impact glass viscosity (as assessed using Hf surrogate) or glass durability. Finally, evaluation of DWPF glass pour samples that had Pu concentrations below the 897 g/m{sup 3} limit showed that Pu concentrations in the glass pour stream were close to targeted compositions in the melter feed indicating that Pu neither volatilized from the melt nor stratified in the melter when processed in the DWPF melter.

  3. HYPERFINE STRUCTURE CONSTANTS OF ENERGETICALLY HIGH-LYING LEVELS OF ODD PARITY OF ATOMIC VANADIUM

    SciTech Connect (OSTI)

    Gzelimen, F.; Yap?c?, B.; Demir, G.; Er, A.; ztrk, I. K.; Ba?ar, G.; Krger, S.; Tamanis, M.; Ferber, R.; Docenko, D.; Ba?ar, G. E-mail: sophie.kroeger@htw-berlin.de

    2014-09-01

    High-resolution Fourier transform spectra of a vanadium-argon plasma have been recorded in the wavelength range of 365-670 nm (15,000-27,400 cm{sup 1}). Optical bandpass filters were used in the experimental setup to enhance the sensitivity of the Fourier transform spectrometer. In total, 138 atomic vanadium spectral lines showing resolved or partially resolved hyperfine structure have been analyzed to determine the magnetic dipole hyperfine structure constants A of the involved energy levels. One of the investigated lines has not been previously classified. As a result, the magnetic dipole hyperfine structure constants A for 90 energy levels are presented: 35 of them belong to the configuration 3d {sup 3}4s4p and 55 to the configuration 3d {sup 4}4p. Of these 90 constants, 67 have been determined for the first time, with 23 corresponding to the configuration 3d {sup 3}4s4p and 44 to 3d {sup 4}4p.

  4. DOE Media Advisory- DOE extends public comment period on Draft Environmental Assessment for Replacement Capability for Disposal of Remote-Handled Low-Level Radioactive Waste Generated at the U.S. Department of Energy’s Idaho Site

    Broader source: Energy.gov [DOE]

    In response to requests from people interested in National Environmental Policy Act activities occurring at the U.S. Department of Energy’s Idaho Operations Office, the department has extended the public comment period that began September 1 on the Draft Environmental Assessment for Replacement Capability for Disposal of Remote-Handled Low-Level Radioactive Waste Generated at the U.S. Department of Energy’s Idaho Site.

  5. Radioactivity in food crops

    SciTech Connect (OSTI)

    Drury, J.S.; Baldauf, M.F.; Daniel, E.W.; Fore, C.S.; Uziel, M.S.

    1983-05-01

    Published levels of radioactivity in food crops from 21 countries and 4 island chains of Oceania are listed. The tabulation includes more than 3000 examples of 100 different crops. Data are arranged alphabetically by food crop and geographical origin. The sampling date, nuclide measured, mean radioactivity, range of radioactivities, sample basis, number of samples analyzed, and bibliographic citation are given for each entry, when available. Analyses were reported most frequently for /sup 137/Cs, /sup 40/K, /sup 90/Sr, /sup 226/Ra, /sup 228/Ra, plutonium, uranium, total alpha, and total beta, but a few authors also reported data for /sup 241/Am, /sup 7/Be, /sup 60/Co, /sup 55/Fe, /sup 3/H, /sup 131/I, /sup 54/Mn, /sup 95/Nb, /sup 210/Pb, /sup 210/Po, /sup 106/Ru, /sup 125/Sb, /sup 228/Th, /sup 232/Th, and /sup 95/Zr. Based on the reported data it appears that radioactivity from alpha emitters in food crops is usually low, on the order of 0.1 Bq.g/sup -1/ (wet weight) or less. Reported values of beta radiation in a given crop generally appear to be several orders of magnitude greater than those of alpha emitters. The most striking aspect of the data is the great range of radioactivity reported for a given nuclide in similar food crops with different geographical origins.

  6. Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)

    SciTech Connect (OSTI)

    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 cracking of Alloy 22 and titanium alloys. The effects of microbial activity and radiation on degradation of Alloy 22 and titanium alloys are also discussed. Further, for titanium alloys, the effects of fluorides, bromides, calcium ions, and galvanic coupling to less noble metals are further considered. It is concluded that, as far as materials degradation is concerned, the materials and design adopted in the U.S. Yucca Mountain Project will provide sufficient safety margins within the 10,000-years regulatory period.

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

    SciTech Connect (OSTI)

    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.

  8. Comments of Santiago Grijalva: High-Level Response to DOE RFI on Smart Grid

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Policy | Department of Energy Santiago Grijalva: High-Level Response to DOE RFI on Smart Grid Policy Comments of Santiago Grijalva: High-Level Response to DOE RFI on Smart Grid Policy High-Level Response to DOE RFI on Smart Grid Policy: This document responds to DOE questions regarding smart grid policy. The approach followed herein is to write concise comments addressing the overall RFI document at a higher level. PDF icon High-Level Response to DOE RFI on Smart Grid Policy More Documents

  9. Graphical and tabular summaries of decay characteristics for once-through PWR, LMFBR, and FFTF fuel cycle materials. [Spent fuel, high-level waste fuel can scrap

    SciTech Connect (OSTI)

    Croff, A.G.; Liberman, M.S.; Morrison, G.W.

    1982-01-01

    Based on the results of ORIGEN2 and a newly developed code called ORMANG, graphical and summary tabular characteristics of spent fuel, high-level waste, and fuel assembly structural material (cladding) waste are presented for a generic pressurized-water reactor (PWR), a liquid-metal fast breeder reactor (LMFBR), and the Fast Flux Test Facility (FFTF). The characteristics include radioactivity, thermal power, and toxicity (water dilution volume). Given are graphs and summary tables containing characteristic totals and the principal nuclide contributors as well as graphs comparing the three reactors for a single material and the three materials for a single reactor.

  10. Civilian Radioactive Waste Management System Requirements Document |

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Department of Energy Civilian Radioactive Waste Management System Requirements Document Civilian Radioactive Waste Management System Requirements Document This document specifies the top-level requirements for the Civilian Radioactive Waste Management System (CRWMS). The document is referred to herein as the CRD, for CRWMS Requirements Document. PDF icon Civilian Radioactive Waste Management System Requirements Document More Documents & Publications FY 2007 Total System Life Cycle Cost,

  11. Evaluation of high-level clouds in cloud resolving model simulations with

    Office of Scientific and Technical Information (OSTI)

    ARM and KWAJEX observations (Journal Article) | SciTech Connect Evaluation of high-level clouds in cloud resolving model simulations with ARM and KWAJEX observations Citation Details In-Document Search Title: Evaluation of high-level clouds in cloud resolving model simulations with ARM and KWAJEX observations In this paper, we evaluate high-level clouds in a cloud resolving model during two convective cases, ARM9707 and KWAJEX. The simulated joint histograms of cloud occurrence and radar

  12. The effect of high-level waste glass composition on spinel liquidus

    Office of Scientific and Technical Information (OSTI)

    temperature (Journal Article) | SciTech Connect Journal Article: The effect of high-level waste glass composition on spinel liquidus temperature Citation Details In-Document Search Title: The effect of high-level waste glass composition on spinel liquidus temperature Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni , Mn, Zn, and Ru. The liquidus temperature (TL) of spinel as the primary crystallization phase is a function of glass composition and the spinel

  13. High Level Computational Chemistry Approaches to the Prediction of Energetic Properties of Chemical Hydrogen Storage Systems

    Broader source: Energy.gov [DOE]

    Presentation on the High Level Computational Chemistry given at the DOE Theory Focus Session on Hydrogen Storage Materials on May 18, 2006.

  14. Conformance Tool High Level Design Document: IEC 61850 Cyber Security Acceleration Project

    SciTech Connect (OSTI)

    Edgar, Thomas W.

    2013-05-01

    This document is the high level design document for the Pacific Northwest National Laboratory (PNNL) IEC 62351-3, 4 and 6 standards conformance test software toolkit.

  15. NERC Presentation: Accommodating High Levels of Variable Generation, October 29, 2010

    Broader source: Energy.gov [DOE]

    North American Electric Reliability Corporation (NERC) presentation to the Electricity Advisory Committee, October 29, 2010, on accommodating high levels of variable electricity eneration. Variable...

  16. Canister storage building evaluation of nuclear safety for solidified high-level waste transfer and storage

    SciTech Connect (OSTI)

    Kidder, R.J., Westinghouse Hanford

    1996-09-17

    This document is issued to evaluate the safety impacts to the Canister Storage Building from transfer and storage of solidified high-level waste.

  17. Coupled Model for Heat and Water Transport in a High Level Waste Repository in Salt

    Broader source: Energy.gov [DOE]

    This report summarizes efforts to simulate coupled thermal-hydrological-chemical (THC) processes occurring within a generic hypothetical high-level waste (HLW) repository in bedded salt.

  18. Some logistical considerations in designing a system of deep boreholes for disposal of high-level radioactive waste.

    SciTech Connect (OSTI)

    Gray, Genetha Anne; Brady, Patrick Vane; Arnold, Bill Walter

    2012-09-01

    Deep boreholes could be a relatively inexpensive, safe, and rapidly deployable strategy for disposing Americas nuclear waste. To study this approach, Sandia invested in a three year LDRD project entitled %E2%80%9CRadionuclide Transport from Deep Boreholes.%E2%80%9D In the first two years, the borehole reference design and backfill analysis were completed and the supporting modeling of borehole temperature and fluid transport profiles were done. In the third year, some of the logistics of implementing a deep borehole waste disposal system were considered. This report describes what was learned in the third year of the study and draws some conclusions about the potential bottlenecks of system implementation.

  19. Thermal-Hydrology Simulations of Disposal of High-Level Radioactive Waste in a Single Deep Borehole

    SciTech Connect (OSTI)

    Hadgu, Teklu; Stein, Emily; Hardin, Ernest; Freeze, Geoffrey A.; Hammond, Glenn Edward

    2015-11-01

    Simulations of thermal-hydrology were carried out for the emplacement of spent nuclear fuel canisters and cesium and strontium capsules using the PFLOTRAN simulator. For the cesium and strontium capsules the analysis looked at disposal options such as different disposal configurations and surface aging of waste to reduce thermal effects. The simulations studied temperature and fluid flux in the vicinity of the borehole. Simulation results include temperature and vertical flux profiles around the borehole at selected depths. Of particular importance are peak temperature increases, and fluxes at the top of the disposal zone. Simulations of cesium and strontium capsule disposal predict that surface aging and/or emplacement of the waste at the top of the disposal zone reduces thermal effects and vertical fluid fluxes. Smaller waste canisters emplaced over a longer disposal zone create the smallest thermal effect and vertical fluid fluxes no matter the age of the waste or depth of emplacement.

  20. ENVIRONMENTAL RADIOACTIVITY

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    NVO-410-15 ENVIRONMENTAL RADIOACTIVITY "* -AT THE -NEVADA TEST SITE July, 1966 through June, 1967 REYNOLDS ELECTRICAL & ENGINEERING CO., INC. ENVIRONMENTAL SCIENCES DEPARTMENT LAS VEGAS, NEVADA 89114 MAY, 1968 PREPARED FOR THE U. S. ATOMIC ENERGY COMMISSION, NEVADA OPERATIONS OFFICE UNDER CONTRACT NO. AT(26-I)-410 ENVIRONMENTAL RADIOAC~TY AT THE NEVADA TEST SITE JULY, 1966 THROUGH JUNE, 1967 MAY, 1968 REYNOLDS ELECTRICAL AND ENGID CO., INC. AN &&#iMW COMPANY ACKNOWLEDGEMENTS The

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

    SciTech Connect (OSTI)

    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.

  2. Evaluation of high-level clouds in cloud resolving model simulations...

    Office of Scientific and Technical Information (OSTI)

    Evaluation of high-level clouds in cloud resolving model simulations with ARM and KWAJEX observations: HIGH CLOUD IN CRM Citation Details In-Document Search This content will ...

  3. Alternatives Generation and Analysis for Heat Removal from High Level Waste Tanks

    SciTech Connect (OSTI)

    WILLIS, W.L.

    2000-06-15

    This document addresses the preferred combination of design and operational configurations to provide heat removal from high-level waste tanks during Phase 1 waste feed delivery to prevent the waste temperature from exceeding tank safety requirement limits. An interim decision for the preferred method to remove the heat from the high-level waste tanks during waste feed delivery operations is presented herein.

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

    SciTech Connect (OSTI)

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

    1998-04-01

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

  5. 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

    SciTech Connect (OSTI)

    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.

  6. Technical survey of DOE programs and facilities applicable to the co-storage of commercial greater-than-Class C Low-Level Radioactive Waste and DOE special Case Waste

    SciTech Connect (OSTI)

    Allred, W.E.

    1995-01-01

    This report presents information on those US Department of Energy (DOE) management programs and facilities, existing and planned, that are potentially capable of storing DOE Special Case Waste (SCW) until a disposal capability is available. Major emphasis is given to the possibility of storing commercial greater-than-Class C low-level radioactive waste (GTCC LLW) together with DOE SCW, as well as with other waste types. In addition to this primary issue, the report gives an in-depth background on SCW and GTCC LLW, and discusses their similarities. Institutional issues concerning these waste types are not addressed in this report.

  7. A Prototype Performance Assessment Model for Generic Deep Borehole Repository for High-Level Nuclear Waste - 12132

    SciTech Connect (OSTI)

    Lee, Joon H.; Arnold, Bill W.; Swift, Peter N.; Hadgu, Teklu; Freeze, Geoff; Wang, Yifeng

    2012-07-01

    A deep borehole repository is one of the four geologic disposal system options currently under study by the U.S. DOE to support the development of a long-term strategy for geologic disposal of commercial used nuclear fuel (UNF) and high-level radioactive waste (HLW). The immediate goal of the generic deep borehole repository study is to develop the necessary modeling tools to evaluate and improve the understanding of the repository system response and processes relevant to long-term disposal of UNF and HLW in a deep borehole. A prototype performance assessment model for a generic deep borehole repository has been developed using the approach for a mined geological repository. The preliminary results from the simplified deep borehole generic repository performance assessment indicate that soluble, non-sorbing (or weakly sorbing) fission product radionuclides, such as I-129, Se-79 and Cl-36, are the likely major dose contributors, and that the annual radiation doses to hypothetical future humans associated with those releases may be extremely small. While much work needs to be done to validate the model assumptions and parameters, these preliminary results highlight the importance of a robust seal design in assuring long-term isolation, and suggest that deep boreholes may be a viable alternative to mined repositories for disposal of both HLW and UNF. (authors)

  8. Long-Term Waste Package Degradation Studies at the Yucca Mountain Potential High-Level Nuclear Waste Repository

    SciTech Connect (OSTI)

    Mon, K. G.; Bullard, B. E.; Longsine, D. E.; Mehta, S.; Lee, J. H.; Monib, A. M.

    2002-02-26

    The Site Recommendation (SR) process for the potential repository for spent nuclear fuel (SNF) and high-level nuclear waste (HLW) at Yucca Mountain, Nevada is underway. Fulfillment of the requirements for substantially complete containment of the radioactive waste emplaced in the potential repository and subsequent slow release of radionuclides from the Engineered Barrier System (EBS) into the geosphere will rely on a robust waste container design, among other EBS components. Part of the SR process involves sensitivity studies aimed at elucidating which model parameters contribute most to the drip shield and waste package degradation characteristics. The model parameters identified included (a) general corrosion rate model parameters (temperature-dependence and uncertainty treatment), and (b) stress corrosion cracking (SCC) model parameters (uncertainty treatment of stress and stress intensity factor profiles in the Alloy 22 waste package outer barrier closure weld regions, the SCC initiation stress threshold, and the fraction of manufacturing flaws oriented favorably for through-wall penetration by SCC). These model parameters were reevaluated and new distributions were generated. Also, early waste package failures due to improper heat treatment were added to the waste package degradation model. The results of these investigations indicate that the waste package failure profiles are governed by the manufacturing flaw orientation model parameters and models used.

  9. Transportation of Spent Nuclear Fuel and High Level Waste to Yucca Mountain: The Next Step in Nevada

    SciTech Connect (OSTI)

    Sweeney, Robin L,; Lechel, David J.

    2003-02-25

    In the U.S. Department of Energy's ''Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada,'' the Department states that certain broad transportation-related decisions can be made. These include the choice of a mode of transportation nationally (mostly legal-weight truck or mostly rail) and in Nevada (mostly rail, mostly legal-weight truck, or mostly heavy-haul truck with use of an associated intermodal transfer station), as well as the choice among alternative rail corridors or heavy-haul truck routes with use of an associated intermodal transfer station in Nevada. Although a rail line does not service the Yucca Mountain site, the Department has identified mostly rail as its preferred mode of transportation, both nationally and in the State of Nevada. If mostly rail is selected for Nevada, the Department would then identify a preference for one of the rail corridors in consultation with affected stakeholders, particularly the State of Nevada. DOE would then select the rail corridor and initiate a process to select a specific rail alignment within the corridor for the construction of a rail line. Five proposed rail corridors were analyzed in the Final Environmental Impact Statement. The assessment considered the impacts of constructing a branch rail line in the five 400-meter (0.25mile) wide corridors. Each corridor connects the Yucca Mountain site with an existing mainline railroad in Nevada.

  10. Annual report, spring 2015. Alternative chemical cleaning methods for high level waste tanks-corrosion test results

    SciTech Connect (OSTI)

    Wyrwas, R. B.

    2015-07-06

    The testing presented in this report is in support of the investigation of the Alternative Chemical Cleaning program to aid in developing strategies and technologies to chemically clean radioactive High Level Waste tanks prior to tank closure. The data and conclusions presented here were the examination of the corrosion rates of A285 carbon steel and 304L stainless steel when interacted with the chemical cleaning solution composed of 0.18 M nitric acid and 0.5 wt. % oxalic acid. This solution has been proposed as a dissolution solution that would be used to remove the remaining hard heel portion of the sludge in the waste tanks. This solution was combined with the HM and PUREX simulated sludge with dilution ratios that represent the bulk oxalic cleaning process (20:1 ratio, acid solution to simulant) and the cumulative volume associated with multiple acid strikes (50:1 ratio). The testing was conducted over 28 days at 50°C and deployed two methods to invest the corrosion conditions; passive weight loss coupon and an active electrochemical probe were used to collect data on the corrosion rate and material performance. In addition to investigating the chemical cleaning solutions, electrochemical corrosion testing was performed on acidic and basic solutions containing sodium permanganate at room temperature to explore the corrosion impacts if these solutions were to be implemented to retrieve remaining actinides that are currently in the sludge of the tank.

  11. SRS Workers Moved Millions of Gallons of High-Level Waste Safely in 2014 |

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Department of Energy SRS Workers Moved Millions of Gallons of High-Level Waste Safely in 2014 SRS Workers Moved Millions of Gallons of High-Level Waste Safely in 2014 February 26, 2015 - 12:00pm Addthis A view of the Savannah River Site, which includes underground waste tanks and facilities. A view of the Savannah River Site, which includes underground waste tanks and facilities. AIKEN, S.C. - EM and its liquid waste contractor safely transferred more than 20 million gallons of high-level

  12. Low-Activity Waste and High-Level Waste Feed Processing Data Quality Objectives

    SciTech Connect (OSTI)

    Patello, Gertrude K. ); Truex, Michael J. ); Wiemers, Karyn D.

    1999-04-15

    This document describes characterization needs for the DOE waste feed processing and disposal management of TWRS Privatization Phase I. The DOE must obtain information to evaluate and minimize risk associated with the private contractor's design phase deliverables (April 2000); authorization to proceed with Part B-2 (August 2000); and start of treatment facility construction (July 2001). Additionally, the DOE must ensure that the contract feed and product specifications are adequate and achievable, and that there is a sufficient basis for negotiating the price for. services. The purpose of this Data Quality Objective (DQO) is to provide data to accomplish the following: (1) update waste characterization information from source tanks to provide an independent assessment that the specifications and Interface Control Documents (ICDS) are adequate for DOE's management of the site M&I contractor and private contractor contracts; (2) provide preliminary information for the private contractor's process and facility designs and DOE's review of the designs in preparation for the authorization to proceed with Phase I Part B-2; (3) provide preliminary information for ILAW and IHLW storage and disposal design/specifications; (4) support update of the ILAW performance assessment (PA) for disposal; (5) help substantiate the ability to (1) comply with U.S. Nuclear Regulatory Commission (NRC) guidelines for incidental waste for LAW and (2) comply with the Office of Civilian Radioactive Waste Management (OCRWM) requirements for disposal of IHLW. The sampling and characterization implemented as a result of this DQO will be used for planning. As a result, a majority of the alternative actions fall into two major categories: either the DOE Privatization Contract is renegotiated or the process/facility designs are adjusted to accommodate increased capacity requirements, new technologies, additional waste stream volumes, etc. Impacts to the DOE are reduced when the need for these fallback positions are realized or eliminated early in the planning process. This DQO replaces earlier separate low-activity waste feed data quality objectives (Truex and Wiemers 1998) and high-level waste feed data quality objectives documents (Wiemers et,al. 1998). This combined DQO updates the data requirements based on the TWRS Privatization Contact issued August 1998 (DOE-RL 1998). Regulatory compliance for TWRS Privatization is addressed in a separate DQO (Wiemers et al. 1998). Additional characterization of the Phase I waste feed will be performed by DOE's contractors: the M&I contractor and the private contractor. Characterization for feed certification and waste acceptance will be completed before transfer of the feed to the private contractor facility. Characterization requirements for staged feed will be identified in other DQOS consistent with the Feed Certification Plans, ICDS 19 and 20, and applicable permits. Newly obtained analytical data and contract changes that have become available in parallel with or subsequent to preparation of this DQO update will be assessed and incorporated into the data needs optimization in the next revision of this DQO. Data available at the time of the tank waste sample request will be considered in the development of the Tank Sampling and Analysis Plan.

  13. Evaluation of high-level clouds in cloud resolving model simulations with ARM and KWAJEX observations

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Liu, Zheng; Muhlbauer, Andreas; Ackerman, Thomas

    2015-11-05

    In this paper, we evaluate high-level clouds in a cloud resolving model during two convective cases, ARM9707 and KWAJEX. The simulated joint histograms of cloud occurrence and radar reflectivity compare well with cloud radar and satellite observations when using a two-moment microphysics scheme. However, simulations performed with a single moment microphysical scheme exhibit low biases of approximately 20 dB. During convective events, two-moment microphysical overestimate the amount of high-level cloud and one-moment microphysics precipitate too readily and underestimate the amount and height of high-level cloud. For ARM9707, persistent large positive biases in high-level cloud are found, which are not sensitivemore » to changes in ice particle fall velocity and ice nuclei number concentration in the two-moment microphysics. These biases are caused by biases in large-scale forcing and maintained by the periodic lateral boundary conditions. The combined effects include significant biases in high-level cloud amount, radiation, and high sensitivity of cloud amount to nudging time scale in both convective cases. The high sensitivity of high-level cloud amount to the thermodynamic nudging time scale suggests that thermodynamic nudging can be a powerful ‘‘tuning’’ parameter for the simulated cloud and radiation but should be applied with caution. The role of the periodic lateral boundary conditions in reinforcing the biases in cloud and radiation suggests that reducing the uncertainty in the large-scale forcing in high levels is important for similar convective cases and has far reaching implications for simulating high-level clouds in super-parameterized global climate models such as the multiscale modeling framework.« less

  14. HLW-OVP-94-00n High Level Waste Management Division HLW System...

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    4-00n High Level Waste Management Division HLW System Plan Revision 3 (U) Westinghouse Savannah River Company Aiken, South Carolina May 31, 1994 Westinghouse Savannah River Company ...

  15. Summary of expenditures of rebates from the low-level radioactive waste surcharge escrow account for calendar year 1989: Report to Congress in response to Public Law 99-240

    SciTech Connect (OSTI)

    Not Available

    1990-06-01

    This response is submitted in response to the Low-Level Radioactive Waste Policy Amendments Act of 1985 (the Act), Public Law 99-240. The report summarizes expenditures made during the calendar year 1989 of surcharge rebates from the July 1, 1986, milestones. Title I of the Act requires the Department of Energy (DOE) to administer a Surcharge Escrow Account. This account consists of a portion of the surcharge fees paid by generators of low-level radioactive waste in nonsited compacts (regional compacts currently without operating disposal sites) and nonmember States (States without disposal sites that are not members of compacts) to the three States with operating disposal facilities (Nevada, South Carolina, and Washington) (sited States) for using their disposal facilities. In administering the Surcharge Escrow Account, the Act requires DOE to: invest the funds in interest-bearing United States Government securities; determine eligibility of rebates of the funds by evaluating State and compact progress toward developing new disposal sites against milestones set forth in the Act; disburse the collected rebates and interest; assess compliance of rebate expenditures with the limitations prescribed in the Act; and submit a report annually to Congress summarizing rebate expenditures by States and regions. 5 tabs.

  16. The Management of the Radioactive Waste Generated by Cernavoda NPP, Romania, an Example of International Cooperation - 13449

    SciTech Connect (OSTI)

    Barariu, Gheorghe

    2013-07-01

    The design criteria and constraints for the development of the management strategy for radioactive waste generated from operating and decommissioning of CANDU Nuclear Units from Cernavoda NPP in Romania, present many specific aspects. The main characteristics of CANDU type waste are its high concentrations of tritium and radiocarbon. Also, the existing management strategy for radioactive waste at Cernavoda NPP provides no treatment or conditioning for radioactive waste disposal. These characteristics embodied a challenging effort, in order to select a proper strategy for radioactive waste management at present, when Romania is an EU member and a signatory country of the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management. The helping of advanced countries in radioactive waste management, directly or into the frame of the international organizations, like IAEA, become solve the aforementioned challenges at adequate level. (authors)

  17. Biofuels and Food Security. A report by the High Level Panel of Experts on

    Office of Scientific and Technical Information (OSTI)

    Food Security and Nutrition (Miscellaneous) | SciTech Connect Miscellaneous: Biofuels and Food Security. A report by the High Level Panel of Experts on Food Security and Nutrition Citation Details In-Document Search Title: Biofuels and Food Security. A report by the High Level Panel of Experts on Food Security and Nutrition In October 2011, the UN Committee on World Food Security (CFS) recommended a ''review of biofuels policies -- where applicable and if necessary -- according to balanced

  18. Mixing Processes in High-Level Waste Tanks (Technical Report) | SciTech

    Office of Scientific and Technical Information (OSTI)

    Connect Technical Report: Mixing Processes in High-Level Waste Tanks Citation Details In-Document Search Title: Mixing Processes in High-Level Waste Tanks Mixing and transport in large waste-tank volumes is controlled by the multidimensional equations describing mass, momentum and energy conservation, and by boundary conditions imposed at walls, structures, and fluid inlets and outlets. For large enclosures, careful scaling arguments show that mixing is generated by free buoyant jets arising

  19. Characterization of High Level Waste from a Hybrid LIFE Engine for Enhanced

    Office of Scientific and Technical Information (OSTI)

    Repository Performance (Technical Report) | SciTech Connect Technical Report: Characterization of High Level Waste from a Hybrid LIFE Engine for Enhanced Repository Performance Citation Details In-Document Search Title: Characterization of High Level Waste from a Hybrid LIFE Engine for Enhanced Repository Performance Authors: Beckett, E ; Fratoni, M Publication Date: 2010-08-25 OSTI Identifier: 1119948 Report Number(s): LLNL-TR-455920 DOE Contract Number: W-7405-ENG-48 Resource Type:

  20. Statement on Launch of the U.S.-Republic of Korea High Level Bilateral

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Commission Pursuant to the Agreement for Cooperation between the United States and Republic of Korea Concerning Peaceful Uses of Nuclear Energy | Department of Energy on Launch of the U.S.-Republic of Korea High Level Bilateral Commission Pursuant to the Agreement for Cooperation between the United States and Republic of Korea Concerning Peaceful Uses of Nuclear Energy Statement on Launch of the U.S.-Republic of Korea High Level Bilateral Commission Pursuant to the Agreement for Cooperation

  1. Emerging Issues and Challenges with Integrating High Levels of Solar into

    Office of Environmental Management (EM)

    the Electrical Generation and Transmission Systems | Department of Energy Electrical Generation and Transmission Systems Emerging Issues and Challenges with Integrating High Levels of Solar into the Electrical Generation and Transmission Systems Emerging Issues and Challenges with Integrating High Levels of Solar into the Electrical Generation and Transmission Systems Increasing the use of grid-flexibility options (improved grid management, demand response, and energy storage) could enable

  2. Radioactive Waste Management

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    1984-02-06

    To establish policies and guidelines by which the Department of Energy (DOE) manages tis radioactive waste, waste byproducts, and radioactively contaminated surplus facilities.

  3. Radioactive Material Transportation Practices

    Broader source: Directives, Delegations, and Requirements [Office of Management (MA)]

    2002-09-23

    Establishes standard transportation practices for Departmental programs to use in planning and executing offsite shipments of radioactive materials including radioactive waste. Does not cancel other directives.

  4. EFFECT OF A HIGH OPACITY ON THE LIGHT CURVES OF RADIOACTIVELY POWERED TRANSIENTS FROM COMPACT OBJECT MERGERS

    SciTech Connect (OSTI)

    Barnes, Jennifer; Kasen, Daniel [Departments of Physics and Astronomy, 366 LeConte Hall, University of California, Berkeley, CA 94720 (United States)

    2013-09-20

    The coalescence of compact objects is a promising astrophysical source of detectable gravitational wave signals. The ejection of r-process material from such mergers may lead to a radioactively powered electromagnetic counterpart signal which, if discovered, would enhance the science returns. As very little is known about the optical properties of heavy r-process elements, previous light-curve models have adopted opacities similar to those of iron group elements. Here we consider the effect of heavier elements, particularly the lanthanides, which increase the ejecta opacity by several orders of magnitude. We include these higher opacities in time-dependent, multi-wavelength radiative transport calculations to predict the broadband light curves of one-dimensional models over a range of parameters (ejecta masses {approx}10{sup -3}-10{sup -1} M{sub Sun} and velocities {approx}0.1-0.3 c). We find that the higher opacities lead to much longer duration light curves which can last a week or more. The emission is shifted toward the infrared bands due to strong optical line blanketing, and the colors at later times are representative of a blackbody near the recombination temperature of the lanthanides (T {approx} 2500 K). We further consider the case in which a second mass outflow, composed of {sup 56}Ni, is ejected from a disk wind, and show that the net result is a distinctive two component spectral energy distribution, with a bright optical peak due to {sup 56}Ni and an infrared peak due to r-process ejecta. We briefly consider the prospects for detection and identification of these transients.

  5. 1989 OCRWM [Office of Civilian Radioactive Waste Management] Bulletin compilation and index

    SciTech Connect (OSTI)

    1990-02-01

    The OCRWM Bulletin is published by the Department of Energy, Office of Civilian Radioactive Waste Management to provide current information about the national program for managing spent fuel and high-level radioactive waste. This document is a compilation of issues from the 1989 calendar year. A table of contents and one index have been provided to assist in finding information contained in this year`s Bulletins. The pages have been numbered consecutively at the bottom for easy reference. 7 figs.

  6. A simulation of the transport and fate of radon-222 derived from thorium-230 low-level waste in the near-surface zone of the Radioactive Waste Management Site in Area 5 of the Nevada Test Site

    SciTech Connect (OSTI)

    Lindstrom, F.T.; Cawlfield, D.E.; Donahue, M.E.; Emer, D.F.; Shott, G.J.

    1993-12-01

    US Department of Energy (DOE) Order 5820.2A (DOE, 1988) requires performance assessments on all new and existing low-level radioactive waste (LLW) disposal sites. An integral part of performance assessment is estimating the fluxes of radioactive gases such as radon-220 and radon-222. Data needs pointed out by mathematical models drive site characterization. They provide a logical means of performing the required flux estimations. Thorium-230 waste, consisting largely of thorium hydroxide and thorium oxides, has been approved for disposal in shallow trenches and pits at the LLW Radioactive Waste Management Site in Area 5 of the Nevada Test Site. A sophisticated gas transport model, CASCADR8 (Lindstrom et al., 1992b), was used to simulate the transport and fate of radon-222 from its source of origin, nine feet below a closure cap of native soil, through the dry alluvial earth, to its point of release into the atmosphere. CASCADR8 is an M-chain gas-phase radionuclide transport and fate model. It has been tailored to the site-specific needs of the dry desert environment of southern Nevada. It is based on the mass balance principle for each radionuclide and uses gas-phase diffusion as well as barometric pressure-induced advection as its main modes of transport. CASCADR8 uses both reversible and irreversible sorption kinetic rules as well as the usual classical Bateman (1910) M-chain decay rules for its kinetic processes. Worst case radon-222 gas-phase concentrations, as well as surface fluxes, were estimated over 40 days. The maximum flux was then used in an exposure assessment model to estimate the total annual dose equivalent received by a person residing in a standard 2500-square-foot house with 10-foot walls. Results are described.

  7. The high level programmer and user interface of the NSLS control system

    SciTech Connect (OSTI)

    Tang, Y.N.; Smith, J.D.; Sathe, S.

    1993-07-01

    This paper presents the major components of the high level software in the NSLS upgraded control system. Both programmer and user interfaces are discussed. The use of the high-speed work stations, fast network communications, UNIX system, X-window and Motif have greatly changed and improved these interfaces.

  8. Integrated High-Level Waste System Planning - Utilizing an Integrated Systems Planning Approach to Ensure End-State Definitions are Met and Executed - 13244

    SciTech Connect (OSTI)

    Ling, Lawrence T.; Chew, David P.

    2013-07-01

    The Savannah River Site (SRS) is a Department of Energy site which has produced nuclear materials for national defense, research, space, and medical programs since the 1950's. As a by-product of this activity, approximately 37 million gallons of high-level liquid waste containing approximately 292 million curies of radioactivity is stored on an interim basis in 45 underground storage tanks. Originally, 51 tanks were constructed and utilized to support the mission. Four tanks have been closed and taken out of service and two are currently undergoing the closure process. The Liquid Waste System is a highly integrated operation involving safely storing liquid waste in underground storage tanks; removing, treating, and dispositioning the low-level waste fraction in grout; vitrifying the higher activity waste at the Defense Waste Processing Facility; and storing the vitrified waste in stainless steel canisters until permanent disposition. After waste removal and processing, the storage and processing facilities are decontaminated and closed. A Liquid Waste System Plan (hereinafter referred to as the Plan) was developed to integrate and document the activities required to disposition legacy and future High-Level Waste and to remove from service radioactive liquid waste tanks and facilities. It establishes and records a planning basis for waste processing in the liquid waste system through the end of the program mission. The integrated Plan which recognizes the challenges of constrained funding provides a path forward to complete the liquid waste mission within all regulatory and legal requirements. The overarching objective of the Plan is to meet all Federal Facility Agreement and Site Treatment Plan regulatory commitments on or ahead of schedule while preserving as much life cycle acceleration as possible through incorporation of numerous cost savings initiatives, elimination of non-essential scope, and deferral of other scope not on the critical path to compliance. There is currently a premium on processing and storage space in the radioactive liquid waste tank system. To enable continuation of risk reduction initiatives, the Plan establishes a processing strategy that provides tank space required to meet, or minimizes the impacts to meeting, programmatic objectives. The Plan also addresses perturbations in funding and schedule impacts. (authors)

  9. EIS-0062: Double-Shell Tanks for Defense High Level Waste Storage, Savannah River Site, Aiken, SC

    Broader source: Energy.gov [DOE]

    This EIS analyzes the impacts of the various design alternatives for the construction of fourteen 1.3 million gallon high-activity radioactive waste tanks. The EIS further evaluates the effects of these alternative designs on tank durability, on the ease of waste retrieval from such tanks, and the choice of technology and timing for long-term storage or disposal of the wastes.

  10. FOSTERING MULTI-LATERAL COOPERATION BETWEEN THE GOVERNMENTS OF DOMINICAN REPUBLIC, COLOMBIA, AND THE UNITED STATES TO ENHANCE THE PROTECTION OF HIGH-ACTIVITY RADIOACTIVE SOURCES

    SciTech Connect (OSTI)

    Butler, Nicholas; McCaw, Erica E.; Wright, Kyle A.; Medina, Maximo

    2009-10-06

    The Global Threat Reduction Initiative (GTRI) reduces and protects vulnerable nuclear and radiological material located at civilian sites worldwide from sabotage, theft or diversion. The GTRI has worked successfully with foreign countries to remove and protect nuclear and radioactive materials including high-activity sources used in medical, commercial, and research applications. There are many barriers to successful bilateral cooperation that must be overcome including language, preconceived perceptions, long distances, and different views on the threat and protection requirements. Successful cooperation is often based on relationships and building trusting relationships takes time. In the case of Dominican Republic, the GTRI first received contact in 2008 from the Government of Dominican Republic. They requested cooperation that was similar to the tri-partite cooperation between Colombia, Mexico and the United States. Throughout the region it was widely known that the GTRIs cooperation with the Government of Colombia was a resounding success resulting in the securing of forty sites; the consolidation of numerous disused/orphan sources at a secure national storage facility; and, the development of a comprehensive approach to security including, inter alia, training and sustainability. The Government of Colombia also showcased this comprehensive approach to thirteen Central American and Caribbean countries at a GTRI regional security conference held in Panama in October 2004. In 2007, Colombia was an integral component of GTRI multi-lateral cooperation initiation in Mexico. As a result, twenty two of Mexicos largest radioactive sites have been upgraded in the past eighteen months. These two endeavors served as catalysts for cooperation opportunities in the Dominican Republic. Representatives from the Colombian government were aware of GTRIs interest in initiating cooperation with the Government of Dominican Republic and to facilitate this cooperation, they traveled to the Dominican Republic and provided briefings and presentations which demonstrated its successful cooperation with the GTRI. Shortly after that visit, the Government of Dominican Republic agreed to move forward and requested that the cooperative efforts in Dominican Republic be performed in a tripartite manner, leveraging the skills, experience, and resources of the Colombians, and the financial and technical support of the United States. As a result, two of Dominican Republics largest radioactive sites had security upgrades in place within 90 days of the cooperation visit in December 2008.

  11. Material for radioactive protection

    DOE Patents [OSTI]

    Taylor, R.S.; Boyer, N.W.

    A boron containing burn resistant, low-level radiation protection material useful, for example, as a liner for radioactive waste disposal and storage, a component for neutron absorber, and a shield for a neutron source is described. The material is basically composed of borax in the range of 25 to 50%, coal tar in the range of 25 to 37.5%, with the remainder being an epoxy resin mix. A preferred composition is 50% borax, 25% coal tar and 25% epoxy resin. The material is not susceptible to burning and is about 1/5 the cost of existing radiation protection material utilized in similar applications.

  12. FOSTERING MULTI-LATERAL COOPERATION BETWEEN THE GOVERNMENTS OF MEXICO, COLOMBIA, AND THE UNITED STATES TO ENHANCE THE PROTECTION OF HIGH-ACTIVITY RADIOACTIVE SOURCES

    SciTech Connect (OSTI)

    Butler, Nicholas; Watson, Erica E.; Wright, Kyle A.

    2009-10-07

    The Global Threat Reduction Initiative (GTRI) reduces and protects vulnerable nuclear and radiological material located at civilian sites worldwide from sabotage, theft or diversion. The GTRI program has worked successfully with foreign countries to remove and protect nuclear and radioactive materials including high-activity sources used in medical, commercial, and research applications. There are many barriers to successful bilateral cooperation that must be overcome including language, preconceived perceptions, long distances, and different views on the threat and protection requirements. Successful cooperation is often based on relationships and building trusting relationships takes time. In the case of Mexico, GTRI first made contact in 2005. The project then lost momentum and stalled. At the same time, GTRIs cooperation with the Republic of Colombia was a resounding success resulting in the securing of forty sites; the consolidation of numerous disused/orphan sources at a secure national storage facility; and, the development of a comprehensive approach to security including, inter alia, training and sustainability. The government of Colombia also showcased this comprehensive approach to thirteen Central American and Caribbean countries at a GTRI regional security conference held in Panama in October 2004. Representatives from the Colombian government were aware of GTRIs interest in initiating cooperation with the Government of Mexico and to facilitate this cooperation, they offered to invite their Mexican counterparts to Colombia to observe its successful cooperation with GTRI. Shortly after that visit, the Government of Mexico agreed to move forward and requested that the cooperative efforts in Mexico be performed in a tripartite manner, leveraging the skills, experience, and resources of the Colombians. As a result, 22 of Mexicos largest radioactive sites have had security upgrades in place within 18 months of cooperation.

  13. Containment at the Source during Waste Volume Reduction of Large Radioactive Components Using Oxylance High-Temperature Cutting Equipment - 13595

    SciTech Connect (OSTI)

    Keeney, G. Neil

    2013-07-01

    As a waste-volume reduction and management technique, highly contaminated Control Element Drive Mechanism (CEDM) housings were severed from the Reactor Pressure Vessel Head (RPVH) inside the San Onofre Unit 2 primary containment utilizing Oxylance high-temperature cutting equipment and techniques. Presented are relevant data concerning: - Radiological profiles of the RPVH and individual CEDMs; - Design overviews of the engineering controls and the specialized confinement housings; - Utilization of specialized shielding; - Observations of apparent metallurgical-contamination coalescence phenomena at high temperatures resulting in positive control over loose-surface contamination conditions; - General results of radiological and industrial hygiene air sampling and monitoring; - Collective dose and personnel contamination event statistics; - Lessons learned. (author)

  14. HIGH LEVEL WASTE (HLW) VITRIFICATION EXPERIENCE IN THE US: APPLICATION OF GLASS PRODUCT/PROCESS CONTROL TO OTHERHLW AND HAZARDOUS WASTES

    SciTech Connect (OSTI)

    Jantzen, C; James Marra, J

    2007-09-17

    Vitrification is currently the most widely used technology for the treatment of high level radioactive wastes (HLW) throughout the world. At the Savannah River Site (SRS) actual HLW tank waste has successfully been processed to stringent product and process constraints without any rework into a stable borosilicate glass waste since 1996. A unique 'feed forward' statistical process control (SPC) has been used rather than statistical quality control (SQC). In SPC, the feed composition to the melter is controlled prior to vitrification. In SQC, the glass product is sampled after it is vitrified. Individual glass property models form the basis for the 'feed forward' SPC. The property models transform constraints on the melt and glass properties into constraints on the feed composition. The property models are mechanistic and depend on glass bonding/structure, thermodynamics, quasicrystalline melt species, and/or electron transfers. The mechanistic models have been validated over composition regions well outside of the regions for which they were developed because they are mechanistic. Mechanistic models allow accurate extension to radioactive and hazardous waste melts well outside the composition boundaries for which they were developed.

  15. Corrosion resistant storage container for radioactive material

    DOE Patents [OSTI]

    Schweitzer, D.G.; Davis, M.S.

    1984-08-30

    A corrosion resistant long-term storage container for isolating high-level radioactive waste material in a repository is claimed. The container is formed of a plurality of sealed corrosion resistant canisters of different relative sizes, with the smaller canisters housed within the larger canisters, and with spacer means disposed between juxtaposed pairs of canisters to maintain a predetermined spacing between each of the canisters. The combination of the plural surfaces of the canisters and the associated spacer means is effective to make the container capable of resisting corrosion, and thereby of preventing waste material from leaking from the innermost canister into the ambient atmosphere.

  16. SRS Crosses Halfway Mark on Closing Another High-Level Waste Tank |

    Energy Savers [EERE]

    Department of Energy SRS Crosses Halfway Mark on Closing Another High-Level Waste Tank SRS Crosses Halfway Mark on Closing Another High-Level Waste Tank February 25, 2016 - 12:20pm Addthis Work is more than halfway complete on grouting Tank 12, the eighth to be operationally closed at Savannah River Site. Work is more than halfway complete on grouting Tank 12, the eighth to be operationally closed at Savannah River Site. AIKEN, S.C. - Savannah River Site (SRS) moved past the halfway mark

  17. Tank waste remediation system phase I high-level waste feed processability assessment report

    SciTech Connect (OSTI)

    Lambert, S.L.; Stegen, G.E., Westinghouse Hanford

    1996-08-01

    This report evaluates the effects of feed composition on the Phase I high-level waste immobilization process and interim storage facility requirements for the high-level waste glass.Several different Phase I staging (retrieval, blending, and pretreatment) scenarios were used to generate example feed compositions for glass formulations, testing, and glass sensitivity analysis. Glass models and data form laboratory glass studies were used to estimate achievable waste loading and corresponding glass volumes for various Phase I feeds. Key issues related to feed process ability, feed composition, uncertainty, and immobilization process technology are identified for future consideration in other tank waste disposal program activities.

  18. Road Map for Development of Crystal-Tolerant High Level Waste Glasses

    Office of Scientific and Technical Information (OSTI)

    (Technical Report) | SciTech Connect Road Map for Development of Crystal-Tolerant High Level Waste Glasses Citation Details In-Document Search Title: Road Map for Development of Crystal-Tolerant High Level Waste Glasses This road map guides the research and development for formulation and processing of crystal-tolerant glasses, identifying near- and long-term activities that need to be completed over the period from 2014 to 2019. The primary objective is to maximize waste loading for Hanford

  19. Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 1, Methodology and results

    SciTech Connect (OSTI)

    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. Although numerous caveats must be placed on the results, the general findings were as follows: 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.

  20. SHAM: High-level seismic tests of piping at the HDR

    SciTech Connect (OSTI)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.; Malcher, L.; Schrammel, D.; Steinhilber, H.; Costello, J.F.

    1988-01-01

    As part of the second phase of vibrational/earthquake investigations at the HDR (Heissdampfreaktor) Test Facility in Kahl/Main, FRG, high-level simulated seismic tests (SHAM) were performed during April--May 1988 on the VKL (Versuchskreislauf) in-plant piping system with two servohydraulic actuators, each capable of generating 40 tons of force. The purpose of these experiments was to study the behavior of piping subjected to seismic excitation levels that exceed design levels manifold and may result in failure/plastification of pipe supports and pipe elements, and to establish seismic margins for piping and pipe supports. The performance of six different dynamic pipe support systems was compared in these tests and the response, operability, and fragility of dynamic supports and of a typical US gate valve were investigated. Data obtained in the tests are used to validate analysis methods. Very preliminary evaluations lead to the observation that, in general, failures of dynamic supports (in particular snubbers) occur only at load levels that substantially exceed the design capacity. Pipe strains at load levels exceeding the design level threefold are quite small, and even when exceeding the design level eightfold are quite tolerable. Hence, under seismic loading, even at extreme levels and in spite of multiple support failures, pipe failure is unlikely. 5 refs., 16 figs.

  1. SETTLING OF SPINEL IN A HIGH-LEVEL WASTE GLASS MELTER

    SciTech Connect (OSTI)

    Pavel Hrma; Pert Schill; Lubomir Nemec

    2002-01-07

    High-level nuclear waste is being vitrified, i.e., converted to a durable glass that can be stored in a safe repository for hundreds of thousands of years. Waste vitrification is accomplished in reactors called melters to which the waste is charged together with glass-forming additives. The mixture is electrically heated to a temperature as high as 1150 decrees C to create a melt that becomes glass on cooling.

  2. Geological problems in radioactive waste isolation

    SciTech Connect (OSTI)

    Witherspoon, P.A.

    1991-01-01

    The problem of isolating radioactive wastes from the biosphere presents specialists in the fields of earth sciences with some of the most complicated problems they have ever encountered. This is especially true for high level waste (HLW) which must be isolated in the underground and away from the biosphere for thousands of years. Essentially every country that is generating electricity in nuclear power plants is faced with the problem of isolating the radioactive wastes that are produced. The general consensus is that this can be accomplished by selecting an appropriate geologic setting and carefully designing the rock repository. Much new technology is being developed to solve the problems that have been raised and there is a continuing need to publish the results of new developments for the benefit of all concerned. The 28th International Geologic Congress that was held July 9--19, 1989 in Washington, DC provided an opportunity for earth scientists to gather for detailed discussions on these problems. Workshop W3B on the subject, Geological Problems in Radioactive Waste Isolation -- A World Wide Review'' was organized by Paul A Witherspoon and Ghislain de Marsily and convened July 15--16, 1989 Reports from 19 countries have been gathered for this publication. Individual papers have been cataloged separately.

  3. Structural integrity and potential failure modes of hanford high-level waste tanks

    SciTech Connect (OSTI)

    Han, F.C.

    1996-09-30

    Structural Integrity of the Hanford High-Level Waste Tanks were evaluated based on the existing Design and Analysis Documents. All tank structures were found adequate for the normal operating and seismic loads. Potential failure modes of the tanks were assessed by engineering interpretation and extrapolation of the existing engineering documents.

  4. Advanced Inverter Functions to Support High Levels of Distributed Solar: Policy and Regulatory Considerations (Brochure)

    SciTech Connect (OSTI)

    Not Available

    2014-11-01

    This paper explains how advanced inverter functions (sometimes called 'smart inverters') contribute to the integration of high levels of solar PV generation onto the electrical grid and covers the contributions of advanced functions to maintaining grid stability. Policy and regulatory considerations associated with the deployment of advanced inverter functions are also introduced.

  5. DEVELOPMENT OF A NEW GLOVE FOR GLOVE BOXES WITH HIGH-LEVEL PERFORMANCES

    SciTech Connect (OSTI)

    Blancher, J.; Poirier, J.M.

    2003-02-27

    This paper describes the results of a joint technological program of COGEMA and MAPA to develop a new generation of glove for glove boxes. The mechanical strength of this glove is twice as high as the best characteristics of gloves available on the market. This new generation of product has both a higher level of performance and better ergonomics.

  6. Radioactive Material Transportation Requirements for the Department of Energy

    SciTech Connect (OSTI)

    John, Mark Earl; Fawcett, Ricky Lee; Bolander, Thane Weston

    2000-07-01

    The Department of Energy (DOE) created the National Transportation Program (NTP) whose goal is to ensure the availability of safe, efficient, and timely transportation of DOE materials. The Integration and Planning Group of the NTP, assisted by Global Technologies Incorporated (GTI), was tasked to identify requirements associated with the transport of DOE Environmental Management (EM) radiological waste/material. A systems engineering approach was used to identify source documents, extract requirements, perform a functional analysis, and set up a transportation requirements management database in RDD-100. Functions and requirements for transporting the following DOE radioactive waste/material are contained in the database: high level radioactive waste (HLW), low-level radioactive waste (LLW), mixed low-level radioactive waste (MLLW), nuclear materials (NM), spent nuclear fuel (SNF), and transuranic waste (TRU waste). The requirements will be used in the development of standard transportation protocols for DOE shipping. The protocols will then be combined into a DOE Transportation Program Management Guide, which will be used to standardize DOE transportation processes.

  7. High Level Trigger Configuration and Handling of Trigger Tables in the CMS Filter Farm

    SciTech Connect (OSTI)

    Bauer, G; Behrens, U; Boyer, V; Branson, J; Brett, A; Cano, E; Carboni, A; Ciganek, M; Cittolin, S; O'dell, V; Erhan, S; Gigi, D; Glege, F; Gomez-Reino, R; Gulmini, M; Gutleber, J; Hollar, J; Lange, D; Kim, J C; Klute, M; Lipeles, E; Perez, J L; Maron, G; Meijers, F; Meschi, E; Moser, R; Mlot, E G; Murray, S; Oh, A; Orsini, L; Paus, C; Petrucci, A; Pieri, M; Pollet, L; Racz, A; Sakulin, H; Sani, M; Schieferdecker, P; Schwick, C; Sumorok, K; Suzuki, I; Tsirigkas, D; Varela, J

    2009-11-22

    The CMS experiment at the CERN Large Hadron Collider is currently being commissioned and is scheduled to collect the first pp collision data in 2008. CMS features a two-level trigger system. The Level-1 trigger, based on custom hardware, is designed to reduce the collision rate of 40 MHz to approximately 100 kHz. Data for events accepted by the Level-1 trigger are read out and assembled by an Event Builder. The High Level Trigger (HLT) employs a set of sophisticated software algorithms, to analyze the complete event information, and further reduce the accepted event rate for permanent storage and analysis. This paper describes the design and implementation of the HLT Configuration Management system. First experiences with commissioning of the HLT system are also reported.

  8. Ocean-current measurements at the Farallon Islands Low-Level Radioactive Waste Disposal Site, 1977-1978. Includes appendix. Final report

    SciTech Connect (OSTI)

    Not Available

    1991-06-01

    The report discusses the results of ocean bottom current measurements obtain from the Farallon Islands Low-Level Waste Disposal Site off the California coast, near San Francisco. The report includes a discussion of the velocity of the currents over the time period and area measured relative to large-scale currents off the California coast, and the possibility for shoreward transport of LLW Materials from the Farallon Islands Site.

  9. Annual Report - FY 2000, Radioactive Waste Shipments to and from the Nevada Test Site, March 2001

    SciTech Connect (OSTI)

    U.S. Department of Energy, Nevada Operations Office

    2001-03-01

    This document reports the low-level radioactive waste, mixed low-level radioactive waste, and Polychlorinated Biphenyl contaminated low-level waste transported to or from the Nevada Test Site during fiscal year 2000.

  10. Development of a Real-Time, High-Speed Distribution Level Data Acquisition System

    SciTech Connect (OSTI)

    Bank, J.; Kroposki, B.

    2012-01-01

    With the development of smart grids and the deployment of their enabling technologies, improved data acquisition will be needed at the distribution level to understand the full impact of these changes. With this in mind, NREL has developed a high-speed measurement and data collection network targeted specifically at the distribution level. This network is based around adaptable, rugged measurement devices designed for deployment at a variety of low and medium voltage locations below the sub-station. Each of these devices is capable of real-time data transmission via an Internet connection. Additionally, several analysis and visualization applications have been developed around the incoming data streams.

  11. IMPACT OF NOBLE METALS AND MERCURY ON HYDROGEN GENERATION DURING HIGH LEVEL WASTE PRETREATMENT AT THE SAVANNAH RIVER SITE

    SciTech Connect (OSTI)

    Stone, M; Tommy Edwards, T; David Koopman, D

    2009-03-03

    The Defense Waste Processing Facility (DWPF) at the Savannah River Site vitrifies radioactive High Level Waste (HLW) for repository internment. The process consists of three major steps: waste pretreatment, vitrification, and canister decontamination/sealing. HLW consists of insoluble metal hydroxides (primarily iron, aluminum, calcium, magnesium, manganese, and uranium) and soluble sodium salts (carbonate, hydroxide, nitrite, nitrate, and sulfate). The pretreatment process in the Chemical Processing Cell (CPC) consists of two process tanks, the Sludge Receipt and Adjustment Tank (SRAT) and the Slurry Mix Evaporator (SME) as well as a melter feed tank. During SRAT processing, nitric and formic acids are added to the sludge to lower pH, destroy nitrite and carbonate ions, and reduce mercury and manganese. During the SME cycle, glass formers are added, and the batch is concentrated to the final solids target prior to vitrification. During these processes, hydrogen can be produced by catalytic decomposition of excess formic acid. The waste contains silver, palladium, rhodium, ruthenium, and mercury, but silver and palladium have been shown to be insignificant factors in catalytic hydrogen generation during the DWPF process. A full factorial experimental design was developed to ensure that the existence of statistically significant two-way interactions could be determined without confounding of the main effects with the two-way interaction effects. Rh ranged from 0.0026-0.013% and Ru ranged from 0.010-0.050% in the dried sludge solids, while initial Hg ranged from 0.5-2.5 wt%, as shown in Table 1. The nominal matrix design consisted of twelve SRAT cycles. Testing included: a three factor (Rh, Ru, and Hg) study at two levels per factor (eight runs), three duplicate midpoint runs, and one additional replicate run to assess reproducibility away from the midpoint. Midpoint testing was used to identify potential quadratic effects from the three factors. A single sludge simulant was used for all tests and was spiked with the required amount of noble metals immediately prior to performing the test. Acid addition was kept effectively constant except to compensate for variations in the starting mercury concentration. SME cycles were also performed during six of the tests.

  12. Method of encapsulating solid radioactive waste material for storage

    DOE Patents [OSTI]

    Bunnell, Lee Roy; Bates, J. Lambert

    1976-01-01

    High-level radioactive wastes are encapsulated in vitreous carbon for long-term storage by mixing the wastes as finely divided solids with a suitable resin, formed into an appropriate shape and cured. The cured resin is carbonized by heating under a vacuum to form vitreous carbon. The vitreous carbon shapes may be further protected for storage by encasement in a canister containing a low melting temperature matrix material such as aluminum to increase impact resistance and improve heat dissipation.

  13. Integrated data base report--1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    1997-12-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and commercial and U.S. government-owned radioactive wastes. Inventories of most of these materials are reported as of the end of fiscal year (FY) 1996, which is September 30, 1996. Commercial SNF and commercial uranium mill tailings inventories are reported on an end-of-calendar year (CY) basis. All SNF and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are SNF, high-level waste, transuranic waste, low-level waste, uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, naturally occurring and accelerator-produced radioactive material, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through FY 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

  14. West Valley demonstration project: alternative processes for solidifying the high-level wastes

    SciTech Connect (OSTI)

    Holton, L.K.; Larson, D.E.; Partain, W.L.; Treat, R.L.

    1981-10-01

    In 1980, the US Department of Energy (DOE) established the West Valley Solidification Project as the result of legislation passed by the US Congress. The purpose of this project was to carry out a high level nuclear waste management demonstration project at the Western New York Nuclear Service Center in West Valley, New York. The DOE authorized the Pacific Northwest Laboratory (PNL), which is operated by Battelle Memorial Institute, to assess alternative processes for treatment and solidification of the WNYNSC high-level wastes. The Process Alternatives Study is the suject of this report. Two pretreatment approaches and several waste form processes were selected for evaluation in this study. The two waste treatment approaches were the salt/sludge separation process and the combined waste process. Both terminal and interim waste form processes were studied.

  15. Release of radionuclides and chelating agents from cement-solidified decontamination low-level radioactive waste collected from the Peach Bottom Atomic Power Station Unit 3

    SciTech Connect (OSTI)

    Akers, D.W.; Kraft, N.C.; Mandler, J.W.

    1994-03-01

    As part of a study being performed for the Nuclear Regulatory Commission (NRC), small-scale waste-form specimens were collected during a low oxidation-state transition-metal ion (LOMI)-nitric permanganate (NP)-LOMI solidification performed in October 1989 at the Peach Bottom Atomic Power Station Unit 3. The purpose of this program was to evaluate the performance of cement-solidified decontamination waste to meet the low-level waste stability requirements defined in the NRC`s ``Technical Position on Waste Form,`` Revision 1. The samples were acquired and tested because little data have been obtained on the physical stability of actual cement-solidified decontamination ion-exchange resin waste forms and on the leachability of radionuclides and chelating agents from those waste forms. The Peach Bottom waste-form specimens were subjected to compressive strength, immersion, and leach testing in accordance with the NRC`s ``Technical Position on Waste Form,`` Revision 1. Results of this study indicate that the specimens withstood the compression tests (>500 psi) before and after immersion testing and leaching, and that the leachability indexes for all radionuclides, including {sup 14}C, {sup 99}{Tc}, and {sup 129}I, are well above the leachability index requirement of 6.0, required by the NRC`s ``Technical Position on Waste Form,`` Revision 1.

  16. Comparison of doubly labeled water with respirometry at low- and high-activity levels

    SciTech Connect (OSTI)

    Westerterp, K.R.; Brouns, F.; Saris, W.H.; ten Hoor, F.

    1988-07-01

    In previous studies the doubly labeled water method for measuring energy expenditure in free-living humans has been validated against respirometry under sedentary conditions. In the present investigation, energy expenditure is measured simultaneously with doubly labeled water and respirometry at low- and high-activity levels. Over 6 days, five subjects were measured doing mainly sedentary activities like desk work; their average daily metabolic rate was 1.40 +/- 0.09 (SD) times sleeping metabolic rate. Four subjects were measured twice over 3.5 days, including 2 days with heavy bicycle ergometer work, resulting in an average daily metabolic rate of 2.61 +/- 0.25 (SD) times sleeping metabolic rate. At the low-activity level, energy expenditures from the doubly labeled water method were on the average 1.4 +/- 3.9% (SD) larger than those from respirometry. At the high-activity level, the doubly labeled water method yielded values that were 1.0 +/- 7.0% (SD) lower than those from respirometry. Results demonstrate the utility of the doubly labeled water method for the determination of energy expenditure in the range of activity levels in daily life.

  17. DELPHI expert panel evaluation of Hanford high level waste tank failure modes and release quantities

    SciTech Connect (OSTI)

    Dunford, G.L.; Han, F.C.

    1996-09-30

    The Failure Modes and Release Quantities of the Hanford High Level Waste Tanks due to postulated accident loads were established by a DELPHI Expert Panel consisting of both on-site and off-site experts in the field of Structure and Release. The Report presents the evaluation process, accident loads, tank structural failure conclusion reached by the panel during the two-day meeting.

  18. Observation of a single spin by transferring its coherence to a high level macroscopic pure state

    SciTech Connect (OSTI)

    Kawamura, Minaru

    2014-12-04

    We discuss about quantum measurement of a single spin in a superconducting RF resonator, where amplification of coherence of the spin is enabled by transferring its coherence to the harmonic oscillator in an non-coherent state with high energy level. This quantum amplification allows that a single spin can induce macroscopic current to permits observation of a single spin state in the number and phase uncertainty relation.

  19. Demonstrating Reliable High Level Waste Slurry Sampling Techniques to Support Hanford Waste Processing

    SciTech Connect (OSTI)

    Kelly, Steven E.

    2013-11-11

    The Hanford Tank Operations Contractor (TOC) and the Hanford Waste Treatment and Immobilization Plant (WTP) contractor are both engaged in demonstrating mixing, sampling, and transfer system capability using simulated Hanford High-Level Waste (HL W) formulations. This work represents one of the remaining technical issues with the high-level waste treatment mission at Hanford. The TOC must demonstrate the ability to adequately mix and sample high-level waste feed to meet the WTP Waste Acceptance Criteria and Data Quality Objectives. The sampling method employed must support both TOC and WTP requirements. To facilitate information transfer between the two facilities the mixing and sampling demonstrations are led by the One System Integrated Project Team. The One System team, Waste Feed Delivery Mixing and Sampling Program, has developed a full scale sampling loop to demonstrate sampler capability. This paper discusses the full scale sampling loops ability to meet precision and accuracy requirements, including lessons learned during testing. Results of the testing showed that the Isolok(R) sampler chosen for implementation provides precise, repeatable results. The Isolok(R) sampler accuracy as tested did not meet test success criteria. Review of test data and the test platform following testing by a sampling expert identified several issues regarding the sampler used to provide reference material used to judge the Isolok's accuracy. Recommendations were made to obtain new data to evaluate the sampler's accuracy utilizing a reference sampler that follows good sampling protocol.

  20. Radioactive waste management in the USSR: A review of unclassified sources. Volume 2

    SciTech Connect (OSTI)

    Bradley, D.J.

    1991-03-01

    The Soviet Union does not currently have an overall radioactive waste management program or national laws that define objectives, procedures, and standards, although such a law is being developed, according to the Soviets. Occupational health and safety does not appear to receive major attention as it does in Western nations. In addition, construction practices that would be considered marginal in Western facilities show up in Soviet nuclear power and waste management operations. The issues involved with radioactive waste management and environmental restoration are being investigated at several large Soviet institutes; however, there is little apparent interdisciplinary integration between them, or interaction with the USSR Academy of Sciences. It is expected that a consensus on technical solutions will be achieved, but it may be slow in coming, especially for final disposal of high-level radioactive wastes and environmental restoration of contaminated areas. Meanwhile, many treatment, solidification, and disposal options for radioactive waste management are being investigated by the Soviets.

  1. Radioactive waste management in the USSR: A review of unclassified sources

    SciTech Connect (OSTI)

    Bradley, D.J.

    1991-03-01

    The Soviet Union does not currently have an overall radioactive waste management program or national laws that define objectives, procedures, and standards, although such a law is being developed, according to the Soviets. Occupational health and safety does not appear to receive major attention as it does in Western nations. In addition, construction practices that would be considered marginal in Western facilities show up in Soviet nuclear power and waste management operations. The issues involved with radioactive waste management and environmental restoration are being investigated at several large Soviet institutes; however, there is little apparent interdisciplinary integration between them, or interaction with the USSR Academy of Sciences. It is expected that a consensus on technical solutions will be achieved, but it may be slow in coming, especially for final disposal of high-level radioactive wastes and environmental restoration of contaminated areas. Meanwhile, many treatment, solidification, and disposal options for radioactive waste management are being investigated by the Soviets.

  2. Development of polyphase ceramics for the immobilization of high-level Defense nuclear waste

    SciTech Connect (OSTI)

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

    1983-02-25

    The report contains two major sections: Section I - An Improved Polyphase Ceramic for High-Level Defense Nucleation Waste reports the work conducted on titanium-silica based ceramics for immobilizing Savannah River Plant waste. Section II - Formulation and Processing of Alumina Based Ceramic Nuclear Waste Forms describes the work conducted on developing a generic alumina and alumina-silica based ceramic waste form capable of immobilizing any nuclear waste with a high aluminum content. Such wastes include the Savannah River Plant wastes, Hanford neutralized purex wastes, and Hanford N-Reactor acid wastes. The design approach and process technology in the two reports demonstrate how the generic high waste loaded ceramic form can be applied to a broad range of nuclear waste compositions. The individual sections are abstracted and indexed separately.

  3. Hanford Site Solid (Radioactive and Hazardous) Waste Program...

    Office of Environmental Management (EM)

    with the Waste 8 Management Programmatic ... include operational low-level radioactive waste (LLW), mixed low- 10 ... Groups of Estimated Criteria-Pollutant Impact ...

  4. DOE Comments on Radioactive Waste | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    PDF icon 1. Summary Comments on Draft Branch Technical Position on a Performance Assessment Methodology for Low-Level Radioactive Waste Disposal Facilities PDF icon 2. Department ...

  5. Regulatory compliance in the design of packages used to transport radioactive materials

    SciTech Connect (OSTI)

    Raske, D.T.

    1993-06-01

    Shipments of radioactive materials within the regulatory jurisdiction of the US Department of Energy (DOE) must meet the package design requirements contained in Title 10 of the Code of Federal Regulations, Part 71, and DOE Order 5480.3. These regulations do not provide design criteria requirements, but only detail the approval standards, structural performance criteria, and package integrity requirements that must be met during transport. The DOE recommended design criterion for high-level Category I radioactive packagings is Section III, Division 1, of the ASME Boiler and Pressure Vessel Code. However, alternative design criteria may be used if all the design requirements are satisfied. The purpose of this paper is to review alternatives to the Code criteria and discuss their applicability to the design of containment vessels in packages for high-level radioactive materials. Issues such as design qualification by physical testing, the use of scale models, and problems encountered using a non-ASME design approach are addressed.

  6. Minor component study for simulated high-level nuclear waste glasses (Draft)

    SciTech Connect (OSTI)

    Li, H.; Langowskim, M.H.; Hrma, P.R.; Schweiger, M.J.; Vienna, J.D.; Smith, D.E.

    1996-02-01

    Hanford Site single-shell tank (SSI) and double-shell tank (DSI) wastes are planned to be separated into low activity (or low-level waste, LLW) and high activity (or high-level waste, HLW) fractions, and to be vitrified for disposal. Formulation of HLW glass must comply with glass processibility and durability requirements, including constraints on melt viscosity, electrical conductivity, liquidus temperature, tendency for phase segregation on the molten glass surface, and chemical durability of the final waste form. A wide variety of HLW compositions are expected to be vitrified. In addition these wastes will likely vary in composition from current estimates. High concentrations of certain troublesome components, such as sulfate, phosphate, and chrome, raise concerns about their potential hinderance to the waste vitrification process. For example, phosphate segregation in the cold cap (the layer of feed on top of the glass melt) in a Joule-heated melter may inhibit the melting process (Bunnell, 1988). This has been reported during a pilot-scale ceramic melter run, PSCM-19, (Perez, 1985). Molten salt segregation of either sulfate or chromate is also hazardous to the waste vitrification process. Excessive (Cr, Fe, Mn, Ni) spinel crystal formation in molten glass can also be detrimental to melter operation.

  7. Probabilistic safety assessment for Hanford high-level waste tank 241-SY-101

    SciTech Connect (OSTI)

    MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W.; Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J.

    1994-05-01

    Los Alamos National Laboratory (Los Alamos) is performing a comprehensive probabilistic safety assessment (PSA), which will include consideration of external events for the 18 tank farms at the Hanford Site. This effort is sponsored by the Department of Energy (DOE/EM, EM-36). Even though the methodology described herein will be applied to the entire tank farm, this report focuses only on the risk from the weapons-production wastes stored in tank number 241-SY-101, commonly known as Tank 101-SY, as configured in December 1992. This tank, which periodically releases ({open_quotes}burps{close_quotes}) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed first because of public safety concerns associated with the potential for release of radioactive tank contents should this gas mixture be ignited during one of the burps. In an effort to mitigate the burping phenomenon, an experiment is being conducted in which a large pump has been inserted into the tank to determine if pump-induced circulation of the tank contents will promote a slow, controlled release of the gases. At the Hanford Site there are 177 underground tanks in 18 separate tank farms containing accumulated liquid/sludge/salt cake radioactive wastes from 50 yr of weapons materials production activities. The total waste volume is about 60 million gal., which contains approximately 120 million Ci of radioactivity.

  8. High-Level Functional and Operational Requirements for the Advanced Fuel Cycle Facilty

    SciTech Connect (OSTI)

    Charles Park

    2006-12-01

    High-Level Functional & Operational Requirements for the AFCF -This document describes the principal functional and operational requirements for the proposed Advanced Fuel Cycle Facility (AFCF). The AFCF is intended to be the world's foremost facility for nuclear fuel cycle research, technology development, and demonstration. The facility will also support the near-term mission to develop and demonstrate technology in support of fuel cycle needs identified by industry, and the long-term mission to retain and retain U.S. leadership in fuel cycle operations. The AFCF is essential to demonstrate a more proliferation-resistant fuel cycle and make long-term improvements in fuel cycle effectiveness, performance and economy.

  9. Fracture toughness measurements on a glass bonded sodalite high-level waste form.

    SciTech Connect (OSTI)

    DiSanto, T.; Goff, K. M.; Johnson, S. G.; O'Holleran, T. P.

    1999-05-19

    The electrometallurgical treatment of metallic spent nuclear fuel produces two high-level waste streams; cladding hulls and chloride salt. Argonne National Laboratory is developing a glass bonded sodalite waste form to immobilize the salt waste stream. The waste form consists of 75 Vol.% crystalline sodalite (containing the salt) with 25 Vol.% of an ''intergranular'' glassy phase. Microindentation fracture toughness measurements were performed on representative samples of this material using a Vickers indenter. Palmqvist cracking was confirmed by post-indentation polishing of a test sample. Young's modulus was measured by an acoustic technique. Fracture toughness, microhardness, and Young's modulus values are reported, along with results from scanning electron microscopy studies.

  10. Idaho High-Level Waste & Facilities Disposition, Final Environmental Impact Statement

    Office of Environmental Management (EM)

    Appendix A Site Evaluation Process A-iii DOE/EIS-0287 Idaho HLW & FD EIS TABLE OF CONTENTS Section Page Appendix A Site Evaluation Process A-1 A.1 Introduction A-1 A.2 Methodology A-1 A.3 High-Level Waste Treatment and Interim Storage Site Selection A-3 A.3.1 Identification of "Must" Criteria A-3 A.3.2 Identification of "Want" Criteria A-3 A.3.3 Identification of Candidate Sites A-3 A.3.4 Evaluation Process A-4 A.3.5 Results of Evaluation Process A-6 A.4 Low-Activity

  11. CREVICE CORROSION & PITTING OF HIGH-LEVEL WASTE CONTAINERS: INTEGRATION OF DETERMINISTIC & PROBABILISTIC MODELS

    SciTech Connect (OSTI)

    JOSEPH C. FARMER AND R. DANIEL MCCRIGHT

    1997-10-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as carbon steel or Monel 400. An integrated predictive model is being developed to account for the effects of localized environmental conditions in the CRM-CAM crevice on the initiation and propagation of pits through the CRM.

  12. Crevice corrosion {ampersand} pitting of high-level waste containers: integration of deterministic {ampersand} probabilistic models

    SciTech Connect (OSTI)

    Farmer, J.C.; McCright, R.D.

    1997-10-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as carbon steel or Monel 400. An integrated predictive model is being developed to account for the effects of localized environmental conditions in the CRM-CAM crevice on the initiation and propagation of pits through the CRM.

  13. Development of the high-level waste high-temperature melter feed preparation flowsheet for vitrification process testing

    SciTech Connect (OSTI)

    Seymour, R.G.

    1995-02-17

    High-level waste (HLW) feed preparation flowsheet development was initiated in fiscal year (FY) 1994 to evaluate alternative flowsheets for preparing melter feed for high-temperature melter (HTM) vitrification testing. Three flowsheets were proposed that might lead to increased processing capacity relative to the Hanford Waste Vitrification Plant (HWVP) and that were flexible enough to use with other HLW melter technologies. This document describes the decision path that led to the selection of flowsheets to be tested in the FY 1994 small-scale HTM tests. Feed preparation flowsheet development for the HLW HTM was based on the feed preparation flowsheet that was developed for the HWVP. This approach allowed the HLW program to build upon the extensive feed preparation flowsheet database developed under the HWVP Project. Primary adjustments to the HWVP flowsheet were to the acid adjustment and glass component additions. Developmental background regarding the individual features of the HLW feed preparation flowsheets is provided. Applicability of the HWVP flowsheet features to the new HLW vitrification mission is discussed. The proposed flowsheets were tested at the laboratory-scale at Pacific Northwest Laboratory. Based on the results of this testing and previously established criteria, a reductant-based flowsheet using glycolic acid and a nitric acid-based flowsheet were selected for the FY 1994 small-scale HTM testing.

  14. High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2

    SciTech Connect (OSTI)

    Cunnane, J.C.; Bates, J.K.; Bradley, C.R.

    1994-03-01

    The objective of this document is to summarize scientific information pertinent to evaluating the extent to which high-level waste borosilicate glass corrosion and the associated radionuclide release processes are understood for the range of environmental conditions to which waste glass may be exposed in service. Alteration processes occurring within the bulk of the glass (e.g., devitrification and radiation-induced changes) are discussed insofar as they affect glass corrosion.This document is organized into three volumes. Volumes I and II represent a tiered set of information intended for somewhat different audiences. Volume I is intended to provide an overview of waste glass corrosion, and Volume 11 is intended to provide additional experimental details on experimental factors that influence waste glass corrosion. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II. Volume I is intended for managers, decision makers, and modelers, the combined set of Volumes I, II, and III is intended for scientists and engineers working in the field of high-level waste.

  15. Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

    SciTech Connect (OSTI)

    Christian, J. H.

    2015-08-01

    There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO₄) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe,Cr)₂O₄), while not detrimental to glass durability, can cause an array of processing problems inside of HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies.

  16. Crystallization in high-level waste glass: A review of glass theory and noteworthy literature

    SciTech Connect (OSTI)

    Christian, J. H.

    2015-08-18

    There is a fundamental need to continue research aimed at understanding nepheline and spinel crystal formation in high-level waste (HLW) glass. Specifically, the formation of nepheline solids (K/NaAlSiO4) during slow cooling of HLW glass can reduce the chemical durability of the glass, which can cause a decrease in the overall durability of the glass waste form. The accumulation of spinel solids ((Fe, Ni, Mn, Zn)(Fe, Cr)2O4), while not detrimental to glass durability, can cause an array of processing problems inside HLW glass melters. In this review, the fundamental differences between glass and solid-crystals are explained using kinetic, thermodynamic, and viscosity arguments, and several highlights of glass-crystallization research, as it pertains to high-level waste vitrification, are described. In terms of mitigating spinel in the melter and both spinel and nepheline formation in the canister, the complexity of HLW glass and the intricate interplay between thermal, chemical, and kinetic factors further complicates this understanding. However, new experiments seeking to elucidate the contributing factors of crystal nucleation and growth in waste glass, and the compilation of data from older experiments, may go a long way towards helping to achieve higher waste loadings while developing more efficient processing strategies. Higher waste loadings and more efficient processing strategies will reduce the overall HLW Hanford Tank Waste Treatment and Immobilization Plant (WTP) vitrification facilities mission life.

  17. THE IMPACT OF OZONE ON THE LOWER FLAMMABLE LIMIT OF HYDROGEN IN VESSELS CONTAINING SAVANNAH RIVER SITE HIGH LEVEL WASTE

    SciTech Connect (OSTI)

    Sherburne, Carol; Osterberg, Paul; Johnson, Tom; Frawely, Thomas

    2013-01-23

    The Savannah River Site, in conjunction with AREVA Federal services, has designed a process to treat dissolved radioactive waste solids with ozone. It is known that in this radioactive waste process, radionuclides radiolytically break down water into gaseous hydrogen and oxygen, which presents a well defined flammability hazard. Flammability limits have been established for both ozone and hydrogen separately; however, there is little information on mixtures of hydrogen and ozone. Therefore, testing was designed to provide critical flammability information necessary to support safety related considerations for the development of ozone treatment and potential scale-up to the commercial level. Since information was lacking on flammability issues at low levels of hydrogen and ozone, a testing program was developed to focus on filling this portion of the information gap. A 2-L vessel was used to conduct flammability tests at atmospheric pressure and temperature using a fuse wire ignition source at 1 percent ozone intervals spanning from no ozone to the Lower Flammable Limit (LFL) of ozone in the vessel, determined as 8.4%(v/v) ozone. An ozone generator and ozone detector were used to generate and measure the ozone concentration within the vessel in situ, since ozone decomposes rapidly on standing. The lower flammability limit of hydrogen in an ozone-oxygen mixture was found to decrease from the LFL of hydrogen in air, determined as 4.2 % (v/v) in this vessel. From the results of this testing, Savannah River was able to develop safety procedures and operating parameters to effectively minimize the formation of a flammable atmosphere.

  18. Repository size for deep geological disposal of partitioning and transmutation high level waste

    SciTech Connect (OSTI)

    Nishihara, Kenji; Nakayama, Shinichi; Oigawa, Hiroyuki

    2007-07-01

    In order to reveal the impact of the partitioning and transmutation (PT) technology on the geological disposal, we investigated the production and disposal of the radioactive wastes from the PT facilities including the dry reprocessing for the spent fuel from accelerator-driven system. After classifying the PT wastes according to the heat generations, the emplacement configurations in the repository were assumed for each group based on the several disposal concepts proposed for the conventional glass waste form. Then, the sizes of the repositories represented by the vault length, emplacement area and excavation volume were estimated. The repository sizes were reduced by PT technology for all disposal concepts. (authors)

  19. A gene stacking approach leads to engineered plants with highly increased galactan levels in Arabidopsis

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Gondolf, Vibe M.; Stoppel, Rhea; Ebert, Berit; Rautengarten, Carsten; Liwanag, April J.M.; Loqué, Dominique; Scheller, Henrik V.

    2014-12-10

    Background: Engineering of plants with a composition of lignocellulosic biomass that is more suitable for downstream processing is of high interest for next-generation biofuel production. Lignocellulosic biomass contains a high proportion of pentose residues, which are more difficult to convert into fuels than hexoses. Therefore, increasing the hexose/pentose ratio in biomass is one approach for biomass improvement. A genetic engineering approach was used to investigate whether the amount of pectic galactan can be specifically increased in cell walls of Arabidopsis fiber cells, which in turn could provide a potential source of readily fermentable galactose. Results: First it was tested ifmore » overexpression of various plant UDP-glucose 4-epimerases (UGEs) could increase the availability of UDP-galactose and thereby increase the biosynthesis of galactan. Constitutive and tissue-specific expression of a poplar UGE and three Arabidopsis UGEs in Arabidopsis plants could not significantly increase the amount of cell wall bound galactose. We then investigated co-overexpression of AtUGE2 together with the β-1,4-galactan synthase GalS1. Co-overexpression of AtUGE2 and GalS1 led to over 80% increase in cell wall galactose levels in Arabidopsis stems, providing evidence that these proteins work synergistically. Furthermore, AtUGE2 and GalS1 overexpression in combination with overexpression of the NST1 master regulator for secondary cell wall biosynthesis resulted in increased thickness of fiber cell walls in addition to the high cell wall galactose levels. Immunofluorescence microscopy confirmed that the increased galactose was present as β-1,4-galactan in secondary cell walls. Conclusions: This approach clearly indicates that simultaneous overexpression of AtUGE2 and GalS1 increases the cell wall galactose to much higher levels than can be achieved by overexpressing either one of these proteins alone. Moreover, the increased galactan content in fiber cells while improving the biomass composition had no impact on plant growth and development and hence on the overall biomass amount. Thus, we could show that the gene stacking approach described here is a promising method to engineer advanced feedstocks for biofuel production.« less

  20. A gene stacking approach leads to engineered plants with highly increased galactan levels in Arabidopsis

    SciTech Connect (OSTI)

    Gondolf, Vibe M.; Stoppel, Rhea; Ebert, Berit; Rautengarten, Carsten; Liwanag, April J.M.; Loqué, Dominique; Scheller, Henrik V.

    2014-12-10

    Background: Engineering of plants with a composition of lignocellulosic biomass that is more suitable for downstream processing is of high interest for next-generation biofuel production. Lignocellulosic biomass contains a high proportion of pentose residues, which are more difficult to convert into fuels than hexoses. Therefore, increasing the hexose/pentose ratio in biomass is one approach for biomass improvement. A genetic engineering approach was used to investigate whether the amount of pectic galactan can be specifically increased in cell walls of Arabidopsis fiber cells, which in turn could provide a potential source of readily fermentable galactose. Results: First it was tested if overexpression of various plant UDP-glucose 4-epimerases (UGEs) could increase the availability of UDP-galactose and thereby increase the biosynthesis of galactan. Constitutive and tissue-specific expression of a poplar UGE and three Arabidopsis UGEs in Arabidopsis plants could not significantly increase the amount of cell wall bound galactose. We then investigated co-overexpression of AtUGE2 together with the β-1,4-galactan synthase GalS1. Co-overexpression of AtUGE2 and GalS1 led to over 80% increase in cell wall galactose levels in Arabidopsis stems, providing evidence that these proteins work synergistically. Furthermore, AtUGE2 and GalS1 overexpression in combination with overexpression of the NST1 master regulator for secondary cell wall biosynthesis resulted in increased thickness of fiber cell walls in addition to the high cell wall galactose levels. Immunofluorescence microscopy confirmed that the increased galactose was present as β-1,4-galactan in secondary cell walls. Conclusions: This approach clearly indicates that simultaneous overexpression of AtUGE2 and GalS1 increases the cell wall galactose to much higher levels than can be achieved by overexpressing either one of these proteins alone. Moreover, the increased galactan content in fiber cells while improving the biomass composition had no impact on plant growth and development and hence on the overall biomass amount. Thus, we could show that the gene stacking approach described here is a promising method to engineer advanced feedstocks for biofuel production.

  1. I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING

    SciTech Connect (OSTI)

    S.M. Frank

    2011-09-01

    Work describe in this report represents the final year activities for the 3-year International Nuclear Energy Research Initiative (I-NERI) project: Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing. Used electrorefiner salt that contained actinide chlorides and was highly loaded with surrogate fission products was processed into three candidate waste forms. The first waste form, a high-loaded ceramic waste form is a variant to the CWF produced during the treatment of Experimental Breeder Reactor-II used fuel at the Idaho National Laboratory (INL). The two other waste forms were developed by researchers at the Korean Atomic Energy Research Institute (KAERI). These materials are based on a silica-alumina-phosphate matrix and a zinc/titanium oxide matrix. The proposed waste forms, and the processes to fabricate them, were designed to immobilize spent electrorefiner chloride salts containing alkali, alkaline earth, lanthanide, and halide fission products that accumulate in the salt during the processing of used nuclear fuel. This aspect of the I-NERI project was to demonstrate 'hot cell' fabrication and characterization of the proposed waste forms. The outline of the report includes the processing of the spent electrorefiner salt and the fabrication of each of the three waste forms. Also described is the characterization of the waste forms, and chemical durability testing of the material. While waste form fabrication and sample preparation for characterization must be accomplished in a radiological hot cell facility due to hazardous radioactivity levels, smaller quantities of each waste form were removed from the hot cell to perform various analyses. Characterization included density measurement, elemental analysis, x-ray diffraction, scanning electron microscopy and the Product Consistency Test, which is a leaching method to measure chemical durability. Favorable results from this demonstration project will provide additional options for fission product immobilization and waste management associated the electrochemical/pyrometallurgical processing of used nuclear fuel.

  2. MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION FINAL REPORT 08R1360-1

    SciTech Connect (OSTI)

    KRUGER AA; MATLACK KS; KOT W; PEGG IL; JOSEPH I; BARDAKCI T; GAN H; GONG W; CHAUDHURI M

    2010-01-04

    This report describes the development and testing of new glass formulations for high aluminum waste streams that achieve high waste loadings while maintaining high processing rates. The testing was based on the compositions of Hanford High Level Waste (HLW) with limiting concentrations of aluminum specified by the Office of River Protection (ORP). The testing identified glass formulations that optimize waste loading and waste processing rate while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts and small scale melt rate screening tests. The results were used to select compositions for subsequent testing in a DuraMelter 100 (DM100) system. These tests were used to determine processing rates for the selected formulations as well as to examine the effects of increased glass processing temperature, and the form of aluminum in the waste simulant. Finally, one of the formulations was selected for large-scale confirmatory testing on the HLW Pilot Melter (DM1200), which is a one third scale prototype of the Hanford Tank Waste Treatment and Immobilization Plant (WTP) HLW melter and off-gas treatment system. This work builds on previous work performed at the Vitreous State Laboratory (VSL) for Department of Energy (DOE) to increase waste loading and processing rates for high-iron HLW waste streams as well as previous tests conducted for ORP on the same high-aluminum waste composition used in the present work and other Hanford HLW compositions. The scope of this study was outlined in a Test Plan that was prepared in response to an ORP-supplied statement of work. It is currently estimated that the number of HLW canisters to be produced in the WTP is about 13,500 (equivalent to 40,500 MT glass). This estimate is based upon the inventory of the tank wastes, the anticipated performance of the sludge treatment processes, and current understanding of the capability of the borosilicate glass waste form. The WTP HLW melter design, unlike earlier DOE melter designs, incorporates an active glass bubbler system. The bubblers create active glass pool convection and thereby improve heat transfer and glass melting rate. The WTP HLW melter has a glass surface area of 3.75 m{sup 2} and depth of {approx}1.1 m. The two melters in the HLW facility together are designed to produce up to 7.5 MT of glass per day at 100% availability. Further increases in HLW waste processing rates can potentially be achieved by increasing the melter operating temperature above 1150 C and by increasing the waste loading in the glass product. Increasing the waste loading also has the added benefit of decreasing the number of canisters for storage. The current estimates and glass formulation efforts have been conservative in terms of achievable waste loadings. These formulations have been specified to ensure that the glasses are homogenous, contain essentially no crystalline phases, are processable in joule-heated, ceramic-lined melters and meet WTP Contract terms. The WTP's overall mission will require the immobilization of tank waste compositions that are dominated by mixtures of aluminum (Al), chromium (Cr), bismuth (Bi), iron (Fe), phosphorous (P), zirconium (Zr), and sulfur (S) compounds as waste-limiting components. Glass compositions for these waste mixtures have been developed based upon previous experience and current glass property models. Recently, DOE has initiated a testing program to develop and characterize HLW glasses with higher waste loadings. Results of this work have demonstrated the feasibility of increases in wasteloading from about 25 wt% to 33-50 wt% (based on oxide loading) in the glass depending on the waste stream. It is expected that these higher waste loading glasses will reduce the HLW canister production requirement by about 25% or more.

  3. High-Level software requirements specification for the TWRS controlled baseline database system

    SciTech Connect (OSTI)

    Spencer, S.G.

    1998-09-23

    This Software Requirements Specification (SRS) is an as-built document that presents the Tank Waste Remediation System (TWRS) Controlled Baseline Database (TCBD) in its current state. It was originally known as the Performance Measurement Control System (PMCS). Conversion to the new system name has not occurred within the current production system. Therefore, for simplicity, all references to TCBD are equivalent to PMCS references. This SRS will reference the PMCS designator from this point forward to capture the as-built SRS. This SRS is written at a high-level and is intended to provide the design basis for the PMCS. The PMCS was first released as the electronic data repository for cost, schedule, and technical administrative baseline information for the TAAS Program. During its initial development, the PMCS was accepted by the customer, TARS Business Management, with no formal documentation to capture the initial requirements.

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

    SciTech Connect (OSTI)

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

    1983-01-31

    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.

  5. Guidelines for development of structural integrity programs for DOE high-level waste storage tanks

    SciTech Connect (OSTI)

    Bandyopadhyay, K.; Bush, S.; Kassir, M.; Mather, B.; Shewmon, P.; Streicher, M.; Thompson, B.; Rooyen, D. van; Weeks, J.

    1997-01-01

    Guidelines are provided for developing programs to promote the structural integrity of high-level waste storage tanks and transfer lines at the facilities of the Department of Energy. Elements of the program plan include a leak-detection system, definition of appropriate loads, collection of data for possible material and geometric changes, assessment of the tank structure, and non-destructive examination. Possible aging degradation mechanisms are explored for both steel and concrete components of the tanks, and evaluated to screen out nonsignificant aging mechanisms and to indicate methods of controlling the significant aging mechanisms. Specific guidelines for assessing structural adequacy will be provided in companion documents. Site-specific structural integrity programs can be developed drawing on the relevant portions of the material in this document.

  6. The effect of high-level waste glass composition on spinel liquidus temperature

    SciTech Connect (OSTI)

    Kruger, A. A.; Riley, Brian J.; Crum, Jarrod V.; Hrma, Pavel; Matyas, Josef

    2012-11-15

    Spinel crystals precipitate in high-level waste glasses containing Fe, Cr, Ni, Mn, Zn, and Ru. The liquidus temperature (T{sub L}d) of spinel as the primary crystallization phase is a function of glass composition, and the spinel solubility (c{sub o}) is a function of both glass composition and temperature (T). Previously reported models of T{sub L} as a function of composition are based on T{sub L} measured directly, which requires laborious experimental procedures. Viewing the curve of c{sub o} versus T as the liquidus line allows a significant broadening of the composition region for model fitting. This paper estimates T{sub L} as a function of composition based on c{sub o} data obtained with the X-ray diffraction technique.

  7. Modeling the corrosion of high-level waste containers: CAM-CRM interface

    SciTech Connect (OSTI)

    Farmer, J. C., LLNL

    1998-06-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 or Monel 400. At the present time, Alloy C-22 and A516 are favored. This publication addresses the development of models to account for corrosion of Alloy C-22 surfaces exposed directly to the Near Field Environment (NFE), as well as to the exacerbated conditions in the CAM-CRM crevice.

  8. Modeling the corrosion of high-level waste containers: CAM-CRM interface

    SciTech Connect (OSTI)

    Farmer, J.C.; Bedrossian, P.J.; McCright, R.D.

    1998-06-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological respository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A516 or Monel 400. At the present time, Alloy C-22 and A516 are favored. This publication addresses the development of models to account for corrosion of Alloy C-22 surfaces exposed directly to the Near Field Environmental (NFE), as well as to the exacerbated conditions in the CAM-CRM crevice.

  9. Glass Property Data and Models for Estimating High-Level Waste Glass Volume

    SciTech Connect (OSTI)

    Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

    2009-10-05

    This report describes recent efforts to develop glass property models that can be used to help estimate the volume of high-level waste (HLW) glass that will result from vitrification of Hanford tank waste. The compositions of acceptable and processable HLW glasses need to be optimized to minimize the waste-form volume and, hence, to save cost. A database of properties and associated compositions for simulated waste glasses was collected for developing property-composition models. This database, although not comprehensive, represents a large fraction of data on waste-glass compositions and properties that were available at the time of this report. Glass property-composition models were fit to subsets of the database for several key glass properties. These models apply to a significantly broader composition space than those previously publised. These models should be considered for interim use in calculating properties of Hanford waste glasses.

  10. Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics

    SciTech Connect (OSTI)

    1995-09-01

    The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

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

    SciTech Connect (OSTI)

    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.%.

  12. Radioactive waste disposal package

    DOE Patents [OSTI]

    Lampe, Robert F. (Bethel Park, PA)

    1986-01-01

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

  13. Turbulent Flame Speeds and NOx Kinetics of HHC Fuels with Contaminants and High Dilution Levels

    SciTech Connect (OSTI)

    Peterson, Eric; Krejci, Michael; Mathieu, Olivier; Vissotski, Andrew; Ravi, Sankat; Plichta, Drew; Sikes, Travis; Levacque, Anthony; Camou, Alejandro; Aul, Christopher

    2013-09-30

    This final report documents the technical results of the 3-year project entitled, “Turbulent Flame Speeds and NOx Kinetics of HHC Fuels with Contaminants and High Dilution Levels,” funded under the NETL of DOE. The research was conducted under six main tasks: 1) program management and planning; 2) turbulent flame speed measurements of syngas mixtures; 3) laminar flame speed measurements with diluents; 4) NOx mechanism validation experiments; 5) fundamental NOx kinetics; and 6) the effect of impurities on NOx kinetics. Experiments were performed using primary constant-volume vessels for laminar and turbulent flame speeds and shock tubes for ignition delay times and species concentrations. In addition to the existing shock- tube and flame speed facilities, a new capability in measuring turbulent flame speeds was developed under this grant. Other highlights include an improved NOx kinetics mechanism; a database on syngas blends for real fuel mixtures with and without impurities; an improved hydrogen sulfide mechanism; an improved ammonia kintics mechanism; laminar flame speed data at high pressures with water addition; and the development of an inexpensive absorption spectroscopy diagnostic for shock-tube measurements of OH time histories. The Project Results for this work can be divided into 13 major sections, which form the basis of this report. These 13 topics are divided into the five areas: 1) laminar flame speeds; 2) Nitrogen Oxide and Ammonia chemical kinetics; 3) syngas impurities chemical kinetics; 4) turbulent flame speeds; and 5) OH absorption measurements for chemical kinetics.

  14. Advanced Inverter Technology for High Penetration Levels of PV Generation in Distribution Systems

    SciTech Connect (OSTI)

    Schauder, C.

    2014-03-01

    This subcontract report was completed under the auspices of the NREL/SCE High-Penetration Photovoltaic (PV) Integration Project, which is co-funded by the U.S. Department of Energy (DOE) Office of Energy Efficiency and Renewable Energy (EERE) and the California Solar Initiative (CSI) Research, Development, Demonstration, and Deployment (RD&D) program funded by the California Public Utility Commission (CPUC) and managed by Itron. This project is focused on modeling, quantifying, and mitigating the impacts of large utility-scale PV systems (generally 1-5 MW in size) that are interconnected to the distribution system. This report discusses the concerns utilities have when interconnecting large PV systems that interconnect using PV inverters (a specific application of frequency converters). Additionally, a number of capabilities of PV inverters are described that could be implemented to mitigate the distribution system-level impacts of high-penetration PV integration. Finally, the main issues that need to be addressed to ease the interconnection of large PV systems to the distribution system are presented.

  15. A COMPARISON OF HANFORD AND SAVANNAH RIVER SITE HIGH-LEVEL WASTES

    SciTech Connect (OSTI)

    HILL RC PHILIP; REYNOLDS JG; RUTLAND PL

    2011-02-23

    This study is a simple comparison of high-level waste from plutonium production stored in tanks at the Hanford and Savannah River sites. Savannah River principally used the PUREX process for plutonium separation. Hanford used the PUREX, Bismuth Phosphate, and REDOX processes, and reprocessed many wastes for recovery of uranium and fission products. Thus, Hanford has 55 distinct waste types, only 17 of which could be at Savannah River. While Hanford and Savannah River wastes both have high concentrations of sodium nitrate, caustic, iron, and aluminum, Hanford wastes have higher concentrations of several key constituents. The factors by which average concentrations are higher in Hanford salt waste than in Savannah River waste are 67 for {sup 241}Am, 4 for aluminum, 18 for chromium, 10 for fluoride, 8 for phosphate, 6 for potassium, and 2 for sulfate. The factors by which average concentrations are higher in Hanford sludges than in Savannah River sludges are 3 for chromium, 19 for fluoride, 67 for phosphate, and 6 for zirconium. Waste composition differences must be considered before a waste processing method is selected: A method may be applicable to one site but not to the other.

  16. PERFORMANCE IMPROVEMENT OF CROSS-FLOW FILTRATION FOR HIGH LEVEL WASTE TREATMENT

    SciTech Connect (OSTI)

    Duignan, M.; Nash, C.; Poirier, M.

    2011-01-12

    In the interest of accelerating waste treatment processing, the DOE has funded studies to better understand filtration with the goal of improving filter fluxes in existing cross-flow equipment. The Savannah River National Laboratory (SRNL) was included in those studies, with a focus on start-up techniques, filter cake development, the application of filter aids (cake forming solid precoats), and body feeds (flux enhancing polymers). This paper discusses the progress of those filter studies. Cross-flow filtration is a key process step in many operating and planned waste treatment facilities to separate undissolved solids from supernate slurries. This separation technology generally has the advantage of self-cleaning through the action of wall shear stress created by the flow of waste slurry through the filter tubes. However, the ability of filter wall self-cleaning depends on the slurry being filtered. Many of the alkaline radioactive wastes are extremely challenging to filtration, e.g., those containing compounds of aluminum and iron, which have particles whose size and morphology reduce permeability. Unfortunately, low filter flux can be a bottleneck in waste processing facilities such as the Savannah River Modular Caustic Side Solvent Extraction Unit and the Hanford Waste Treatment Plant. Any improvement to the filtration rate would lead directly to increased throughput of the entire process. To date increased rates are generally realized by either increasing the cross-flow filter axial flowrate, limited by pump capacity, or by increasing filter surface area, limited by space and increasing the required pump load. SRNL set up both dead-end and cross-flow filter tests to better understand filter performance based on filter media structure, flow conditions, filter cleaning, and several different types of filter aids and body feeds. Using non-radioactive simulated wastes, both chemically and physically similar to the actual radioactive wastes, the authors performed several tests to demonstrate increases in filter performance. With the proper use of filter flow conditions and filter enhancers, filter flow rates can be increased over rates currently realized today.

  17. An Overview of Project Planning for Hot-Isostatic Pressure Treatment of High-Level Waste Calcine for the Idaho Cleanup Project - 12289

    SciTech Connect (OSTI)

    Nenni, Joseph A.; Thompson, Theron J.

    2012-07-01

    The Calcine Disposition Project is responsible for retrieval, treatment by hot-isostatic pressure, packaging, and disposal of highly radioactive calcine stored at the Idaho Nuclear Technology and Engineering Center at the Idaho National Laboratory Site in southeast Idaho. In the 2009 Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement the Department of Energy documented the selection of hot-isostatic pressure as the technology to treat the calcine. The Record of Decision specifies that the treatment results in a volume-reduced, monolithic waste form suitable for transport outside of Idaho by a target date of December 31, 2035. That target date is specified in the 1995 Idaho Settlement Agreement to treat and prepare the calcine for transport out of Idaho in exchange for allowing storage of Navy spent nuclear fuel at the INL Site. The project is completing the design of the calcine-treatment process and facility to comply with Record of Decision, Settlement Agreement, Idaho Department of Environmental Quality, and Department of Energy requirements. A systems engineering approach is being used to define the project mission and requirements, manage risks, and establish the safety basis for decision making in compliance with DOE O 413.3B, 'Program and Project Management for the Acquisition of Capital Assets'. The approach draws heavily on 'design-for-quality' tools to systematically add quality, predict design reliability, and manage variation in the earliest possible stages of design when it is most efficient. Use of these tools provides a standardized basis for interfacing systems to interact across system boundaries and promotes system integration on a facility-wide basis. A mass and energy model was developed to assist in the design of process equipment, determine material-flow parameters, and estimate process emissions. Data generated from failure modes and effects analysis and reliability, availability, maintainability, and inspectability analysis were incorporated into a time and motion model to validate and verify the capability to complete treatment of the calcine within the required schedule. The Calcine Disposition Project systems engineering approach, including use of industry-proven design-for-quality tools and quantitative assessment techniques, has strengthened the project's design capability to meet its intended mission in a safe, cost-effective, and timely manner. Use of these tools has been particularly helpful to the project in early design planning to manage variation; improve requirements and high-consequence risk management; and more effectively apply alternative, interface, failure mode, RAMI, and time and motion analyses at the earliest possible stages of design when their application is most efficient and cost effective. The project is using these tools to design and develop HIP treatment of highly radioactive calcine to produce a volume-reduced, monolithic waste form with immobilization of hazardous and radioactive constituents. (authors)

  18. Emerging Issues and Challenges with Integrating High Levels of Solar into the Distribution System

    Broader source: Energy.gov [DOE]

    Wide use of advanced inverters could double the electricity-distribution system’s hosting capacity for distributed PV at low costs—from about 170 GW to 350 GW (see Palmintier et al. 2016). At the distribution system level, increased variable generation due to high penetrations of distributed PV (typically rooftop and smaller ground-mounted systems) could challenge the management of distribution voltage, potentially increase wear and tear on electromechanical utility equipment, and complicate the configuration of circuit-breakers and other protection systems—all of which could increase costs, limit further PV deployment, or both. However, improved analysis of distribution system hosting capacity—the amount of distributed PV that can be interconnected without changing the existing infrastructure or prematurely wearing out equipment—has overturned previous rule-of-thumb assumptions such as the idea that distributed PV penetrations higher than 15% require detailed impact studies. For example, new analysis suggests that the hosting capacity for distributed PV could rise from approximately 170 GW using traditional inverters to about 350 GW with the use of advanced inverters for voltage management, and it could be even higher using accessible and low-cost strategies such as careful siting of PV systems within a distribution feeder and additional minor changes in distribution operations. Also critical to facilitating distributed PV deployment is the improvement of interconnection processes, associated standards and codes, and compensation mechanisms so they embrace PV’s contributions to system-wide operations. Ultimately SunShot-level PV deployment will require unprecedented coordination of the historically separate distribution and transmission systems along with incorporation of energy storage and “virtual storage,” which exploits improved management of electric vehicle charging, building energy systems, and other large loads. Additional analysis and innovation are needed in every area to realize the potential of this integrated vision.

  19. PILOT-SCALE TEST RESULTS OF A THIN FILM EVAPORATOR SYSTEM FOR MANAGEMENT OF LIQUID HIGH-LEVEL WASTES AT THE HANFORD SITE WASHINGTON USA -11364

    SciTech Connect (OSTI)

    CORBETT JE; TEDESCH AR; WILSON RA; BECK TH; LARKIN J

    2011-02-14

    A modular, transportable evaporator system, using thin film evaporative technology, is planned for deployment at the Hanford radioactive waste storage tank complex. This technology, herein referred to as a wiped film evaporator (WFE), will be located at grade level above an underground storage tank to receive pumped liquids, concentrate the liquid stream from 1.1 specific gravity to approximately 1.4 and then return the concentrated solution back into the tank. Water is removed by evaporation at an internal heated drum surface exposed to high vacuum. The condensed water stream will be shipped to the site effluent treatment facility for final disposal. This operation provides significant risk mitigation to failure of the aging 242-A Evaporator facility; the only operating evaporative system at Hanford maximizing waste storage. This technology is being implemented through a development and deployment project by the tank farm operating contractor, Washington River Protection Solutions (WRPS), for the Office of River Protection/Department of Energy (ORPIDOE), through Columbia Energy and Environmental Services, Inc. (Columbia Energy). The project will finalize technology maturity and install a system at one of the double-shell tank farms. This paper summarizes results of a pilot-scale test program conducted during calendar year 2010 as part of the ongoing technology maturation development scope for the WFE.

  20. End of FY10 report - used fuel disposition technical bases and lessons learned : legal and regulatory framework for high-level waste disposition in the United States.

    SciTech Connect (OSTI)

    Weiner, Ruth F.; Blink, James A.; Rechard, Robert Paul; Perry, Frank; Jenkins-Smith, Hank C.; Carter, Joe; Nutt, Mark; Cotton, Tom

    2010-09-01

    This report examines the current policy, legal, and regulatory framework pertaining to used nuclear fuel and high level waste management in the United States. The goal is to identify potential changes that if made could add flexibility and possibly improve the chances of successfully implementing technical aspects of a nuclear waste policy. Experience suggests that the regulatory framework should be established prior to initiating future repository development. Concerning specifics of the regulatory framework, reasonable expectation as the standard of proof was successfully implemented and could be retained in the future; yet, the current classification system for radioactive waste, including hazardous constituents, warrants reexamination. Whether or not consideration of multiple sites are considered simultaneously in the future, inclusion of mechanisms such as deliberate use of performance assessment to manage site characterization would be wise. Because of experience gained here and abroad, diversity of geologic media is not particularly necessary as a criterion in site selection guidelines for multiple sites. Stepwise development of the repository program that includes flexibility also warrants serious consideration. Furthermore, integration of the waste management system from storage, transportation, and disposition, should be examined and would be facilitated by integration of the legal and regulatory framework. Finally, in order to enhance acceptability of future repository development, the national policy should be cognizant of those policy and technical attributes that enhance initial acceptance, and those policy and technical attributes that maintain and broaden credibility.

  1. Control of radioactive waste-glass melters

    SciTech Connect (OSTI)

    Bickford, D.F. ); Hrma, P. ); Bowan, B.W. II )

    1990-01-01

    Slurries of simulated high level radioactive waste and glass formers have been isothermally reacted and analyzed to identify the sequence of the major chemical reactions in waste vitrification, their effect on glass production rate, and the development of leach resistance. Melting rates of waste batches have been increased by the addition of reducing agents (formic acid, sucrose) and nitrates. The rate increases are attributable in part to exothermic reactions which occur at critical stages in the vitrification process. Nitrates must be balanced by adequate reducing agents to avoid the formation of persistent foam, which would destabilize the melting process. The effect of foaming on waste glass production rates is analyzed, and melt rate limitations defined for waste-glass melters, based upon measurable thermophysical properties. Minimum melter residence times required to homogenize glass and assure glass quality are much smaller than those used in current practice. Thus, melter size can be reduced without adversely affecting glass quality. Physical chemistry and localized heat transfer of the waste-glass melting process are examined, to refine the available models for predicting and assuring glass production rate. It is concluded that the size of replacement melters and future waste processing facilities can be significantly decreased if minimum heat transfer requirements for effective melting are met by mechanical agitation. A new class of waste glass melters has been designed, and proof of concept tests completed on simulated High Level Radioactive Waste slurry. Melt rates have exceeded 155 kg m{sup {minus}2} h{sup {minus}1} with slurry feeds (32 lb ft{sup {minus}2} h{sup {minus}1}), and 229 kg kg m{sup {minus}2} h{sup {minus}1} with dry feed (47 lb ft{sup {minus}2} h{sup {minus}1}). This is about 8 times the melt rate possible in conventional waste- glass melters of the same size. 39 refs., 5 figs., 9 tabs.

  2. High Level Waste System Impacts from Small Column Ion Exchange Implementation

    SciTech Connect (OSTI)

    McCabe, D. J.; Hamm, L. L.; Aleman, S. E.; Peeler, D. K.; Herman, C. C.; Edwards, T. B.

    2005-08-18

    The objective of this task is to identify potential waste streams that could be treated with the Small Column Ion Exchange (SCIX) and perform an initial assessment of the impact of doing so on the High-Level Waste (HLW) system. Design of the SCIX system has been performed as a backup technology for decontamination of High-Level Waste (HLW) at the Savannah River Site (SRS). The SCIX consists of three modules which can be placed in risers inside underground HLW storage tanks. The pump and filter module and the ion exchange module are used to filter and decontaminate the aqueous tank wastes for disposition in Saltstone. The ion exchange module contains Crystalline Silicotitanate (CST in its engineered granular form is referred to as IONSIV{reg_sign} IE-911), and is selective for removal of cesium ions. After the IE-911 is loaded with Cs-137, it is removed and the column is refilled with a fresh batch. The grinder module is used to size-reduce the cesium-loaded IE-911 to make it compatible with the sludge vitrification system in the Defense Waste Processing Facility (DWPF). If installed at the SRS, this SCIX would need to operate within the current constraints of the larger HLW storage, retrieval, treatment, and disposal system. Although the equipment has been physically designed to comply with system requirements, there is also a need to identify which waste streams could be treated, how it could be implemented in the tank farms, and when this system could be incorporated into the HLW flowsheet and planning. This document summarizes a preliminary examination of the tentative HLW retrieval plans, facility schedules, decontamination factor targets, and vitrified waste form compatibility, with recommendations for a more detailed study later. The examination was based upon four batches of salt solution from the currently planned disposition pathway to treatment in the SCIX. Because of differences in capabilities between the SRS baseline and SCIX, these four batches were combined into three batches for a total of about 3.2 million gallons of liquid waste. The chemical and radiological composition of these batches was estimated from the SpaceMan Plus{trademark} model using the same data set and assumptions as the baseline plans.

  3. Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

    SciTech Connect (OSTI)

    Raske, D.T.; Wang, Z.

    1992-07-01

    The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material.

  4. Transport of Radioactive Material by Alpha Recoil

    SciTech Connect (OSTI)

    Icenhour, A.S.

    2005-05-19

    The movement of high-specific-activity radioactive particles (i.e., alpha recoil) has been observed and studied since the early 1900s. These studies have been motivated by concerns about containment of radioactivity and the protection of human health. Additionally, studies have investigated the potential advantage of alpha recoil to effect separations of various isotopes. This report provides a review of the observations and results of a number of the studies.

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

    SciTech Connect (OSTI)

    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.

  6. Studies Related to Chemical Mechanisms of Gas Formation in Hanford High-Level Nuclear Wastes

    SciTech Connect (OSTI)

    E. Kent Barefield; Charles L. Liotta; Henry M. Neumann

    2002-04-08

    The objective of this work is to develop a more detailed mechanistic understanding of the thermal reactions that lead to gas production in certain high-level waste storage tanks at the Hanford, Washington site. Prediction of the combustion hazard for these wastes and engineering parameters for waste processing depend upon both a knowledge of the composition of stored wastes and the changes that they undergo as a result of thermal and radiolytic decomposition. Since 1980 when Delagard first demonstrated that gas production (H2and N2O initially, later N2 and NH3)in the affected tanks was related to oxidative degradation of metal complexants present in the waste, periodic attempts have been made to develop detailed mechanisms by which the gases were formed. These studies have resulted in the postulation of a series of reactions that account for many of the observed products, but which involve several reactions for which there is limited, or no, precedent. For example, Al(OH)4 has been postulated to function as a Lewis acid to catalyze the reaction of nitrite ion with the metal complexants, NO is proposed as an intermediate, and the ratios of gaseous products may be a result of the partitioning of NO between two or more reactions. These reactions and intermediates have been the focus of this project since its inception in 1996.

  7. HIGH-LEVEL WASTE FEED CERTIFICATION IN HANFORD DOUBLE-SHELL TANKS

    SciTech Connect (OSTI)

    THIEN MG; WELLS BE; ADAMSON DJ

    2010-01-14

    The ability to effectively mix, sample, certify, and deliver consistent batches of High Level Waste (HLW) feed from the Hanford Double Shell Tanks (DST) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. DOE's River Protection Project (RPP) mission modeling and WTP facility modeling assume that individual 3785 cubic meter (l million gallon) HLW feed tanks are homogenously mixed, representatively sampled, and consistently delivered to the WTP. It has been demonstrated that homogenous mixing ofHLW sludge in Hanford DSTs is not likely achievable with the baseline design thereby causing representative sampling and consistent feed delivery to be more difficult. Inconsistent feed to the WTP could cause additional batch-to-batch operational adjustments that reduce operating efficiency and have the potential to increase the overall mission length. The Hanford mixing and sampling demonstration program will identify DST mixing performance capability, will evaluate representative sampling techniques, and will estimate feed batch consistency. An evaluation of demonstration program results will identify potential mission improvement considerations that will help ensure successful mission completion. This paper will discuss the history, progress, and future activities that will define and mitigate the mission risk.

  8. ESTIMATING HIGH LEVEL WASTE MIXING PERFORMANCE IN HANFORD DOUBLE SHELL TANKS

    SciTech Connect (OSTI)

    THIEN MG; GREER DA; TOWNSON P

    2011-01-13

    The ability to effectively mix, sample, certify, and deliver consistent batches of high level waste (HLW) feed from the Hanford double shell tanks (DSTs) to the Waste Treatment and Immobilization Plant (WTP) presents a significant mission risk with potential to impact mission length and the quantity of HLW glass produced. The Department of Energy's (DOE's) Tank Operations Contractor (TOC), Washington River Protection Solutions (WRPS) is currently demonstrating mixing, sampling, and batch transfer performance in two different sizes of small-scale DSTs. The results of these demonstrations will be used to estimate full-scale DST mixing performance and provide the key input to a programmatic decision on the need to build a dedicated feed certification facility. This paper discusses the results from initial mixing demonstration activities and presents data evaluation techniques that allow insight into the performance relationships of the two small tanks. The next steps, sampling and batch transfers, of the small scale demonstration activities are introduced. A discussion of the integration of results from the mixing, sampling, and batch transfer tests to allow estimating full-scale DST performance is presented.

  9. TWRS retrieval and disposal mission, immobilized high-level waste storage plan

    SciTech Connect (OSTI)

    Calmus, R.B.

    1998-01-07

    This project plan has a two fold purpose. First, it provides a plan specific to the Hanford Tank Waste Remediation System (TWRS) Immobilized High-Level Waste (EMW) Storage Subproject for the Washington State Department of Ecology (Ecology) that meets the requirements of Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) milestone M-90-01 (Ecology et al. 1996) and is consistent with the project plan content guidelines found in Section 11.5 of the Tri-Party Agreement action plan. Second, it provides an upper tier document that can be used as the basis for future subproject line item construction management plans. The planning elements for the construction management plans are derived from applicable U.S. Department of Energy (DOE) planning guidance documents (DOE Orders 4700.1 (DOE 1992a) and 430.1 (DOE 1995)). The format and content of this project plan are designed to accommodate the plan`s dual purpose. A cross-check matrix is provided in Appendix A to explain where in the plan project planning elements required by Section 11.5 of the Tri-Party Agreement are addressed.

  10. US Department of Energy Storage of Spent Fuel and High Level Waste

    SciTech Connect (OSTI)

    Sandra M Birk

    2010-10-01

    ABSTRACT This paper provides an overview of the Department of Energy's (DOE) spent nuclear fuel (SNF) and high level waste (HLW) storage management. Like commercial reactor fuel, DOE's SNF and HLW were destined for the Yucca Mountain repository. In March 2010, the DOE filed a motion with the Nuclear Regulatory Commission (NRC) to withdraw the license application for the repository at Yucca Mountain. A new repository is now decades away. The default for the commercial and DOE research reactor fuel and HLW is on-site storage for the foreseeable future. Though the motion to withdraw the license application and delay opening of a repository signals extended storage, DOE's immediate plans for management of its SNF and HLW remain the same as before Yucca Mountain was designated as the repository, though it has expanded its research and development efforts to ensure safe extended storage. This paper outlines some of the proposed research that DOE is conducting and will use to enhance its storage systems and facilities.

  11. Design Improvements and Analysis of Innovative High-Level Waste Pipeline Unplugging Technologies - 12171

    SciTech Connect (OSTI)

    Pribanic, Tomas; Awwad, Amer; Crespo, Jairo; McDaniel, Dwayne; Varona, Jose; Gokaltun, Seckin; Roelant, David

    2012-07-01

    Transferring high-level waste (HLW) between storage tanks or to treatment facilities is a common practice performed at the Department of Energy (DoE) sites. Changes in the chemical and/or physical properties of the HLW slurry during the transfer process may lead to the formation of blockages inside the pipelines resulting in schedule delays and increased costs. To improve DoE's capabilities in the event of a pipeline plugging incident, FIU has continued to develop two novel unplugging technologies: an asynchronous pulsing system and a peristaltic crawler. The asynchronous pulsing system uses a hydraulic pulse generator to create pressure disturbances at two opposite inlet locations of the pipeline to dislodge blockages by attacking the plug from both sides remotely. The peristaltic crawler is a pneumatic/hydraulic operated crawler that propels itself by a sequence of pressurization/depressurization of cavities (inner tubes). The crawler includes a frontal attachment that has a hydraulically powered unplugging tool. In this paper, details of the asynchronous pulsing system's ability to unplug a pipeline on a small-scale test-bed and results from the experimental testing of the second generation peristaltic crawler are provided. The paper concludes with future improvements for the third generation crawler and a recommended path forward for the asynchronous pulsing testing. (authors)

  12. Modeling the corrosion of high-level waste containers: CAM-CRM interface

    SciTech Connect (OSTI)

    Bedrossian, P J; Farmer, J C; McCright, R D

    1998-06-01

    A key component of the Engineered Barrier System (EBS) being designed for containment of spent-fuel and high-level waste at the proposed geological repository at Yucca Mountain, Nevada is a two-layer canister. In this particular design, the inner barrier is made of a corrosion resistant material (CRM) such as Alloy 825, 625 or C-22, while the outer barrier is made of a corrosion-allowance material (CAM) such as A5 16 or Monel 400. At the present time, Alloy C-22 and A516 are favored. This publication addresses the development of models to account for corrosion of Alloy C-22 surfaces exposed directly to the Near Field Environment (NFE), as well as to the exacerbated conditions in the CAM-CRM crevice. [5]. Haynes International has published corrosion rates of Alloys 625 and C-22 in artificial crevice solutions (5-10 wt. % FeCl,) at various temperatures (25, 50 and 75 C) [6,7]. In this case, the observed rates for Alloy C-22 appear to be due to passive dissolution. It is believed that Alloy C-22 must be at an electrochemical potential above the repassivation potential to initiate localized corrosion.

  13. Alternate approaches to verifying the structural adequacy of the Defense High Level Waste Shipping Cask

    SciTech Connect (OSTI)

    Zimmer, A.; Koploy, M.

    1991-12-01

    In the early 1980s, the US Department of Energy/Defense Programs (DOE/DP) initiated a project to develop a safe and efficient transportation system for defense high level waste (DHLW). A long-standing objective of the DHLW transportation project is to develop a truck cask that represents the leading edge of cask technology as well as one that fully complies with all applicable DOE, Nuclear Regulatory Commission (NRC), and Department of Transportation (DOT) regulations. General Atomics (GA) designed the DHLW Truck Shipping Cask using state-of-the-art analytical techniques verified by model testing performed by Sandia National Laboratories (SNL). The analytical techniques include two approaches, inelastic analysis and elastic analysis. This topical report presents the results of the two analytical approaches and the model testing results. The purpose of this work is to show that there are two viable analytical alternatives to verify the structural adequacy of a Type B package and to obtain an NRC license. It addition, this data will help to support the future acceptance by the NRC of inelastic analysis as a tool in packaging design and licensing.

  14. ANNUAL RADIOACTIVE WASTE TANK INSPECTION PROGRAM - 2011

    SciTech Connect (OSTI)

    West, B.; Waltz, R.

    2012-06-21

    Aqueous radioactive wastes from Savannah River Site (SRS) separations and vitrification processes are contained in large underground carbon steel tanks. Inspections made during 2011 to evaluate these vessels and other waste handling facilities along with evaluations based on data from previous inspections are the subject of this report. The 2011 inspection program revealed that the structural integrity and waste confinement capability of the Savannah River Site waste tanks were maintained. All inspections scheduled per SRR-LWE-2011-00026, HLW Tank Farm Inspection Plan for 2011, were completed. Ultrasonic measurements (UT) performed in 2011 met the requirements of C-ESR-G-00006, In-Service Inspection Program for High Level Waste Tanks, Rev. 3, and WSRC-TR-2002-00061, Rev.6. UT inspections were performed on Tanks 25, 26 and 34 and the findings are documented in SRNL-STI-2011-00495, Tank Inspection NDE Results for Fiscal Year 2011, Waste Tanks 25, 26, 34 and 41. A total of 5813 photographs were made and 835 visual and video inspections were performed during 2011. A potential leaksite was discovered at Tank 4 during routine annual inspections performed in 2011. The new crack, which is above the allowable fill level, resulted in no release to the environment or tank annulus. The location of the crack is documented in C-ESR-G-00003, SRS High Level Waste Tank Leaksite Information, Rev.6.

  15. Process Flow Chart for Immobilizing of Radioactive High Concentration Sodium Hydroxide Product from the Sodium Processing Facility at the BN-350 Nuclear power plant in Aktau, Kazakhstan

    SciTech Connect (OSTI)

    Burkitbayev, M.; Omarova, K.; Tolebayev, T.; Galkin, A.; Bachilova, N.; Blynskiy, A.; Maev, V.; Wells, D.; Herrick, A.; Michelbacher, J.

    2008-07-01

    This paper describes the results of a joint research investigations carried out by the group of Kazakhstan, British and American specialists in development of a new material for immobilization of radioactive 35% sodium hydroxide solutions from the sodium coolant processing facility of the BN-350 nuclear power plant. The resulting solid matrix product, termed geo-cement stone, is capable of isolating long lived radionuclides from the environment. The physico-mechanical properties of geo-cement stone have been investigated and the flow chart for its production verified in a full scale experiments. (author)

  16. EIS-0287: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002)

    Broader source: Energy.gov [DOE]

    This EIS analyzes the potential environmental consequences of alternatives for managing high-level waste (HLW) calcine, mixed transuranic waste/sodium bearing waste (SBW) and newly generated liquid...

  17. Operating Experience Level 3, Update to Requalification Test Failure of Certain High Efficiency Particulate Air (HEPA) Filters

    Broader source: Energy.gov [DOE]

    Operating Experience Level 3 (OE-3) document provides information regarding the previous requalification test failure and subsequent successful requalification, of certain high efficiency particulate air (HEPA) Filters models manufactured by Flanders Corporation.

  18. Radioactivity of spent TRIGA fuel

    SciTech Connect (OSTI)

    Usang, M. D. Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P.

    2015-04-29

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  19. Advanced Inverter Functions to Support High Levels of Distributed Solar: Policy and Regulatory Considerations (Brochure), NREL (National Renewable Energy Laboratory)

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ADVANCED INVERTER FUNCTIONS TO SUPPORT HIGH LEVELS OF DISTRIBUTED SOLAR POLICY AND REGULATORY CONSIDERATIONS The use of advanced inverters in the design of solar photovoltaic (PV) systems can address some of the challenges to the integration of high levels of distributed solar generation on the electricity system. Although the term "advanced inverters" seems to imply a special type of inverter, some of the inverters currently deployed with PV systems can already provide advanced

  20. Evaluation of Flygt Propeller Xixers for Double Shell Tank (DST) High Level Waste Auxiliary Solids Mobilization

    SciTech Connect (OSTI)

    PACQUET, E.A.

    2000-07-20

    The River Protection Project (RPP) is planning to retrieve radioactive waste from the single-shell tanks (SST) and double-shell tanks (DST) underground at the Hanford Site. This waste will then be transferred to a waste treatment plant to be immobilized (vitrified) in a stable glass form. Over the years, the waste solids in many of the tanks have settled to form a layer of sludge at the bottom. The thickness of the sludge layer varies from tank to tank, from no sludge or a few inches of sludge to about 15 ft of sludge. The purpose of this technology and engineering case study is to evaluate the Flygt{trademark} submersible propeller mixer as a potential technology for auxiliary mobilization of DST HLW solids. Considering the usage and development to date by other sites in the development of this technology, this study also has the objective of expanding the knowledge base of the Flygt{trademark} mixer concept with the broader perspective of Hanford Site tank waste retrieval. More specifically, the objectives of this study delineated from the work plan are described.