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1

Water borne transport of high level nuclear waste in very deep borehole disposal of high level nuclear waste  

E-Print Network (OSTI)

The purpose of this report is to examine the feasibility of the very deep borehole experiment and to determine if it is a reasonable method of storing high level nuclear waste for an extended period of time. The objective ...

Cabeche, Dion Tunick

2011-01-01T23:59:59.000Z

2

Locations of Spent Nuclear Fuel and High-Level Radioactive Waste  

Energy.gov (U.S. Department of Energy (DOE))

Map of the United States of America showing the locations of spent nuclear fuel and high-level radioactive waste.

3

Sulfate Retention in High Level Nuclear Waste Glasses  

Science Conference Proceedings (OSTI)

Symposium, Materials Solutions for the Nuclear Renaissance ... Atomistic Simulations of Radiation Effects in Ceramics for Nuclear Waste Disposal ... Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Applications.

4

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

Energy.gov (U.S. Department of Energy (DOE))

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...

5

Liquidus Temperature Studies for High Level Nuclear Waste Glasses  

Science Conference Proceedings (OSTI)

... of Fluidized Bed Steam Reforming (FBSR) with Hanford Low Activity Wastes ... Level Waste at the Defense Waste Processing Facility through Sludge Batch 7b.

6

Feasibility of lateral emplacement in very deep borehole disposal of high level nuclear waste  

E-Print Network (OSTI)

The U.S. Department of Energy recently filed a motion to withdraw the Nuclear Regulatory Commission license application for the High Level Waste Repository at Yucca Mountain in Nevada. As the U.S. has focused exclusively ...

Gibbs, Jonathan Sutton

2010-01-01T23:59:59.000Z

7

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

DOE Patents (OSTI)

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.

Boatner, Lynn A. (Oak Ridge, TN); Sales, Brian C. (Oak Ridge, TN)

1989-01-01T23:59:59.000Z

8

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

DOE Patents (OSTI)

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

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

1984-04-11T23:59:59.000Z

9

Resolution of the nuclear criticality safety issue for the Hanford site high-level waste tanks  

SciTech Connect

This paper describes the approach used to resolve the Nuclear Criticality Safety Issue for the Hanford Site high-level waste tanks. Although operational controls have been in place at the Hanford Site throughout its operating life to minimize the amount of fissile material discarded as waste, estimates of the total amount of plutonium that entered the waste tanks range from 500 to 1,000 kg. Nuclear criticality safety concerns were heightened in 1991 based on a review of waste analysis results and a subsequent U.S. Department of Energy 1399 review of the nuclear criticality program. Although the DOE review team concluded that there was no imminent risk of a criticality at the Hanford Site tank farms, the team also stated its concern regarding the lack of definitive knowledge of the fissile material inventory and distribution within the waste tanks and the lack of sufficient management support for the overall criticality safety program. An in-depth technical review of the nuclear criticality safety of the waste tanks was conducted to develop a defensible technical basis to ensure that waste tanks are subcritical. The review covered all relevant aspects of nuclear criticality safety including neutronics and chemical and physical phenomena of the waste form under aging waste conditions as well as during routine waste management operations. This paper provides a review of the technical basis to support the conclusion that given current plutonium inventories and operating conditions, a nuclear criticality is incredible. The DOE has been requested to close the Nuclear Criticality Safety Issue. The Defense Nuclear Facilities Safety Board is currently reviewing the technicalbasis.

Bratzel, D.R.

1997-01-07T23:59:59.000Z

10

Hydrogen generation rates in Savannah River Site high-level nuclear waste  

DOE Green Energy (OSTI)

High-level nuclear waste (HLW) is stored at the Savannah River Site (SRS) as alkaline, high-nitrate slurries in underground carbon steel tanks. Hydrogen is continuously generated in the waste tanks as a result of the radiolysis of water. Hydrogen generation rates have recently been measured in several waste tanks containing different types of waste. The measured rates ranged from 1.1 to 6.7 cubic feet per million Btu of decay heat. The measured rates are consistent with laboratory data which show that the hydrogen generation rate depends on the nitrate concentration and the decay heat content of the waste. Sampling at different locations indicated that the hydrogen is uniformly distributed radially within the tank.

Hobbs, D.T.; Norris, P.W.; Pucko, S.A.; Bibler, N.E.; Walker, D.D.; d'Entremont, P.D.

1992-01-01T23:59:59.000Z

11

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

Science Conference Proceedings (OSTI)

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

DOE

2002-12-01T23:59:59.000Z

12

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

SciTech Connect

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.

Dees, L.A.

1994-08-15T23:59:59.000Z

13

Modeling and simulation in analyzing geological repositories for high level nuclear waste  

Science Conference Proceedings (OSTI)

Nuclear energy is very often used to generate electricity. But first the energy must be released from atoms which can be done in two ways: nuclear fusion and nuclear fission. Nuclear power plants use nuclear fission to produce electrical energy. Electrical ... Keywords: modeling, nuclear energy, nuclear waste, nuclear waste storage, simulation

Dietmar P. F. Möller

2007-07-01T23:59:59.000Z

14

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

Science Conference Proceedings (OSTI)

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

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

1983-11-01T23:59:59.000Z

15

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

SciTech Connect

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

Bassett, G.W. Jr. [Chicago Univ., IL (United States). Dept. of Economics; Hemphill, R.; Kohout, E. [Argonne National Lab., IL (United States)

1992-07-01T23:59:59.000Z

16

Risk-informing decisions about high-level nuclear waste repositories  

E-Print Network (OSTI)

Performance assessments (PAs) are important sources of information for societal decisions in high-level radioactive waste (HLW) management, particularly in evaluating safety cases for proposed HLW repository development. ...

Ghosh, Suchandra Tina, 1973-

2004-01-01T23:59:59.000Z

17

Seismic design and analysis considerations for high level nuclear waste repositories  

SciTech Connect

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

Hossain, Q.A.

1993-09-30T23:59:59.000Z

18

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

SciTech Connect

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

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

1980-05-01T23:59:59.000Z

19

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

SciTech Connect

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

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

1980-01-01T23:59:59.000Z

20

Expected environments in high-level nuclear waste and spent fuel repositories in salt  

SciTech Connect

The purpose of this report is to describe the expected environments associated with high-level waste (HLW) and spent fuel (SF) repositories in salt formations. These environments include the thermal, fluid, pressure, brine chemistry, and radiation fields predicted for the repository conceptual designs. In this study, it is assumed that the repository will be a room and pillar mine in a rock-salt formation, with the disposal horizon located approx. 2000 ft (610 m) below the surface of the earth. Canistered waste packages containing HLW in a solid matrix or SF elements are emplaced in vertical holes in the floor of the rooms. The emplacement holes are backfilled with crushed salt or other material and sealed at some later time. Sensitivity studies are presented to show the effect of changing the areal heat load, the canister heat load, the barrier material and thickness, ventilation of the storage room, and adding a second row to the emplacement configuration. The calculated thermal environment is used as input for brine migration calculations. The vapor and gas pressure will gradually attain the lithostatic pressure in a sealed repository. In the unlikely event that an emplacement hole will become sealed in relatively early years, the vapor space pressure was calculated for three scenarios (i.e., no hole closure - no backfill, no hole closure - backfill, and hole closure - no backfill). It was assumed that the gas in the system consisted of air and water vapor in equilibrium with brine. A computer code (REPRESS) was developed assuming that these changes occur slowly (equilibrium conditions). The brine chemical environment is outlined in terms of brine chemistry, corrosion, and compositions. The nuclear radiation environment emphasized in this report is the stored energy that can be released as a result of radiation damage or crystal dislocations within crystal lattices.

Claiborne, H.C.; Rickertsen, L.D., Graham, R.F.

1980-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

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

SciTech Connect

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

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

1984-04-01T23:59:59.000Z

22

Studies Related to Chemical Mechanisms of Gas Formation in Hanford High-Level Nuclear Wastes  

DOE Green Energy (OSTI)

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.

E. Kent Barefield; Charles L. Liotta; Henry M. Neumann

2002-04-08T23:59:59.000Z

23

Technology of high-level nuclear waste disposal. Advances in the science and engineering of the management of high-level nuclear wastes. Volume 2  

SciTech Connect

The twenty papers in this volume are divided into three parts: site exploration and characterization; repository development and design; and waste package development and design. These papers represent the status of technology that existed in 1981 and 1982. Individual papers were processed for inclusion in the Energy Data Base.

Hofmann, P.L. (ed.)

1982-01-01T23:59:59.000Z

24

Performance Confirmation for the Candidate Yucca Mountain High-Level Nuclear Waste Repository: Final Report -- December 2001  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy (DOE) secretary, the President, and Congress may soon consider whether to allow the DOE to proceed into the formal licensing process with the Yucca Mountain high-level nuclear waste repository site. A long-term research and development (R&D) and performance confirmation program can provide substantially more confidence that the correct decisions are being made at each stage of repository development. It is necessary, however, to make sure that the confirmation program will, ...

2001-12-13T23:59:59.000Z

25

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

SciTech Connect

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

Arthur, E.D.

1993-10-01T23:59:59.000Z

26

Some Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository Study (The Yucca Mountain Project)  

Science Conference Proceedings (OSTI)

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.

F. Hua; P. Pasupathi; N. Brown; K. Mon

2005-09-19T23:59:59.000Z

27

High-Level Waste Melter Review  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy (DOE) is faced with a massive cleanup task in resolving the legacy of environmental problems from years of manufacturing nuclear weapons. One of the major activities within this task is the treatment and disposal of the extremely large amount of high-level radioactive (HLW) waste stored at the Hanford Site in Richland, Washington. The current planning for the method of choice for accomplishing this task is to vitrify (glassify) this waste for disposal in a geologic repository. This paper describes the results of the DOE-chartered independent review of alternatives for solidification of Hanford HLW that could achieve major cost reductions with reasonable long-term risks, including recommendations on a path forward for advanced melter and waste form material research and development. The potential for improved cost performance was considered to depend largely on increased waste loading (fewer high-level waste canisters for disposal), higher throughput, or decreased vitrification facility size.

Ahearne, J.; Gentilucci, J.; Pye, L. D.; Weber, T.; Woolley, F.; Machara, N. P.; Gerdes, K.; Cooley, C.

2002-02-26T23:59:59.000Z

28

Design of a mixing system for simulated high-level nuclear waste melter feed slurries  

SciTech Connect

The Nuclear Waste Treatment Program development program consists of coordinated nonradioactive and radioactive testing combined with numerical modeling of the process to provide a complete basis for design and operation of a vitrification facility. The radioactive demonstration tests of equipment and processes are conducted before incorporation in radioactive pilot-scale melter systems for final demonstration. The mixing system evaluation described in this report was conducted as part of the nonradioactive testing. The format of this report follows the sequence in which the design of a large-scale mixing system is determined. The initial program activity was concerned with gaining an understanding of the theoretical foundation of non-Newtonian mixing systems. Section 3 of this report describes the classical rheological models that are used to describe non-Newtonian mixing systems. Since the results obtained here are only valid for the slurries utilized, Section 4, Preparation of Simulated Hanford and West Valley Slurries, describes how the slurries were prepared. The laboratory-scale viscometric and physical property information is summarized in Section 5, Laboratory Rheological Evaluations. The bench-scale mixing evaluations conducted to define the effects of the independent variables described above on the degree of mixing achieved with each slurry are described in Section 6. Bench-scale results are scaled-up to establish engineering design requirements for the full-scale mixing system in Section 7. 24 refs., 37 figs., 44 tabs.

Peterson, M.E.; McCarthy, D.; Muhlstein, K.D.

1986-03-01T23:59:59.000Z

29

The High-Level Radioactive Waste Act (Manitoba, Canada)  

Energy.gov (U.S. Department of Energy (DOE))

Manitoba bars the storage of high-level radioactive wastes from spent nuclear fuel, not intended for research purposes, that was produced at a nuclear facility or in a nuclear reactor outside the...

30

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

SciTech Connect

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

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

1995-01-31T23:59:59.000Z

31

Numerical simulation of high-level radioactive nuclear waste glass production  

SciTech Connect

Vitrification of radioactive waste has become an international approach for converting highly radioactive wastes into a durable solid prior to placing them in a permanent disposal repository. The technology for the process is not new. The conversion melter is a direct descendant of all electric melters used for manufacturing of some commercial glass types. Therefore, the vitrification process of radioactive wastes inherits typical problems of all electric furnaces and creates some other specific problems such as noble metal sedimentation. The noble metals and nickel sulfides in the melter are heavier than molten glass and have a low solubility. In a reducing condition, these metals amalgamate and tend to settle on the melter floor. The metal deposit resulting from this settling has a potential to short circuit the melter. The objective of this paper is to identify the typical problems that have been encountered in the waste melter operations and to address how these problems can be tackled using state-of-the-art numerical simulation techniques. It is believed that the large amount of pilot-scale melter experience throughout the world, combined with the knowledge gained from state-of-the-art computer modeling techniques would give assurance that the existing and future radioactive wastes can be effectively converted into a durable glass material and safely placed in a permanent repository.

Choi, I.G. (Westinghouse Savannah River Co., Aiken, SC (United States)); Ungan, A. (Purdue Univ., Indianapolis, IN (United States). Dept. of Mechanical Engineering)

1991-01-01T23:59:59.000Z

32

Numerical simulation of high-level radioactive nuclear waste glass production  

SciTech Connect

Vitrification of radioactive waste has become an international approach for converting highly radioactive wastes into a durable solid prior to placing them in a permanent disposal repository. The technology for the process is not new. The conversion melter is a direct descendant of all electric melters used for manufacturing of some commercial glass types. Therefore, the vitrification process of radioactive wastes inherits typical problems of all electric furnaces and creates some other specific problems such as noble metal sedimentation. The noble metals and nickel sulfides in the melter are heavier than molten glass and have a low solubility. In a reducing condition, these metals amalgamate and tend to settle on the melter floor. The metal deposit resulting from this settling has a potential to short circuit the melter. The objective of this paper is to identify the typical problems that have been encountered in the waste melter operations and to address how these problems can be tackled using state-of-the-art numerical simulation techniques. It is believed that the large amount of pilot-scale melter experience throughout the world, combined with the knowledge gained from state-of-the-art computer modeling techniques would give assurance that the existing and future radioactive wastes can be effectively converted into a durable glass material and safely placed in a permanent repository.

Choi, I.G. [Westinghouse Savannah River Co., Aiken, SC (United States); Ungan, A. [Purdue Univ., Indianapolis, IN (United States). Dept. of Mechanical Engineering

1991-12-31T23:59:59.000Z

33

An International Initiative on Long-Term Behavior of High-Level Nuclear Waste Glass  

Science Conference Proceedings (OSTI)

Nations using borosilicate glass as an immobilization material for radioactive waste have reinforced the importance of scientific collaboration to obtain a consensus on the mechanisms controlling the longterm dissolution rate of glass. This goal is deemed to be crucial for the development of reliable performance assessment models for geological disposal. The collaborating laboratories all conduct fundamental and/or applied research using modern materials science techniques. This paper briefly reviews the radioactive waste vitrification programs of the six participant nations and summarizes the current state of glass corrosion science, emphasizing the common scientific needs and justifications for on-going initiatives.

Gin, Stephane [French Atomic Energy Commission (CEA); Abdelouas, Abdesselam [SUBATECH Laboratory (France); Criscenti, Louise J [Sandia National Laboratory (SNL); Ebert, William L [Argonne National Laboratory (ANL); Ferrand, K [Belgian Nuclear Research Centre, SCK-CEN; Geisler, T [Rheinische Friedrich-Wilhelms-Universität Bonn, Bonn, Germany; Harrison, Michael T [National Nuclear Laboratory (NNL); Inagaki, Y [Kyushu University, Japan; Mitsui, S [Japan Atomic Energy Agency (JAEA); Mueller, K T [Pacific Northwest National Laboratory (PNNL); Marra, James C [Savannah River National Laboratory (SRNL), Aiken, S.C.; Pantano, Carlo G [Pennsylvania State University, State College, PA; Pierce, Eric M [ORNL; Ryan, Joseph V [Pacific Northwest National Laboratory (PNNL); Schofield, J M [AMEC, Harwell Oxford Didcot Oxfordshire, United Kingdom; Steefel, Carl I [Lawrence Berkeley National Laboratory (LBNL); Vienna, John D. [Pacific Northwest National Laboratory (PNNL)

2013-01-01T23:59:59.000Z

34

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

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.

Rechard, R.P. [ed.

1995-03-01T23:59:59.000Z

35

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

Science Conference Proceedings (OSTI)

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.

Not Available

1994-09-01T23:59:59.000Z

36

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

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.

Rechard, R.P. [ed.

1993-12-01T23:59:59.000Z

37

Closure development for high-level nuclear waste containers for the tuff repository; Phase 1, Final report  

SciTech Connect

This report summarizes Phase 1 activities for closure development of the high-level nuclear waste package task for the tuff repository. Work was conducted under U.S. Department of Energy (DOE) Contract 9172105, administered through the Lawrence Livermore National Laboratory (LLNL), as part of the Yucca Mountain Project (YMP), funded through the DOE Office of Civilian Radioactive Waste Management (OCRWM). The goal of this phase was to select five closure processes for further evaluation in later phases of the program. A decision tree methodology was utilized to perform an objective evaluation of 15 potential closure processes. Information was gathered via a literature survey, industrial contacts, and discussions with project team members, other experts in the field, and the LLNL waste package task staff. The five processes selected were friction welding, electron beam welding, laser beam welding, gas tungsten arc welding, and plasma arc welding. These are felt to represent the best combination of weldment material properties and process performance in a remote, radioactive environment. Conceptual designs have been generated for these processes to illustrate how they would be implemented in practice. Homopolar resistance welding was included in the Phase 1 analysis, and developments in this process will be monitored via literature in Phases 2 and 3. Work was conducted in accordance with the YMP Quality Assurance Program. 223 refs., 20 figs., 9 tabs.

Robitz, E.S. Jr.; McAninch, M.D. Jr.; Edmonds, D.P. [Babcock and Wilcox Co., Lynchburg, VA (USA). Nuclear Power Div.]|[Babcock and Wilcox Co., Alliance, OH (USA). Research and Development Div.

1990-09-01T23:59:59.000Z

38

Materials Degradation Issues in the U.S. High-Level Nuclear Waste Repository  

DOE Green Energy (OSTI)

This paper reviews the state-of-the-art understanding of the degradation processes by the Yucca Mountain Project (YMP) 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 first 10,000-years after repository closure. This paper provides an overview of the degradation of the waste packages and drip shields in the repository after permanent closure of the facility. The degradation modes 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 the degradation of Alloy 22 and titanium alloys are also discussed. Further, for titanium alloys, the effects of fluorides, bromides, and galvanic coupling to less noble metals are considered. It is concluded that the materials and design adopted will provide sufficient safety margins for at least 10,000-years after repository closure.

K.G. Mon; F. Hua

2005-04-12T23:59:59.000Z

39

INCORPORATION OF MONO SODIUM TITANATE AND CRYSTALLINE SILICOTITANATE FEEDS IN HIGH LEVEL NUCLEAR WASTE GLASS  

Science Conference Proceedings (OSTI)

Four series of glass compositions were selected, fabricated, and characterized as part of a study to determine the impacts of the addition of Crystalline Silicotitanate (CST) and Monosodium Titanate (MST) from the Small Column Ion Exchange (SCIX) process on the Defense Waste Processing Facility (DWPF) glass waste form and the applicability of the DWPF process control models. All of the glasses studied were considerably more durable than the benchmark Environmental Assessment (EA) glass. The measured Product Consistency Test (PCT) responses were compared with the predicted values from the current DWPF durability model. One of the KT01-series and two of the KT03-series glasses had measured PCT responses that were outside the lower bound of the durability model. All of the KT04 glasses had durabilities that were predictable regardless of heat treatment or compositional view. In general, the measured viscosity values of the KT01, KT03, and KT04-series glasses are well predicted by the current DWPF viscosity model. The results of liquidus temperature (T{sub L}) measurements for the KT01-series glasses were mixed with regard to the predictability of the T{sub L} for each glass. All of the measured T{sub L} values were higher than the model predicted values, although most fell within the 95% confidence intervals. Overall, the results of this study show a reasonable ability to incorporate the anticipated SCIX streams into DWPF-type glass compositions with TiO{sub 2} concentrations of 4-5 wt % in glass.

Fox, K.; Johnson, F.; Edwards, T.

2010-11-23T23:59:59.000Z

40

Transportation of Spent Nuclear Fuel and High Level Waste to Yucca Mountain: The Next Step in Nevada  

Science Conference Proceedings (OSTI)

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.

Sweeney, Robin L,; Lechel, David J.

2003-02-25T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system  

Science Conference Proceedings (OSTI)

A new approach, termed SMILE (Small Module Inductively Loaded Energy), for the vitrification of high level nuclear wastes (HLW) is described. Present vitrification systems liquefy the HLW solids and associated frit material in large high temperature melters. The molten mix is then poured into small ({approximately}1 m{sup 3}) disposal canisters, where it solidifies and cools. SMILE eliminates the separate, large high temperature melter. Instead, the BLW solids and frit melt inside the final disposal containers, using inductive heating. The contents then solidify and cool in place. The SMILE modules and the inductive heating process are designed so that the outer stainless can of the module remains at near ambient temperature during the process cycle. Module dimensions are similar to those of present disposal containers. The can is thermally insulated from the high temperature inner container by a thin layer of refractory alumina firebricks. The inner container is a graphite crucible lined with a dense alumina refractory that holds the HLW and fiit materials. After the SMILE module is loaded with a slurry of HLW and frit solids, an external multi-turn coil is energized with 30-cycle AC current. The enclosing external coil is the primary of a power transformer, with the graphite crucible acting as a single turn ``secondary.`` The induced current in the ``secondary`` heats the graphite, which in turn heats the HLW and frit materials. The first stage of the heating process is carried out at an intermediate temperature to drive off remnant liquid water and water of hydration, which takes about 1 day. The small fill/vent tube to the module is then sealed off and the interior temperature raised to the vitrification range, i.e., {approximately}1200C. Liquefaction is complete after approximately 1 day. The inductive heating then ceases and the module slowly loses heat to the environment, allowing the molten material to solidify and cool down to ambient temperature.

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

1996-03-01T23:59:59.000Z

42

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  

Science Conference Proceedings (OSTI)

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.

Rechard, R.P. [ed.

1993-12-01T23:59:59.000Z

43

PAIRWISE BLENDING OF HIGH LEVEL WASTE (HLW)  

Science Conference Proceedings (OSTI)

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.

CERTA, P.J.

2006-02-22T23:59:59.000Z

44

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  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Contents Contents CR-iii TABLE OF CONTENTS Section Page 8. Transportation Modes, Routes, Affected Environment, and Impacts............................................ CR8-1 8.1 General Opposition to Transporting Spent Nuclear Fuel and High-Level Radioactive Waste ............................................................................................................ CR8-6 8.2 Number of Shipments ..................................................................................................... CR8-37 8.3 Transportation Modes and Routes .................................................................................. CR8-41 8.3.1 State Highway 127, Hoover Dam, Nevada Department of Transportation Alternatives ..............................................................................................................

45

Mathematical analysis of hydrogen mixing and diffusion in the vapor space of a high-level nuclear waste tank  

DOE Green Energy (OSTI)

This paper presents mathematical analyses of the possible accumulation of radiolytically produced hydrogen in the vapor space in a tank storing liquid high-level radioactive waste. Under normal operating conditions, these tanks are continuously ventilated with air to ensure that the concentration of hydrogen never reaches its lower flammability limit (4%). These scenarios are considered in which it is postulated that hydrogen may accumulate and present a flammability hazard. These scenarios are stratification due to gravity, slow mixing when the ventilation system is operating, and slow mixing when the ventilation system is not operating. In all three cases, the analyses indicate that the accumulation of hydrogen is not likely and thus does not present a flammability problem so long as controls are in place to dilute its concentration to less than 4%.

Bibler, N.E. (ed.); Wallace, R.M.

1991-01-01T23:59:59.000Z

46

High-Level Waste Melter Study Report  

SciTech Connect

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.

Perez, 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.

2001-07-13T23:59:59.000Z

47

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

SciTech Connect

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.

1988-12-01T23:59:59.000Z

48

Site characterization plan conceptual design report for a high-level nuclear waste repository in salt, vertical emplacement mode: Volume 1  

Science Conference Proceedings (OSTI)

This Conceptual Design Report describes the conceptual design of a high-level nuclear waste repository in salt at a proposed site in Deaf Smith County, Texas. Waste receipt, processing, packing, and other surface facility operations are described. Operations in the shafts underground are described, including waste hoisting, transfer, and vertical emplacement. This report specifically addresses the vertical emplacement mode, the reference design for the repository. Waste retrieval capability is described. The report includes a description of the layout of the surface, shafts, and underground. Major equipment items are identified. The report includes plans for decommissioning and sealing of the facility. The report discusses how the repository will satisfy performance objectives. Chapters are included on basis for design, design analyses, and data requirements for completion of future design efforts. 105 figs., 52 tabs.

Not Available

1987-12-01T23:59:59.000Z

49

High-level radioactive wastes. Supplement 1  

SciTech Connect

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.

McLaren, L.H. (ed.) [ed.

1984-09-01T23:59:59.000Z

50

High-level radioactive waste management alternatives  

SciTech Connect

A summary of a comprehensive overview study of potential alternatives for long-term management of high-level radioactive waste is presented. The concepts studied included disposal in geologic formations, disposal in seabeds, disposal in ice caps, disposal into space, and elimination by transmutation. (TFD)

1974-05-01T23:59:59.000Z

51

Handbook of high-level radioactive waste transportation  

Science Conference Proceedings (OSTI)

The High-Level Radioactive Waste Transportation Handbook serves as a reference to which state officials and members of the general public may turn for information on radioactive waste transportation and on the federal government`s system for transporting this waste under the Civilian Radioactive Waste Management Program. The Handbook condenses and updates information contained in the Midwestern High-Level Radioactive Waste Transportation Primer. It is intended primarily to assist legislators who, in the future, may be called upon to enact legislation pertaining to the transportation of radioactive waste through their jurisdictions. The Handbook is divided into two sections. The first section places the federal government`s program for transporting radioactive waste in context. It provides background information on nuclear waste production in the United States and traces the emergence of federal policy for disposing of radioactive waste. The second section covers the history of radioactive waste transportation; summarizes major pieces of legislation pertaining to the transportation of radioactive waste; and provides an overview of the radioactive waste transportation program developed by the US Department of Energy (DOE). To supplement this information, a summary of pertinent federal and state legislation and a glossary of terms are included as appendices, as is a list of publications produced by the Midwestern Office of The Council of State Governments (CSG-MW) as part of the Midwestern High-Level Radioactive Waste Transportation Project.

Sattler, L.R.

1992-10-01T23:59:59.000Z

52

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  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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 DOE/EIS-0250 Errata Sheet Since release of the Final EIS for Yucca Mountain on February 14, 2002 as part of the Site Recommendation documentation required under the Nuclear Waste Policy Act, as amended, the Department of Energy (DOE) has identified a variety of errors in the document. These errors were found to include: editing errors - errors in editorial style, rounding, and unit conversions data entry errors, errors in typing a number transcription errors - errors in transcribing information from one part of the document to another, failures to update the text from the most current analyses at the time of the

53

AN EVALUATION OF HYDROGEN INDUCED CRACKING SUSCEPTIBILITY OF TITANIUM ALLOYS IN US HIGH-LEVEL NUCLEAR WASTE REPOSITORY ENVIRONMENTS  

DOE Green Energy (OSTI)

This paper evaluates hydrogen-induced cracking (HIC) susceptibility of titanium alloys in environments anticipated in the Yucca Mountain nuclear waste repository with particular emphasis on the. effect of the oxide passive film on the hydrogen absorption process of titanium alloys being evaluated. The titanium alloys considered in this review include Ti 2, 5 , 7, 9, 11, 12, 16, 17, 18, 24 and 29. In general, the concentration of hydrogen in a titanium alloy can increase due to absorption of atomic hydrogen produced from passive general corrosion of that alloy or galvanic coupling of it to a less noble metal. It is concluded that under the exposure conditions anticipated in the Yucca Mountain repository, the HIC of titanium drip shield will not occur because there will not be sufficient hydrogen in the metal even after 10,000 years of emplacement. Due to the conservatisms adopted in the current evaluation, this assessment is considered very conservative.

G. De; K. Mon; G. Gordon; D. Shoesmith; F. Hua

2006-02-21T23:59:59.000Z

54

High Level Waste Corporate Board Charter  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

on 24 July 2008 1 on 24 July 2008 1 Office of Environmental Management High-Level Waste Corporate Board Charter Purpose This Charter establishes the High- Level Waste (HLW) Corporate Board, (hereinafter referred to as the 'Board') within the Office of Environmental Management (EM). The Board will serve as a consensus building body to integrate the Department of Energy (DOE) HLW management and disposition activities across the EM program and, with the coordination and cooperation of other program offices, across the DOE complex. The Board will identify the need for and develop policies, planning, standards and guidance and provide the integration necessary to implement an effective and efficient national HLW program. The Board will also evaluate the implications of HLW issues and their

55

Infrared Thermography in High Level Waste  

Science Conference Proceedings (OSTI)

The Savannah River Site is a Department of Energy, government-owned, company-operated industrial complex built in the 1950s to produce materials used in nuclear weapons. Five reactors were built to support the production of nuclear weapons material. Irradiated materials were moved from the reactors to one of the two chemical separation plants. In these facilities, known as ''canyons,'' the irradiated fuel and target assemblies were chemically processed to separate useful products from waste. Unfortunately, the by-product waste of nuclear material production was a highly radioactive liquid that had to be stored and maintained. In 1993 a strategy was developed to implement predictive maintenance technologies in the Liquid Waste Disposition Project Division responsible for processing the liquid waste. Responsibilities include the processing and treatment of 51 underground tanks designed to hold 750,000 to1,300,000 gallons of liquid waste and operation of a facility that vitrifies highly radioactive liquid waste into glass logs. Electrical and mechanical equipment monitored at these facilities is very similar to that found in non-nuclear industrial plants. Annual inspections are performed on electrical components, roof systems, and mechanical equipment. Troubleshooting and post installation and post-maintenance infrared inspections are performed as needed. In conclusion, regardless of the industry, the use of infrared thermography has proven to be an efficient and effective method of inspection to help improve plant safety and reliability through early detection of equipment problems.

GLEATON, DAVIDT.

2004-08-24T23:59:59.000Z

56

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

SciTech Connect

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

Golding, D.; White, A. [Clark Univ., Worcester, MA (United States). Center for Technology, Environment, and Development

1990-12-01T23:59:59.000Z

57

Draft Supplemental Environmental Impact Statement for a Geologice Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mounta  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

v v COVER SHEET RESPONSIBLE AGENCY: U.S. Department of Energy (DOE) TITLE: Draft 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 - Nevada Rail Transportation Corridor (DOE/EIS-0250F-S2D; the Nevada Rail Corridor SEIS), and Draft Environmental Impact Statement for a Rail Alignment for the Construction and Operation of a Railroad in Nevada to a Geologic Repository at Yucca Mountain, Nye County, Nevada (DOE/EIS-0369D; the Rail Alignment EIS) CONTACTS: For more information about this document, write or call: For general information on the DOE NEPA process, write or call: U.S. Department of Energy Office of Civilian Radioactive Waste Management

58

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

UPCOMING UPCOMING EVENTS: The Low-Level 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 Board meeting will be held at DOE- RL on 6 November 2008. Meeting details will be presented here and e- mailed to those persons with an interest to participate. Topics for discussion include but are not limited to: ï‚· Results of the Tank Integrity Workshop ï‚· Strategic Initiative Briefing ï‚· Performance Assessment Guide Proposal NEWS ITEMS 3 June 2008: WASHINGTON, DC - The U.S. Department of Energy today announced submittal of a License Application to the U.S. Nuclear Regulatory Commission seeking authorization to construct America's first repository for spent nuclear fuel and high-level radioactive waste at Yucca Mountain, Nevada. (http://www.ocrwm.doe.gov) 8

59

DOCUMENTATION OF NATIONAL WEATHER CONDITIONS AFFECTING LONG-TERM DEGRADATION OF COMMERCIAL SPENT NUCLEAR FUEL AND DOE SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE  

SciTech Connect

The U.S. Department of Energy (DOE) is preparing a proposal to construct, operate 2nd monitor, and eventually close a repository at Yucca Mountain in Nye County, Nevada, for the geologic disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). As part of this effort, DOE has prepared a viability assessment and an assessment of potential consequences that may exist if the repository is not constructed. The assessment of potential consequences if the repository is not constructed assumes that all SNF and HLW would be left at the generator sites. These include 72 commercial generator sites (three commercial facility pairs--Salem and Hope Creek, Fitzpatrick and Nine Mile Point, and Dresden and Morris--would share common storage due to their close proximity to each other) and five DOE sites across the country. DOE analyzed the environmental consequences of the effects of the continued storage of these materials at these sites in a report titled Continued Storage Analysis Report (CSAR; Reference 1 ) . The CSAR analysis includes a discussion of the degradation of these materials when exposed to the environment. This document describes the environmental parameters that influence the degradation analyzed in the CSAR. These include temperature, relative humidity, precipitation chemistry (pH and chemical composition), annual precipitation rates, annual number of rain-days, and annual freeze/thaw cycles. The document also tabulates weather conditions for each storage site, evaluates the degradation of concrete storage modules and vaults in different regions of the country, and provides a thermal analysis of commercial SNF in storage.

W. L. Poe, Jr.; P.F. Wise

1998-11-01T23:59:59.000Z

60

Spent fuel and high-level radioactive waste transportation report  

SciTech Connect

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.

1989-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Spent fuel and high-level radioactive waste transportation report  

SciTech Connect

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.

1990-11-01T23:59:59.000Z

62

Spent Fuel and High-Level Radioactive Waste Transportation Report  

SciTech Connect

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.

1992-03-01T23:59:59.000Z

63

Progress in evaluation of radionuclide geochemical information developed by DOE high-level nuclear waste repository site projects: report for January-March 1985. Volume 2  

Science Conference Proceedings (OSTI)

Geochemical information relevant to the retention of radionuclides by the Hanford Site (in basalt) and the Yucca Mountain site (in tuff), candidate high-level nuclear waste geologic repositories being developed by US Department of Energy (DOE) projects, is being evaluated by Oak Ridge National Laboratory (ORNL) for the US Nuclear Regulatory Commission (NRC). Our evaluation of the sorption of technetium by basalt/groundwater systems was essentially completed this quarter and the results summarized; we conclude that the experimental methodology and results reported by the DOE for the Hanford Site have not conclusively established that significant retardation of technetium migration may be provided by phases present in the basalts of the Hanford Site. We have shown that sodium boltwoodite is the saturating uranium solid phase in two basalt/groundwater systems. Because thermodynamic data are not available for sodium boltwoodite, calculated solubilities for uranium are erroneous in these systems. Results of radionuclide solubility/speciation calculations, published by the DOE for the Yucca Mountain site, were evaluated this quarter under our geochemical modeling task. We express concerns relative to the inherent limitations of such calculations. Samples of Yucca Mountain tuff and J-13 well water were received for use in our planned radionuclide sorption/solubility experiments. These Yucca Mountain materials will be used to evaluate radionuclide sorption and apparent concentration limit values published by the Nevada Nuclear Waste Storage Investigations (NNWSI) project. 40 refs., 5 figs., 16 tabs.

Kelmers, A.D.; Seeley, F.G.; Arnold, W.D.; Blencoe, J.G.; Meyer, R.E.; Jacobs, G.K.; Whatley, S.K.

1985-09-01T23:59:59.000Z

64

Defense High Level Waste Disposal Container System Description Document  

Science Conference Proceedings (OSTI)

The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents.

N. E. Pettit

2001-07-13T23:59:59.000Z

65

Social impacts of hazardous and nuclear facilities and events: Implications for Nevada and the Yucca Mountain high-level nuclear waste repository; [Final report  

SciTech Connect

Social impacts of a nuclear waste repository are described. Various case studies are cited such as Rocky Flats Plant, the Feed Materials Production Center, and Love Canal. The social impacts of toxic contamination, mitigating environmental stigma and loss of trust are also discussed.

Freudenburg, W.R. [Wisconsin Univ., Madison, WI (United States)] Wisconsin Univ., Madison, WI (United States); Carter, L.F.; Willard, W. [Washington State Univ., Pullman, WA (United States)] Washington State Univ., Pullman, WA (United States); Lodwick, D.G. [Miami Univ., Oxford, OH (United States)] Miami Univ., Oxford, OH (United States); Hardert, R.A. [Arizona State Univ., Tempe, AZ (United States)] Arizona State Univ., Tempe, AZ (United States); Levine, A.G. [State Univ. of New York, Buffalo, NY (United States). Dept. of Sociology] State Univ. of New York, Buffalo, NY (United States). Dept. of Sociology; Kroll-Smith, S. [New Orleans Univ., LA (United States)] New Orleans Univ., LA (United States); Couch, S.R. [Pennsylvania State Univ., University Park, PA (United States)] Pennsylvania State Univ., University Park, PA (United States); Edelstein, M.R. [Ramapo College, Mahwah, NJ (United States)] Ramapo College, Mahwah, NJ (United States)

1992-05-01T23:59:59.000Z

66

Decontamination of high-level waste canisters  

SciTech Connect

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

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

1980-12-01T23:59:59.000Z

67

EPRI Review of Geologic Disposal for Used Fuel and High Level Radioactive Waste: Volume I--The U.S. Site Selection Process Prior to the Nuclear Waste Policy Amendments Act  

Science Conference Proceedings (OSTI)

U.S. efforts to site and construct a deep geologic repository for used fuel and high level radioactive waste (HLW) proceeded in fits and starts over a three decade period from the late 1950s until 1982, when the U.S. Congress enacted the Nuclear Waste Policy Act (NWPA). This legislation codified a national approach for developing a deep geologic repository. Amendment of the NWPA in 1987 resulted in a number of dramatic changes in direction for the U.S. program, most notably the selection of Yucca Mountai...

2010-05-27T23:59:59.000Z

68

A literature review of coupled thermal-hydrologic-mechanical-chemical processes pertinent to the proposed high-level nuclear waste repository at Yucca Mountain  

SciTech Connect

A literature review has been conducted to determine the state of knowledge available in the modeling of coupled thermal (T), hydrologic (H), mechanical (M), and chemical (C) processes relevant to the design and/or performance of the proposed high-level waste (HLW) repository at Yucca Mountain, Nevada. The review focuses on identifying coupling mechanisms between individual processes and assessing their importance (i.e., if the coupling is either important, potentially important, or negligible). The significance of considering THMC-coupled processes lies in whether or not the processes impact the design and/or performance objectives of the repository. A review, such as reported here, is useful in identifying which coupled effects will be important, hence which coupled effects will need to be investigated by the US Nuclear Regulatory Commission in order to assess the assumptions, data, analyses, and conclusions in the design and performance assessment of a geologic reposit``. Although this work stems from regulatory interest in the design of the geologic repository, it should be emphasized that the repository design implicitly considers all of the repository performance objectives, including those associated with the time after permanent closure. The scope of this review is considered beyond previous assessments in that it attempts with the current state-of-knowledge) to determine which couplings are important, and identify which computer codes are currently available to model coupled processes.

Manteufel, R.D.; Ahola, M.P.; Turner, D.R.; Chowdhury, A.H. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

1993-07-01T23:59:59.000Z

69

Update to Assessment of Direct Disposal in Unsaturated Tuff of Spent Nuclear Fuel and High-Level Waste Owned by U.S. Department of Energy  

SciTech Connect

The overall purpose of this study is to provide information and guidance to the Office of Environmental Management of the U.S. Department of Energy (DOE) about the level of characterization necessary to dispose of DOE-owned spent nuclear fuel (SNF). The disposal option modeled was codisposal of DOE SNF with defense high-level waste (DHLW). A specific goal was to demonstrate the influence of DOE SNF, expected to be minor, in a predominately commercial repository using modeling conditions similar to those currently assumed by the Yucca Mountain Project (YMP). A performance assessment (PA) was chosen as the method of analysis. The performance metric for this analysis (referred to as the 1997 PA) was dose to an individual; the time period of interest was 100,000 yr. Results indicated that cumulative releases of 99Tc and 237Np (primary contributors to human dose) from commercial SNF exceed those of DOE SNF both on a per MTHM and per package basis. Thus, if commercial SNF can meet regulatory performance criteria for dose to an individual, then the DOE SNF can also meet the criteria. This result is due in large part to lower burnup of the DOE SNF (less time for irradiation) and to the DOE SNF's small percentage of the total activity (1.5%) and mass (3.8%) of waste in the potential repository. Consistent with the analyses performed for the YMP, the 1997 PA assumed all cladding as failed, which also contributed to the relatively poor performance of commercial SNF compared to DOE SNF.

P. D. Wheatley (INEEL POC); R. P. Rechard (SNL)

1998-09-01T23:59:59.000Z

70

West Valley Demonstration Project High-Level Waste Management  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

DRAFT_19507_1 DRAFT_19507_1 High-Level Waste Management Bryan Bower, DOE Director - WVDP DOE High-Level Waste Corporate Board Meeting Savannah River Site April 1, 2008 West Valley Demonstration Project West Valley Demonstration Project DRAFT_19507_2 West Valley High-Level Waste To solidify the radioactive material from approximately 600,000 gallons of high-level radioactive waste into a durable, high-quality glass, both a pretreatment system to remove salts and sulfates from the waste and a vitrification system/process were designed. To solidify the radioactive material from approximately 600,000 gallons of high-level radioactive waste into a durable, high-quality glass, both a pretreatment system to remove salts and sulfates from the waste and a vitrification system/process were designed.

71

Defense High Level Waste Disposal Container System Description  

Science Conference Proceedings (OSTI)

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.

NONE

2000-10-12T23:59:59.000Z

72

Development of a High Level Waste Tank Inspection System  

SciTech Connect

The Westinghouse Savannah River Technology Center was requested by it`s sister site, West Valley Nuclear Service (WVNS), to develop a remote inspection system to gather wall thickness readings of their High Level Waste Tanks. WVNS management chose to take a proactive approach to gain current information on two tanks t hat had been in service since the early 70`s. The tanks contain high level waste, are buried underground, and have only two access ports to an annular space between the tank and the secondary concrete vault. A specialized remote system was proposed to provide both a visual surveillance and ultrasonic thickness measurements of the tank walls. A magnetic wheeled crawler was the basis for the remote delivery system integrated with an off-the-shelf Ultrasonic Data Acquisition System. A development program was initiated for Savannah River Technology Center (SRTC) to design, fabricate, and test a remote system based on the Crawler. The system was completed and involved three crawlers to perform the needed tasks, an Ultrasonic Crawler, a Camera Crawler, and a Surface Prep Crawler. The crawlers were computer controlled so that their operation could be done remotely and their position on the wall could be tracked. The Ultrasonic Crawler controls were interfaced with ABB Amdata`s I-PC, Ultrasonic Data Acquisition System so that thickness mapping of the wall could be obtained. A second system was requested by Westinghouse Savannah River Company (WSRC), to perform just ultrasonic mapping on their similar Waste Storage Tanks; however, the system needed to be interfaced with the P-scan Ultrasonic Data Acquisition System. Both remote inspection systems were completed 9/94. Qualifications tests were conducted by WVNS prior to implementation on the actual tank and tank development was achieved 10/94. The second inspection system was deployed at WSRC 11/94 with success, and the system is now in continuous service inspecting the remaining high level waste tanks at WSRC.

Appel, D.K.; Loibl, M.W. [Westinghouse Savannah River Company, SC (United States); Meese, D.C. [Westinghouse West Valley Nuclear Services, West Valley, NY (United States)

1995-03-21T23:59:59.000Z

73

Final Systems Development Report for the Clark County Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Yucca Mountain, NV  

SciTech Connect

The Systems Development Report represents the third major step in the Clark County Socioeconomic Impact Assessment of the Proposed High-Level Nuclear Waste Repository at Yucca Mound Nevada. The first of these steps was to forge a Research Design that would serve as a guide for the overall research process. The second step was the construction of the Base Case, the purpose of which was to describe existing conditions in Clark County in the specified analytic areas of Economic-Demographic/Fiscal, Emergency Planning and Management, Transportation and Sociocultural analysis. The base case description will serve as a basis for assessing changes in these topic areas that might result from the Yucca Mountain project. These changes will be assessed by analyzing conditions with and without repository development in the county. Prior to performing such assessments, however, the snapshot type of data found in the base case must be operationalized or systematized to allow for more dynamic data utilization. In other words, a data system that can be used to analyze the consequences of the introduction of different variables (or variable values) in the Clark County context must be constructed. Such a system must be capable of being updated through subsequent data collection and monitoring efforts to both provide a rolling base case and supply information necessary to construct trend analyses. For example, during the Impact Assessment phase of the study process, the without repository analysis is accomplished by analyzing growth for the county given existing conditions and likely trends. These data are then compared to the with Yucca Mountain project conditions anticipated for the county. Similarly, once the emergency planning management and response needs associated with the repository are described, these needs will be juxtaposed against existing (and various future) capacity(ies) in order to determine the nature and magnitude of impacts in this analytic area. Analogous tasks will be performed for the other analytic areas detailed in the Base Case and outlined below.

NONE

1992-06-18T23:59:59.000Z

74

High-level waste tank farm set point document  

Science Conference Proceedings (OSTI)

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.

Anthony, J.A. III

1995-01-15T23:59:59.000Z

75

Immobilization of Nuclear Wastes  

Science Conference Proceedings (OSTI)

Oct 20, 2010 ... Glassy and Glass Composite Nuclear Wasteforms: Michael Ojovan1; Bill Lee2; ... wastes which should be solidified for safe storage and disposal. ... has been vitrifying the Department of Energy's High Level Waste (HLW) at ...

76

National high-level waste systems analysis report  

SciTech Connect

This report documents the assessment of budgetary impacts, constraints, and repository availability on the storage and treatment of high-level waste and on both existing and pending negotiated milestones. The impacts of the availabilities of various treatment systems on schedule and throughput at four Department of Energy sites are compared to repository readiness in order to determine the prudent application of resources. The information modeled for each of these sites is integrated with a single national model. The report suggests a high-level-waste model that offers a national perspective on all high-level waste treatment and storage systems managed by the Department of Energy.

Kristofferson, K.; Oholleran, T.P.; Powell, R.H.

1995-09-01T23:59:59.000Z

77

Studies of Mercury in High Level Waste Systems  

Science Conference Proceedings (OSTI)

During nuclear weapons production, nuclear reactor target and fuel rods were processed in F- and H-Canyons. For the target rods, a caustic dissolution of the aluminum cladding was performed prior to nitric acid dissolution of the uranium metal targets in the large canyon dissolvers. To dissolve the aluminum cladding and the U-Al fuel, mercury in the form of soluble mercury (II) nitrate was added as a catalyst to accelerate the dissolution of the aluminum. F-Canyon began to process plutonium-containing residues that were packaged in aluminum cans and thus required the use of mercury as a dissolution catalyst. Following processing to remove uranium and plutonium using the solvent extraction process termed the Plutonium-Uranium Recovery by Extraction (PUREX) process, the acidic waste solutions containing fission products and other radionuclides were neutralized with sodium hydroxide. The mercury used in canyon processing is fractionated between the sludge and supernate that is transferred from the canyons to the tank farm. The sludge component of the waste is currently vitrified in the Defense Waste Processing Facility (DWPF). The vitrified waste canisters are to be sent to the federal repository for High Level Waste. The mercury in the sludge, presumably in an oxide or hydroxide form is reduced to elemental mercury by the chemical additions and high temperatures, steam stripped and collected in the Mercury Collection Tank. The mercury in the dilute supernate is in the form of mercuric ion and is soluble. During evaporation, the mercuric ion is reduced to elemental mercury, vaporizes into the overheads system and is collected as a metallic liquid in the Mercury Removal Tank.

Wilmarth, W.R.

2003-09-03T23:59:59.000Z

78

Systems study of the feasibility of high-level nuclear waste fractionation for thermal stress control in a geologic repository: appendices  

Science Conference Proceedings (OSTI)

This study assesses the benefits and costs of fractionating the cesium and strontium (Cs/Sr) components in commercial high-level waste (HLW) to a separate waste stream for the purpose of reducing geologic-repository thermal stresses in the region of the HLW. The major conclusion is that the Cs/Sr fractionation concept offers the prospect of a substantial total system cost advantage for HLW disposal if reduced HLW package temperatures in a basalt repository are desired. However there is no cost advantage if currently designated maximum design temperatures are acceptable. Aging the HLW for 50 to 100 years can accomplish similar results at equivalent or lower costs. Volume II contains appendices for: (1) thermal analysis supplement; (2) fractionation process experimental results supplement; (3) cost analysis supplement; and (4) radiological risk analysis supplement.

McKee, R.W.; Elder, H.K.; McCallum, R.F.; Silviera, D.J.; Swanson, J.L.; Wiles, L.E.

1983-06-01T23:59:59.000Z

79

ATW system impact on high-level waste  

SciTech Connect

This report discusses the Accelerator Transmutation of Waste (ATW) concept which aims at destruction of key long-lived radionuclides in high-level nuclear waste (HLW), both fission products and actinides. This focus makes it different from most other transmutation concepts which concentrate primarily on actinide burning. The ATW system uses an accelerator-driven, sub-critical assembly to create an intense thermal neutron environment for radionuclide transmutation. This feature allows rapid transmutation under low-inventory system conditions, which in turn, has a direct impact on the size of chemical separations and materials handling components of the system. Inventories in ATW are factors of eight to thirty times smaller than reactor systems of equivalent thermal power. Chemical separations systems are relatively small in scale and can be optimized to achieve high decontamination factors and minimized waste streams. The low-inventory feature also directly impacts material amounts remaining in the system at its end of life. In addition to its low-inventory operation, the accelerator-driven neutron source features of ATW are key to providing a sufficient level of neutrons to allow transmutation of long-lived fission products.

Arthur, E.D.

1992-01-01T23:59:59.000Z

80

ATW system impact on high-level waste  

Science Conference Proceedings (OSTI)

This report discusses the Accelerator Transmutation of Waste (ATW) concept which aims at destruction of key long-lived radionuclides in high-level nuclear waste (HLW), both fission products and actinides. This focus makes it different from most other transmutation concepts which concentrate primarily on actinide burning. The ATW system uses an accelerator-driven, sub-critical assembly to create an intense thermal neutron environment for radionuclide transmutation. This feature allows rapid transmutation under low-inventory system conditions, which in turn, has a direct impact on the size of chemical separations and materials handling components of the system. Inventories in ATW are factors of eight to thirty times smaller than reactor systems of equivalent thermal power. Chemical separations systems are relatively small in scale and can be optimized to achieve high decontamination factors and minimized waste streams. The low-inventory feature also directly impacts material amounts remaining in the system at its end of life. In addition to its low-inventory operation, the accelerator-driven neutron source features of ATW are key to providing a sufficient level of neutrons to allow transmutation of long-lived fission products.

Arthur, E.D.

1992-12-01T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

High-Level Liquid Waste Tank Integrity Workshop - 2008  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Liquid Waste Tank Integrity Liquid Waste Tank Integrity Workshop - 2008 Karthik Subramanian Bruce Wiersma November 2008 High Level Waste Corporate Board Meeting karthik.subramanian@srnl.doe.gov bruce.wiersma@srnl.doe.gov 2 Acknowledgements * Bruce Wiersma (SRNL) * Kayle Boomer (Hanford) * Michael T. Terry (Facilitator) * SRS - Liquid Waste Organization * Hanford Tank Farms * DOE-EM 3 Background * High level radioactive waste (HLW) tanks provide critical interim confinement for waste prior to processing and permanent disposal * Maintaining structural integrity (SI) of the tanks is a critical component of operations 4 Tank Integrity Workshop - 2008 * Discuss the HLW tank integrity technology needs based upon the evolving waste processing and tank closure requirements along with its continued storage mission

82

Proposed methodology for completion of scenario analysis for the Basalt Waste Isolation Project. [Assessment of post-closure performance for a proposed repository for high-level nuclear waste  

Science Conference Proceedings (OSTI)

This report presents the methodology to complete an assessment of postclosure performance, considering all credible scenarios, including the nominal case, for a proposed repository for high-level nuclear waste at the Hanford Site, Washington State. The methodology consists of defensible techniques for identifying and screening scenarios, and for then assessing the risks associated with each. The results of the scenario analysis are used to comprehensively determine system performance and/or risk for evaluation of compliance with postclosure performance criteria (10 CFR 60 and 40 CFR 191). In addition to describing the proposed methodology, this report reviews available methodologies for scenario analysis, discusses pertinent performance assessment and uncertainty concepts, advises how to implement the methodology (including the organizational requirements and a description of tasks) and recommends how to use the methodology in guiding future site characterization, analysis, and engineered subsystem design work. 36 refs., 24 figs., 1 tab.

Roberds, W.J.; Plum, R.J.; Visca, P.J.

1984-11-01T23:59:59.000Z

83

Analysis of Surface Leaching Processes in Vitrified High-Level Nuclear Wastes Using In-Situ Raman Imaging and Atomistic Modeling - Final Report  

SciTech Connect

The in situ analysis of surface conditions of vitrified nuclear wastes can provide an important check of the burial status of radioactive objects without risk of radiation exposure. Raman spectroscopy was initially chosen as the most promising method for testing the surface conditions of glasses undergoing chemical corrosion, and was used extensively during the first year. However, it was determined that infrared reflection spectroscopy was better suited to this particular need and was used for the remaining two years to investigate the surface corrosion behavior of model silicate glasses for extension to nuclear waste glasses. The developed methodology is consistent with the known theory of optical propagation of dielectric media and uses the Kramers-Kronig formalism. The results show that it is possible to study the corrosion of glass by analyzing the glass surface using reflection fast Fourier infrared measurements and the newly developed ''dispersion analysis method.'' The data show how this analysis can be used to monitor the corrosion behavior of vitrified waste glasses over extended periods of storage.

Simmons, Joseph H.

2001-04-24T23:59:59.000Z

84

High Level Radioactive Waste- Doing Something about It  

Science Conference Proceedings (OSTI)

Symposium, Materials Issues in Nuclear Waste Management in the 21st Century. Presentation Title ... Metal Organic Frameworks for Clean Energy Applications.

85

RETENTION OF SULFATE IN HIGH LEVEL RADIOACTIVE WASTE GLASS  

SciTech Connect

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.

Fox, K.

2010-09-07T23:59:59.000Z

86

PLUTONIUM/HIGH-LEVEL VITRIFIED WASTE BDBE DOSE CALCULATION  

Science Conference Proceedings (OSTI)

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.

D.C. Richardson

2003-03-19T23:59:59.000Z

87

West Valley demonstration project: alternative processes for solidifying the high-level wastes  

SciTech Connect

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.

Holton, L.K.; Larson, D.E.; Partain, W.L.; Treat, R.L.

1981-10-01T23:59:59.000Z

88

Corrosion and failure processes in high-level waste tanks  

Science Conference Proceedings (OSTI)

A large amount of radioactive waste has been stored safely at the Savannah River and Hanford sites over the past 46 years. The aim of this report is to review the experimental corrosion studies at Savannah River and Hanford with the intention of identifying the types and rates of corrosion encountered and indicate how these data contribute to tank failure predictions. The compositions of the High-Level Wastes, mild steels used in the construction of the waste tanks and degradation-modes particularly stress corrosion cracking and pitting are discussed. Current concerns at the Hanford Site are highlighted.

Mahidhara, R.K.; Elleman, T.S.; Murty, K.L. [North Carolina State Univ., Raleigh, NC (United States)

1992-11-01T23:59:59.000Z

89

Criticality Safety Evaluation of Hanford Site High Level Waste Storage Tanks  

Science Conference Proceedings (OSTI)

This criticality safety evaluation covers operations for waste in underground storage tanks at the high-level waste tank farms on the Hanford site. This evaluation provides the bases for criticality safety limits and controls to govern receipt, transfer, and long-term storage of tank waste. Justification is provided that a nuclear criticality accident cannot occur for tank farms operations, based on current fissile material and operating conditions.

ROGERS, C.A.

2000-02-17T23:59:59.000Z

90

Break Throughs in High-Level Waste Vitrification for the Hanford ...  

Science Conference Proceedings (OSTI)

... Throughs in High-Level Waste Vitrification for the Hanford Waste Vitrification Plant ... Waste at the Defense Waste Processing Facility through Sludge Batch 7b .

91

High-Level Waste Corporate Board Meeting Agenda  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

High-Level Waste Corporate Board High-Level Waste Corporate Board Meeting Agenda Loews Hotel 1065 West Peachtree St, Atlanta, Georgia November 18, 2010 Time Topic Speaker 7:30 AM Closed Session - ratify Charter Board members 8:30 AM Welcome, Introduction, 2011 focus for HLW Corp Board Shirley Olinger 8:50 AM Introduction to Tc/I in Hanford Flowsheet ï‚· Show flowsheet w/ split locations ï‚· Describe recycle of LAW concept ï‚· Discuss baseline assumptions ï‚· Describe subsequent talks using flowsheet figure Gary Smith 9:15 AM Waste Treatment & Immobilization Plant (WTP) ï‚· Tc/I split factors (w/ and w/o recycle) ï‚· Water management (w/ and w/o recycle) Albert Kruger 9:45 AM WTP Melter/Offgas Systems Decontamination Factors ï‚· Re as a stimulant for Tc ï‚· Issues that limit Tc incorporation in LAW glass

92

EIS-0287: Idaho High-Level Waste and Facilities Disposition Final...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002) EIS-0287: Idaho High-Level Waste and Facilities Disposition Final...

93

High-level waste management technology program plan  

Science Conference Proceedings (OSTI)

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.

Harmon, H.D.

1995-01-01T23:59:59.000Z

94

High-Level Waste Corporate Board Performance Assessment Subcommittee  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Level Level Waste Corporate Board Performance Assessment Subcommittee John E. Marra, Ph.D. Associate Laboratory Director November 6, 2008 Richland, WA DOE-EM HLW Corporate Board Meeting Background - Performance Assessment Process Performance assessments are the fundamental risk assessment tool used by the DOE to evaluate and communicate the effectiveness and long-term impact of waste management and cleanup decisions. This includes demonstrations of compliance, NEPA analyses, and decisions about technologies and 2 analyses, and decisions about technologies and waste forms. Background - Process Perception EM-2 'Precepts' for Improved High-Level Waste Management (HLW Corporate Board Meeting - April 2008) Improved Performance Assessments (PA) The PA process is not consistently applied amongst the 3 The PA process is not consistently applied amongst the major HLW sites PA

95

Northeast High-Level Radioactive Waste Transportation Task Force Agenda  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Northeast High-Level Radioactive Waste Transportation Task Force Northeast High-Level Radioactive Waste Transportation Task Force Spring Meeting - May 15, 2012 Hilton Knoxville 501 West Church Avenue, Knoxville, TN 37902-2591 Agenda (Draft #1 - 4/18/12) ______________________________________________________________________________ Tuesday, May 15 - 9:00 AM - 3:30 PM / (need meeting room name) 8:00 a.m. Continental Breakfast - served in meeting room 9:00 a.m. Task Force Business Meeting - John Giarrusso, MEMA and Rich Pinney, NJDEP Co-chairs presiding  Welcome: Introductions; Agenda Review; Announcements  2012 funding  Co-Chair Election  Rules of Procedure  Membership: members & alternates appointment status  Legislative Liaisons  Staff Regional Meeting Attendance

96

Nondestructive examination of DOE high-level waste storage tanks  

SciTech Connect

A number of DOE sites have buried tanks containing high-level waste. Tanks of particular interest am double-shell inside concrete cylinders. A program has been developed for the inservice inspection of the primary tank containing high-level waste (HLW), for testing of transfer lines and for the inspection of the concrete containment where possible. Emphasis is placed on the ultrasonic examination of selected areas of the primary tank, coupled with a leak-detection system capable of detecting small leaks through the wall of the primary tank. The NDE program is modelled after ASME Section XI in many respects, particularly with respects to the sampling protocol. Selected testing of concrete is planned to determine if there has been any significant degradation. The most probable failure mechanisms are corrosion-related so that the examination program gives major emphasis to possible locations for corrosion attack.

Bush, S.; Bandyopadhyay, K.; Kassir, M.; Mather, B.; Shewmon, P.; Streicher, M.; Thompson, B.; van Rooyen, D.; Weeks, J.

1995-05-01T23:59:59.000Z

97

Life Extension of Aging High-Level Waste Tanks  

Science Conference Proceedings (OSTI)

The Double Shell Tanks (DSTs) play a critical role in the Hanford High-Level Waste Treatment Complex, and therefore activities are underway to protect and better understand these tanks. The DST Life Extension Program is focused on both tank life extension and on evaluation of tank integrity. Tank life extension activities focus on understanding tank failure modes and have produced key chemistry and operations controls to minimize tank corrosion and extend useful tank life. Tank integrity program activities have developed and applied key technologies to evaluate the condition of the tank structure and predict useful tank life. Program results to date indicate that DST useful life can be extended well beyond the original design life and allow the existing tanks to fill a critical function within the Hanford High-Level Waste Treatment Complex. In addition the tank life may now be more reliably predicted, facilitating improved planning for the use and possible future replacement of these tanks.

Bryson, D.; Callahan, V.; Ostrom, M.; Bryan, W.; Berman, H.

2002-02-26T23:59:59.000Z

98

DOE/EIS-0250D; Draft 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  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

U.S. Department of Energy (DOE) U.S. Department of Energy (DOE) TITLE: Draft 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 CONTACT: For more information on this Environmental Impact Statement (EIS), write or call: Wendy R. Dixon, EIS Project Manager Yucca Mountain Site Characterization Office Office of Civilian Radioactive Waste Management U.S. Department of Energy P.O. Box 30307, Mail Stop 010 North Las Vegas, Nevada 89036-0307 Telephone: (800) 967-3477 The EIS is also available on the Internet at the Yucca Mountain Project website at http://www.ymp.gov and on the DOE National Environmental Policy Act (NEPA) website at http://tis.eh.doe.gov/nepa/. For general information on the DOE NEPA process, write or call:

99

Permitting plan for the high-level waste interim storage  

SciTech Connect

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.

Deffenbaugh, M.L.

1997-04-23T23:59:59.000Z

100

High-level waste melter alternatives assessment report  

SciTech Connect

This document describes the Tank Waste Remediation System (TWRS) High-Level Waste (HLW) Program`s (hereafter referred to as HLW Program) Melter Candidate Assessment Activity performed in fiscal year (FY) 1994. The mission of the TWRS Program is to store, treat, and immobilize highly radioactive Hanford Site waste (current and future tank waste and encapsulated strontium and cesium isotopic sources) in an environmentally sound, safe, and cost-effective manner. The goal of the HLW Program is to immobilize the HLW fraction of pretreated tank waste into a vitrified product suitable for interim onsite storage and eventual offsite disposal at a geologic repository. Preparation of the encapsulated strontium and cesium isotopic sources for final disposal is also included in the HLW Program. As a result of trade studies performed in 1992 and 1993, processes planned for pretreatment of tank wastes were modified substantially because of increasing estimates of the quantity of high-level and transuranic tank waste remaining after pretreatment. This resulted in substantial increases in needed vitrification plant capacity compared to the capacity of original Hanford Waste Vitrification Plant (HWVP). The required capacity has not been finalized, but is expected to be four to eight times that of the HWVP design. The increased capacity requirements for the HLW vitrification plant`s melter prompted the assessment of candidate high-capacity HLW melter technologies to determine the most viable candidates and the required development and testing (D and T) focus required to select the Hanford Site HLW vitrification plant melter system. An assessment process was developed in early 1994. This document describes the assessment team, roles of team members, the phased assessment process and results, resulting recommendations, and the implementation strategy.

Calmus, R.B.

1995-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

High-Level Waste Systems Plan. Revision 7  

Science Conference Proceedings (OSTI)

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

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

1996-10-01T23:59:59.000Z

102

High level waste interim storge architecture selection - decision report  

SciTech Connect

The U.S. Department of Energy (DOE) has embarked upon a course to acquire Hanford Site tank waste treatment and immobilization services using privatized facilities (RL 1996a). This plan contains a two-phased approach. Phase I is a proof-of-principle/connnercial demonstration- scale effort and Phase II is a fiill-scale production effort. In accordance with the planned approach, interim storage and disposal of various products from privatized facilities are to be DOE fumished. The high-level waste (BLW) interim storage options, or alternative architectures, were identified and evaluated to provide the framework from which to select the most viable method of Phase I BLW interim storage (Calmus 1996). This evaluation, hereafter referred to as the Alternative Architecture Evaluation, was performed to established performance and risk criteria (technical merit, cost, schedule, etc.). Based on evaluation results, preliminary architectures and path forward reconunendations were provided for consideration in the architecture decision- maldng process. The decision-making process used for selection of a Phase I solidified BLW interim storage architecture was conducted in accordance with an approved Decision Plan (see the attachment). This decision process was based on TSEP-07,Decision Management Procedure (WHC 1995). The established decision process entailed a Decision Board, consisting of Westinghouse Hanford Company (VY`HC) management staff, and included appointment of a VTHC Decision Maker. The Alternative Architecture Evaluation results and preliminary recommendations were presented to the Decision Board members for their consideration in the decision-making process. The Alternative Architecture Evaluation was prepared and issued before issuance of @C-IP- 123 1, Alternatives Generation and Analysis Procedure (WI-IC 1996a), but was deemed by the Board to fully meet the intent of WHC-IP-1231. The Decision Board members concurred with the bulk of the Alternative Architecture Evaluation results and recommendations. However, the Board required changes to some criteria definitions and weightings in establishing its own recommendation basis. This report documents information presented to the Decision Board, and the Decision Board`s recommendations and basis for these recommendations. The Board`s recommendations were fully adopted by the WHC Decision Maker, R. J. Murkowski, Manager, TWRS Storage and Disposal. The Decision Board`s recommendation is as follows. The Phase I BLW Interim storage concept architecture will use Vaults 2 and 3 of the Hanford Site Spent Nuclear Fuel Canister Storage Building, being located in the Hanford Site 200 East Area, and include features to faciliate addition of one or more vaults at a later date.

Calmus, R.B.

1996-09-27T23:59:59.000Z

103

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

SciTech Connect

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.

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)

1990-12-01T23:59:59.000Z

104

The thermal conductivity of filler materials and permeability of a cement sealant for deep borehole repositories for high level nuclear waste  

E-Print Network (OSTI)

The Department of Energy is contractually obligated to begin the removal of spent nuclear fuel from reactor sites by the year 2020 at the risk of increased liabilities. The Blue Ribbon Commission on America's Nuclear Future ...

Salazar, Alex, III

2013-01-01T23:59:59.000Z

105

Spent Fuel and High-Level Waste Requirements (Maine) | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Spent Fuel and High-Level Waste Requirements (Maine) Spent Fuel and High-Level Waste Requirements (Maine) Spent Fuel and High-Level Waste Requirements (Maine) < Back Eligibility Agricultural Commercial Construction Fed. Government Fuel Distributor General Public/Consumer Industrial Installer/Contractor Institutional Investor-Owned Utility Local Government Low-Income Residential Multi-Family Residential Municipal/Public Utility Nonprofit Residential Retail Supplier Rural Electric Cooperative Schools State/Provincial Govt Systems Integrator Transportation Tribal Government Utility Program Info State Maine Program Type Safety and Operational Guidelines Provider Public Utilities Commission All proposed nuclear power generation facilities must be certified by the Public Utilities Commission under this statute prior to construction and

106

Potential for erosion corrosion of SRS high level waste tanks  

Science Conference Proceedings (OSTI)

SRS high-level radioactive waste tanks will not experience erosion corrosion to any significant degree during slurry pump operations. Erosion corrosion in carbon steel structures at reported pump discharge velocities is dominated by electrochemical (corrosion) processes. Interruption of those processes, as by the addition of corrosion inhibitors, sharply reduces the rate of metal loss from erosion corrosion. The well-inhibited SRS waste tanks have a near-zero general corrosion rate, and therefore will be essentially immune to erosion corrosion. The experimental data on carbon steel erosion corrosion most relevant to SRS operations was obtained at the Hanford Site on simulated Purex waste. A metal loss rate of 2.4 mils per year was measured at a temperature of 102 C and a slurry velocity comparable to calculated SRS slurry velocities on ground specimens of the same carbon steel used in SRS waste tanks. Based on these data and the much lower expected temperatures, the metal loss rate of SRS tanks under waste removal and processing conditions should be insignificant, i.e. less than 1 mil per year.

Zapp, P.E.

1994-01-01T23:59:59.000Z

107

Workshop on the source term for radionuclide migration from high-level waste or spent nuclear fuel under realistic repository conditions: proceedings  

SciTech Connect

Sixteen papers were presented at the workshop. The fourteen full-length papers included in the proceedings were processed separately. Only abstracts were included for the following two papers: Data Requirements Based on Performance Assessment Analyses of Conceptual Waste Packages in Salt Repositories, and The Potential Effects of Radiation on the Source Term in a Salt Repository. (LM)

Hunter, T.O.; Muller, A.B. (eds.)

1985-07-01T23:59:59.000Z

108

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

3 June 2008 3 June 2008 UPCOMING EVENTS: Next High-Level Waste Corporate Board meeting will be held at DOE-ID on 24 July 2008. Meeting details will be presented here and e-mailed to those persons with an interest to participate. Topics for discussion include: * Strategic Planning Initiative * Technology Development / Needs Collection / Prioritization * Waste Acceptance Product Specification This meeting will include a members-only executive session OTHER NEWS DOE SELECTS WASHINGTON RIVER PROTECTION SOLUTIONS, LLC FOR TANK OPERATIONS CONTRACT AT HANFORD SITE WASHINGTON, DC - The U.S. Department of Energy (DOE) today announced that Washington River Protection Solutions (WRPS), LLC has been selected as the tank operations contractor to store, retrieve and treat Hanford tank

109

Control of high level radioactive waste-glass melters  

DOE Green Energy (OSTI)

A necessary step in Defense Waste Processing Facility (DWPF) melter feed preparation for the immobilization of High Level Radioactive Waste (HLW) is reduction of Hg(II) to Hg(0), permitting steam stripping of the Hg. Denitrition and associated NOx evolution is a secondary effect of the use of formic acid as the mercury-reducing agent. Under certain conditions the presence of transition or noble metals can result in significant formic acid decomposition, with associated CO{sub 2} and H{sub 2} evolution. These processes can result in varying redox properties of melter feed, and varying sequential gaseous evolution of oxidants and hydrogen. Electrochemical methods for monitoring the competing processes are discussed. Laboratory scale techniques have been developed for simulating the large-scale reactions, investigating the relative effectiveness of the catalysts, and the effectiveness of catalytic poisons. The reversible nitrite poisoning of formic acid catalysts is discussed.

Bickford, D.F.; Coleman, C.J.; Hsu, C.L.W.; Eibling, R.E.

1990-01-01T23:59:59.000Z

110

High level radioactive waste vitrification process equipment component testing  

Science Conference Proceedings (OSTI)

Remote operability and maintainability of vitrification equipment were assessed under shielded-cell conditions. The equipment tested will be applied to immobilize high-level and transuranic liquid waste slurries that resulted from plutonium production for defense weapons. Equipment tested included: a turntable for handling waste canisters under the melter; a removable discharge cone in the melter overflow section; a thermocouple jumper that extends into a shielded cell; remote instrument and electrical connectors; remote, mechanical, and heat transfer aspects of the melter glass overflow section; a reamer to clean out plugged nozzles in the melter top; a closed circuit camera to view the melter interior; and a device to retrieve samples of the glass product. A test was also conducted to evaluate liquid metals for use in a liquid metal sealing system.

Siemens, D.H.; Heath, W.O.; Larson, D.E.; Craig, S.N.; Berger, D.N.; Goles, R.W.

1985-04-01T23:59:59.000Z

111

4.5 Meter high level waste canister study  

SciTech Connect

The Tank Waste Remediation System (TWRS) Storage and Disposal Project has established the Immobilized High-Level Waste (IBLW) Storage Sub-Project to provide the capability to store Phase I and II BLW products generated by private vendors. A design/construction project, Project W-464, was established under the Sub-Project to provide the Phase I capability. Project W-464 will retrofit the Hanford Site Canister Storage Building (CSB) to accommodate the Phase I I-ILW products. Project W-464 conceptual design is currently being performed to interim store 3.0 m-long BLW stainless steel canisters with a 0.61 in diameter, DOE is considering using a 4.5 in canister of the same diameter to reduce permanent disposal costs. This study was performed to assess the impact of replacing the 3.0 in canister with the 4.5 in canister. The summary cost and schedule impacts are described.

Calmus, R.B., Westinghouse Hanford, Richland, WA

1997-10-01T23:59:59.000Z

112

Review of High Level Waste Tanks Ultrasonic Inspection Data  

SciTech Connect

A review of the data collected during ultrasonic inspection of the Type I high level waste tanks has been completed. The data was analyzed for relevance to the possibility of vapor space corrosion and liquid/air interface corrosion. The review of the Type I tank UT inspection data has confirmed that the vapor space general corrosion is not an unusually aggressive phenomena and correlates well with predicted corrosion rates for steel exposed to bulk solution. The corrosion rates are seen to decrease with time as expected. The review of the temperature data did not reveal any obvious correlations between high temperatures and the occurrences of leaks. The complex nature of temperature-humidity interaction, particularly with respect to vapor corrosion requires further understanding to infer any correlation. The review of the waste level data also did not reveal any obvious correlations.

Wiersma, B

2006-03-09T23:59:59.000Z

113

DESIGN ANALYSIS FOR THE DEFENSE HIGH-LEVEL WASTE DISPOSAL CONTAINER  

SciTech Connect

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.

G. Radulesscu; J.S. Tang

2000-06-07T23:59:59.000Z

114

Process Design Concepts for Stabilization of High Level Waste Calcine  

Science Conference Proceedings (OSTI)

The current baseline assumption is that packaging ¡§as is¡¨ and direct disposal of high level waste (HLW) calcine in a Monitored Geologic Repository will be allowed. The fall back position is to develop a stabilized waste form for the HLW calcine, that will meet repository waste acceptance criteria currently in place, in case regulatory initiatives are unsuccessful. A decision between direct disposal or a stabilization alternative is anticipated by June 2006. The purposes of this Engineering Design File (EDF) are to provide a pre-conceptual design on three low temperature processes under development for stabilization of high level waste calcine (i.e., the grout, hydroceramic grout, and iron phosphate ceramic processes) and to support a down selection among the three candidates. The key assumptions for the pre-conceptual design assessment are that a) a waste treatment plant would operate over eight years for 200 days a year, b) a design processing rate of 3.67 m3/day or 4670 kg/day of HLW calcine would be needed, and c) the performance of waste form would remove the HLW calcine from the hazardous waste category, and d) the waste form loadings would range from about 21-25 wt% calcine. The conclusions of this EDF study are that: (a) To date, the grout formulation appears to be the best candidate stabilizer among the three being tested for HLW calcine and appears to be the easiest to mix, pour, and cure. (b) Only minor differences would exist between the process steps of the grout and hydroceramic grout stabilization processes. If temperature control of the mixer at about 80„aC is required, it would add a major level of complexity to the iron phosphate stabilization process. (c) It is too early in the development program to determine which stabilizer will produce the minimum amount of stabilized waste form for the entire HLW inventory, but the volume is assumed to be within the range of 12,250 to 14,470 m3. (d) The stacked vessel height of the hot process vessels in the hydroceramic grout process (i.e., 21 m) appears to be about the same as that estimated by the Direct Cementitious Waste Process in 1998, for which a conceptual design was developed. Some of the conceptual design efforts in the 1998 study may be applicable to the stabilizer processes addressed in this EDF. (e) The gamma radiation fields near the process vessels handling HLW calcine would vary from a range of about 300-350 R/hr at a distance of 2.5 cm from the side of the vessels to a range of about 50-170 R/hr at a distance of 100 cm from the side of the vessels. The calculations were made for combined calcine, which was defined as the total HLW calcine inventory uniformly mixed. (f) The gamma radiation fields near the stabilized waste in canisters would range from about 25-170 R/hr at 2.5 cm from the side of the canister and 5-35 R/hr at 100 cm from the side of the canister, depending on the which bin set was the source of calcine.

T. R. Thomas; A. K. Herbst

2005-06-01T23:59:59.000Z

115

HIGH LEVEL WASTE SLUDGE BATCH 4 VARIABILITY STUDY  

Science Conference Proceedings (OSTI)

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.

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

2006-10-02T23:59:59.000Z

116

Defense High-Level Waste Leaching Mechanisms Program. Final report  

SciTech Connect

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.

Mendel, J.E. (compiler)

1984-08-01T23:59:59.000Z

117

High Level Waste Corporate Board Newsletter - 06/03/09  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

UPCOMING EVENTS: UPCOMING EVENTS: Tank Waste Corporate Board Oak Ridge National Laboratory Oak Ridge, Tennessee 28 - 29 July 2009 The Board meeting will be preceded by a tour of the Radiochemical Engineering and Development Center on the afternoon of Tuesday, 28 July, and the meeting is planned for a full day on Wednesday, 29 July. Agenda Items include: ï‚· Future Directions for DOE Office of Nuclear Energy ï‚· Robotic Arm for Tank Cleaning ï‚· AREVA Mobile Hot Cell ï‚· Integrated Project Team Report ï‚· DOE Nuclear Safety Research and Development Coordinating Committee ï‚· Melton Valley Clean-Up: Lessons Learned ï‚· Chemical Cleaning of Waste Tanks at Savannah River - F Tank Farm Closure Project ï‚· Structural Integrity of Single Shell Tanks ï‚· Report from the Performance

118

High Level Waste System Impacts from Acid Dissolution of Sludge  

DOE Green Energy (OSTI)

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.

KETUSKY, EDWARD

2006-04-20T23:59:59.000Z

119

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

SciTech Connect

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.

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

2012-01-10T23:59:59.000Z

120

Alternatives Generation and Analysis for Heat Removal from High Level Waste Tanks  

Science Conference Proceedings (OSTI)

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.

WILLIS, W.L.

2000-06-15T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

Hanford high-level waste melter system evaluation data packages  

SciTech Connect

The Tank Waste Remediation System is selecting a reference melter system for the Hanford High-Level Waste vitrification plant. A melter evaluation was conducted in FY 1994 to narrow down the long list of potential melter technologies to a few for testing. A formal evaluation was performed by a Melter Selection Working Group (MSWG), which met in June and August 1994. At the June meeting, MSWG evaluated 15 technologies and selected six for more thorough evaluation at the Aug. meeting. All 6 were variations of joule-heated or induction-heated melters. Between the June and August meetings, Hanford site staff and consultants compiled data packages for each of the six melter technologies as well as variants of the baseline technologies. Information was solicited from melter candidate vendors to supplement existing information. This document contains the data packages compiled to provide background information to MSWG in support of the evaluation of the six technologies. (A separate evaluation was performed by Fluor Daniel, Inc. to identify balance of plant impacts if a given melter system was selected.)

Elliott, M.L.; Shafer, P.J.; Lamar, D.A.; Merrill, R.A.; Grunewald, W.; Roth, G.; Tobie, W.

1996-03-01T23:59:59.000Z

122

High Level Waste Feed Certification in Hanford Double Shell Tanks  

SciTech Connect

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 (1 million gallon) HLW feed tanks are homogenously mixed, representatively sampled, and consistently delivered to the WTP. It has been demonstrated that homogenous mixing of HLW 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 reduces operating efficiency and has 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.

Thien, Micheal G.; Wells, Beric E.; Adamson, Duane J.

2010-03-01T23:59:59.000Z

123

Remote Handling Equipment for a High-Level Waste Waste Package Closure System  

SciTech Connect

High-level waste will be placed in sealed waste packages inside a shielded closure cell. The Idaho National Laboratory (INL) has designed a system for closing the waste packages including all cell interior equipment and support systems. This paper discusses the material handling aspects of the equipment used and operations that will take place as part of the waste package closure operations. Prior to construction, the cell and support system will be assembled in a full-scale mockup at INL.

Kevin M. Croft; Scott M. Allen; Mark W. Borland

2006-04-01T23:59:59.000Z

124

JET MIXING ANALYSIS FOR SRS HIGH-LEVEL WASTE RECOVERY  

Science Conference Proceedings (OSTI)

The process of recovering the waste in storage tanks at the Savannah River Site (SRS) typically requires mixing the contents of the tank to ensure uniformity of the discharge stream. Mixing is accomplished with one to four slurry pumps located within the tank liquid. The slurry pump may be fixed in position or they may rotate depending on the specific mixing requirements. The high-level waste in Tank 48 contains insoluble solids in the form of potassium tetraphenyl borate compounds (KTPB), monosodium titanate (MST), and sludge. Tank 48 is equipped with 4 slurry pumps, which are intended to suspend the insoluble solids prior to transfer of the waste to the Fluidized Bed Steam Reformer (FBSR) process. The FBSR process is being designed for a normal feed of 3.05 wt% insoluble solids. A chemical characterization study has shown the insoluble solids concentration is approximately 3.05 wt% when well-mixed. The project is requesting a Computational Fluid Dynamics (CFD) mixing study from SRNL to determine the solids behavior with 2, 3, and 4 slurry pumps in operation and an estimate of the insoluble solids concentration at the suction of the transfer pump to the FBSR process. The impact of cooling coils is not considered in the current work. The work consists of two principal objectives by taking a CFD approach: (1) To estimate insoluble solids concentration transferred from Tank 48 to the Waste Feed Tank in the FBSR process and (2) To assess the impact of different combinations of four slurry pumps on insoluble solids suspension and mixing in Tank 48. For this work, several different combinations of a maximum of four pumps are considered to determine the resulting flow patterns and local flow velocities which are thought to be associated with sludge particle mixing. Two different elevations of pump nozzles are used for an assessment of the flow patterns on the tank mixing. Pump design and operating parameters used for the analysis are summarized in Table 1. The baseline pump orientations are chosen by the previous work [Lee et. al, 2008] and the initial engineering judgement for the conservative flow estimate since the modeling results for the other pump orientations are compared with the baseline results. As shown in Table 1, the present study assumes that each slurry pump has 900 gpm flowrate for the tank mixing analysis, although the Standard Operating Procedure for Tank 48 currently limits the actual pump speed and flowrate to a value less than 900 gpm for a 29 inch liquid level. Table 2 shows material properties and weight distributions for the solids to be modeled for the mixing analysis in Tank 48.

Lee, S.

2011-07-05T23:59:59.000Z

125

Preconceptual design study for solidifying high-level waste: Appendices A, B and C West Valley Demonstration Project  

SciTech Connect

This report presents a preconceptual design study for processing radioactive high-level liquid waste presently stored in underground tanks at Western New York Nuclear Service Center (WNYNSC) near West Valley, New York, and for incorporating the radionculides in that waste into a solid. The high-level liquid waste accumulated from the operation of a chemical reprocessing plant by the Nuclear Fuel Services, Inc. from 1966 to 1972. The high-level liquid waste consists of approximately 560,000 gallons of alkaline waste from Purex process operations and 12,000 gallons of acidic (nitric acid) waste from one campaign of processing thoria fuels by a modified Thorex process (during this campaign thorium was left in the waste). The alkaline waste contains approximately 30 million curies and the acidic waste contains approximately 2.5 million curies. The reference process described in this report is concerned only with chemically processing the high-level liquid waste to remove radionuclides from the alkaline supernate and converting the radionuclide-containing nonsalt components in the waste into a borosilicate glass.

Hill, O.F. (comp.)

1981-04-01T23:59:59.000Z

126

High level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 6  

SciTech Connect

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 6) outlines the standards and requirements for the sections on: Environmental Restoration and Waste Management, Research and Development and Experimental Activities, and Nuclear Safety.

Not Available

1994-04-01T23:59:59.000Z

127

DOE-EA-0179; Waste Form Selection for Savannah River Plant High-Level Waste  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

48326 (F.R.) 48326 (F.R.) NOTICES DEPARTMENT OF ENERGY Compliance With the National Environmental Policy Act Proposed Finding of No Significant Impact, Selection of Borosilicate Glass as the Defense Waste Processing Facility Waste Form for High -Level Radioactive Wastes Savanah River Plant, Aiken, South Carolina Thursday, July 29, 1982 *32778 AGENCY: Energy Department. ACTION: Notice. SUMMARY: The Department of Energy (DOE) has prepared an environmental assessment (DOE/EA- 0179) on the proposed selection of borosilicate glass as the Defense Waste Processing Facility (DWPF) waste form for the immobilization of the high -level radioactive wastes generated and stored at the DOE Savannah River Plant (SRP), Aiken, South Carolina. DOE recently decided to immobilize

128

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

Science Conference Proceedings (OSTI)

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.

SK Sundaram; ML Elliott; D Bickford

1999-11-19T23:59:59.000Z

129

Title: An Advanced Solution for the Storage, Transportation and Disposal of Vitrified High Level Waste  

NLE Websites -- All DOE Office Websites (Extended Search)

Presented at Global 99, Jackson, Wyoming, August 29 - September 2, 1999 Presented at Global 99, Jackson, Wyoming, August 29 - September 2, 1999 1 AN ADVANCED SOLUTION FOR THE STORAGE, TRANSPORTATION AND DISPOSAL OF SPENT FUEL AND VITRIFIED HIGH LEVEL WASTE William J. Quapp Teton Technologies, Inc. 860 W. Riverview Dr. Idaho Falls, ID 83401 208-535-9001 ABSTRACT For future nuclear power deployment in the US, certain changes in the back end of the fuel cycle, i.e., disposal of high level waste and spent fuel, must become a real options. However, there exists another problem from the front end of the fuel cycle which has until recently, received less attention. Depleted uranium hexafluoride is a by-product of the enrichment process and has accumulated for over 50 years. It now represents a potential environmental problem. This paper describes a

130

Pyrochemical treatment of Idaho Chemical Processing Plant high-level waste calcine  

SciTech Connect

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Engineering Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1951 to recover uranium, krypton-85, and isolated fission products for interim treatment and immobilization. The acidic radioactive high-level liquid waste (HLLW) is routinely stored in stainless steel tanks and then, since 1963, calcined to form a dry granular solid. The resulting high-level waste (HLW) calcine is stored in seismically hardened stainless steel bins that are housed in underground concrete vaults. A research and development program has been established to determine the feasibility of treating ICPP HLW calcine using pyrochemical technology.This technology is described.

Todd, T.A.; DelDebbio, J.A.; Nelson, L.O.; Sharpsten, M.R.

1993-06-01T23:59:59.000Z

131

Modeling of Boehmite Leaching from Actual Hanford High-Level Waste Samples  

SciTech Connect

The Department of Energy plans to vitrify approximately 60,000 metric tons of high level waste sludge from underground storage tanks at the Hanford Nuclear Reservation. To reduce the volume of high level waste requiring treatment, a goal has been set to remove about 90 percent of the aluminum, which comprises nearly 70 percent of the sludge. Aluminum in the form of gibbsite and sodium aluminate can be easily dissolved by washing the waste stream with caustic, but boehmite, which comprises nearly half of the total aluminum, is more resistant to caustic dissolution and requires higher treatment temperatures and hydroxide concentrations. In this work, the dissolution kinetics of aluminum species during caustic leaching of actual Hanford high level waste samples is examined. The experimental results are used to develop a shrinking core model that provides a basis for prediction of dissolution dynamics from known process temperature and hydroxide concentration. This model is further developed to include the effects of particle size polydispersity, which is found to strongly influence the rate of dissolution.

Peterson, Reid A.; Lumetta, Gregg J.; Rapko, Brian M.; Poloski, Adam P.

2007-06-27T23:59:59.000Z

132

Polysiloxane Encapsulation of High Level Calcine Waste for Transportation or Disposal  

SciTech Connect

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.

Loomis, Guy George

2000-03-01T23:59:59.000Z

133

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

SciTech Connect

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.

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

2000-01-01T23:59:59.000Z

134

West Valley Demonstration Project Prepares to Relocate High-Level Waste |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

West Valley Demonstration Project Prepares to Relocate High-Level West Valley Demonstration Project Prepares to Relocate High-Level Waste West Valley Demonstration Project Prepares to Relocate High-Level Waste December 24, 2013 - 12:00pm Addthis The West Valley Demonstration Project’s high-level waste canisters will be relocated to interim onsite storage. The West Valley Demonstration Project's high-level waste canisters will be relocated to interim onsite storage. The first group of eight concrete storage casks for the West Valley Demonstration Project’s high-level waste. The first group of eight concrete storage casks for the West Valley Demonstration Project's high-level waste. Site subcontractor American DND completed demolition of the contaminated 01-14 Building in 2013. Site subcontractor American DND completed demolition of the contaminated

135

West Valley Demonstration Project Prepares to Relocate High-Level Waste |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

West Valley Demonstration Project Prepares to Relocate High-Level West Valley Demonstration Project Prepares to Relocate High-Level Waste West Valley Demonstration Project Prepares to Relocate High-Level Waste December 24, 2013 - 12:00pm Addthis The West Valley Demonstration Project’s high-level waste canisters will be relocated to interim onsite storage. The West Valley Demonstration Project's high-level waste canisters will be relocated to interim onsite storage. The first group of eight concrete storage casks for the West Valley Demonstration Project’s high-level waste. The first group of eight concrete storage casks for the West Valley Demonstration Project's high-level waste. Site subcontractor American DND completed demolition of the contaminated 01-14 Building in 2013. Site subcontractor American DND completed demolition of the contaminated

136

High-Level Waste Corporate Board Performance Assessment Subcommittee  

NLE Websites -- All DOE Office Websites (Extended Search)

Level Waste Corporate Board Performance Assessment Community of Practice John E. Marra, Ph.D. Associate Laboratory Director 21 May 2009 Denver, CO Office of Waste Processing...

137

CERMET High Level Waste Forms - Oak Ridge National Laboratory  

>30% waste loading, reducing waste volume by 50% as compared to baseline glasses, while achieving performance equal to or better than such glasses.

138

EIS-0287: Idaho High-Level Waste & Facilities Disposition | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

7: Idaho High-Level Waste & Facilities Disposition 7: Idaho High-Level Waste & Facilities Disposition EIS-0287: Idaho High-Level Waste & Facilities Disposition SUMMARY 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 waste at the Idaho National Engineering and Environmental Laboratory (INEEL) in liquid and solid forms. This EIS also analyzes alternatives for the final disposition of HLW management facilities at the INEEL after their missions are completed. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD January 12, 2010 EIS-0287: Amended Record of Decision Idaho High-Level Waste and Facilities Disposition January 4, 2010

139

EIS-0287: Idaho High-Level Waste and Facilities Disposition Final  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Idaho High-Level Waste and Facilities Disposition Final Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002) EIS-0287: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, EIS-0287 (September 2002) 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 waste at the Idaho National Engineering and Environmental Laboratory (INEEL) in liquid and solid forms. This EIS also analyzes alternatives for the final disposition of HLW management facilities at the INEEL after their missions are completed. Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement, DOE/EIS-0287 (September 2002)

140

Characteristics Data Base: Programmer's guide to the High-Level Waste Data Base  

SciTech Connect

The High-Level Waste Data Base is a menu-driven PC data base developed as part of OCRWM's technical data base on the characteristics of potential repository wastes, which also includes spent fuel and other materials. This programmer's guide completes the documentation for the High-Level Waste Data Base, the user's guide having been published previously. 3 figs.

Jones, K.E. (DataPhile, Inc., Knoxville, TN (USA)); Salmon, R. (Oak Ridge National Lab., TN (USA))

1990-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

EM Waste Acceptance Product Specification (WAPS) for Vitrified High-Level Waste Forms  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EM Waste Acceptance Product EM Waste Acceptance Product Specification (WAPS) for Vitrified High-Level Waste Forms Presentation to the HLW Corporate Board July 24, 2008 By Tony Kluk/Ken Picha 2 Background * Originally Waste Acceptance Preliminary Specifications were Office of Civilian Radioactive Waste Management (RW) documents and project specific: - Defense Waste Processing Facility (PE-03, July 1989) - West Valley Demonstration Project (PE-04, January 1990) * Included many of same specifications as current version of WAPS * First version of RW Waste Acceptance System Requirements Document in January 1993 (included requirements for both SNF and HLW) * EM decided to extract requirements for HLW and put into the WAPS document 3 Background (Cont'd) * Lists technical specifications for acceptance of borosilicate HLW

142

Development and application of a conceptual approach for defining high-level waste  

SciTech Connect

This paper presents a conceptual approach to defining high-level radioactive waste (HLW) and a preliminary quantitative definition obtained from an example implementation of the conceptual approach. On the basis of the description of HLW in the Nuclear Waste Policy Act of 1982, we have developed a conceptual model in which HLW has two attributes: HLW is (1) highly radioactive and (2) requires permanent isolation via deep geologic disposal. This conceptual model results in a two-dimensional waste categorization system in which one axis, related to ''requires permanent isolation,'' is associated with long-term risks from waste disposal and the other axis, related to ''highly radioactive,'' is associated with short-term risks from waste management and operations; this system also leads to the specification of categories of wastes that are not HLW. Implementation of the conceptual model for defining HLW was based primarily on health and safety considerations. Wastes requiring permanent isolation via deep geologic disposal were defined by estimating the maximum concentrations of radionuclides that would be acceptable for disposal using the next-best technology, i.e., greater confinement disposal (GCD) via intermediate-depth burial or engineered surface structures. Wastes that are highly radioactive were defined by adopting heat generation rate as the appropriate measure and examining levels of decay heat that necessitate special methods to control risks from operations in a variety of nuclear fuel-cycle situations. We determined that wastes having a power density >200 W/m/sup 3/ should be considered highly radioactive. Thus, in the example implementation, the combination of maximum concentrations of long-lived radionuclides that are acceptable for GCD and a power density of 200 W/m/sup 3/ provides boundaries for defining wastes that are HLW.

Croff, A.G.; Forsberg, C.W.; Kocher, D.C.; Cohen, J.J.; Smith, C.F.; Miller, D.E.

1986-01-01T23:59:59.000Z

143

Safety analysis report vitrified high level waste type B shipping cask  

Science Conference Proceedings (OSTI)

This Safety Analysis Report describes the design, analyses, and principle features of the Vitrified High Level Waste (VHLW) Cask. In preparing this report a detailed evaluation of the design has been performed to ensure that all safety, licensing, and operational goals for the cask and its associated Department of Energy program can be met. The functions of this report are: (1) to fully document that all functional and regulatory requirements of 10CFR71 can be met by the package; and (2) to document the design and analyses of the cask for review by the Nuclear Regulatory Commission. The VHLW Cask is the reusable shipping package designed by GNSI under Department of Energy contract DE-AC04-89AL53-689 for transportation of Vitrified High Level Waste, and to meet the requirements for certification under 10CFR71 for a Type B(U) package. The VHLW cask has been designed as packaging for transport of canisters of Vitrified High Level Waste solidified at Department of Energy facilities.

NONE

1995-03-01T23:59:59.000Z

144

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

Science Conference Proceedings (OSTI)

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.

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

1990-04-01T23:59:59.000Z

145

EIS-0303: Savannah River Site High-Level Waste Tank Closure | Department of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

03: Savannah River Site High-Level Waste Tank Closure 03: Savannah River Site High-Level Waste Tank Closure EIS-0303: Savannah River Site High-Level Waste Tank Closure SUMMARY 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 selected the preferred alternative identified in the Final EIS, Stabilize Tanks-Fill with Grout, to guide development and implementation of closure of the high-level waste tanks and associated equipment at the Savannah River Site. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD March 5, 2012 EIS-0303: Supplement Analysis Savannah River Site High-Level Waste Tank Closure, SC July 8, 2011 EIS-0303: Notice of Intent to Prepare an Environmental Impact Statement

146

Development of Crystal-Tolerant High-Level Waste Glasses  

SciTech Connect

Twenty five glasses were formulated. They were batched from HLW AZ-101 simulant or raw chemicals and melted and tested with a series of tests to elucidate the effect of spinel-forming components (Ni, Fe, Cr, Mn, and Zn), Al, and noble metals (Rh2O3 and RuO2) on the accumulation rate of spinel crystals in the glass discharge riser of the high-level waste (HLW) melter. In addition, the processing properties of glasses, such as the viscosity and TL, were measured as a function of temperature and composition. Furthermore, the settling of spinel crystals in transparent low-viscosity fluids was studied at room temperature to access the shape factor and hindered settling coefficient of spinel crystals in the Stokes equation. The experimental results suggest that Ni is the most troublesome component of all the studied spinel-forming components producing settling layers of up to 10.5 mm in just 20 days in Ni-rich glasses if noble metals or a higher concentration of Fe was not introduced in the glass. The layer of this thickness can potentially plug the bottom of the riser, preventing glass from being discharged from the melter. The noble metals, Fe, and Al were the components that significantly slowed down or stopped the accumulation of spinel at the bottom. Particles of Rh2O3 and RuO2, hematite and nepheline, acted as nucleation sites significantly increasing the number of crystals and therefore decreasing the average crystal size. The settling rate of ?10-?m crystal size around the settling velocity of crystals was too low to produce thick layers. The experimental data for the thickness of settled layers in the glasses prepared from AZ-101 simulant were used to build a linear empirical model that can predict crystal accumulation in the riser of the melter as a function of concentration of spinel-forming components in glass. The developed model predicts the thicknesses of accumulated layers quite well, R2 = 0.985, and can be become an efficient tool for the formulation of the crystal-tolerant HLW glasses for higher waste loading. A physical modeling effort revealed that the Stokes and Richardson-Zaki equations can be used to adequately predict the accumulation rate of spinel crystals of different sizes and concentrations in the glass discharge riser of HLW melters. The determined shape factor for the glass beads was only 0.73% lower than the theoretical shape factor for a perfect sphere. The shape factor for the spinel crystals matched the theoretically predicted value to within 10% and was smaller than that of the beads, given the larger drag force caused by the larger surface area-to-volume ratio of the octahedral crystals. In the hindered settling experiments, both the glass bead and spinel suspensions were found to follow the predictions of the Richardson-Zaki equation with the exponent n = 3.6 and 2.9 for glass beads and spinel crystals, respectively.

Matyas, Josef; Vienna, John D.; Schaible, Micah J.; Rodriguez, Carmen P.; Crum, Jarrod V.; Arrigoni, Alyssa L.; Tate, Rachel M.

2010-12-17T23:59:59.000Z

147

Demonstrating Reliable High Level Waste Slurry Sampling Techniques to Support Hanford Waste Processing - 14194  

SciTech Connect

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.

Kelly, Steven E.

2013-11-11T23:59:59.000Z

148

Life Estimation of High Level Waste Tank Steel for H-Tank Farm ...  

the tanks is not considered in the analysis. Life Estimation of High Level Waste Tank ... conservative scenario in which the concrete vault has completely

149

Parametric Analyses of Heat Removal from High Level Waste Tanks  

Science Conference Proceedings (OSTI)

The general thermal hydraulics program GOTH-SNF was used to predict the thermal response of the waste in tanks 241-AY-102 and 241-AZ-102 when mixed by two 300 horsepower mixer pumps. This mixing was defined in terms of a specific waste retrieval scenario. Both dome and annulus ventilation system flow are necessary to maintain the waste within temperature control limits during the mixing operation and later during the sludge-settling portion of the scenario are defined.

TRUITT, J.B.

2000-06-05T23:59:59.000Z

150

WTP: Challenges and Major Breakthroughs in High Level Waste ...  

Science Conference Proceedings (OSTI)

Abstract Scope, The US DOE has developed glass property-composition models to control glass compositions for HLW vitrification at Hanford Waste Treatment ...

151

DOE G 435.1-1 Chapter 2, High-Level Waste Requirements  

Directives, Delegations, and Requirements

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.

1999-07-09T23:59:59.000Z

152

High level waste facilities -- Continuing operation or orderly shutdown  

SciTech Connect

Two options for Environmental Impact Statement No action alternatives describe operation of the radioactive liquid waste facilities at the Idaho Chemical Processing Plant at the Idaho National Engineering and Environmental Laboratory. The first alternative describes continued operation of all facilities as planned and budgeted through 2020. Institutional control for 100 years would follow shutdown of operational facilities. Alternatively, the facilities would be shut down in an orderly fashion without completing planned activities. The facilities and associated operations are described. Remaining sodium bearing liquid waste will be converted to solid calcine in the New Waste Calcining Facility (NWCF) or will be left in the waste tanks. The calcine solids will be stored in the existing Calcine Solids Storage Facilities (CSSF). Regulatory and cost impacts are discussed.

Decker, L.A.

1998-04-01T23:59:59.000Z

153

Source terms for analysis of accidents at a high level waste repository  

SciTech Connect

This paper describes an approach to identifying source terms from possible accidents during the preclosure phase of a high-level nuclear waste repository. A review of the literature on repository safety analyses indicated that source term estimation is in a preliminary stage, largely based on judgement-based scoping analyses. The approach developed here was to partition the accident space into domains defined by certain threshold values of temperature and impact energy density which may arise in potential accidents and specify release fractions of various radionuclides, present in the waste form, in each domain. Along with a more quantitative understanding of accident phenomenology, this approach should help in achieving a clearer perspective on scenarios important to preclosure safety assessments of geologic repositories. 18 refs., 3 tabs.

Mubayi, V.; Davis, R.E.; Youngblood, R.

1989-01-01T23:59:59.000Z

154

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

SciTech Connect

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)

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

2008-07-01T23:59:59.000Z

155

Preliminary estimates of cost savings for defense high level waste vitrification options  

SciTech Connect

The potential for realizing cost savings in the disposal of defense high-level waste through process and design modificatins has been considered. Proposed modifications range from simple changes in the canister design to development of an advanced melter capable of processing glass with a higher waste loading. Preliminary calculations estimate the total disposal cost (not including capital or operating costs) for defense high-level waste to be about $7.9 billion dollars for the reference conditions described in this paper, while projected savings resulting from the proposed process and design changes could reduce the disposal cost of defense high-level waste by up to $5.2 billion.

Merrill, R.A.; Chapman, C.C.

1993-09-01T23:59:59.000Z

156

Fracturing of simulated high-level waste glass in canisters  

SciTech Connect

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

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

1981-09-01T23:59:59.000Z

157

Development and Deployment of Advanced Corrosion Monitoring Systems for High-Level Waste Tanks  

SciTech Connect

This paper describes the results of a collaborative technology development program, sponsored by the Tanks Focus Area, to use electrochemical noise (EN) for corrosion monitoring in underground storage tanks. These tanks, made of carbon or stainless steels, contain high-level radioactive liquid waste (HLW) generated by weapons production or radioactive liquid waste from nuclear fuel reprocessing activities at several Department of Energy (DOE) sites. The term EN is used to describe low frequency fluctuations in current and voltage measurements associated with corrosion. In their most basic form, EN-based corrosion monitoring systems measure and record these fluctuations over time from electrodes immersed in the environment of interest--in this case, radioactive tank waste. The resulting EN signals have characteristic patterns for different corrosion mechanisms. In recent years, engineers and scientists from several DOE sites, in collaboration with several private companies, have conducted laboratory studies and field applications to correlate the EN signals with corrosion mechanisms active in the radioactive waste tanks. The participating DOE sites are Hanford, Savannah River, Oak Ridge Reservation and the Idaho National Engineering and Environmental Laboratory. The commercial vendors have included HiLine Engineering and Fabrication, Inc., EIC Laboratories, Inc., and AEA Technologies. Successful deployment of the EN technology will yield improved information of waste tank corrosion conditions, better tank management, and lower overall cost.

Terry, M. T.; Edgemon, G. L.; Mickalonis, J. I.; Mizia, R. E.

2002-02-26T23:59:59.000Z

158

Development and deployment of advanced corrosion monitoring systems for high-level waste tanks.  

Science Conference Proceedings (OSTI)

This paper describes the results of a collaborative technology development program, sponsored by the Tanks Focus Area, to use electrochemical noise (EN) for corrosion monitoring in underground storage tanks. These tanks, made of carbon or stainless steels, contain high-level radioactive liquid waste (HLW) generated by weapons production or radioactive liquid waste from nuclear fuel reprocessing activities at several Department of Energy (DOE) sites. The term EN is used to describe low frequency fluctuations in current and voltage measurements associated with corrosion. In their most basic form, EN-based corrosion monitoring systems measure and record these fluctuations over time from electrodes immersed in the environment of interest - in this case, radioactive tank waste. The resulting EN signals have characteristic patterns for different corrosion mechanisms. In recent years, engineers and scientists from several DOE sites, in collaboration with several private companies, have conducted laboratory studies and field applications to correlate the EN signals with corrosion mechanisms active in the radioactive waste tanks. The participating DOE sites are Hanford, Savannah River, Oak Ridge Reservation and the Idaho National Engineering and Environmental Laboratory. The commercial vendors have included HiLine Engineering and Fabrication, Inc., EIC Laboratories, Inc., and M A Technologies. Successful deployment of the EN technology will yield improved information of waste tank corrosion conditions, better tank management, and lower overall cost.

Terry, M. T. (Michael T.); Edgemon, G. L. (Glenn L.); Mickalonis, J. I. (John I.); Mizia, R. E. (Ronald E.)

2002-01-01T23:59:59.000Z

159

Iron Phosphate Glass as Potential Waste Matrix for High-Level Radioactive Waste  

Science Conference Proceedings (OSTI)

Recently, Iron Phosphate Glass (IPG) is investigated as the alternative final waste form for High-Level Radioactive Waste (HLW) in U.S. This study is aimed to investigate feasibility of IPG to HLW arising from commercial reprocessing in Japan. In order to evaluate favorable preparation conditions, maximum waste loading and property of IPG, the melting tests were carried. From the results of melting tests, the favorable preparation conditions was with matrix of Fe/P 0.43 (mole ratio in products) and melting at 1200{sup o} for 4h. The products of 10-20mass% waste loading of simulated HLW were glassy and had no crystal peaks, however the product of 30mass% waste loading showed some crystal peaks by XRD analysis. IPG and Borosilicate glass (BG) had about the same thermal properties. As a result, IPG had enough potential for high waste loading and the extremely good chemical durability for consideration as a waste form for Japanese HLW.

Fukui, T.; Ishinomori, T.; Endo, Y.; Sazarashi, M.; Ono, S.; Suzuki, K.

2003-02-25T23:59:59.000Z

160

Final Environmental Impact Statement (Supplement to ERDA-1537, September 1977) Waste Management Operations Double-Shell Tanks for Defense High-Level Radioactive Waste Storage Savannah River Plant  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Do Do E/EIS-0062 FINAL ENVIRONMENTAL IMPACT mATEIUIENT (Supplement to ERDA-1537, September 1977) Waste ~ Management Operations Savannah River Plant ! Aiken, South Carolina Double-Shell Tanks for Defense High-Level Radioactive Waste Storage April 1980 U.S. DEPARTMENT OF ENERGY WASHINGTON. D.C.20545 1980 WL 94273 (F.R.) NOTICES DEPARTMENT OF ENERGY Office of Deputy Assistant Secretary for Nuclear Waste Management Double-Shell Tanks for Defense High-Level Radioactive Waste Storage, Savannah River Plant, Aiken, S.C. Wednesday, July 9, 1980 *46154 Record of Decision Decision. The decision has been made to complete the construction of the 14 double-shell tanks and use them to store defense high-level radioactive waste at the Savannah River Plant (SRP). Background. The SRP, located near Aiken, South Carolina, is a major installation of the

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161

Coupled Model for Heat and Water Transport in a High Level Waste Repository  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Coupled Model for Heat and Water Transport in a High Level Waste Coupled Model for Heat and Water Transport in a High Level Waste Repository in Salt Coupled Model for Heat and Water Transport in a High Level Waste Repository in Salt 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; chemical processes of the system allow precipitation and dissolution of salt with elevated temperatures that drive water and water vapor flow around hot waste packages. Characterizing salt backfill processes is an important objective of the exercise. An evidence-based algorithm for mineral dehydration is also applied in the modeling. The Finite Element Heat and Mass transfer code (FEHM) is used to simulate coupled thermal,

162

Swedish nuclear waste efforts  

SciTech Connect

After the introduction of a law prohibiting the start-up of any new nuclear power plant until the utility had shown that the waste produced by the plant could be taken care of in an absolutely safe way, the Swedish nuclear utilities in December 1976 embarked on the Nuclear Fuel Safety Project, which in November 1977 presented a first report, Handling of Spent Nuclear Fuel and Final Storage of Vitrified Waste (KBS-I), and in November 1978 a second report, Handling and Final Storage of Unreprocessed Spent Nuclear Fuel (KBS II). These summary reports were supported by 120 technical reports prepared by 450 experts. The project engaged 70 private and governmental institutions at a total cost of US $15 million. The KBS-I and KBS-II reports are summarized in this document, as are also continued waste research efforts carried out by KBS, SKBF, PRAV, ASEA and other Swedish organizations. The KBS reports describe all steps (except reprocessing) in handling chain from removal from a reactor of spent fuel elements until their radioactive waste products are finally disposed of, in canisters, in an underground granite depository. The KBS concept relies on engineered multibarrier systems in combination with final storage in thoroughly investigated stable geologic formations. This report also briefly describes other activities carried out by the nuclear industry, namely, the construction of a central storage facility for spent fuel elements (to be in operation by 1985), a repository for reactor waste (to be in operation by 1988), and an intermediate storage facility for vitrified high-level waste (to be in operation by 1990). The R and D activities are updated to September 1981.

Rydberg, J.

1981-09-01T23:59:59.000Z

163

PLUTONIUM SOLUBILITY IN HIGH-LEVEL WASTE ALKALI BOROSILICATE GLASS  

SciTech Connect

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.

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

2011-01-04T23:59:59.000Z

164

Disposition of actinides released from high-level waste glass  

SciTech Connect

A series of static leach tests was conducted using glasses developed for vitrifying tank wastes at the Savannah River Site to monitor the disposition of actinide elements upon corrosion of the glasses. In these tests, glasses produced from SRL 131 and SRL 202 frits were corroded at 90{degrees}C in a tuff groundwater. Tests were conducted using crushed glass at different glass surface area-to-solution volume (S/V) ratios to assess the effect of the S/V on the solution chemistry, the corrosion of the glass, and the disposition of actinide elements. Observations regarding the effects of the S/V on the solution chemistry and the corrosion of the glass matrix have been reported previously. This paper highlights the solution analyses performed to assess how the S/V used in a static leach test affects the disposition of actinide elements between fractions that are suspended or dissolved in the solution, and retained by the altered glass or other materials.

Ebert, W.L.; Bates, J.K.; Buck, E.C.; Gong, M.; Wolf, S.F.

1994-05-01T23:59:59.000Z

165

PSA results for Hanford high level waste Tank 101-SY  

DOE Green Energy (OSTI)

Los Alamos National Laboratory has performed a comprehensive probabilistic safety assessment (PSA) that includes consideration of external events for 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 (``burps``) a gaseous mixture of hydrogen, nitrous oxide, ammonia, and nitrogen, was analyzed 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 underway 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. This PSA for Tank 101-SY, which did not consider the pump experiment or future tank-remediation activities, involved three distinct tasks. First, the accident sequence analysis identified and quantified those potential accidents whose consequences result in tank material release. Second, characteristics and release paths for the airborne and liquid radioactive source terms were determined. Finally, the consequences, primarily onsite and offsite potential health effects resulting from radionuclide release, were estimated, and overall risk curves were constructed. An overview of each of these tasks and a summary of the overall results of the analysis are presented in the following sections.

MacFarlane, D.R.; Bott, T.F.; Brown, L.F.; Stack, D.W. [Los Alamos National Lab., NM (United States); Kindinger, J.; Deremer, R.K.; Medhekar, S.R.; Mikschl, T.J. [PLG, Inc., Newport Beach, CA (United States)

1993-10-01T23:59:59.000Z

166

Annotated bibliography for the design of waste packages for geologic disposal of spent fuel and high-level waste  

SciTech Connect

This bibliography identifies documents that are pertinent to the design of waste packages for geologic disposal of nuclear waste. The bibliography is divided into fourteen subject categories so that anyone wishing to review the subject of leaching, for example, can turn to the leaching section and review the abstracts of reports which are concerned primarily with leaching. Abstracts are also cross referenced according to secondary subject matter so that one can get a complete list of abstracts for any of the fourteen subject categories. All documents which by their title alone appear to deal with the design of waste packages for the geologic disposal of spent fuel or high-level waste were obtained and reviewed. Only those documents which truly appear to be of interest to a waste package designer were abstracted. The documents not abstracted are listed in a separate section. There was no beginning date for consideration of a document for review. About 1100 documents were reviewed and about 450 documents were abstracted.

Wurm, K.J.; Miller, N.E.

1982-11-01T23:59:59.000Z

167

Survey of waste package designs for disposal of high-level waste/spent fuel in selected foreign countries  

SciTech Connect

This report presents the results of a survey of the waste package strategies for seven western countries with active nuclear power programs that are pursuing disposal of spent nuclear fuel or high-level wastes in deep geologic rock formations. Information, current as of January 1989, is given on the leading waste package concepts for Belgium, Canada, France, Federal Republic of Germany, Sweden, Switzerland, and the United Kingdom. All but two of the countries surveyed (France and the UK) have developed design concepts for their repositories, but none of the countries has developed its final waste repository or package concept. Waste package concepts are under study in all the countries surveyed, except the UK. Most of the countries have not yet developed a reference concept and are considering several concepts. Most of the information presented in this report is for the current reference or leading concepts. All canisters for the wastes are cylindrical, and are made of metal (stainless steel, mild steel, titanium, or copper). The canister concepts have relatively thin walls, except those for spent fuel in Sweden and Germany. Diagrams are presented for the reference or leading concepts for canisters for the countries surveyed. The expected lifetimes of the conceptual canisters in their respective disposal environment are typically 500 to 1,000 years, with Sweden's copper canister expected to last as long as one million years. Overpack containers that would contain the canisters are being considered in some of the countries. All of the countries surveyed, except one (Germany) are currently planning to utilize a buffer material (typically bentonite) surrounding the disposal package in the repository. Most of the countries surveyed plan to limit the maximum temperature in the buffer material to about 100{degree}C. 52 refs., 9 figs.

Schneider, K.J.; Lakey, L.T.; Silviera, D.J.

1989-09-01T23:59:59.000Z

168

HIGH ALUMINUM HLW (HIGH LEVEL WASTE ) GLASSES FOR HANFORDS WTP (WASTE TREATMENT PROJECT)  

Science Conference Proceedings (OSTI)

This paper presents the results of glass formulation development and melter testing to identify high waste loading glasses to treat high-Al high level waste (HLW) at Hanford. Previous glass formulations developed for this HLW had high waste loadings but their processing rates were lower that desired. The present work was aimed at improving the glass processing rate while maintaining high waste loadings. Glass formulations were designed, prepared at crucible-scale and characterized to determine their properties relevant to processing and product quality. Glass formulations that met these requirements were screened for melt rates using small-scale tests. The small-scale melt rate screening included vertical gradient furnace (VGF) and direct feed consumption (DFC) melter tests. Based on the results of these tests, modified glass formulations were developed and selected for larger scale melter tests to determine their processing rate. Melter tests were conducted on the DuraMelter 100 (DMIOO) with a melt surface area of 0.11 m{sup 2} and the DuraMelter 1200 (DMI200) HLW Pilot Melter with a melt surface area of 1.2 m{sup 2}. The newly developed glass formulations had waste loadings as high as 50 wt%, with corresponding Al{sub 2}O{sub 3} concentration in the glass of 26.63 wt%. The new glass formulations showed glass production rates as high as 1900 kg/(m{sup 2}.day) under nominal melter operating conditions. The demonstrated glass production rates are much higher than the current requirement of 800 kg/(m{sup 2}.day) and anticipated future enhanced Hanford Tank Waste Treatment and Immobilization Plant (WTP) requirement of 1000 kg/(m{sup 2}.day).

KRUGER AA; BOWAN BW; JOSEPH I; GAN H; KOT WK; MATLACK KS; PEGG IL

2010-01-04T23:59:59.000Z

169

EIS-0023: Long-Term Management of Defense High-Level Radioactive Wastes  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

023: Long-Term Management of Defense High-Level Radioactive 023: Long-Term Management of Defense High-Level Radioactive Wastes (Research and Development Program for Immobilization) Savannah River Plant, Aiken, South Carolina EIS-0023: Long-Term Management of Defense High-Level Radioactive Wastes (Research and Development Program for Immobilization) Savannah River Plant, Aiken, South Carolina SUMMARY 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-level radioactive wastes resulting from chemical separations operations for defense radionuclides production at the DOE Savannah River Plant (SRP) near Aiken, South Carolina. PUBLIC COMMENT OPPORTUNITIES None available at this time.

170

Impact of Alkali Source on Vitrification of SRS High Level Waste  

SciTech Connect

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

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

2005-09-08T23:59:59.000Z

171

Progress in resolving Savannah River Site high-level waste tank safety issues  

SciTech Connect

At the Savannah River Site (SRS), near Aiken, South Carolina, approximately 35 million gallons of high-level radioactive waste are stored in 51 underground, carbon steel waste tanks. These tanks and associated facilities are distributed between the F and H areas, two processing areas at SRS, and are called the F- and H-area high-level waste tank farms. Within the last few years, issues have been raised about the safety of high-level waste tank farms throughout the DOE complex, including those at SRS. Plans for resolution of these issues were reported at the Waste Management 192 conference. This paper addresses progress made at SRS since 1992. Most of the efforts for resolving the six safety issues identified at SRS have concentrated on (1) preparing the tanks for waste removal and (2) completing construction, testing, and starting up three key facilities. These facilities will transform the waste into forms suitable for final disposal, specifically borosilicate glass and saltstone (grout). Removing the waste from the tanks and processing it is needed to resolve three of the safety issues. Two facilities -- In-Tank Precipitation and the Defense Waste Processing Facility -- are undergoing non-radioactive simulant testing (``cold runs``) at this time. The third facility -- Sludge Processing -- began testing with actual waste in October 1993. In Tank Precipitation is scheduled to be operating by the end of 1994.

d`Entremont, P.D.

1993-12-31T23:59:59.000Z

172

EPRI Review of Geologic Disposal for Used Fuel and High Level Radioactive Waste: Volume IV - Lessons Learned  

Science Conference Proceedings (OSTI)

The effective termination of the Yucca Mountain program by the U.S. Administration in 2009 has further delayed the construction and operation of a permanent disposal facility for used fuel and high level radioactive waste (HLW) in the United States. In concert with this decision, the President directed the Energy Secretary to establish the Blue Ribbon Commission on America's Nuclear Future to review and provide recommendations on options for managing used fuel and HLW. EPRI is uniquely positioned to prov...

2010-09-29T23:59:59.000Z

173

High-Level Waste Corporate Board, Dr. Inᅢᄅs Triay  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Office of Environmental Management Office of Environmental Management High-Level Waste Corporate Board April 1, 2008 safety v performance v cleanup v closure M E Environmental Management Environmental Management What Are Corporate Issues? * They usually occur at multiple sites * They usually have an impact that exceeds their initial point of application. Thus, they impact: - Policies - Planning - Standards & Guidance - EM's relationship with other agencies both internal and external to DOE safety v performance v cleanup v closure M E Environmental Management Environmental Management Current Corporate Issues * Performance Assessment * Quality Assurance * Methods to Determine the Waste Inventory * Chemical Processing * Waste Forms * Actual Disposition of Waste * Waste Treatment safety v

174

Alternatives Generation and Analysis for Phase 1 High Level Waste Feed Tanks Selection  

Science Conference Proceedings (OSTI)

A recent revision of the US. Department of Energy privatization contract for the immobilization of high-level waste (HLW) at Hanford necessitates the investigation of alternative waste feed sources to meet contractual feed requirements. This analysis identifies wastes to be considered as HLW feeds and develops and conducts alternative analyses to comply with established criteria. A total of 12,426 cases involving 72 waste streams are evaluated and ranked in three cost-based alternative models. Additional programmatic criteria are assessed against leading alternative options to yield an optimum blended waste feed stream.

CRAWFORD, T.W.

1999-08-16T23:59:59.000Z

175

Data acquisition and control system for the High-Level Waste Tank Farm at Hanford, Washington  

SciTech Connect

The High-Level Nuclear Waste Storage Tank 241-SY-101 periodically releases flammable gasses. Mitigation experiments to release the gasses continuously to avoid a catastrophic build-up are planned for FY93 and beyond. Los Alamos has provided a data acquisition and control system (DACS) to monitor and control mitigation experiments on SY-101. The DACS consists of a data acquisition trailer to house the electronic components and computers in a friendly environment, a computer system running process control software for monitoring and controlling the tests, signal conditioners to convert the instrument signals to a usable form for the DACS, programmable logic controllers to process sensor signals and take action quickly, a fast data acquisition system for recording transient data, and a remote monitoring system to monitor the progress of the experiment. Equipment to monitor the release of the gasses was also provided. The first experiment involves a mixer pump to mix the waste and allow the gasses to be released at the surface of the liquid as the gas is being formed. The initial tests are scheduled for July 1993.

Hoida, H.W.; Hatcher, C.R.; Trujillo, L.T.; Holt, D.H. [Los Alamos National Lab., NM (United States); Vargo, G.F. [Westinghouse Hanford Co., Richland, WA (United States); Martin, J.; Stastny, G.; Echave, R. [EG and G Energy Measurements, Inc., Las Vegas, NV (United States); Eldridge, K. [Eldridge Engineering, Spokane, WA (United States)

1993-08-01T23:59:59.000Z

176

Supplement Analysis for the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement  

Science Conference Proceedings (OSTI)

In October 2002, DOE issued the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (Final EIS) (DOE 2002) that provided an analysis of the potential environmental consequences of alternatives/options for the management and disposition of Sodium Bearing Waste (SBW), High-Level Waste (HL W) calcine, and HLW facilities at the Idaho Nuclear Technology and Engineering Center (INTEC) located at the Idaho National Engineering and Environmental Laboratory (INEEL), now known as the Idaho National Laboratory (INL) and referred to hereafter as the Idaho Site. Subsequent to the issuance of the Final EIS, DOE included the requirement for treatment of SBW in the Request for Proposals for Environmental Management activities on the Idaho Site. The new Idaho Cleanup Project (ICP) Contractor identified Steam Reforming as their proposed method to treat SBW; a method analyzed in the Final EIS as an option to treat SBW. The proposed Steam Reforming process for SBW is the same as in the Final EIS for retrieval, treatment process, waste form and transportation for disposal. In addition, DOE has updated the characterization data for both the HLW Calcine (BBWI 2005a) and SBW (BBWI 2004 and BBWI 2005b) and identified two areas where new calculation methods are being used to determine health and safety impacts. Because of those changes, DOE has prepared this supplement analysis to determine whether there are ''substantial changes in the proposed action that are relevant to environmental concerns'' or ''significant new circumstances or information'' within the meaning of the Council of Environmental Quality and DOE National Environmental Policy Act (NEPA) Regulations (40 CFR 1502.9 (c) and 10 CFR 1021.314) that would require preparation of a Supplemental EIS. Specifically, this analysis is intended to determine if: (1) the Steam Reforming Option identified in the Final EIS adequately bounds impacts from the Steam Reforming Process proposed by the new ICP Contractor using the new characterization data, (2) the new characterization data is significantly different than the data presented in the Final EIS, (3) the new calculation methods present a significant change to the impacts described in the Final EIS, and (4) would the updated characterization data cause significant changes in the environmental impacts for the action alternatives/options presented in the Final EIS. There are no other aspects of the Final EIS that require additional review because DOE has not identified any additional new significant circumstances or information that would warrant such a review.

N /A

2005-06-30T23:59:59.000Z

177

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

NLE Websites -- All DOE Office Websites (Extended Search)

HLW is the highly radioactive material resulting from the reprocessing of SNF. Under the Nuclear Waste Policy Act of 1982, the federal government is responsible for the disposal...

178

US Department of Energy Storage of Spent Fuel and High Level Waste  

DOE Green Energy (OSTI)

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.

Sandra M Birk

2010-10-01T23:59:59.000Z

179

DEPARTMENT OF ENERGY Disposal of Hanford Defense High-Level, Transuranic, and Tank Wastes, Hanford  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Disposal of Hanford Defense High-Level, Transuranic, and Tank Wastes, Hanford Disposal of Hanford Defense High-Level, Transuranic, and Tank Wastes, Hanford Site, Richland, Washington; Record of Decision (ROO). This Record of Decision has been prepared pursuant to the Council on Environme~tal Quality ~egulations for Implementing the Procedural Provisions of the National Environmental Pol icy Act (NEPAl (40 CFR Parts 1500-1508) and the Department of Energy NEPA Guidelines (52 FR 47662, December 15, 1987). It is based on DOE's "Environmental Impact Statement for the Oi sposal of Hanford Defense High-Level, Transuranic, and Tank Wastes'' (OOE/EIS-0113) and consideration of ~11 public and agency comments received on the Environmental Impact Statement (EIS). fJECISION The decision is to implement the ''Preferred Alternative'' as discussed in

180

Separating and Stabilizing Phosphate from High-Level Radioactive Waste: Process Development and Spectroscopic Monitoring  

SciTech Connect

Removing phosphate from alkaline high-level waste sludges at the Department of Energy's Hanford Site in Washington State is necessary to increase the waste loading in the borosilicate glass waste form that will be used to immobilize the highly radioactive fraction of these wastes. We are developing a process which first leaches phosphate from the high-level waste solids with aqueous sodium hydroxide, and then isolates the phosphate by precipitation with calcium oxide. Tests with actual tank waste confirmed that this process is an effective method of phosphate removal from the sludge and offers an additional option for managing the phosphorus in the Hanford tank waste solids. The presence of vibrationally active species, such as nitrate and phosphate ions, in the tank waste processing streams makes the phosphate removal process an ideal candidate for monitoring by Raman or infrared spectroscopic means. As a proof-of-principle demonstration, Raman and Fourier transform infrared (FTIR) spectra were acquired for all phases during a test of the process with actual tank waste. Quantitative determination of phosphate, nitrate, and sulfate in the liquid phases was achieved by Raman spectroscopy, demonstrating the applicability of Raman spectroscopy for the monitoring of these species in the tank waste process streams.

Lumetta, Gregg J.; Braley, Jenifer C.; Peterson, James M.; Bryan, Samuel A.; Levitskaia, Tatiana G.

2012-05-09T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


181

Radioactive waste from transmutation of technetium: a model for anticipating characteristics of high level waste from transmutation  

SciTech Connect

At this early stage in the conceptualization of fuel treatment and radioisotope transmutation for the disposition of nuclear wastes, it is possible to anticipate some characteristics of the waste stream resulting from the deployment of advanced technologies. Fission products and actinides cannot be completely destroyed by transmutation even with continuous purification and recycle. This is demonstrated for technetium in this analysis, but is true for all radioisotopes. Also, some of the reaction products are themselves long-lived radioactive isotopes. The purification and recycle steps produce nuclear wastes that must be planned for geologic disposal. Five radioisotopes have been identified to be produced in abundance by transmutation of technetium using fast neutrons. Four of these isotopes may be more benign than the original technetium-99 because of their longer half lives. However, one isotope, molybdenum-93 with a half life of four thousand years, may be troublesome. All of the isotopes arising from the transmutation process that end up in high level waste must be examined in terms of their behavior in geologic disposal. In selecting goals for chemical separations, the technologists must consider the entire cycle of separation and transmutation before applying the performance expected in a single separation to implications concerning a repository. A separation efficiency of 0.95 can translate into the disposal of as much as 30 to 60 percent of the technetium in the repository if down stream losses are not controlled. In this case, the treatment may have little impact on anticipated off site radiation from technetium. The destruction of technetium through continuous recycle requires the cost of increased neutron dose and increased space in reactors that must be considered in design of fuel treatment systems. (authors)

Seitz, M.G. [Booz Allen Hamilton, Washington DC (United States)

2007-07-01T23:59:59.000Z

182

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

Science Conference Proceedings (OSTI)

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.

R.A. Levich; J.S. Stuckless

2006-09-25T23:59:59.000Z

183

Solvent extraction in the treatment of acidic high-level liquid waste : where do we stand?  

SciTech Connect

During the last 15 years, a number of solvent extraction/recovery processes have been developed for the removal of the transuranic elements, {sup 90}Sr and {sup 137}Cs from acidic high-level liquid waste. These processes are based on the use of a variety of both acidic and neutral extractants. This chapter will present an overview and analysis of the various extractants and flowsheets developed to treat acidic high-level liquid waste streams. The advantages and disadvantages of each extractant along with comparisons of the individual systems are discussed.

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

1998-06-18T23:59:59.000Z

184

Implementation of seismic design and evaluation guidelines for the Department of Energy high-level waste storage tanks and appurtenances  

SciTech Connect

In the fall of 1992, a draft of the Seismic Design and Evaluation Guidelines for the Department of Energy (DOE) High-level Waste Storage Tanks and Appurtenances was issued. The guidelines were prepared by the Tanks Seismic Experts Panel (TSEP) and this task was sponsored by DOE, Environmental Management. The TSEP is comprised of a number of consultants known for their knowledge of seismic ground motion and expertise in the analysis of structures, systems and components subjected to seismic loads. The development of these guidelines was managed by staff from Brookhaven National Laboratory, Engineering Research and Applications Division, Department of Nuclear Energy. This paper describes the process used to incorporate the Seismic Design and Evaluation Guidelines for the DOE High-Level Waste Storage Tanks and Appurtenances into the design criteria for the Multi-Function Waste Tank Project at the Hanford Site. This project will design and construct six new high-level waste tanks in the 200 Areas at the Hanford Site. This paper also discusses the vehicles used to ensure compliance to these guidelines throughout Title 1 and Title 2 design phases of the project as well as the strategy used to ensure consistent and cost-effective application of the guidelines by the structural analysts. The paper includes lessons learned and provides recommendations for other tank design projects which might employ the TSEP guidelines.

Conrads, T.J.

1993-06-01T23:59:59.000Z

185

Nuclear waste solidification  

DOE Patents (OSTI)

High level liquid waste solidification is achieved on a continuous basis by atomizing the liquid waste and introducing the atomized liquid waste into a reaction chamber including a fluidized, heated inert bed to effect calcination of the atomized waste and removal of the calcined waste by overflow removal and by attrition and elutriation from the reaction chamber, and feeding additional inert bed particles to the fluidized bed to maintain the inert bed composition.

Bjorklund, William J. (Richland, WA)

1977-01-01T23:59:59.000Z

186

Alternatives generation and analysis for the phase 1 high-level waste pretreatment process selection  

Science Conference Proceedings (OSTI)

This report evaluates the effects of enhanced sludge washing and sludge washing without caustic leaching during the preparation of the Phase 1 high-level waste feeds. The pretreatment processing alternatives are evaluated against their ability to satisfy contractual, cost minimization, and other criteria. The information contained in this report is consistent with, and supplemental to, the Tank Waste Remediation System Operation and Utilization Plan (Kirkbride et al. 1997).

Manuel, A.F.

1997-10-02T23:59:59.000Z

187

Thermal analysis of Yucca Mountain commercial high-level waste packages  

Science Conference Proceedings (OSTI)

The thermal performance of commercial high-level waste packages was evaluated on a preliminary basis for the candidate Yucca Mountain repository site. The purpose of this study is to provide an estimate for waste package component temperatures as a function of isolation time in tuff. Several recommendations are made concerning the additional information and modeling needed to evaluate the thermal performance of the Yucca Mountain repository system.

Altenhofen, M.K. [Altenhofen (M.K.), Richland, WA (United States); Eslinger, P.W. [Pacific Northwest Lab., Richland, WA (United States)

1992-10-01T23:59:59.000Z

188

Automated Sampling and Sample Pneumatic Transport of High Level Tank Wastes at the Hanford Waste Treatment Plant  

Science Conference Proceedings (OSTI)

This paper describes the development work, and design and engineering tasks performed, to provide a fully automated sampling system for the Waste Treatment Plant (WTP) project at the Hanford Site in southeastern Washington State, USA. WTP is being built to enable the emptying and immobilization of highly active waste resulting from processing of irradiated nuclear fuel since the 1940's. The Hanford Tank Wastes are separated into Highly Level Waste (HLW), and Low Active Waste (LAW) fractions, which are separately immobilized by vitrification into borosilicate glass. Liquid samples must be taken of the waste and Glass Forming Chemicals (GFCs) before vitrification, and analyzed to insure the glass products will comply with specifications established in the WTP contract. This paper describes the non-radioactive testing of the sampling of the HLW and LAW melter feed simulants that was performed ahead of final equipment design. These trials were essential to demonstrate the effectiveness and repeatability of the integrated sampling system to collect representative samples, free of cross-contamination. Based on existing tried and proven equipment, the system design is tailored to meet the WTP project's specific needs. The design provides sampling capabilities from 47 separate sampling points and includes a pneumatic transport system to move the samples from the 3 separate facilities to the centralized analytical laboratory. The physical and rheological compositions of the waste simulants provided additional challenges in terms of the sample delivery, homogenization, and sample capture equipment design requirements. The activity levels of the actual waste forms, specified as 486 E9 Bq/liter (Cs-137), 1.92 E9 Bq/liter (Co-60), and 9.67 E9 Bq/liter (Eu-154), influenced the degree of automation provided, and justified the minimization of manual intervention needed to obtain and deliver samples from the process facilities to the analytical laboratories. Maintaining high integrity primary and secondary confinement, including during the cross-site transportation of the samples, is a key requirement that is achieved and assured at all times. (authors)

Phillips, C.; Richardson, J. E. [BNG America, 2345 Stevens Drive, Richland, WA, 99354 (United States)

2006-07-01T23:59:59.000Z

189

Nuclear Waste Management  

NLE Websites -- All DOE Office Websites (Extended Search)

Waste Management's Yucca Mountain Project and the Office of Nuclear Energy's Advanced Fuel Cycle Initiative (AFCI) and Global Nuclear Energy Partnership (GNEP) programs. Efforts...

190

Glass Property Data and Models for Estimating High-Level Waste Glass Volume  

SciTech Connect

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.

Vienna, John D.; Fluegel, Alexander; Kim, Dong-Sang; Hrma, Pavel R.

2009-10-05T23:59:59.000Z

191

Design of equipment used for high-level waste vitrification at the West Valley Demonstration Project  

SciTech Connect

The equipment as designed, started, and operated for high-level radioactive waste vitrification at the West Valley Demonstration Project in western New York State is described. Equipment for the processes of melter feed make-up, vitrification, canister handling, and off-gas treatment are included. For each item of equipment the functional requirements, process description, and hardware descriptions are presented.

Vance, R.F.; Brill, B.A.; Carl, D.E. [and others

1997-06-01T23:59:59.000Z

192

Structural integrity and potential failure modes of hanford high-level waste tanks  

Science Conference Proceedings (OSTI)

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.

Han, F.C.

1996-09-30T23:59:59.000Z

193

Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)  

SciTech Connect

The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

Burgard, K.C.

1998-04-09T23:59:59.000Z

194

Conceptual design report for immobilized high-level waste interim storage facility (Phase 1)  

SciTech Connect

The Hanford Site Canister Storage Building (CSB Bldg. 212H) will be utilized to interim store Phase 1 HLW products. Project W-464, Immobilized High-Level Waste Interim Storage, will procure an onsite transportation system and retrofit the CSB to accommodate the Phase 1 HLW products. The Conceptual Design Report establishes the Project W-464 technical and cost basis.

Burgard, K.C.

1998-06-02T23:59:59.000Z

195

A One System Integrated Approach to Simulant Selection for Hanford High Level Waste Mixing and Sampling Tests  

SciTech Connect

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 capabilities using simulated Hanford High-Level Waste (HLW) formulations. This represents one of the largest remaining technical issues with the high-level waste treatment mission at Hanford. Previous testing has focused on very specific TOC or WTP test objectives and consequently the simulants were narrowly focused on those test needs. A key attribute in the Defense Nuclear Facilities Safety Board (DNFSB) Recommendation 2010-2 is to ensure testing is performed with a simulant that represents the broad spectrum of Hanford waste. The One System Integrated Project Team is a new joint TOC and WTP organization intended to ensure technical integration of specific TOC and WTP systems and testing. A new approach to simulant definition has been mutually developed that will meet both TOC and WTP test objectives for the delivery and receipt of HLW. The process used to identify critical simulant characteristics, incorporate lessons learned from previous testing, and identify specific simulant targets that ensure TOC and WTP testing addresses the broad spectrum of Hanford waste characteristics that are important to mixing, sampling, and transfer performance are described.

Thien, Mike G. [Washington River Protection Solutions, LLC, Richland, WA (United States); Barnes, Steve M. [URS, Richland, WA (United States)

2013-01-17T23:59:59.000Z

196

Separation of strontium-90 from Hanford high-level radioactive waste  

SciTech Connect

Current guidelines for disposing of high-level radioactive wastes stored in underground tanks at the US Department of Energy`s Hanford Site call for vitrifying high-level waste (HLW) in borosilicate glass and disposing the glass canisters in a deep geologic repository. Disposition of the low-level waste (LLW) is yet to be determined, but it will likely be immobilized in a glass matrix and disposed of on site. To lower the radiological risk associated with the LLW form, methods are being developed to separate {sup 90}Sr from the bulk waste material so this isotope can be routed to the HLW stream. A solvent extraction method is being investigated to separate {sup 90}Sr from acid-dissolved Hanford tank wastes. Results of experiments with actual tank waste indicate that this method can be used to achieve separation of {sup 90}Sr from the bulk waste components. Greater than 99% of the {sup 90}Sr was removed from an acidic dissolved sludge solution by extraction with di-tbutylcyclohexano-18-crown-6 in 1-octanol (the SREX process). The major sludge components were not extracted.

Lumetta, G.J.; Wagner, M.J.; Jones, E.O.

1993-10-01T23:59:59.000Z

197

Progress in resolving Hanford Site high-level waste tank safety issues  

DOE Green Energy (OSTI)

Interim storage of alkaline, high-level radioactive waste, from two generations of spent fuel reprocessing and waste management activities, has resulted in the accumulation of 238 million liters of waste in Hanford Site single and double-shell tanks. Before the 1990`s, the stored waste was believed to be: (1) chemically unreactive under its existing storage conditions and plausible accident scenarios; and (2) chemically stable. This paradigm was proven incorrect when detailed evaluation of tank contents and behavior revealed a number of safety issues and that the waste was generating flammable and noxious gases. In 1990, the Waste Tank Safety Program was formed to focus on identifying safety issues and resolving the ferrocyanide, flammable gas, organic, high heat, noxious vapor, and criticality issues. The tanks of concern were placed on Watch Lists by safety issue. This paper summarizes recent progress toward resolving Hanford Site high-level radioactive waste tank safety issues, including modeling, and analyses, laboratory experiments, monitoring upgrades, mitigation equipment, and developing a strategy to screen tanks for safety issues.

Babad, H.; Eberlein, S.J.; Johnson, G.D.; Meacham, J.E.; Osborne, J.W.; Payne, M.A.; Turner, D.A.

1995-02-01T23:59:59.000Z

198

THE RETRIEVAL KNOWLEDGE CENTER EVALUATION OF LOW TANK LEVEL MIXING TECHNOLOGIES FOR DOE HIGH LEVEL WASTE TANK RETRIEVAL 10516  

Science Conference Proceedings (OSTI)

The Department of Energy (DOE) Complex has over two-hundred underground storage tanks containing over 80-million gallons of legacy waste from the production of nuclear weapons. The majority of the waste is located at four major sites across the nation and is planned for treatment over a period of almost forty years. The DOE Office of Technology Innovation & Development within the Office of Environmental Management (DOE-EM) sponsors technology research and development programs to support processing advancements and technology maturation designed to improve the costs and schedule for disposal of the waste and closure of the tanks. Within the waste processing focus area are numerous technical initiatives which included the development of a suite of waste removal technologies to address the need for proven equipment and techniques to remove high level radioactive wastes from the waste tanks that are now over fifty years old. In an effort to enhance the efficiency of waste retrieval operations, the DOE-EM Office of Technology Innovation & Development funded an effort to improve communications and information sharing between the DOE's major waste tank locations as it relates to retrieval. The task, dubbed the Retrieval Knowledge Center (RKC) was co-lead by the Savannah River National Laboratory (SRNL) and the Pacific Northwest National Laboratory (PNNL) with core team members representing the Oak Ridge and Idaho sites, as well as, site contractors responsible for waste tank operations. One of the greatest challenges to the processing and closure of many of the tanks is complete removal of all tank contents. Sizeable challenges exist for retrieving waste from High Level Waste (HLW) tanks; with complications that are not normally found with tank retrieval in commercial applications. Technologies currently in use for waste retrieval are generally adequate for bulk removal; however, removal of tank heels, the materials settled in the bottom of the tank, using the same technology have proven to be difficult. Through the RKC, DOE-EM funded an evaluation of adaptable commercial technologies that could assist with the removal of the tank heels. This paper will discuss the efforts and results of developing the RKC to improve communications and discussion of tank waste retrieval through a series of meetings designed to identify technical gaps in retrieval technologies at the DOE Hanford and Savannah River Sites. This paper will also describe the results of an evaluation of commercially available technologies for low level mixing as they might apply to HLW tank heel retrievals.

Fellinger, A.

2009-12-08T23:59:59.000Z

199

I-NERI-2007-004-K, DEVELOPMENT AND CHARACTERIZATION OF NEW HIGH-LEVEL WASTE FORMS FOR ACHIEVING WASTE MINIMIZATION FROM PYROPROCESSING  

SciTech Connect

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.

S.M. Frank

2011-09-01T23:59:59.000Z

200

A COMPARISON OF HANFORD AND SAVANNAH RIVER SITE HIGH-LEVEL WASTES  

SciTech Connect

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.

HILL RC PHILIP; REYNOLDS JG; RUTLAND PL

2011-02-23T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


201

Characterization and reaction behavior of ferrocyanide simulants and Hanford Site high-level ferrocyanide waste  

Science Conference Proceedings (OSTI)

Nonradioactive waste simulants and initial ferrocyanide tank waste samples were characterized to assess potential safety concerns associated with ferrocyanide high-level radioactive waste stored at the Hanford Site in underground single-shell tanks (SSTs). Chemical, physical, thermodynamic, and reaction properties of the waste simulants were determined and compared to properties of initial samples of actual ferrocyanide wastes presently in the tanks. The simulants were shown to not support propagating reactions when subjected to a strong ignition source. The simulant with the greatest ferrocyanide concentration was shown to not support a propagating reaction that would involve surrounding waste because of its high water content. Evaluation of dried simulants indicated a concentration limit of about 14 wt% disodium mononickel ferrocyanide, below which propagating reactions could not occur in the ambient temperature bulk tank waste. For postulated localized hot spots where dried waste is postulated to be at an initial temperature of 130 C, a concentration limit of about 13 wt% disodium mononickel ferrocyanide was determined, below which propagating reactions could not occur. Analyses of initial samples of the presently stored ferrocyanide waste indicate that the waste tank ferrocyanide concentrations are considerably lower than the limit for propagation for dry waste and that the water content is near that of the as-prepared simulants. If the initial trend continues, it will be possible to show that runaway ferrocyanide reactions are not possible under present tank conditions. The lower ferrocyanide concentrations in actual tank waste may be due to tank waste mixing and/or degradation from radiolysis and/or hydrolysis, which may have occurred over approximately 35 years of storage.

Jeppson, D.W.; Simpson, B.C.

1994-02-01T23:59:59.000Z

202

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

SciTech Connect

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.

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

2011-10-01T23:59:59.000Z

203

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

SciTech Connect

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.

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

1999-01-01T23:59:59.000Z

204

An Istrument for Measuring the TRU Concentration in High-Level Liquid Waste  

Science Conference Proceedings (OSTI)

An online monitor has been designed, built, and tested, which is capable of measuring the residual transuranic concentrations in processed high-level wastes with a detection limit of 370 Bq/ml (10 nCi/ml) in less than six hours. The monitor measures the neutrons produced by the transuranics, primarily via (?,n) reactions, in the presence of gamma-ray fields up to 1 Sv/h (100 R/h). The optimum design was determined by Monte Carlo modeling and then tempered with practical engineering and cost considerations. Correct operation of the monitor was demonstrated in a hot cell utilizing an actual sample of high-level waste. Results of that demonstration are given, and suggestions for improvements in the next generation system are discussed.

Brodzinski, Ronald L.; Craig, R. A.; Fink, Samuel D.; Hensley, Walter K.; Holt, Noah O.; Knopf, Michael A.; Lepel, Elwood A.; Mullen, O Dennis; Salaymeh, Saleem R.; Samuel, Todd J.; Smart, John E.; Tinker, Michael R.; Walker, Darrell D.

2005-02-01T23:59:59.000Z

205

High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 3  

Science Conference Proceedings (OSTI)

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. Volume III contains a bibliography of glass corrosion studies, including studies that are not cited in Volumes I and II.

Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)] [and others

1994-03-01T23:59:59.000Z

206

High-level waste borosilicate glass: A compendium of corrosion characteristics. Volume 2  

Science Conference Proceedings (OSTI)

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.

Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)] [and others

1994-03-01T23:59:59.000Z

207

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

Science Conference Proceedings (OSTI)

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.

Larson, D.E. (ed.)

1980-09-01T23:59:59.000Z

208

Nuclear Waste Policy Act.doc  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Civilian Radioactive Civilian Radioactive Waste Management Washington, D.C. 20585 March 2004 i THE NUCLEAR WASTE POLICY ACT OF 1982 1 An Act to provide for the development of repositories for the disposal of high-level radioactive waste and spent nuclear fuel, to establish a program of research, development, and demonstration regarding the disposal of high-level radioactive waste and spent nuclear fuel, and for other purposes. Be it enacted by the Senate and House of Representatives of the United States of America in Congress assembled, SHORT TITLE AND TABLE OF CONTENTS Section 1. This Act may be cited as the "Nuclear Waste Policy Act of 1982". Sec. 1. Short title and table of contents...........................................................................i

209

RESORCINOL-FORMALDEHYDE ION EXCHANGE RESIN CHEMISTRY FOR HIGH LEVEL WASTE TREATMENT  

Science Conference Proceedings (OSTI)

A principal goal at the Savannah River Site is to safely dispose of the large volume of liquid nuclear waste held in many storage tanks. In-tank ion exchange technology is being considered for cesium removal using a polymer resin made of resorcinol formaldehyde that has been engineered into microspheres. The waste under study is generally lower in potassium and organic components than Hanford waste; therefore, the resin performance was evaluated with actual dissolved salt waste. The ion exchange performance and resin chemistry results are discussed.

Nash, C.; Duignan, M.

2010-01-14T23:59:59.000Z

210

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

Science Conference Proceedings (OSTI)

Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of high-level radioactive-waste disposal containers. The waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The copper-based alloy materials are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The austenitic materials are Types 304L and 316L stainless steels and Alloy 825. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr, and they must be retrievable from the disposal site during the first 50 yr after emplacement. The containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This volume surveys the available data on the phase stability of both groups of candidate alloys. The austenitic alloys are reviewed in terms of the physical metallurgy of the iron-chromium-nickel system, martensite transformations, carbide formation, and intermetallic-phase precipitation. The copper-based alloys are reviewed in terms of their phase equilibria and the possibility of precipitation of the minor alloying constituents. For the austenitic materials, the ranking based on phase stability is: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is: CDA 102 (oxygen-free copper) (best), and then both CDA 715 and CDA 613. 75 refs., 24 figs., 6 tabs.

Bullen, D.B.; Gdowski, G.E. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

1988-08-01T23:59:59.000Z

211

Savannah River Site high-level waste safety issues: The need for final disposal of the wastes  

DOE Green Energy (OSTI)

Using new criteria developed by the High-Level Waste Tank Safety Task Force, the Savannah River Site (SRS) identified six safety issues in the SRS tank farms. None of the safety issues were priority 1, the most significant issues handled by the Task Force. This paper discusses the safety issues and the programs for resolving each of them.

d`Entremont, P.D.; Hobbs, D.T.

1991-12-31T23:59:59.000Z

212

Savannah River Site high-level waste safety issues: The need for final disposal of the wastes  

DOE Green Energy (OSTI)

Using new criteria developed by the High-Level Waste Tank Safety Task Force, the Savannah River Site (SRS) identified six safety issues in the SRS tank farms. None of the safety issues were priority 1, the most significant issues handled by the Task Force. This paper discusses the safety issues and the programs for resolving each of them.

d'Entremont, P.D.; Hobbs, D.T.

1991-01-01T23:59:59.000Z

213

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

DOE Green Energy (OSTI)

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.

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

2009-10-01T23:59:59.000Z

214

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

Science Conference Proceedings (OSTI)

Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The 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, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-level waste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials (CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)), which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs.

Gdowski, G.E.; Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

1988-08-01T23:59:59.000Z

215

Lesson 7 - Waste from Nuclear Power Plants | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

7 - Waste from Nuclear Power Plants 7 - Waste from Nuclear Power Plants Lesson 7 - Waste from Nuclear Power Plants This lesson takes a look at the waste from electricity production at nuclear power plants. It considers the different types of waste generated, as well as how we deal with each type of waste. Specific topics covered include: Nuclear Waste Some radioactive Types of radioactive waste Low-level waste High-level waste Disposal and storage Low-level waste disposal Spent fuel storage Waste isolation Reprocessing Decommissioning Lesson 7 - Waste.pptx More Documents & Publications National Report Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management Third National Report for the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management

216

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

Science Conference Proceedings (OSTI)

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.

Cunnane, J.C. [comp.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States)

1994-03-01T23:59:59.000Z

217

Nuclear Waste Policy Act Signed | National Nuclear Security Administra...  

National Nuclear Security Administration (NNSA)

> Nuclear Waste Policy Act Signed Nuclear Waste Policy Act Signed January 07, 1983 Washington, DC Nuclear Waste Policy Act Signed President Reagan signs the Nuclear Waste...

218

Accelerator-driven transmutation of high-level waste from the defense and commercial sectors  

SciTech Connect

This is the final report of a three-year, Laboratory-Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The major goal has been to develop accelerator transmutation of waste (ATW) system designs that will thoroughly and rapidly transmute nuclear waste, including plutonium from dismantled weapons and spent reactor fuel, while generating useful electrical power and without producing a long-lived radioactive waste stream. We have identified and quantified the unique qualities of subcritical nuclear systems and their capabilities in bringing about the complete destruction of plutonium. Although the 1191 subcritical systems involved in our most effective designs radically depart from traditional nuclear reactor concepts, they are based on extrapolations of existing technologies. Overall, care was taken to retain the highly desired features that nuclear technology has developed over the years within a conservative design envelope. We believe that the ATW systems designed in this project will enable almost complete destruction of nuclear waste (conversion to stable species) at a faster rate and without many of the safety concerns associated with the possible reactor approaches.

Bowman, C.; Arthur, E.; Beard, C. [and others

1996-09-01T23:59:59.000Z

219

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

Science Conference Proceedings (OSTI)

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.

Ross, B. (ed.)

1989-02-01T23:59:59.000Z

220

RADIOACTIVE HIGH LEVEL WASTE TANK PITTING PREDICTIONS: AN INVESTIGATION INTO CRITICAL SOLUTION CONCENTRATIONS  

Science Conference Proceedings (OSTI)

A series of cyclic potentiodynamic polarization tests was performed on samples of ASTM A537 carbon steel in support of a probability-based approach to evaluate the effect of chloride and sulfate on corrosion the steel?s susceptibility to pitting corrosion. Testing solutions were chosen to systemically evaluate the influence of the secondary aggressive species, chloride, and sulfate, in the nitrate based, high-level wastes. The results suggest that evaluating the combined effect of all aggressive species, nitrate, chloride, and sulfate, provides a consistent response for determining corrosion susceptibility. The results of this work emphasize the importance for not only nitrate concentration limits, but also chloride and sulfate concentration limits.

Hoffman, E.

2012-11-08T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

Collaboration, Automation, and Information Management at Hanford High Level Radioactive Waste (HLW) Tank Farms - 14210  

Science Conference Proceedings (OSTI)

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.

Aurah, Mirwaise Y.; Roberts, Mark A.

2013-12-12T23:59:59.000Z

222

DOE management of high-level waste at the Hanford Site  

SciTech Connect

Approximately 60 million gallons of high-level radioactive waste--caustic liquids, slurries, saltcakes, and sludges--are stored in underground tanks at the Department of Energy`s Hanford Site. At least one-third of the tanks are known to have leaked waste into the enviroranent, and there are many unresolved tank safety issues. In order to resolve the environmental and safety concerns, the Department plans to retrieve the waste, immobilize it, and dispose of it in a permanent geologic repository. Processing all of the tank waste in this manner could cost $40 billion, including $1.2 billion to construct the Hanford Waste Vitrification Plant. The purpose of our audit was to examine the reasons for cost estimate increases and schedule delays on the Hanford vitrification program. We also wanted to report on outstanding technical, safety, and environmental issues that could make the project even more costly and further delay its completion. We found that the Department managed the Hanford remediation system as a number of separate projects not fully integrated into one major system acquisition. Total costs have, therefore, been obscured, and the Department has not yet clearly defined system requirements or developed overall cost and schedule baselines. This lack of visibility could result in additional cost growth and schedule delays. We also noted a vast array of technical uncertainties, including tank safety and inadequate information about the makeup of tank waste, that could significantly affect the program`s cost and ultimate success. To increase visibility of program cost and schedule, we are recommending that all separate projects relating to tank waste be included in a single major system acquisition, and that the Department complete its ongoing baselining effort to the extent practical before making major funding commitments. Management concurred with our finding and recommendations.

Not Available

1993-04-14T23:59:59.000Z

223

Statement of work for the immobilized high-level waste transportation system, Project W-464  

SciTech Connect

The objective of this Statement of Work (SOW) is to present the scope, the deliverables, the organization, the technical and schedule expectations for the development of a Package Design Criteria (PDC), cost and schedule estimate for the acquisition of a transportation system for the Immobilized High-Level Waste (IHLW). This transportation system which includes the truck, the trailer, and a shielded cask will be used for on-site transportation of the IHLW canisters from the private vendor vitrification facility to the Hanford Site interim storage facility, i.e., vaults 2 and 3 of the Canister Storage Building (CSB). This Statement of Work asks Waste Management Federal Services, Inc., Northwest Operations, to provide Project W-464 with a Design Criteria Document, plus a life-cycle schedule and cost estimate for the acquisition of a transportation system (shielded cask, truck, trailer) for IHLW on-site transportation.

Mouette, P.

1998-06-24T23:59:59.000Z

224

Preliminary Waste Form Compliance Plan for the Idaho National Engineering and Environmental Laboratory High-Level Waste  

SciTech Connect

The Department of Energy (DOE) has specific technical and documentation requirements for high-level waste (HLW) that is to be placed in a federal repository. This document describes in general terms the strategy to be used at the Idaho National Engineering and Environmental Laboratory (INEEL) to demonstrate that vitrified HLW, if produced at the INEEL, meets these requirements. Waste form, canister, quality assurance, and documentation specifications are discussed. Compliance strategy is given, followed by an overview of how this strategy would be implemented for each specification.

B. A. Staples; T. P. O' Holleran

1999-05-01T23:59:59.000Z

225

Comparison of borosilicate glass and synthetic minerals as media for the immobilization of high-level radioactive waste  

Science Conference Proceedings (OSTI)

In this paper, the structure and properties of the different solid forms currently being developed for high-level radioactive waste disposal are compared. Good capacity to accept all the elements in the waste and flexibility of composition range to accommodate variations in the waste, are primarily discussed. 13 refs.

Tempest, P.A.

1981-03-01T23:59:59.000Z

226

Chapter 19 - Nuclear Waste Fund  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Waste Fund 19-1 Nuclear Waste Fund 19-1 CHAPTER 19 NUCLEAR WASTE FUND 1. INTRODUCTION. a. Purpose. This chapter establishes the financial, accounting, and budget policies and procedures for civilian and defense nuclear waste activities, as authorized in Public Law 97-425, the Nuclear Waste Policy Act, as amended, referred to hereafter as the Act. b. Applicability. This chapter applies to all Departmental elements, including the National Nuclear Security Administration, and activities that are funded by the Nuclear Waste Fund (NWF) or the Defense Nuclear Waste Disposal appropriation. c. Background. The Act established the Office of Civilian Radioactive Waste Management (OCRWM) and assigned it responsibility for the management

227

The Nuclear Waste Policy Act, as amended with appropriations acts appended  

Science Conference Proceedings (OSTI)

The Nuclear Waste Policy Act of 1982 provides for the development of repositories for the disposal of high-level radioactive waste and spent nuclear fuel, to establish a program of research, development and demonstration regarding the disposal of high-level radioactive waste and spent nuclear fuel. Titles 1 and 2 cover these subjects. Also included in this Act are: Title 3: Other provisions relating to radioactive waste; Title 4: Nuclear waste negotiation; Title 5: Nuclear waste technical review board; and Title 6: High-level radioactive waste. An appendix contains excerpts from appropriations acts from fiscal year 1984--1994.

Not Available

1994-03-01T23:59:59.000Z

228

Nuclear waste management. Quarterly progress report, October-December 1979  

SciTech Connect

Progress and activities are reported on the following: high-level waste immobilization, alternative waste forms, nuclear waste materials characterization, TRU waste immobilization programs, TRU waste decontamination, krypton solidification, thermal outgassing, iodine-129 fixation, monitoring of unsaturated zone transport, well-logging instrumentation development, mobile organic complexes of fission products, waste management system and safety studies, assessment of effectiveness of geologic isolation systems, waste/rock interactions technology, spent fuel and fuel pool integrity program, and engineered barriers. (DLC)

Platt, A.M.; Powell, J.A. (comps.)

1980-04-01T23:59:59.000Z

229

Nuclear Waste Programs semiannual progress report, April--September 1992  

Science Conference Proceedings (OSTI)

This document reports on the work done by the Nuclear Waste Programs of the Chemical Technology Division (CMT), Argonne National Laboratory, in the period April--September 1992. In these programs, studies are underway on the performance of waste glass and spent fuel in projected nuclear repository conditions to provide input to the licensing of the nation`s high-level waste repositories.

Bates, J.K.; Bradley, C.R.; Buck, E.C. [and others

1994-05-01T23:59:59.000Z

230

Nuclear waste programs; Semiannual progress report, October 1991--March 1992  

SciTech Connect

This document reports on the work done by the Nuclear Waste Programs of the Chemical Technology Division (CMT), Argonne National Laboratory, in the period October 1991-March 1992. In these programs, studies are underway on the performance of waste glass and spent fuel in projected nuclear repository conditions to provide input to the licensing of the nation`s high-level waste repositories

Bates, J.K.; Bradley, C.R.; Buck, E.C.; Dietz, N.L.; Ebert, W.L.; Emery, J.W.; Feng, X.; Finn, P.A.; Gerding, T.J.; Hoh, J.C. [and others

1993-11-01T23:59:59.000Z

231

New high-level waste management technology for IFR pyroprocessing wastes  

SciTech Connect

The pyrochemical electrorefining process for recovery of actinides in spent fuel from the Integral Fast Reactor accumulates fission product wastes as chlorides dissolved in molten LiCI-KCI and as metals, some of which are in molten cadmium. Pyrochemical processes are being developed to recover uranium and transuranium elements for return to the reactor, and to separate and immobilize fission products in suitable waste forms. Solvent cadmium is recycled within the process. Electrolyte salt is treated in a series of salt/cadmium extraction steps; it is also returned to the process. Salt-borne fission products are concentrated on a zeolite bed that is converted to a stable, leach-resistant mineral. Rare earth fission products from the salt, noble metal fission products, and cladding hulls are dispersed in a metal matrix.

Ackerman, J.P.; Johnson, T.R.

1993-09-01T23:59:59.000Z

232

Hydro-mechanical behaviour of bentonite-based materials used for high-level radioactive waste disposal.  

E-Print Network (OSTI)

??This study deals with the hydro-mechanical behaviour of compacted bentonite-based materials used as sealing materials in high-level radioactive waste repositories. The pure MX80 bentontie, mixtures… (more)

Wang, Qiong

2012-01-01T23:59:59.000Z

233

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

SciTech Connect

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.

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

2011-08-01T23:59:59.000Z

234

Unsaturated flow and transport through fractured rock related to high-level waste repositories; Final report, Phase 3  

SciTech Connect

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.

Evans, D.D.; Rasmussen, T.C. [Arizona Univ., Tucson, AZ (USA). Dept. of Hydrology and Water Resources

1991-01-01T23:59:59.000Z

235

Idaho High-Level Waste & Facilities Disposition, Final Environmental Impact Statement  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Appendix A 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 Waste Disposal Site Selection A-6 A.4.1 Identification of "Must" Criteria A-7 A.4.2 Identification of "Want" Criteria A-8 A.4.3 Identification of Candidate Sites A-8 A.4.4 Evaluation Process A-8 A.4.5 Results of Evaluation Process A-9 A.4.6 Final Selection of a Low-Activity Waste Disposal Facility

236

Analysis of vapor samples from the Organice PISA high level waste tanks  

SciTech Connect

Analyses for organic materials in vapor samples taken from the eight High Level Waste tanks (26F, 33F, 46F, 11H, 22H, 32H, 39H, and 43H) have been completed. Of these tanks, 26F, 33F, and 43H are designated 'organic'' tanks. Samples were collected on solvent desorption (SD) tubes (containing activated charcoal) from various heights above the tank waste. Tank ventilation was stopped for one hour prior to sampling and was not reinitiated until sample collection was complete. The results indicate that the concentration of organic materials is extremely low in all samples. Some organic materials were found in the vapor samples but in nanogram/liter (ng/L) quantities. These materials were present in the samples within the practical quantitation limit (PQL), which represents a practical and routinely achievable detection limit with a relatively good certainty that any reported value is reliable. Because of the low levels and the fact that no field background analysis was run (laboratory background analyses were run), the researchers cannot absolutely determine whether the materials were actually taken from the waste tanks or whether they are from environmental background. In any case, the quantities of material found are several orders magnitude below that which would comprise a flammability concern.

Swingle, R.F. II

2000-06-01T23:59:59.000Z

237

Chromium Speciation and Mobility in a High Level Nuclear Waste...  

NLE Websites -- All DOE Office Websites (Extended Search)

is a significant groundwater contaminant at the U.S. Department of Energy (DOE) Hanford Site in southeastern WA (Poston et al., 2001) where Pu was produced during WWII and...

238

Program on Technology Innovation: Advanced Fuel Cycles--Impact on High-Level Waste Disposal  

Science Conference Proceedings (OSTI)

The aim of advanced fuel cycles is to improve the sustainability of nuclear energy by enhancing the effectiveness of natural uranium resource utilization and by mitigating waste disposal issues, while keeping the costs of energy products, in particular electricity, economically viable. In addition, this aim has to be achieved under conditions that minimize the risks of diversion of separated fissile materials and their possible misuse for non-peaceful ends. The report presents results from recently publi...

2007-12-21T23:59:59.000Z

239

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

SciTech Connect

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.

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

2010-05-01T23:59:59.000Z

240

An Instrument for Measuring the TRU Concentration in High-Level Liquid Waste  

Science Conference Proceedings (OSTI)

An online monitor has been designed, built, and tested that is capable of measuring the residual transuranic concentrations in processed high-level wastes with a detection limit of 370 Bq/ml (10 nCi/ml) in less than six hours. The monitor measures the ({alpha},n) neutrons in the presence of gamma-ray fields up to 1 Sv/h (100 R/h). The optimum design was determined by Monte Carlo modeling and then tempered with practical engineering and cost considerations. A multiplicity counter is used in data acquisition to reject the large fraction of coincident and highly variable cosmic-ray-engendered background events and results in a S/N ratio {approx}1.

Brodzinski, Ronald L.; Craig, R A.; Fink, Samuel D.; Hensley, Walter K.; Holt, Noah OA; Knopf, Michael A.; Lepel, Elwood A.; Mullen, O Dennis; Salaymeh, Saleem R.; Samuel, Todd J.; Smart, John E.; Tinker, Mike R.; Walker, D

2005-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

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

Science Conference Proceedings (OSTI)

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.

Farmer, J. C., LLNL

1998-06-01T23:59:59.000Z

242

The effect of high-level waste glass composition on spinel liquidus temperature  

Science Conference Proceedings (OSTI)

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.

Kruger, A. A. [Department of Energy, Office of River Protection, Richland, Washington (United States); Riley, Brian J. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Crum, Jarrod V. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hrma, Pavel [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Matyas, Josef [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

2012-11-15T23:59:59.000Z

243

Test methods for selection of materials of construction for high-level radioactive waste vitrification. Revision  

Science Conference Proceedings (OSTI)

Candidate materials of construction were evaluated for a facility at the Department of Energy's Savannah River Plant to vitrify high-level radioactive waste. Limited operating experience was available under the corrosive conditions of the complex vitrification process. The objective of the testing program was to provide a high degree of assurance that equipment will meet or exceed design lifetimes. To meet this objective in reasonable time and minimum cost, a program was designed consisting of a combination of coupon immersion and electrochemical laboratory tests and pilot-scale tests. Stainless steels and nickel-based alloys were tested. Alloys that were most resistant to general and local attack contained nickel, molybdenum (>9%), and chromium (where Cr + Mo > 30%). Alloy C-276 was selected as the reference material for process equipment. Stellite 6 was selected for abrasive service in the presence of formic acid. Alloy 690 and ALLCORR were selected for specific applications.

Bickford, D F; Corbett, R A; Morrison, W S

1986-01-01T23:59:59.000Z

244

[Study of institutional issues relating to transportation of high level waste]. Final technical report  

SciTech Connect

This is the ``seventh`` and final Quarterly Report under the scope of work for cooperative agreement between the Western Interstate Energy Board and the US Department of Energy. The report covers the period January--March 1993. The cooperative agreement was to expire in June 1992, but DOE granted an extension through March 24, 1993. Since this is the last Quarterly Report under the expired cooperative agreement, most tasks are noted as being completed. Two final items, however, will soon be sent to DOE -- final minutes from the March 9--11 High- Level Radioactive Waste Committee meeting, and the Year-End Technical Report. Some highlights from the quarter: The Committee decided on a preferred format for the revised Spent Fuel and High-Level Radioactive Waste Transportation Primer. The document would be 100- 200 pages, accompanied by a series of white papers on key transportation elements. A 25--30 page handbook for educating western state elected officials would also be prepared. On March 24, the Committee sent a letter to DOE commenting on the Near-Site Transportation Infrastructure report findings. The Committee is concerned that infrastructure limitations may limit the rail shipping option in many instances, even after upgrades have been implemented. The NSTI findings may also have significant relevance to the decision to develop multi-purpose canisters. On April 1, the Committee sent DOE the white paper, Transportation Implications of Various NWPA Program Options, which determined that DOE cannot develop a national transportation system by 1998 for shipments to an MRS or other federal storage facility.

Not Available

1993-06-25T23:59:59.000Z

245

Nuclear waste solutions  

DOE Patents (OSTI)

High efficiency removal of technetium values from a nuclear waste stream is achieved by addition to the waste stream of a precipitant contributing tetraphenylphosphonium cation, such that a substantial portion of the technetium values are precipitated as an insoluble pertechnetate salt.

Walker, Darrel D. (1684 Partridge Dr., Aiken, SC 29801); Ebra, Martha A. (129 Hasty Rd., Aiken, SC 29801)

1987-01-01T23:59:59.000Z

246

DOUBLE SHELL TANK (DST) INTEGRITY PROJECT HIGH LEVEL WASTE CHEMISTRY OPTIMIZATION  

SciTech Connect

The U.S. Department of Energy's Office (DOE) of River Protection (ORP) has a continuing program for chemical optimization to better characterize corrosion behavior of High-Level Waste (HLW). The DOE controls the chemistry in its HLW to minimize the propensity of localized corrosion, such as pitting, and stress corrosion cracking (SCC) in nitrate-containing solutions. By improving the control of localized corrosion and SCC, the ORP can increase the life of the Double-Shell Tank (DST) carbon steel structural components and reduce overall mission costs. The carbon steel tanks at the Hanford Site are critical to the mission of safely managing stored HLW until it can be treated for disposal. The DOE has historically used additions of sodium hydroxide to retard corrosion processes in HLW tanks. This also increases the amount of waste to be treated. The reactions with carbon dioxide from the air and solid chemical species in the tank continually deplete the hydroxide ion concentration, which then requires continued additions. The DOE can reduce overall costs for caustic addition and treatment of waste, and more effectively utilize waste storage capacity by minimizing these chemical additions. Hydroxide addition is a means to control localized and stress corrosion cracking in carbon steel by providing a passive environment. The exact mechanism that causes nitrate to drive the corrosion process is not yet clear. The SCC is less of a concern in the newer stress relieved double shell tanks due to reduced residual stress. The optimization of waste chemistry will further reduce the propensity for SCC. The corrosion testing performed to optimize waste chemistry included cyclic potentiodynamic volarization studies. slow strain rate tests. and stress intensity factor/crack growth rate determinations. Laboratory experimental evidence suggests that nitrite is a highly effective:inhibitor for pitting and SCC in alkaline nitrate environments. Revision of the corrosion control strategies to a nitrite-based control, where there is no constant depletion mechanism as with hydroxide, should greatly enhance tank lifetime, tank space availability, and reduce downstream reprocessing costs by reducing chemical addition to the tanks.

WASHENFELDER DJ

2008-01-22T23:59:59.000Z

247

Composition of simulants used in the evaluation of electrochemical processes for the treatment of high-level wastes  

SciTech Connect

Four simulants are being used in the evaluation of electrochemical processes for the treatment of high-level wastes (HLW). These simulants represent waste presently stored at the Hanford, Idaho Falls, Oak Ridge, and Savannah River sites. Three of the simulants are highly alkaline salt solutions (Hanford, Oak Ridge, and Savannah River), and one is highly acidic (Idaho Falls).

Hobbs, D.T.

1994-06-27T23:59:59.000Z

248

ESTIMATING HIGH LEVEL WASTE MIXING PERFORMANCE IN HANFORD DOUBLE SHELL TANKS  

SciTech Connect

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.

THIEN MG; GREER DA; TOWNSON P

2011-01-13T23:59:59.000Z

249

HIGH-LEVEL WASTE FEED CERTIFICATION IN HANFORD DOUBLE-SHELL TANKS  

SciTech Connect

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.

THIEN MG; WELLS BE; ADAMSON DJ

2010-01-14T23:59:59.000Z

250

PERFORMANCE OF A BURIED RADIOACTIVE HIGH LEVEL WASTE GLASS AFTER 24 YEARS  

SciTech Connect

A radioactive high level waste glass was made in 1980 with Savannah River Site (SRS) Tank 15 waste. This glass was buried in the SRS burial ground for 24 years but lysimeter data was only available for the first 8 years. The glass was exhumed and analyzed in 2004. The glass was predicted to be very durable and laboratory tests confirmed the durability response. The laboratory results indicated that the glass was very durable as did analysis of the lysimeter data. Scanning electron microscopy of the glass burial surface showed no significant glass alteration consistent with the results of the laboratory and field tests. No detectable Pu, Am, Cm, Np, or Ru leached from the glass into the surrounding sediment. Leaching of {beta}/{delta} from {sup 90}Sr and {sup 137}Cs in the glass was diffusion controlled. Less than 0.5% of the Cs and Sr in the glass leached into the surrounding sediment, with >99% of the leached radionuclides remaining within 8 centimeters of the glass pellet.

Jantzen, C; Daniel Kaplan, D; Ned Bibler, N; David Peeler, D; John Plodinec, J

2008-05-05T23:59:59.000Z

251

TWRS retrieval and disposal mission, immobilized high-level waste storage plan  

SciTech Connect

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.

Calmus, R.B.

1998-01-07T23:59:59.000Z

252

Hanford high level waste (HLW) tank mixer pump safe operating envelope reliability assessment  

DOE Green Energy (OSTI)

The US Department of Energy and its contractor, Westinghouse Corp., are responsible for the management and safe storage of waste accumulated from processing defense reactor irradiated fuels for plutonium recovery at the Hanford Site. These wastes, which consist of liquids and precipitated solids, are stored in underground storage tanks pending final disposition. Currently, 23 waste tanks have been placed on a safety watch list because of their potential for generating, storing, and periodically releasing various quantities of hydrogen and other gases. Tank 101-SY in the Hanford SY Tank Farm has been found to release hydrogen concentrations greater than the lower flammable limit (LFL) during periodic gas release events. In the unlikely event that an ignition source is present during a hydrogen release, a hydrogen burn could occur with a potential to release nuclear waste materials. To mitigate the periodic gas releases occurring from Tank 101-SY, a large mixer pump currently is being installed in the tank to promote a sustained release of hydrogen gas to the tank dome space. An extensive safety analysis (SA) effort was undertaken and documented to ensure the safe operation of the mixer pump after it is installed in Tank 101-SY.1 The SA identified a need for detailed operating, alarm, and abort limits to ensure that analyzed safety limits were not exceeded during pump operations.

Fischer, S.R. [Los Alamos National Lab., NM (United States); Clark, J. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)

1993-10-01T23:59:59.000Z

253

COMBINED RETENTION OF MOLYBDENUM AND SULFUR IN SIMULATED HIGH LEVEL WASTE GLASS  

SciTech Connect

This study was undertaken to investigate the effect of elevated sulfate and molybdenum concentrations in nuclear waste glasses. A matrix of 24 glasses was developed and the glasses were tested for acceptability based on visual observations, canister centerline-cooled heat treatments, and chemical composition analysis. Results from the chemical analysis of the rinse water from each sample were used to confirm the presence of SO{sup 2-}{sub 4} and MoO{sub 3} on the surface of glasses as well as other components which might form water soluble compounds with the excess sulfur and molybdenum. A simple, linear model was developed to show acceptable concentrations of SO{sub 4}{sup 2-} and MoO{sub 3} in an example waste glass composition. This model was constructed for scoping studies only and is not ready for implementation in support of actual waste vitrification. Several other factors must be considered in determining the limits of sulfate and molybdenum concentrations in the waste vitrification process, including but not limited to, impacts on refractory and melter component corrosion, effects on the melter off-gas system, and impacts on the chemical durability and crystallization of the glass product.

Fox, K.

2009-10-16T23:59:59.000Z

254

Idaho National Engineering Laboratory High-Level Waste Roadmap. Revision 2  

SciTech Connect

The Idaho National Engineering Laboratory (INEL) High-Level Waste (HLW) Roadmap takes a strategic look at the entire HLW life-cycle starting with generation, through interim storage, treatment and processing, transportation, and on to final disposal. The roadmap is an issue-based planning approach that compares ``where we are now`` to ``where we want and need to be.`` The INEL has been effectively managing HLW for the last 30 years. Calcining operations are continuing to turn liquid HLW into a more manageable form. Although this document recognizes problems concerning HLW at the INEL, there is no imminent risk to the public or environment. By analyzing the INEL current business operations, pertinent laws and regulations, and committed milestones, the INEL HLW Roadmap has identified eight key issues existing at the INEL that must be resolved in order to reach long-term objectives. These issues are as follows: A. The US Department of Energy (DOE) needs a consistent policy for HLW generation, handling, treatment, storage, and disposal. B. The capability for final disposal of HLW does not exist. C. Adequate processes have not been developed or implemented for immobilization and disposal of INEL HLW. D. HLW storage at the INEL is not adequate in terms of capacity and regulatory requirements. E. Waste streams are generated with limited consideration for waste minimization. F. HLW is not adequately characterized for disposal nor, in some cases, for storage. G. Research and development of all process options for INEL HLW treatment and disposal are not being adequately pursued due to resource limitations. H. HLW transportation methods are not selected or implemented. A root-cause analysis uncovered the underlying causes of each of these issues.

1993-08-01T23:59:59.000Z

255

Canister design for deep borehole disposal of nuclear waste  

E-Print Network (OSTI)

The objective of this thesis was to design a canister for the disposal of spent nuclear fuel and other high-level waste in deep borehole repositories using currently available and proven oil, gas, and geothermal drilling ...

Hoag, Christopher Ian

2006-01-01T23:59:59.000Z

256

Pioneering Nuclear Waste Disposal  

NLE Websites -- All DOE Office Websites (Extended Search)

18 18 19 T he WIPP's first waste receipt, 11 years later than originally planned, was a monumental step forward in the safe management of nuclear waste. Far from ending, however, the WIPP story has really just begun. For the next 35 years, the DOE will face many challenges as it manages a complex shipment schedule from transuranic waste sites across the United States and continues to ensure that the repository complies with all regulatory requirements. The DOE will work to maintain the highest level of safety in waste handling and trans- portation. Coordination with sites Disposal operations require coordination with sites that will ship transuranic waste to the WIPP and include periodic certification of waste characterization and handling practices at those facilities. During the WIPP's

257

THE APPARENT SOLUBILITY OF ALUMINUM(III) IN HANFORD HIGH-LEVEL WASTE TANKS  

SciTech Connect

The solubility of aluminum in Hanford nuclear waste impacts on the process ability of the waste by a number of proposed treatment options. For many years, Hanford staff has anecdotally noted that aluminum appears to be considerably more soluble in Hanford waste than the simpler electrolyte solutions used as analogues. There has been minimal scientific study to confirm these anecdotal observations, however. The present study determines the apparent solubility product for gibbsite in 50 tank samples. The ratio of hydroxide to aluminum in the liquid phase for the samples is calculated and plotted as a function of total sodium molarity. Total sodium molarity is used as a surrogate for ionic strength, because the relative ratios of mono, di and trivalent anions are not available for all of the samples. These results were compared to the simple NaOH-NaAl(OH{sub 4})H{sub 2}O system, and the NaOH-NaAl(OH{sub 4})NaCl-H{sub 2}O system data retrieved from the literature. The results show that gibbsite is apparently more soluble in the samples than in the simple systems whenever the sodium molarity is greater than two. This apparent enhanced solubility cannot be explained solely by differences in ionic strength. The change in solubility with ionic strength in simple systems is small compared to the difference between aluminum solubility in Hanford waste and the simple systems. The reason for the apparent enhanced solubility is unknown, but could include. kinetic or thermodynamic factors that are not present in the simple electrolyte systems. Any kinetic explanation would have to explain why the samples are always supersaturated whenever the sodium molarity is above two. Real waste characterization data should not be used to validate thermodynamic solubility models until it can be confirmed that the apparent enhanced gibbsite solubility is a thermodynamic effect and not a kinetic effect.

REYNOLDS JG

2012-06-20T23:59:59.000Z

258

Nuclear waste package fabricated from concrete  

Science Conference Proceedings (OSTI)

After the United States enacted the Nuclear Waste Policy Act in 1983, the Department of Energy must design, site, build and operate permanent geologic repositories for high-level nuclear waste. The Department of Energy has recently selected three sites, one being the Hanford Site in the state of Washington. At this particular site, the repository will be located in basalt at a depth of approximately 3000 feet deep. The main concern of this site, is contamination of the groundwater by release of radionuclides from the waste package. The waste package basically has three components: the containment barrier (metal or concrete container, in this study concrete will be considered), the waste form, and other materials (such as packing material, emplacement hole liners, etc.). The containment barriers are the primary waste container structural materials and are intended to provide containment of the nuclear waste up to a thousand years after emplacement. After the containment barriers are breached by groundwater, the packing material (expanding sodium bentonite clay) is expected to provide the primary control of release of radionuclide into the immediate repository environment. The loading conditions on the concrete container (from emplacement to approximately 1000 years), will be twofold; (1) internal heat of the high-level waste which could be up to 400/sup 0/C; (2) external hydrostatic pressure up to 1300 psi after the seepage of groundwater has occurred in the emplacement tunnel. A suggested container is a hollow plain concrete cylinder with both ends capped. 7 refs.

Pfeiffer, P.A.; Kennedy, J.M.

1987-03-01T23:59:59.000Z

259

Waste Form Performance Modeling [Nuclear Waste Management using...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

260

Particle Generation by Laser Ablation in Support of Chemical Analysis of High Level Mixed Waste from Plutonium Production Operations  

Science Conference Proceedings (OSTI)

Investigate particles produced by laser irradiation and their analysis by Laser Ablation Inductively Coupled Plasma Mass Spectroscopy (LA/ICP-MS), with a view towards optimizing particle production for analysis of high level waste materials and waste glass. LA/ICP-MS has considerable potential to increase the safety and speed of analysis required for the remediation of high level wastes from cold war plutonium production operations. In some sample types, notably the sodium nitrate-based wastes at Hanford and elsewhere, chemical analysis using typical laser conditions depends strongly on the details of sample history composition in a complex fashion, rendering the results of analysis uncertain. Conversely, waste glass materials appear to be better behaved and require different strategies to optimize analysis.

J. Thomas Dickinson; Michael L. Alexander

2001-11-30T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


261

EPRI Review of Geologic Disposal for Used Fuel and High Level Radioactive Waste: Volume II--U.S. Regulations for Geologic Disposal  

Science Conference Proceedings (OSTI)

U.S. efforts to site and construct a deep geologic repository for used fuel and high level radioactive waste (HLW) proceeded sporadically over a three-decade period from the late 1950s until 1982, when the U.S. Congress enacted the Nuclear Waste Policy Act (NWPA) codifying a national approach for developing a deep geologic repository. Amendment of the NWPA in 1987 resulted in a number of dramatic changes in direction for the U.S. program, most notably the selection of Yucca Mountain as the only site of t...

2010-06-29T23:59:59.000Z

262

Development of Cementitious Waste Forms for Nuclear Waste ...  

Science Conference Proceedings (OSTI)

Symposium, Materials Solutions for the Nuclear Renaissance. Presentation Title, Development of Cementitious Waste Forms for Nuclear Waste Immobilization.

263

Preliminary Technology Maturation Plan for Immobilization of High-Level Waste in Glass Ceramics  

Science Conference Proceedings (OSTI)

A technology maturation plan (TMP) was developed for immobilization of high-level waste (HLW) raffinate in a glass ceramics waste form using a cold-crucible induction melter (CCIM). The TMP was prepared by the following process: 1) define the reference process and boundaries of the technology being matured, 2) evaluate the technology elements and identify the critical technology elements (CTE), 3) identify the technology readiness level (TRL) of each of the CTE’s using the DOE G 413.3-4, 4) describe the development and demonstration activities required to advance the TRLs to 4 and 6 in order, and 5) prepare a preliminary plan to conduct the development and demonstration. Results of the technology readiness assessment identified five CTE’s and found relatively low TRL’s for each of them: • Mixing, sampling, and analysis of waste slurry and melter feed: TRL-1 • Feeding, melting, and pouring: TRL-1 • Glass ceramic formulation: TRL-1 • Canister cooling and crystallization: TRL-1 • Canister decontamination: TRL-4 Although the TRL’s are low for most of these CTE’s (TRL-1), the effort required to advance them to higher values. The activities required to advance the TRL’s are listed below: • Complete this TMP • Perform a preliminary engineering study • Characterize, estimate, and simulate waste to be treated • Laboratory scale glass ceramic testing • Melter and off-gas testing with simulants • Test the mixing, sampling, and analyses • Canister testing • Decontamination system testing • Issue a requirements document • Issue a risk management document • Complete preliminary design • Integrated pilot testing • Issue a waste compliance plan A preliminary schedule and budget were developed to complete these activities as summarized in the following table (assuming 2012 dollars). TRL Budget Year MSA FMP GCF CCC CD Overall $M 2012 1 1 1 1 4 1 0.3 2013 2 2 1 1 4 1 1.3 2014 2 3 1 1 4 1 1.8 2015 2 3 2 2 4 2 2.6 2016 2 3 2 2 4 2 4.9 2017 2 3 3 2 4 2 9.8 2018 3 3 3 3 4 3 7.9 2019 3 3 3 3 4 3 5.1 2020 3 3 3 3 4 3 14.6 2021 3 3 3 3 4 3 7.3 2022 3 3 3 3 4 3 8.8 2023 4 4 4 4 4 4 9.1 2024 5 5 5 5 5 5 6.9 2025 6 6 6 6 6 6 6.9 CCC = canister cooling and crystallization; FMP = feeding, melting, and pouring; GCF = glass ceramic formulation; MSA = mixing, sampling, and analyses. This TMP is intended to guide the development of the glass ceramics waste form and process to the point where it is ready for industrialization.

Vienna, John D.; Crum, Jarrod V.; Sevigny, Gary J.; Smith, G L.

2012-09-30T23:59:59.000Z

264

An approximate-reasoning-based method for screening high-level waste tanks for flammable gas  

DOE Green Energy (OSTI)

The in situ retention of flammable gas produced by radiolysis and thermal decomposition in high-level waste can pose a safety problem if the gases are released episodically into the dome space of a storage tank. Screening efforts at Hanford have been directed at identifying tanks in which this situation could exist. Problems encountered in screening motivated an effort to develop an improved screening methodology. Approximate reasoning (AR) is a formalism designed to emulate the kinds of complex judgments made by subject matter experts. It uses inductive logic structures to build a sequence of forward-chaining inferences about a subject. AR models incorporate natural language expressions known as linguistic variables to represent evidence. The use of fuzzy sets to represent these variables mathematically makes it practical to evaluate quantitative and qualitative information consistently. The authors performed a pilot study to investigate the utility of AR for flammable gas screening. They found that the effort to implement such a model was acceptable and that computational requirements were reasonable. The preliminary results showed that important judgments about the validity of observational data and the predictive power of models could be made. These results give new insights into the problems observed in previous screening efforts.

Eisenhawer, S.W.; Bott, T.F.; Smith, R.E.

1998-07-01T23:59:59.000Z

265

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

SciTech Connect

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.

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

2003-02-27T23:59:59.000Z

266

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

Science Conference Proceedings (OSTI)

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

Not Available

1983-06-01T23:59:59.000Z

267

EPRI Review of Geologic Disposal for Used Fuel and High-Level Radioactive Waste: Volume III - Review of National Repository Programs  

Science Conference Proceedings (OSTI)

The effective termination of the Yucca Mountain program by the U.S. Administration in 2009 has left the U.S. program for management of used fuel and high level radioactive waste (HLW) in a state of uncertainty. In concert with this major policy reset and in response to the resulting policy vacuum, the President directed the Energy Secretary to establish the Blue Ribbon Commission on America's Nuclear Future (BRC) "...to conduct a comprehensive review of policies for managing the back end of the nuclear f...

2010-12-21T23:59:59.000Z

268

WEB RESOURCE: Nuclear Waste Disposal  

Science Conference Proceedings (OSTI)

May 10, 2007 ... The complete "Yucca Mountain Resource Book" is also available for download at this site. Citation: Nuclear Waste Disposal. 2007. Nuclear ...

269

Nuclear Waste Fund Activities Management Team | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

Waste Fund Activities Management Team Waste Fund Activities Management Team Nuclear Waste Fund Activities Management Team The Nuclear Waste Fund Activities Management Team has responsibility to: Manage the investments and expenditures of the Nuclear Waste Fund; Support correspondence regarding Nuclear Waste Policy Act issues raised by congressional, Inspector General, Government Accounting Office and Freedom of Information Act inquiries; and, Manage the annual fee adequacy assessment process. Applicable Documents Nuclear Waste Policy Act of 1982 Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste Standard Contract Amendment for New Reactors FY 2007 Total System Life Cycle Cost, Pub 2008 FY 2007 Fee Adequacy, Pub 2008 2009 Letter to Congress OCRWM Financial Statements for Annual Report for Years Ended

270

INTEC High-Level Waste Studies Universal Solvent Extraction Feasibility Study  

SciTech Connect

This report summarizes a feasibility study that has been conducted on the Universal Solvent Extraction (UNEX) Process for treatment and disposal of 4.3 million liters of INEEL sodium-bearing waste located at the Idaho Nuclear Technology and Engineering Center. This feasibility study covers two scenarios of treatment. The first, the UNEX Process, partitions the Cs/Sr from the SBW and creates remote-handled LLW and contact-handled TRU waste forms. Phase one of this study, covered in the 30% review documents, dealt with defining the processes and defining the major unit operations. The second phase of the project, contained in the 60% review, expanded on the application of the UNEX processes and included facility requirements and definitions. Two facility options were investigated for the UNEX process, resulting in a 2 x 2 matrix of process/facility scenarios as follows: Option A, UNEX at Greenfield Facility, Option B, Modified UNEX at Greenfield Facility, Option C, UNEX at NWCF, th is document, covers life-cycle costs for all options presented along with results and conclusions determined from the study.

J. Banaee; C. M. Barnes; T. Battisti (ANL-W) [ANL-W; S. Herrmann (ANL-W) [ANL-W; S. J. Losinski; S. McBride (ANL-W) [ANL-W

2000-09-01T23:59:59.000Z

271

Nuclear Waste Assessment System for Technical Evaluation (NUWASTE)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NWTRB NWTRB www.nwtrb.gov U.S. Nuclear Waste Technical Review Board U.S. Nuclear Waste Technical Review Board: Roles and Priorities Presented by: Nigel Mote, Executive Director, U.S. Nuclear Waste Technical Review Board May 14, 2013 Hyatt Regency Buffalo, Buffalo, NY. Presented to: National Transportation Stakeholders' Forum NWTRB www.nwtrb.gov U.S. Nuclear Waste Technical Review Board The Board's Statutory Mandate * The 1987 amendments to the Nuclear Waste Policy Act (NWPA) established the U.S. Nuclear Waste Technical Review Board. * The Board evaluates the technical and scientific validity of DOE activities related to implementing the NWPA, including: - transportation, packaging, and storage of spent nuclear fuel (SNF) and high-level radioactive waste (HLW)

272

Pioneering Nuclear Waste Disposal  

NLE Websites -- All DOE Office Websites (Extended Search)

PIONEERING NUCLEAR WASTE DISPOSAL U.S. Department of Energy Carlsbad Area Office February 2000 DOECAO-00-3124 T h e W a s t e I s o l a t i o n P i l o t P l a n t ii Table of...

273

High-level waste storage tank farms/242-A evaporator standards/requirements identification document (S/RID), Vol. 6  

SciTech Connect

The scope of the Environmental Restoration and Waste Management (EM) Functional Area includes the programmatic controls associated with the management and operation of the Hanford Tank Farm Facility. The driving management organization implementing the programmatic controls is the Tank Farms Waste Management (WM)organization whose responsibilities are to ensure that performance objectives are established; and that measurable criteria for attaining objectives are defined and reflected in programs, policies and procedures. Objectives for the WM Program include waste minimization, establishment of effective waste segregation methods, waste treatment technology development, radioactive (low-level, high-level) hazardous and mixed waste transfer, treatment, and storage, applicability of a corrective action program, and management and applicability of a decontamination and decommissioning (D&D) program in future years.

Not Available

1994-04-01T23:59:59.000Z

274

INTERNATIONAL STUDY OF ALUMINUM IMPACTS ON CRYSTALLIZATION IN U.S. HIGH LEVEL WASTE GLASS  

SciTech Connect

The objective of this task was to develop glass formulations for (Department of Energy) DOE waste streams with high aluminum concentrations 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. Liquidus temperatures and crystallization behavior were carefully characterized to support model development for higher waste loading glasses. The experimental work, characterization, and data interpretation necessary to meet these objectives were performed among three partnering laboratories: the V.G. Khlopin Radium Institute (KRI), Pacific Northwest National Laboratory (PNNL) and Savannah River National Laboratory (SRNL). Projected glass compositional regions that bound anticipated Defense Waste Processing Facility (DWPF) and Hanford high level waste (HLW) glass regions of interest were developed and used to generate glass compositions of interest for meeting the objectives of this study. A thorough statistical analysis was employed to allow for a wide range of waste glass compositions to be examined while minimizing the number of glasses that had to be fabricated and characterized in the laboratory. The glass compositions were divided into two sets, with 45 in the test matrix investigated by the U.S. laboratories and 30 in the test matrix investigated by KRI. Fabrication and characterization of the US and KRI-series glasses were generally handled separately. This report focuses mainly on the US-series glasses. Glasses were fabricated and characterized by SRNL and PNNL. Crystalline phases were identified by X-ray diffraction (XRD) in the quenched and canister centerline cooled (CCC) glasses and were generally iron oxides and spinels, which are not expected to impact durability of the glass. Nepheline was detected in five of the glasses after the CCC heat treatment. Chemical composition measurements for each of the glasses were conducted following an analytical plan. A review of the individual oxides for each glass revealed that there were no errors in batching significant enough to impact the outcome of the study. A comparison of the measured compositions of the replicates indicated an acceptable degree of repeatability as the percent differences for most of the oxides were less than 5% and percent differences for all of the oxides were less than 10 wt%. Chemical durability was measured using the Product Consistency Test (PCT). All but two of the study glasses had normalized leachate for boron (NL [B]) values that were well below that of the Environmental Assessment (EA) reference glass. The two highest NL [B] values were for the CCC versions of glasses US-18 and US-27 (10.498 g/L and 15.962 g/L, respectively). Nepheline crystallization was identified by qualitative XRD in five of the US-series glasses. Each of these five glasses (US-18, US-26, US-27, US-37 and US-43) showed a significant increase in NL [B] values after the CCC heat treatment. This reduction in durability can be attributed to the formation of nepheline during the slow cooling cycle and the removal of glass formers from the residual glass network. The liquidus temperature (T{sub L}) of each glass in the study was determined by both optical microscopy and XRD methods. The correlation coefficient of the measured XRD TL data versus the measured optical TL data was very good (R{sup 2} = 0.9469). Aside from a few outliers, the two datasets aligned very well across the entire temperature range (829 C to 1312 C for optical data and 813 C to 1310 C for XRD crystal fraction data). The data also correlated well with the predictions of a PNNL T{sub L} model. The correlation between the measured and calculated data had a higher degree of merit for the XRD crystal fraction data than for the optical data (higher R{sup 2} value of 0.9089 versus 0.8970 for the optical data). The SEM-EDS analysis of select samples revealed the presence of undissolved RuO{sub 2} in all glasses due to the low solubility of RuO{sub 2} in borosilicate glass. These

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

2008-09-23T23:59:59.000Z

275

Immobilized High Level Waste (HLW) Interim Storage Alternative Generation and analysis and Decision Report 2nd Generation Implementing Architecture  

SciTech Connect

Two alternative approaches were previously identified to provide second-generation interim storage of Immobilized High-Level Waste (IHLW). One approach was retrofit modification of the Fuel and Materials Examination Facility (FMEF) to accommodate IHLW. The results of the evaluation of the FMEF as the second-generation IHLW interim storage facility and subsequent decision process are provided in this document.

CALMUS, R.B.

2000-09-14T23:59:59.000Z

276

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID)  

SciTech Connect

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 3) presents the standards and requirements for the following sections: Safeguards and Security, Engineering Design, and Maintenance.

Not Available

1994-04-01T23:59:59.000Z

277

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 2  

Science Conference Proceedings (OSTI)

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Document (S/RID) is contained in multiple volumes. This document (Volume 2) presents the standards and requirements for the following sections: Quality Assurance, Training and Qualification, Emergency Planning and Preparedness, and Construction.

Not Available

1994-04-01T23:59:59.000Z

278

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 4  

SciTech Connect

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 4) presents the standards and requirements for the following sections: Radiation Protection and Operations.

Not Available

1994-04-01T23:59:59.000Z

279

High-level waste storage tank farms/242-A evaporator Standards/Requirements Identification Document (S/RID), Volume 5  

SciTech Connect

The High-Level Waste Storage Tank Farms/242-A Evaporator Standards/Requirements Identification Document (S/RID) is contained in multiple volumes. This document (Volume 5) outlines the standards and requirements for the Fire Protection and Packaging and Transportation sections.

1994-04-01T23:59:59.000Z

280

I-NERI Annual Technical Progress Report 2007-004-K Development and Characterization of New High-Level Waste Forms for Achieving Waste Minimization from Pyroprocessing  

Science Conference Proceedings (OSTI)

The current method for the immobilization of fission products that accumulate in electrorefiner salt during the electrochemical processing of used metallic nuclear fuel is to encapsulate the electrorefiner salt in a glass-bonded sodalite ceramic waste form. This process was developed by Argonne National Laboratory in the USA and is currently performed at the Idaho National Laboratory for the treatment of Experimental Breeder Reactor-II (EBR-II) used fuel. This process utilizes a “once-through” option for the disposal of spent electrorefiner salt; where, after the treatment of the EBR-II fuel, the electrorefiner salt containing the active fission products will be disposed of in the ceramic waste form (CWF). The CWF produced will have low fission product loading of approximately 2 to 5 weight percent due to the limited fuel inventory currently being processed. However; the design and implementation of advanced electrochemical processing facilities to treat used fuel would process much greater quantities fuel. With an advanced processing facility, it would be necessary to selectively remove fission products from the electrorefiner salt for salt recycle and to concentrate the fission products to reduce the volume of high-level waste from the treatment facility. The Korean Atomic Energy Research Institute and the Idaho National Laboratory have been collaborating on I-NERI research projects for a number of years to investigate both aspects of selective fission product separation from electrorefiner salt, and to develop advanced waste forms for the immobilization of the collected fission products. The first joint KAERI/INL I-NERI project titled: 2006-002-K, Separation of Fission Products from Molten LiCl-KCl Salt Used for Electrorefining of Metal Fuels, was successfully completed in 2009 by concentrating and isolating fission products from actual electrorefiner salt used for the treated used EBR-II fuel. Two separation methods were tested and from these tests were produced concentrated salt products that acted as the feed material for development of advanced waste forms investigated in this proposal. Accomplishments from the first year activities associated with this I-NERI project included the down selection of candidate waste forms to immobilize fission products separated from electrorefiner salt, and the design of equipment to fabricate actual waste forms in the Hot Fuels Examination Facility (HFEF) at the INL. Reported in this document are accomplishments from the second year (FY10) work performed at the INL, and includes the testing of waste form fabrication equipment, repeating the fission product precipitation experiment, and initial waste form fabrication efforts.

S. Frank

2010-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


281

IMPACT OF ELIMINATING MERCURY REMOVAL PRETREATMENT ON THE PERFORMANCE OF A HIGH LEVEL RADIOACTIVE WASTE MELTER OFFGAS SYSTEM  

DOE Green Energy (OSTI)

The Defense Waste Processing Facility at the Savannah River Site processes high-level radioactive waste from the processing of nuclear materials that contains dissolved and precipitated metals and radionuclides. Vitrification of this waste into borosilicate glass for ultimate disposal at a geologic repository involves chemically modifying the waste to make it compatible with the glass melter system. Pretreatment steps include removal of excess aluminum by dissolution and washing, and processing with formic and nitric acids to: (1) adjust the reduction-oxidation (redox) potential in the glass melter to reduce radionuclide volatility and improve melt rate; (2) adjust feed rheology; and (3) reduce by steam stripping the amount of mercury that must be processed in the melter. Elimination of formic acid pretreatment has been proposed to eliminate the production of hydrogen in the pretreatment systems; alternative reductants would be used to control redox. However, elimination of formic acid would result in significantly more mercury in the melter feed; the current specification is no more than 0.45 wt%, while the maximum expected prior to pretreatment is about 2.5 wt%. An engineering study has been undertaken to estimate the effects of eliminating mercury removal on the melter offgas system performance. A homogeneous gas-phase oxidation model and an aqueous phase model were developed to study the speciation of mercury in the DWPF melter offgas system. The model was calibrated against available experimental data and then applied to DWPF conditions. The gas-phase model predicted the Hg{sub 2}{sup 2-}/Hg{sup 2+} ratio accurately, but some un-oxidized Hg{sup 0} remained. The aqueous model, with the addition of less than 1 mM Cl{sub 2} showed that this remaining Hg{sup 0} would be oxidized such that the final Hg{sub 2}{sup 2+}/Hg{sup 2+} ratios matched the experimental data. The results of applying the model to DWPF show that due to excessive shortage of chloride, only 6% of the mercury fed is expected to be chlorinated, mostly as Hg{sub 2}Cl{sub 2}, while the remaining mercury would exist either as elemental mercury (90%) or HgO (4%).

Zamecnik, J; Alexander Choi, A

2009-03-17T23:59:59.000Z

282

Use of depleted uranium metal as cask shielding in high-level waste storage, transport, and disposal systems  

SciTech Connect

The US DOE has amassed over 555,000 metric tons of depleted uranium from its uranium enrichment operations. Rather than dispose of this depleted uranium as waste, this study explores a beneficial use of depleted uranium as metal shielding in casks designed to contain canisters of vitrified high-level waste. Two high-level waste storage, transport, and disposal shielded cask systems are analyzed. The first system employs a shielded storage and disposal cask having a separate reusable transportation overpack. The second system employs a shielded combined storage, transport, and disposal cask. Conceptual cask designs that hold 1, 3, 4 and 7 high-level waste canisters are described for both systems. In all cases, cask design feasibility was established and analyses indicate that these casks meet applicable thermal, structural, shielding, and contact-handled requirements. Depleted uranium metal casting, fabrication, environmental, and radiation compatibility considerations are discussed and found to pose no serious implementation problems. About one-fourth of the depleted uranium inventory would be used to produce the casks required to store and dispose of the nearly 15,400 high-level waste canisters that would be produced. This study estimates the total-system cost for the preferred 7-canister storage and disposal configuration having a separate transportation overpack would be $6.3 billion. When credits are taken for depleted uranium disposal cost, a cost that would be avoided if depleted uranium were used as cask shielding material rather than disposed of as waste, total system net costs are between $3.8 billion and $5.5 billion.

Yoshimura, H.R.; Ludwigsen, J.S.; McAllaster, M.E. [and others

1996-09-01T23:59:59.000Z

283

Pioneering Nuclear Waste Disposal  

NLE Websites -- All DOE Office Websites (Extended Search)

2 2 3 T he journey to the WIPP began nearly 60 years before the first barrels of transuranic waste arrived at the repository. The United States produced the world's first sig- nificant quantities of transuranic material during the Manhattan Project of World War II in the early 1940s. The government idled its plutonium- producing reactors and warhead manu- facturing plants at the end of the Cold War and scheduled most of them for dismantlement. However, the DOE will generate more transuranic waste as it cleans up these former nuclear weapons facilities. The WIPP is a cor- nerstone of the effort to clean up these facilities by providing a safe repository to isolate transuranic waste in disposal rooms mined out of ancient salt beds, located 2,150 feet below ground. The need for the WIPP

284

A QUARTER CENTURY OF NUCLEAR WASTE MANAGEMENT IN JAPAN  

Science Conference Proceedings (OSTI)

This paper is entitled ''A QUARTER CENTURY OF NUCLEAR WASTE MANAGEMENT IN JAPAN''. Since the first statement on the strategy for radioactive waste management in Japan was made by the Atomic Energy Commission (AEC) in 1976, a quarter century has passed, in which much experience has been accumulated both in technical and social domains. This paper looks back in this 25-year history of radioactive waste management in Japan by highlighting activities related to high-level radioactive waste (HLW) disposal.

Masuda, S.

2002-02-25T23:59:59.000Z

285

Method of preparing nuclear wastes for tansportation and interim storage  

SciTech Connect

Nuclear waste is formed into a substantially water-insoluble solid for temporary storage and transportation by mixing the calcined waste with at least 10 weight percent powdered anhydrous sodium silicate to form a mixture and subjecting the mixture to a high humidity environment for a period of time sufficient to form cementitious bonds by chemical reaction. The method is suitable for preparing an interim waste form from dried high level radioactive wastes.

Bandyopadhyay, Gautam (Naperville, IL); Galvin, Thomas M. (Darien, IL)

1984-01-01T23:59:59.000Z

286

FRIT DEVELOPMENT FOR HIGH LEVEL WASTE SLUDGE BATCH 5: COMPOSITIONAL TRENDS FOR VARYING ALUMINUM CONCENTRATIONS  

Science Conference Proceedings (OSTI)

The objective of this study was to experimentally measure the properties and performance of a series of glasses with compositions that could represent Sludge Batch 5 (SB5) as processed at the Defense Waste Processing Facility (DWPF). The data was used to provide recommendations to the Liquid Waste Organization (LWO) regarding blending and washing strategies in preparing SB5 based on acceptability of the glass compositions. These data were also used to guide frit optimization efforts as the SB5 composition was finalized. Glass compositions for this study were developed by combining a series of SB5 composition projections with a group of frits. Three composition projections for SB5 were developed using a model-based approach at Savannah River National Laboratory (SRNL). These compositions, referred to as SB5 Cases B, C and D, projected removal of 25, 50 and 75% (respectively) of the aluminum in Tank 51 through the low temperature aluminum dissolution process. The frits for this study (Frits 530 through 537) were selected based on their predicted operating windows (i.e., ranges of waste loadings over which the predicted properties of the glasses were acceptable) and their potential (based on historical trends) to provide acceptable melt rates for SB5. Six additional glasses were designed to evaluate alternatives for uranium in DWPF-type glasses used for variability studies and some scoping studies. Since special measures are necessary when working with uranium-containing glasses in the laboratory, it is desirable as a cost and time saving measure to find an alternative for uranium to support frit optimization efforts. Hafnium and neodymium were investigated as potential surrogates for uranium, and other glasses were made by simply excluding the radioactive components and renormalizing the glass composition. The study glasses were fabricated and characterized at SRNL. Chemical composition analyses suggested only minor difficulties in meeting the targeted compositions for some of the oxides for some of the glasses. Although minor differences were observed, they did not have a significant impact on the conclusions made in this study. Several of the study compositions showed retention of more than 0.5 wt% SO{sub 4}{sup 2-} in glass. Trevorite (a spinel) was the only crystalline phase that was positively identified in a few of the study glasses after the canister centerline cooled (CCC) heat treatment. Spinels are not of concern as they have been shown to have little impact on the durability of high level waste glasses. The crystallization behavior of the surrogate glasses was generally the same as that of their U{sub 3}O{sub 8}-containing counterparts. There are two pairs that were exceptions: SB5-04 (amorphous) and SB5-24 (possible trevorite), along with SB5-07 (amorphous) and SB5-25 (trevorite). In these cases, the surrogate glasses (SB5-24 and SB5-25) appear to be more conservative (more prone to crystallization) than their U{sub 3}O{sub 8}-containing counterparts. Chemical durability was quantified using the Product Consistency Test (PCT). The normalized leachate (NL) values for B, Li, Na and Si for all of the study glasses were well below those of the Environmental Assessment (EA) benchmark glass, regardless of heat treatment or compositional view. This indicates that all of the glasses had very acceptable durability performance. The highest NL [B] for the study glasses was 0.914 g/L (the quenched version of glass SB5-13), normalized using the measured, bias-correct composition. There was little practical impact of the CCC heat treatment on the PCT responses of the study glasses. The measured PCT responses were predictable by the current {Delta}G{sub p} models. In general, the PCT responses for the surrogate glasses or the glasses without U{sub 3}O{sub 8} were quite similar to their U{sub 3}O{sub 8}-containing counterparts. The average percent error in NL [B] normalized by the measured, bias-corrected compositions for the surrogate glasses compared with their radioactive counterparts was 8.8%. The largest difference in NL

Fox, K; Tommy Edwards; David Best; Irene Reamer; Phyllis Workman

2008-08-28T23:59:59.000Z

287

Natural analogues of nuclear waste glass corrosion.  

SciTech Connect

This report reviews and summarizes studies performed to characterize the products and processes involved in the corrosion of natural glasses. Studies are also reviewed and evaluated on how well the corrosion of natural glasses in natural environments serves as an analogue for the corrosion of high-level radioactive waste glasses in an engineered geologic disposal system. A wide range of natural and experimental corrosion studies has been performed on three major groups of natural glasses: tektite, obsidian, and basalt. Studies of the corrosion of natural glass attempt to characterize both the nature of alteration products and the reaction kinetics. Information available on natural glass was then compared to corresponding information on the corrosion of nuclear waste glasses, specifically to resolve two key questions: (1) whether one or more natural glasses behave similarly to nuclear waste glasses in laboratory tests, and (2) how these similarities can be used to support projections of the long-term corrosion of nuclear waste glasses. The corrosion behavior of basaltic glasses was most similar to that of nuclear waste glasses, but the corrosion of tektite and obsidian glasses involves certain processes that also occur during the corrosion of nuclear waste glasses. The reactions and processes that control basalt glass dissolution are similar to those that are important in nuclear waste glass dissolution. The key reaction of the overall corrosion mechanism is network hydrolysis, which eventually breaks down the glass network structure that remains after the initial ion-exchange and diffusion processes. This review also highlights some unresolved issues related to the application of an analogue approach to predicting long-term behavior of nuclear waste glass corrosion, such as discrepancies between experimental and field-based estimates of kinetic parameters for basaltic glasses.

Abrajano, T.A. Jr.; Ebert, W.L.; Luo, J.S.

1999-01-06T23:59:59.000Z

288

Expected brine movement at potential nuclear waste repository salt sites  

SciTech Connect

The BRINEMIG brine migration code predicts rates and quantities of brine migration to a waste package emplaced in a high-level nuclear waste repository in salt. The BRINEMIG code is an explicit time-marching finite-difference code that solves a mass balance equation and uses the Jenks equation to predict velocities of brine migration. Predictions were made for the seven potentially acceptable salt sites under consideration as locations for the first US high-level nuclear waste repository. Predicted total quantities of accumulated brine were on the order of 1 m/sup 3/ brine per waste package or less. Less brine accumulation is expected at domal salt sites because of the lower initial moisture contents relative to bedded salt sites. Less total accumulation of brine is predicted for spent fuel than for commercial high-level waste because of the lower temperatures generated by spent fuel. 11 refs., 36 figs., 29 tabs.

McCauley, V.S.; Raines, G.E.

1987-08-01T23:59:59.000Z

289

Evolved Gas Analysis for High-alumina HLW (High Level Waste) Feed  

Science Conference Proceedings (OSTI)

Using the thermogravimetry coupled with gas chromatography-mass spectrometer, ... Tungstic Acid for Sorption of Uranium from Natural and Waste Waters and ...

290

Evaluation of the Candidate High-Level Radioactive Waste Repository at Yucca Mountain Using Total System Performance Assessment: Phase 5  

Science Conference Proceedings (OSTI)

A successful license application for the candidate spent-fuel and high level waste (HLW) repository at Yucca Mountain depends on a robust demonstration of long-term safety. This report presents EPRI's independent review to identify any conservatisms in the U.S. Depawrtment of Energy's (DOE's) Phase 5 Yucca Mountain Total System Performance Assessment (TSPA). The review specifically identifies key facility components, makes recommendations regarding technical development work priorities, and evaluates ove...

2000-11-21T23:59:59.000Z

291

Evaluation of the Proposed High-Level Radioactive Waste Repository at Yucca Mountain Using Total System Performance Assessment: Phase 6  

Science Conference Proceedings (OSTI)

A successful license application for the candidate spent-fuel and high-level waste repository at Yucca Mountain depends on a robust demonstration of long-term safety. This report presents EPRI's evaluation of, and makes a case for, the suitability of the Yucca Mountain repository using a Total System Performance Assessment (TSPA). The report discusses factors that make the Yucca Mountain repository system suitable for continued development and initiation of the licensing process. Information in this Phas...

2002-02-28T23:59:59.000Z

292

Evaluation of high-level waste vitrification feed preparation chemistry for an NCAW simulant, FY 1994: Alternate flowsheets (DRAFT)  

SciTech Connect

High-level radioactive waste stored in tanks at the U.S. Department of Energy`s (DOE`s) Hanford Site will be pretreated to concentrate radioactive constituents and fed to the vitrification plant A flowsheet for feed preparation within the vitrification plant (based on the Hanford Waste Vitrification Plant (HWVP) design) called for HCOOH addition during the feed preparation step to adjust rheology and glass redox conditions. However, the potential for generating H{sub 2} and NH{sub 3} during treatment of high-level waste (HLW) with HCOOH was identified at Pacific Northwest Laboratory (PNL). Studies at the University of Georgia, under contract with Savannah River Technology Center (SRTC) and PNL, have verified the catalytic role of noble metals (Pd, Rh, Ru), present in the waste, in the generation of H{sub 2} and NH{sub 3}. Both laboratory-scale and pilot-scale studies at SRTC have documented the H{sub 2} and NH{sub 3} generation phenomenal Because H{sub 2} and NH{sub 3} may create hazardous conditions in the vessel vapor space and offgas system of a vitrification plant, reducing the H{sub 2} generation rate and the NH{sub 3} generation to the lowest possible levels consistent with desired melter feed characteristics is important. The Fiscal Year 1993 and 1994 studies were conducted with simulated (non-radioactive), pre-treated neutralized current acid waste (NCAW). Neutralized current acid waste is a high-level waste originating from the plutonium/uranium extraction (PUREX) plant that has been partially denitrated with sugar, neutralized with NaOH, and is presently stored in double-shell tanks. The non-radioactive simulant used for the present study includes all of the trace components found in the waste, or substitutes a chemically similar element for radioactive or very toxic species. The composition and simulant preparation steps were chosen to best simulate the chemical processing characteristics of the actual waste.

Smith, H.D.; Merz, M.D.; Wiemers, K.D.; Smith, G.L.

1996-02-01T23:59:59.000Z

293

Waste canister for storage of nuclear wastes  

DOE Patents (OSTI)

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

Duffy, James B. (Fullerton, CA)

1977-01-01T23:59:59.000Z

294

General and Localized Corrosion of Outer Barrier of High-Level Waste Container in Yucca Mountain  

DOE Green Energy (OSTI)

As described in the License Application Design Selection Report, the recommended waste, package design is Engineering Design Alternative II (CRWMS M&O 1999). This design includes a double-wall waste package (WP) underneath a protective drip shield (DS). purpose and scope of the process-level model described here is to account for both general and localized corrosion of the waste package outer barrier (WPOB), which assumed to be Alloy 22 (UNS N06022-21Cr-13Mo-4Fe-3W-2C-Ni) (ASTM 1997a). This model will include several sub-models, which will account for dry oxidation (DOX), humid air corrosion (HAC), general corrosion (GC) in the aqueous phase, and localized corrosion (LC) the aqueous phase. This model serves as a feed to the waste package degradation (WAPDEG) code for performance, assessment.

Farmer, J.; McCright, D.; Gdowski, G.; Wang, F.; Summers, T.; Bedrossian, P.; Horn, J.; Lian, T.; Estill, J.; Lingenfelter, A.; Halsey, W.

2000-05-02T23:59:59.000Z

295

State of the Art Report on High-Level Waste Tank Closure  

Science Conference Proceedings (OSTI)

This report includes strategies for treating the incidental waste left in the emptied tanks as non-retrievable heels and methods and materials for physically stabilizing the void space in the tanks to prevent future subsidence.

Langton, C.A.

2002-06-18T23:59:59.000Z

296

High Level Waste Tank Farm Replacement Project for the Idaho Chemical Processing Plant at the Idaho National Engineering Laboratory. Environmental Assessment  

Science Conference Proceedings (OSTI)

The Department of Energy (DOE) has prepared an environmental assessment (EA), DOE/EA-0831, for the construction and operation of the High-Level Waste Tank Farm Replacement (HLWTFR) Project for the Idaho Chemical Processing Plant located at the Idaho National Engineering Laboratory (INEL). The HLWTFR Project as originally proposed by the DOE and as analyzed in this EA included: (1) replacement of five high-level liquid waste storage tanks with four new tanks and (2) the upgrading of existing tank relief piping and high-level liquid waste transfer systems. As a result of the April 1992 decision to discontinue the reprocessing of spent nuclear fuel at INEL, DOE believes that it is unlikely that the tank replacement aspect of the project will be needed in the near term. Therefore, DOE is not proposing to proceed with the replacement of the tanks as described in this-EA. The DOE`s instant decision involves only the proposed upgrades aspect of the project described in this EA. The upgrades are needed to comply with Resource Conservation and Recovery Act, the Idaho Hazardous Waste Management Act requirements, and the Department`s obligations pursuant to the Federal Facilities Compliance Agreement and Consent Order among the Environmental Protection Agency, DOE, and the State of Idaho. The environmental impacts of the proposed upgrades are adequately covered and are bounded by the analysis in this EA. If DOE later proposes to proceed with the tank replacement aspect of the project as described in the EA or as modified, it will undertake appropriate further review pursuant to the National Environmental Policy Act.

Not Available

1993-06-01T23:59:59.000Z

297

Assessment of chemical vulnerabilities in the Hanford high-level waste tanks  

SciTech Connect

The purpose of this report is to summarize results of relevant data (tank farm and laboratory) and analysis related to potential chemical vulnerabilities of the Hanford Site waste tanks. Potential chemical safety vulnerabilities examined include spontaneous runaway reactions, condensed phase waste combustibility, and tank headspace flammability. The major conclusions of the report are the following: Spontaneous runaway reactions are not credible; condensed phase combustion is not likely; and periodic releases of flammable gas can be mitigated by interim stabilization.

Meacham, J.E. [and others

1996-02-15T23:59:59.000Z

298

Treatment of high-level wastes from the IFR fuel cycle  

SciTech Connect

The Integral Fast Reactor (IFR) is being developed as a future commercial power source that promises to have important advantages over present reactors, including improved resource conservation and waste management. The spent metal alloy fuels from an IFR will be processed in an electrochemical cell operating at 500{degree}C with a molten chloride salt electrolyte and cadmium metal anode. After the actinides have been recovered from several batches of core and blanket fuels, the salt cadmium in this electrorefiner will be treated to separate fission products from residual transuranic elements. This treatment produces a waste salt that contains the alkali metal, alkaline earth, and halide fission products; some of the rare earths; and less than 100 nCi/g of alpha activity. The treated metal wastes contain the rest of the fission products (except T, Kr, and Xe) small amounts of uranium, and only trace amounts of transuranic elements. The current concept for the salt waste form is an aluminosilicate matrix, and the concept for the metal waste form is a corrosion-resistant metal alloy. The processes and equipment being developed to treat and immobilize the salt and metal wastes are described.

Johnson, T.R.; Lewis, M.A.; Newman, A.E.; Laidler, J.J.

1992-01-01T23:59:59.000Z

299

Treatment of high-level wastes from the IFR fuel cycle  

Science Conference Proceedings (OSTI)

The Integral Fast Reactor (IFR) is being developed as a future commercial power source that promises to have important advantages over present reactors, including improved resource conservation and waste management. The spent metal alloy fuels from an IFR will be processed in an electrochemical cell operating at 500{degree}C with a molten chloride salt electrolyte and cadmium metal anode. After the actinides have been recovered from several batches of core and blanket fuels, the salt cadmium in this electrorefiner will be treated to separate fission products from residual transuranic elements. This treatment produces a waste salt that contains the alkali metal, alkaline earth, and halide fission products; some of the rare earths; and less than 100 nCi/g of alpha activity. The treated metal wastes contain the rest of the fission products (except T, Kr, and Xe) small amounts of uranium, and only trace amounts of transuranic elements. The current concept for the salt waste form is an aluminosilicate matrix, and the concept for the metal waste form is a corrosion-resistant metal alloy. The processes and equipment being developed to treat and immobilize the salt and metal wastes are described.

Johnson, T.R.; Lewis, M.A.; Newman, A.E.; Laidler, J.J.

1992-08-01T23:59:59.000Z

300

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

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.

Weiner, Ruth F.; Blink, James A. (Lawrence Livermore National Laboratory, Livermore, CA); Rechard, Robert Paul; Perry, Frank (Los Alamos National Laboratory, Los Alamos, NM); Jenkins-Smith, Hank C. (University of Oklahoma, Norman, OK); Carter, Joe (Savannah River Nuclear Solutions, Aiken, SC); Nutt, Mark (Argonne National Laboratory, Argonne, IL); Cotton, Tom (Complex Systems Group, Washington DC)

2010-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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301

Program on Technology Innovation: Advanced Fuel Cycles - Impact on High-Level Waste Disposal  

Science Conference Proceedings (OSTI)

This report presents the results of a dynamic simulation analysis for deployment of advanced light water reactors (LWRs) and fast burner reactors, as proposed by the Global Nuclear Energy Partnership (GNEP) program. Conditions for the analysis were selected for their potential to challenge the nuclear fuel simulation codes that were used, due to the large variations in nuclear fuel composition for the burner reactors before equilibrium conditions are approached. The analysis was performed in a U.S. conte...

2008-09-30T23:59:59.000Z

302

Evaluation of Flygt Propeller Xixers for Double Shell Tank (DST) High Level Waste Auxiliary Solids Mobilization  

Science Conference Proceedings (OSTI)

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.

PACQUET, E.A.

2000-07-20T23:59:59.000Z

303

PERFORMANCE IMPROVEMENT OF CROSS-FLOW FILTRATION FOR HIGH LEVEL WASTE TREATMENT  

Science Conference Proceedings (OSTI)

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.

Duignan, M.; Nash, C.; Poirier, M.

2011-01-12T23:59:59.000Z

304

MELT RATE ENHANCEMENT FOR HIGH ALUMINUM HLW (HIGH LEVEL WASTE) GLASS FORMULATION FINAL REPORT 08R1360-1  

Science Conference Proceedings (OSTI)

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.

KRUGER AA; MATLACK KS; KOT W; PEGG IL; JOSEPH I; BARDAKCI T; GAN H; GONG W; CHAUDHURI M

2010-01-04T23:59:59.000Z

305

Nuclear Waste Policy Act | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Waste Policy Act Nuclear Waste Policy Act Document on the Nuclear Waste Policy Act of 1982 An Act to provide for the development of repositories for the disposal of...

306

PROPOSAL FOR INSTALLATION OF EQUIPMENT FOR TREATMENT OF ORNL HIGH LEVEL WASTE  

SciTech Connect

It is proposed that neutralized ORNL storage tank supernate waste be used as a feed for a waste processing development facility. An electrolytic cell can be installed to remove nitrate and ruthenium from the wastes prior to discharge to the soil disposal pits. This eliminates the hazards of nitrate and ruthenium in the ORNL soil disposal pits. In addition, a continuous ion exchange unit can be installed to study the isolation of fission products from a portion of the waste stream. The total waste stream can be processed by the ion exchange if desired. A six to eight inch diameter continuous ion exchange contactor is considered sufficient to remove cesium, strontium, and rare earths from 5,000 to 10,000 gallons of waste per day. The estimated installed cost of such a unit is about ,000. The labor and chemical operating cost at 100% overhead is estimated at .50 per day. Three types of electrolytic cells were considered. All of the cells remove about 90% of the ruthenium by plating on the cathode. It is expected that the degree of ruthenium removed may be increased by further laboratory development. The first cell type is the Ionics cell which converts NaNO/sub 3/ to NaOH and HNO/sub 3/ for recycle. The second is the nitrate reduction (or destruction) cell which produces caustic and ammonia gas, as studied by KAPL. The acid-base cell is favored because of a factor of 8 lower power consumption, and the nitric acid may be reused as well as sodium hydroxide. An acid-base cell with rectifier for 5,000 gallons per day of 0.3 M NaNO/sub 3/ waste is estimated at about ,000. The operating cost with 100% overhead, including labor, electricity, and cell maintenance is estimated at .72 per day. Operations may be credited with per day of caustic and per day of nitric acid. If it is desired to isolate the ruthenium and not remove the nitrate, a simplified electrolytic cell may be used. Cell cost is estimited at approximates 80% of the acid-base cell and about one tenth the cost for a rectifier. About 90% of the ruthenium may be conveniently removed. Technetium is not considered a radioactive hazard, but recovery is considered because of its value and expected ease of recovery. Technetium is not scavenged and has a very great affinity for anion resin. At current prices it is estimated that 0.25 g per day of technetium is being discharged at ORNL, which at present prices is 000/day. (auth)

Higgins, I.R.

1957-10-24T23:59:59.000Z

307

Chemotoxicity of nuclear waste repositories  

Science Conference Proceedings (OSTI)

In this paper published studies on chemotoxicity of nuclear waste repositories are reviewed. According to these studies, radiotoxicity is of primary concern. However, there also is a risk, primarily from genotoxic chemical substances, which could eventually reach the biosphere. Possible chemotoxic effects should be studied as an integral part of the risk assessment and risk management of repositories for nuclear waste.

Buchheim, B. (Nordostschweizerische Kraftwerke-AG, Parkstrasse 23, CH-5401 Baden (CH)); Persson, L. (Swedish Radiation Protection Inst., P.O. Box 60204, S-104 01 Stockholm (SE))

1992-03-01T23:59:59.000Z

308

Used Fuel and High-Level Radioactive Waste Extended Storage Collaboration Program  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) convened a workshop of over 40 representatives of the nuclear industry, federal government, national laboratories, and suppliers of used-fuel dry-storage systems to discuss the potential issues associated with extended dry storage of used fuel, that is, storage considerably beyond the term of current and recently proposed U.S. Nuclear Regulatory Commission (NRC) regulations. The workshop was held November 18-19, 2009, at EPRI's offices in Washington, DC.

2010-03-02T23:59:59.000Z

309

Evaluation of alternatives for high-level and transuranic radioactive- waste disposal standards  

Science Conference Proceedings (OSTI)

The remand of the US Environmental Protection Agency`s long-term performance standards for radioactive-waste disposal provides an opportunity to suggest modifications that would make the regulation more defensible and remove inconsistencies yet retain the basic structure of the original rule. Proposed modifications are in three specific areas: release and dose limits, probabilistic containment requirements, and transuranic-waste disposal criteria. Examination of the modifications includes discussion of the alternatives, demonstration of methods of development and implementation, comparison of the characteristics, attributes, and deficiencies of possible options within each area, and analysis of the implications for performance assessments. An additional consideration is the impact on the entire regulation when developing or modifying the individual components of the radiological standards.

Klett, R.D. [Sandia National Labs., Albuquerque, NM (United States); Gruebel, M.M. [Tech. Reps., Inc., Albuquerque, NM (United States)

1992-12-01T23:59:59.000Z

310

SOLUBILITY OF URANIUM AND PLUTONIUM IN ALKALINE SAVANNAH RIVER SITE HIGH LEVEL WASTE SOLUTIONS  

Science Conference Proceedings (OSTI)

Five actual Savannah River Site tank waste samples and three chemically-modified samples were tested to determine solubility limits for uranium and plutonium over a one year time period. Observed final uranium concentrations ranged from 7 mg U/L to 4.5 g U/L. Final plutonium concentrations ranged from 4 {micro}g Pu/L to 12 mg Pu/L. Actinide carbonate complexation is believed to result in the dramatic solubility increases observed for one sample over long time periods. Clarkeite, NaUO{sub 2}(O)OH {center_dot} H{sub 2}O, was found to be the dominant uranium solid phase in equilibrium with the waste supernate in most cases.

King, W.; Hobbs, D.; Wilmarth, B.; Edwards, T.

2010-03-10T23:59:59.000Z

311

Consequence assessment for the high-level waste tanks probabilistic risk assessment  

SciTech Connect

At the US DOE Hanford Site, there are 177 underground tanks in 18 separate tank farms containing accumulated liquid radioactive wastes from 50 yr of weapons materials production activities. The total volume is about 60 million gallons containing approximately 120 Curies of radioactivity. The radioactive material consists primarily of {sup 137}Cs, {sup 90}Sr, and transuranics. Risk concerns with the tanks are associated with possible energy releases because of the presence of flammable gases, organic liquids, reactive chemical compounds, and radioactive decay heat. Because of the high concentration of radioactivity in the wastes and because a large number of the older single-shell tanks have some history or evidence of leaking, there is a public perception that they pose a serious risk to the onsite workers and the offsite public. The tank farm probabilistic safety assessment (PSA) was performed for two reasons: (1) to develop a baseline estimate of the risks these wastes pose to the workers and the public for the present tank contents and configurations and (2) to provide a relative ranking of the risks associated with individual groups of tanks. The latter information would be helpful in planning the order of the tank remediation work by indicating which tanks pose the greatest risk; the former could help allay concerns.

MacFarlane, D.R. [Los Alamos National Lab., NM (United States); Kindinger, J.; Deremer, R.K. [PLG, Inc., Newport Beach, CA (United States)

1995-12-31T23:59:59.000Z

312

Small businesses selected for nuclear waste services  

NLE Websites -- All DOE Office Websites (Extended Search)

Small businesses selected for nuclear waste services Small businesses selected for nuclear waste clean-up services Northern New Mexico businesses compete for up to 200 million in...

313

Audit of the Replacement High Level Waste Evaporator at Savannah River, ER-B-95-04  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

OFFICE OF INSPECTOR GENERAL REPORT ON THE AUDIT OF THE REPLACEMENT HIGH LEVEL WASTE EVAPORATOR AT THE SAVANNAH RIVER SITE The Office of Audit Services wants to make the distribution of its audit reports as customer friendly and cost effective as possible. Therefore, this report will be available electroni- cally through the Internet five to seven days after publication at the following alternative addresses: Department of Energy Headquarters Gopher gopher.hr.doc.gov Department of Energy Headquarters Anonymous FTP vm1.hqadmin.doe.gov U.S. Department of Energy Human Resources and Administration

314

Idaho Nuclear Technology and Engineering Center (INTEC) Sodium Bearing Waste - Waste Incidental to Reprocessing Determination  

SciTech Connect

U.S. Department of Energy Manual 435.1-1, Radioactive Waste Management, Section I.1.C, requires that all radioactive waste subject to Department of Energy Order 435.1 be managed as high-level radioactive waste, transuranic waste, or low-level radioactive waste. Determining the radiological classification of the sodium-bearing waste currently in the Idaho Nuclear Technology and Engineering Center Tank Farm Facility inventory is important to its proper treatment and disposition. This report presents the technical basis for making the determination that the sodium-bearing waste is waste incidental to spent fuel reprocessing and should be managed as mixed transuranic waste. This report focuses on the radiological characteristics of the sodiumbearing waste. The report does not address characterization of the nonradiological, hazardous constituents of the waste in accordance with Resource Conservation and Recovery Act requirements.

Jacobson, Victor Levon

2002-08-01T23:59:59.000Z

315

HIGH LEVEL WASTE MECHANCIAL SLUDGE REMOVAL AT THE SAVANNAH RIVER SITE F TANK FARM CLOSURE PROJECT  

SciTech Connect

The Savannah River Site F-Tank Farm Closure project has successfully performed Mechanical Sludge Removal (MSR) using the Waste on Wheels (WOW) system for the first time within one of its storage tanks. The WOW system is designed to be relatively mobile with the ability for many components to be redeployed to multiple waste tanks. It is primarily comprised of Submersible Mixer Pumps (SMPs), Submersible Transfer Pumps (STPs), and a mobile control room with a control panel and variable speed drives. In addition, the project is currently preparing another waste tank for MSR utilizing lessons learned from this previous operational activity. These tanks, designated as Tank 6 and Tank 5 respectively, are Type I waste tanks located in F-Tank Farm (FTF) with a capacity of 2,840 cubic meters (750,000 gallons) each. The construction of these tanks was completed in 1953, and they were placed into waste storage service in 1959. The tank's primary shell is 23 meters (75 feet) in diameter, and 7.5 meters (24.5 feet) in height. Type I tanks have 34 vertically oriented cooling coils and two horizontal cooling coil circuits along the tank floor. Both Tank 5 and Tank 6 received and stored F-PUREX waste during their operating service time before sludge removal was performed. DOE intends to remove from service and operationally close (fill with grout) Tank 5 and Tank 6 and other HLW tanks that do not meet current containment standards. Mechanical Sludge Removal, the first step in the tank closure process, will be followed by chemical cleaning. After obtaining regulatory approval, the tanks will be isolated and filled with grout for long-term stabilization. Mechanical Sludge Removal operations within Tank 6 removed approximately 75% of the original 95,000 liters (25,000 gallons). This sludge material was transferred in batches to an interim storage tank to prepare for vitrification. This operation consisted of eleven (11) Submersible Mixer Pump(s) mixing campaigns and multiple intraarea transfers utilizing STPs from July 2006 to August 2007. This operation and successful removal of sludge material meets requirement of approximately 19,000 to 28,000 liters (5,000 to 7,500 gallons) remaining prior to the Chemical Cleaning process. Removal of the last 35% of sludge was exponentially more difficult, as less and less sludge was available to mobilize and the lighter sludge particles were likely removed during the early mixing campaigns. The removal of the 72,000 liters (19,000 gallons) of sludge was challenging due to a number factors. One primary factor was the complex internal cooling coil array within Tank 6 that obstructed mixer discharge jets and impacted the Effective Cleaning Radius (ECR) of the Submersible Mixer Pumps. Minimal access locations into the tank through tank openings (risers) presented a challenge because the available options for equipment locations were very limited. Mechanical Sludge Removal activities using SMPs caused the sludge to migrate to areas of the tank that were outside of the SMP ECR. Various SMP operational strategies were used to address the challenge of moving sludge from remote areas of the tank to the transfer pump. This paper describes in detail the Mechanical Sludge Removal activities and mitigative solutions to cooling coil obstructions and other challenges. The performance of the WOW system and SMP operational strategies were evaluated and the resulting lessons learned are described for application to future Mechanical Sludge Removal operations.

Jolly, R; Bruce Martin, B

2008-01-15T23:59:59.000Z

316

Noble Metals and Spinel Settling in High Level Waste Glass Melters  

SciTech Connect

In the continuing effort to support the Defense Waste Processing Facility (DWPF), the noble metals issue is addressed. There is an additional concern about the amount of noble metals expected to be present in the future batches that will be considered for vitrification in the DWPF. Several laboratory, as well as melter-scale, studies have been completed by various organizations (mainly PNNL, SRTC, and WVDP in the USA). This letter report statuses the noble metals issue and focuses at the settling of noble metals in melters.

Sundaram, S. K.; Perez, Joseph M.

2000-09-30T23:59:59.000Z

317

Systems engineering management and implementation plan for Project W-464, immobilized high-level waste storage  

SciTech Connect

The Systems Engineering Management and Implementation Plan (SEMIP) for TWRS Project W-46 describes the project implementation of the Tank Waste Remediation System Systems Engineering Management Plan. (TWRS SEMP), Rev. 1. The SEMIP outlines systems engineering (SE) products and processes to be used by the project for technical baseline development. A formal graded approach is used to determine the products necessary for requirements, design, and operational baseline completion. SE management processes are defined, and roles and responsibilities for management processes and major technical baseline elements are documented.

Wecks, M.D.

1998-04-15T23:59:59.000Z

318

Laboratory Report on Performance Evaluation of Key Constituents during Pre-Treatment of High Level Waste Direct Feed  

SciTech Connect

The analytical capabilities of the 222-S Laboratory are tested against the requirements for an optional start up scenario of the Waste Treatment and Immobilization Plant on the Hanford Site. In this case, washed and in-tank leached sludge would be sent directly to the High Level Melter, bypassing Pretreatment. The sludge samples would need to be analyzed for certain key constituents in terms identifying melter-related issues and adjustment needs. The analyses on original tank waste as well as on washed and leached material were performed using five sludge samples from tanks 241-AY-102, 241-AZ-102, 241-AN-106, 241-AW-105, and 241-SY-102. Additionally, solid phase characterization was applied to determine the changes in mineralogy throughout the pre-treatment steps.

Huber, Heinz J.

2013-06-24T23:59:59.000Z

319

Evaluation of alternative chemical additives for high-level waste vitrification feed preparation processing  

SciTech Connect

During the development of the feed processing flowsheet for the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), research had shown that use of formic acid (HCOOH) could accomplish several processing objectives with one chemical addition. These objectives included the decomposition of tetraphenylborate, chemical reduction of mercury, production of acceptable rheological properties in the feed slurry, and controlling the oxidation state of the glass melt pool. However, the DEPF research had not shown that some vitrification slurry feeds had a tendency to evolve hydrogen (H{sub 2}) and ammonia (NH{sub 3}) as the result of catalytic decomposition of CHOOH with noble metals (rhodium, ruthenium, palladium) in the feed. Testing conducted at Pacific Northwest Laboratory and later at the Savannah River Technical Center showed that the H{sub 2} and NH{sub 3} could evolve at appreciable rates and quantities. The explosive nature of H{sub 2} and NH{sub 3} (as ammonium nitrate) warranted significant mitigation control and redesign of both facilities. At the time the explosive gas evolution was discovered, the DWPF was already under construction and an immediate hardware fix in tandem with flowsheet changes was necessary. However, the Hanford Waste Vitrification Plant (HWVP) was in the design phase and could afford to take time to investigate flowsheet manipulations that could solve the problem, rather than a hardware fix. Thus, the HWVP began to investigate alternatives to using HCOOH in the vitrification process. This document describes the selection, evaluation criteria, and strategy used to evaluate the performance of the alternative chemical additives to CHOOH. The status of the evaluation is also discussed.

Seymour, R.G.

1995-06-07T23:59:59.000Z

320

Nuclear Waste Management using Electrometallurgical Technology - Nuclear  

NLE Websites -- All DOE Office Websites (Extended Search)

Technology Technology Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Waste Management using Electrometallurgical Technology Bookmark and Share The NE system engineering activities involve the conceptual design, through the manufacturing and qualification testing of the Mk-IV and Mk-V electrorefiner and the cathode processor. These first-of-a-kind large scale

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


321

Characterization, propagation and analysis of aleatory and epistemic uncertainty in the 2008 performance assessment for the proposed repository for high-level radioactive waste at Yucca Mountain, Nevada  

Science Conference Proceedings (OSTI)

The 2008 performance assessment (PA) for the proposed repository for high-level radioactive waste at Yucca Mountain (YM), Nevada, illustrates the conceptual structure of risk assessments for complex systems. The 2008 YM PA is based on the following three ...

Clifford W. Hansen; Jon C. Helton; Cédric J. Sallaberry

2010-09-01T23:59:59.000Z

322

Nuclear Waste Management Policy in France  

Science Conference Proceedings (OSTI)

Technical Paper / New Directions in Nuclear Energy with Emphasis on Fuel Cycles / Radioactive Waste Management

Jean F. Lefevre

323

Accident analysis for high-level waste management alternatives in the US Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement  

SciTech Connect

A comparative generic accident analysis was performed for the programmatic alternatives for high-level waste (HLW) management in the US Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement (EM PEIS). The key facilities and operations of the five major HLW management phases were considered: current storage, retrieval, pretreatment, treatment, and interim canister storage. A spectrum of accidents covering the risk-dominant accidents was analyzed. Preliminary results are presented for HLW management at the Hanford site. A comparison of these results with those previously advanced shows fair agreement.

Folga, S.; Mueller, C.; Roglans-Ribas, J.

1994-02-01T23:59:59.000Z

324

Advancements in nuclear waste assay.  

E-Print Network (OSTI)

??The research described in this thesis is directed at advancing the state of the practice of the non-destructive gamma-ray assay of nuclear waste containers. A… (more)

Curtis, Deborah Claire

2008-01-01T23:59:59.000Z

325

Pioneering Nuclear Waste Disposal  

NLE Websites -- All DOE Office Websites (Extended Search)

Department of Energy (DOE) is closing the circle on the generation, management, and disposal of transuranic waste. But the WIPP story is not just about radioactive waste. It is...

326

Evaluation of the post-emplacement environment of high level radioactive waste packages at Yucca Mountain, Nevada  

SciTech Connect

Evaluation of the post-emplacement environment around high level radioactive waste containers is required by federal regulations. The information derived from this evaluation will be used to determine the service performance of the waste containers, the chemical and hydrological conditions that may influence radionuclide release and transport if containers are breached, and retrievability of the waste containers prior to closure of the repository. Laboratory studies, numerical simulations, and field experiments and tests are used to provide data necessary for this evaluation. Results obtained to date demonstrate that the post-emplacement environment in the welded tuff at Yucca Mountain, Nevada maintains relatively benign chemical features (i.e., near neutral pH, low concentrations of dissolved species) for most scenarios. The hydrological environment appears to be one of low flow volume and rates for the expected condition of an unsaturated medium. Emplacement borehole stability will be a function of fracture density and orientation, which may be influenced by microcrack development. Field studies and numerical simulations are in progress that will extend the results of laboratory studies to long time periods. The extent to which chemical, hydrological and mechanical processes can be adequately coupled through numerical simulations remains a matter of concern. 18 refs., 4 figs., 1 tab.

Glassley, W.

1989-03-01T23:59:59.000Z

327

U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network (OSTI)

FOR YUCCA MOUNTAIN, NEVADA CAPT Raymond L. Clark, U.S. Public Health Service Team Leader for the Yucca disposal system in Yucca Mountain, Nevada. These standards are found in Part 197 of Title 40 of the Code for the potential spent nuclear fuel and high-level radioactive waste disposal system in Yucca Mountain, Nevada

328

DOE/EIS-0303D; High-Level Waste Tank Closure Draft Environmental Impact Statement (November 2000)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

EIS-0303D EIS-0303D DRAFT November 2000 Summary S-iii COVER SHEET RESPONSIBLE AGENCY: U.S. Department of Energy (DOE) TITLE: Savannah River Site, High-Level Waste Tank Closure Draft Environmental Impact Statement (DOE/EIS-0303D), Aiken, SC. CONTACT: For additional information or to submit comments on this environmental impact statement (EIS), write or call: Andrew R. Grainger, NEPA Compliance Officer U.S. Department of Energy, Savannah River Operations Office Building 742A, Room 183 Aiken, South Carolina 29802 Attention: Tank Closure EIS Local and Nationwide Telephone: (800) 881-7292 Email: nepa@srs.gov The EIS is also available on the internet at: http://tis.eh.doe.gov/nepa/docs/docs.htm For general information on the process that DOE follows in complying with the National Environmental

329

Chemical Speciation of Strontium, Americium, and Curium in High Level Waste: Predictive Modeling of Phase Partitioning During Tank Processing  

DOE Green Energy (OSTI)

The objective of this research project is to measure the effects of organic chelate complexation on the speciation and solubility of Sr and trivalent actinides under strongly basic, high carbonate conditions, similar to those present in high- level waste tanks at U.S. Department of Energy storage sites. We proposed, (1) extension to important chelates not previously studied; (2) studies of completing metal ions; and (3) specific studies using Am(III)/Cm(III). The chelate complexation studies would extend our previous research on EDTA, HEDTA, NTA, and IDA to citrate and oxalate. In addition, we propose to address the possible formation of mixed ligand- ligand complexes for Eu(III) in EDTA-HEDTA, EDTA-NTA, HEDTA-NTA, and ligand-carbonate solutions. The fundamental data on chemical speciation and solubility will be used to develop accurate thermodynamic models which are valid to high ionic strength.

CHOPPIN, GREGORY R.

2003-06-01T23:59:59.000Z

330

Seismic design and evaluation guidelines for the Department of Energy high-level waste storage tanks and appurtenances  

Science Conference Proceedings (OSTI)

This document provides guidelines for the design and evaluation of underground high-level waste storage tanks due to seismic loads. Attempts were made to reflect the knowledge acquired in the last two decades in the areas of defining the ground motion and calculating hydrodynamic loads and dynamic soil pressures for underground tank structures. The application of the analysis approach is illustrated with an example. The guidelines are developed for specific design of underground storage tanks, namely double-shell structures. However, the methodology discussed is applicable for other types of tank structures as well. The application of these and of suitably adjusted versions of these concepts to other structural types will be addressed in a future version of this document.

Bandyopadhyay, K.; Cornell, A.; Costantino, C.; Kennedy, R.; Miller, C.; Veletsos, A.

1993-01-01T23:59:59.000Z

331

Radioactive Waste Management (Minnesota)  

Energy.gov (U.S. Department of Energy (DOE))

This section regulates the transportation and disposal of high-level radioactive waste in Minnesota, and establishes a Nuclear Waste Council to monitor the federal high-level radioactive waste...

332

State of Nevada, Agency for Nuclear Projects/Nuclear Waste Project Office narrative report, July 1--September 30, 1991  

Science Conference Proceedings (OSTI)

The Agency for Nuclear Projects/Nuclear Waste Project Office (NWPO) is the State of Nevada agency designated by State law to monitor and oversee US Department of Energy (DOE) activities relative to the possible siting, construction, operation and closure of a high-level nuclear waste repository at Yucca Mountain and to carry out the State of Nevada`s responsibilities under the Nuclear Waste Policy Act of 1982. During the reporting period the NWPO continued to work toward the five objectives designed to implement the Agency`s oversight responsibilities. (1) Assure that the health and safety of Nevada`s citizens are adequately protected with regard to any federal high-level radioactive waste program within the State. (2) Take the responsibilities and perform the duties of the State of Nevada as described in the Nuclear Waste Policy Act of 1982 (Public Law 97-425) and the Nuclear Waste Policy Amendments Act of 1987. (3) Advise the Governor, the State Commission on Nuclear Projects and the Nevada State Legislature on matters concerning the potential disposal of high-level radioactive waste in the State. (4) Work closely and consult with affected local governments and State agencies. (5) Monitor and evaluate federal planning and activities regarding high-level radioactive waste disposal. Plan and conduct independent State studies regarding the proposed repository.

NONE

1991-12-31T23:59:59.000Z

333

Effective thermal conductivity measurements relevant to deep borehole nuclear waste disposal  

E-Print Network (OSTI)

The objective of this work was to measure the effective thermal conductivity of a number of materials (particle beds, and fluids) proposed for use in and around canisters for disposal of high level nuclear waste in deep ...

Shaikh, Samina

2007-01-01T23:59:59.000Z

334

An evaluation of the feasibility of disposal of nuclear waste in very deep boreholes  

E-Print Network (OSTI)

Deep boreholes, 3 to 5 km into igneous rock, such as granite, are evaluated for next- generation repository use in the disposal of spent nuclear fuel and other high level waste. The primary focus is on the stability and ...

Anderson, Victoria Katherine, 1980-

2004-01-01T23:59:59.000Z

335

Pioneering Nuclear Waste Disposal  

NLE Websites -- All DOE Office Websites (Extended Search)

T h e W a s t e I s o l a t i o n P i l o t P l a n t DOE 1980. Final Environmental Impact Statement, Waste Isolation Pilot Plant. DOE/EIS-0026, Washington, DC, Office of Environmental Management, U.S. Department of Energy. DOE 1981. Waste Isolation Pilot Plant (WIPP): Record of Decision. Federal Register, Vol. 46, No. 18, p. 9162, (46 Federal Register 9162), January 28, 1981. U.S. Department of Energy. DOE 1990. Final Supplement Environmental Impact Statement, Waste Isolation Pilot Plant. DOE/EIS-0026-FS, Washington, DC, Office of Environmental Management, U.S. Department of Energy. DOE 1990. Record of Decision: Waste Isolation Pilot Plant. Federal Register, Vol. 55, No. 121, 25689-25692, U.S. Department of Energy. DOE 1994. Comparative Study of Waste Isolation Pilot Plant (WIPP) Transportation Alternatives.

336

Some logistical considerations in designing a system of deep boreholes for disposal of high-level radioactive waste.  

Science Conference Proceedings (OSTI)

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.

Gray, Genetha Anne; Brady, Patrick Vane [Sandia National Laboratories, Albuquerque, NM; Arnold, Bill Walter [Sandia National Laboratories, Albuquerque, NM

2012-09-01T23:59:59.000Z

337

Some logistical considerations in designing a system of deep boreholes for disposal of high-level radioactive waste.  

SciTech Connect

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.

Gray, Genetha Anne; Brady, Patrick Vane [Sandia National Laboratories, Albuquerque, NM; Arnold, Bill Walter [Sandia National Laboratories, Albuquerque, NM

2012-09-01T23:59:59.000Z

338

The Social and Ethical Aspects of Nuclear Waste  

E-Print Network (OSTI)

issues in nuclear energy: Radioactive waste. La Grange Park,into radioactive waste management ( Nuclear Energy Agency,Nuclear Energy Agency, Radioactive Waste Management

Marshall, Alan

2005-01-01T23:59:59.000Z

339

Materials Science of Nuclear Waste Management I  

Science Conference Proceedings (OSTI)

Mar 6, 2013 ... Separation of the nuclear waste stream into actinides and fission products offers new opportunities for development of ceramic waste forms.

340

Materials and Processes to Immobilize Nuclear Waste  

Science Conference Proceedings (OSTI)

Oct 8, 2012 ... While borosilicate glass is widely regarded as baseline technology for nuclear waste immobilisation, there are a wide range of such wastes that ...

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


341

Pioneering Nuclear Waste Disposal  

NLE Websites -- All DOE Office Websites (Extended Search)

request for further delays After the EPA certified that the WIPP met the standards for disposal of transuranic waste in May 1998, then-New Mexico Attorney General Tom Udall...

342

Precipitation-adsorption process for the decontamination of nuclear waste supernates  

DOE Patents (OSTI)

High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

Lee, L.M.; Kilpatrick, L.L.

1982-05-19T23:59:59.000Z

343

Precipitation-adsorption process for the decontamination of nuclear waste supernates  

DOE Patents (OSTI)

High-level nuclear waste supernate is decontaminated of cesium by precipitation of the cesium and potassium with sodium tetraphenyl boron. Simultaneously, strontium-90 is removed from the waste supernate sorption of insoluble sodium titanate. The waste solution is then filtered to separate the solution decontaminated of cesium and strontium.

Lee, Lien-Mow (North Augusta, SC); Kilpatrick, Lester L. (Aiken, SC)

1984-01-01T23:59:59.000Z

344

IMPACT OF NOBLE METALS AND MERCURY ON HYDROGEN GENERATION DURING HIGH LEVEL WASTE PRETREATMENT AT THE SAVANNAH RIVER SITE  

DOE Green Energy (OSTI)

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.

Stone, M; Tommy Edwards, T; David Koopman, D

2009-03-03T23:59:59.000Z

345

Final Report - High Level Waste Vitrification System Improvements, VSL-07R1010-1, Rev 0, dated 04/16/07  

SciTech Connect

This report describes work conducted to support the development and testing of new glass formulations that extend beyond those that have been previously investigated for the Hanford Waste Treatment and Immobilization Plant (WTP). The principal objective was to investigate maximization of the incorporation of several waste components that are expected to limit waste loading and, consequently, high level waste (HLW) processing rates and canister count. The work was performed with four waste compositions specified by the Office of River Protection (ORP); these wastes contain high concentrations of bismuth, chromium, aluminum, and aluminum plus sodium. The tests were designed to identify glass formulations that maximize waste loading while meeting all processing and product quality requirements. The work included preparation and characterization of crucible melts in support of subsequent DuraMelter 100 (DM100) tests designed to examine the effects of enhanced glass formulations, increased glass processing temperature, increased crystallinity, and feed solids content on waste processing rate and product quality.

Kruger, Albert A.; Gan, H.; Pegg, I. L.; Gong, W.; Champman, C. C.; Joseph, I.; Matlack, K. S.

2013-11-13T23:59:59.000Z

346

Risk assessment for the off-site transportation of high-level waste for the U.S. Department of Energy waste management programmatic environmental impact statement  

Science Conference Proceedings (OSTI)

This report describes the human health risk assessment conducted for the transportation of high-level waste (HLW) in support of the US Department of Energy Waste Management Programmatic Environmental Impact Statement (WM PEIS). The assessment considers risks to collective populations and individuals under both routine and accident transportation conditions for truck and rail shipment modes. The report discusses the scope of the HLW transportation assessment, describes the analytical methods used for the assessment, defines the alternatives considered in the WM PEIS, and details important assessment assumptions. Results are reported for five alternatives. In addition, to aid in the understanding and interpretation of the results, specific areas of uncertainty are described, with an emphasis on how the uncertainties may affect comparisons of the alternatives.

Monette, F.A.; Biwer, B.M.; LePoire, D.J.; Chen, S.Y. [Argonne National Lab., IL (United States). Environmental Assessment Div.

1996-12-01T23:59:59.000Z

347

Review of scenario selection approaches for performance assessment of high-level waste repositories and related issues.  

SciTech Connect

The selection of scenarios representing plausible realizations of the future conditions-with associated probabilities of occurrence-that can affect the long-term performance of a high-level radioactive waste (HLW) repository is the commonly used method for treating the uncertainty in the prediction of the future states of the system. This method, conventionally referred to as the ``scenario approach,`` while common is not the only method to deal with this uncertainty; other method ``ch as the environmental simulation approach (ESA), have also been proposed. Two of the difficulties with the scenario approach are the lack of uniqueness in the definition of the term ``scenario`` and the lack of uniqueness in the approach to formulate scenarios, which relies considerably on subjective judgments. Consequently, it is difficult to assure that a complete and unique set of scenarios can be defined for use in a performance assessment. Because scenarios are key to the determination of the long-term performance of the repository system, this lack of uniqueness can present a considerable challenge when attempting to reconcile the set of scenarios, and their level of detail, obtained using different approaches, particularly among proponents and regulators of a HLW repository.

Banano, E.J. [Beta Corporation International, Albuquerque, NM (United States); Baca, R.G. [Southwest Research Inst., San Antonio, TX (United States). Center for Nuclear Waste Regulatory Analyses

1995-08-01T23:59:59.000Z

348

CRYSTALLIZATION IN HIGH-LEVEL WASTE GLASSES U.S. DEPARTMENT OF ENERGY OFFICE OF RIVER PROTECTION WTP ENGINEERING DIVISION  

SciTech Connect

Various circumstances influence crystallization in glassmaking, for example: (1) crystals nucleate and grow before the glass-forming melt occurs; (2) crystals grow or dissolve in flowing melt and during changing temperature; (3) crystals move under the influence of gravity; (4) crystals agglomerate and interact with gas bubbles; (5) high-level wastes (HLW) are mixtures of a large number of components in unusual proportions; (6) melter processing of HLW and the slow cooling of HLW glass in canisters provides an opportunity for a variety of crystalline forms to precipitate; (7) settling of crystals in a HLW glass melter may produce undesirable sludge at the melter bottom; and (8) crystallization of the glass product may increase, but also ruin chemical durability. The conclusions are: (1) crystal growth and dissolution typically proceed in a convective medium at changing temperature; (2) to represent crystallization or dissolution the kinetics must be expressed in the form of rate equations, such as dC/dt = f(C,T) and the temperature dependence of kinetic coefficients and equilibrium concentrations must be accounted for; and (3) non-equilibrium phenomena commonly occur - metastable crystallization, periodic distribution of crystals; and dendritic crystal growth.

KRUGER AA; HRMA PR

2009-08-19T23:59:59.000Z

349

Seismic design and evaluation guidelines for the Department of Energy High-Level Waste Storage Tanks and Appurtenances  

Science Conference Proceedings (OSTI)

This document provides seismic design and evaluation guidelines for underground high-level waste storage tanks. The guidelines reflect the knowledge acquired in the last two decades in defining seismic ground motion and calculating hydrodynamic loads, dynamic soil pressures and other loads for underground tank structures, piping and equipment. The application of the guidelines is illustrated with examples. The guidelines are developed for a specific design of underground storage tanks, namely double-shell structures. However, the methodology discussed is applicable for other types of tank structures as well. The application of these and of suitably adjusted versions of these concepts to other structural types will be addressed in a future version of this document. The original version of this document was published in January 1993. Since then, additional studies have been performed in several areas and the results are included in this revision. Comments received from the users are also addressed. Fundamental concepts supporting the basic seismic criteria contained in the original version have since then been incorporated and published in DOE-STD-1020-94 and its technical basis documents. This information has been deleted in the current revision.

Bandyopadhyay, K.; Cornell, A.; Costantino, C.; Kennedy, R.; Miller, C.; Veletsos, A.

1995-10-01T23:59:59.000Z

350

Unit Process Modeling [Nuclear Waste Management using Electrometallurg...  

NLE Websites -- All DOE Office Websites (Extended Search)

Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form Modeling Departments Engineering...

351

Effect of aluminum and silicon reactants and process parameters on glass-ceramic waste form characteristics for immobilization of high-level fluorinel-sodium calcined waste  

SciTech Connect

In this report, the effects of aluminum and silicon reactants, process soak time and the initial calcine particle size on glass-ceramic waste form characteristics for immobilization of the high-level fluorinel-sodium calcined waste stored at the Idaho Chemical Processing Plant (ICPP) are investigated. The waste form characteristics include density, total and normalized elemental leach rates, and microstructure. Glass-ceramic waste forms were prepared by hot isostatically pressing (HIPing) a pre-compacted mixture of pilot plant fluorinel-sodium calcine, Al, and Si metal powders at 1050{degrees}C, 20,000 psi for 4 hours. One of the formulations with 2 wt % Al was HIPed for 4, 8, 16 and 24 hours at the same temperature and pressure. The calcine particle size range include as calcined particle size smaller than 600 {mu}m (finer than {minus}30 mesh, or 215 {mu}m Mass Median Diameter, MMD) and 180 {mu}m (finer than 80 mesh, or 49 {mu}m MMD).

Vinjamuri, K.

1993-06-01T23:59:59.000Z

352

The Social and Ethical Aspects of Nuclear Waste  

E-Print Network (OSTI)

people feel toward nuclear weapons seem to have generalizedwaste left over from nuclear weapons and nuclear powerfor nuclear waste facility planners to derail weapons/waste

Marshall, Alan

2005-01-01T23:59:59.000Z

353

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network (OSTI)

in the assay of waste materials generated in the decommissioning of nuclear installations in which one would

354

Environmental Degradation of Nuclear Waste Storage Canister ...  

Science Conference Proceedings (OSTI)

Abstract Scope, The interaction between mild steel nuclear waste storage containers ... Durable and Highly Efficient Energy-harvesting Electrochromic Window ...

355

The U.S. Nuclear Waste Technical Review Board Status Update  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

NWTRB NWTRB www.nwtrb.gov U.S. Nuclear Waste Technical Review Board The U S Nuclear Waste Technical Review Board The U.S. Nuclear Waste Technical Review Board Status Update Presented to: National Transportation Stakeholders Forum Presented By: National Transportation Stakeholders Forum Mark Abkowitz May 11, 2011 The Board's Statutory Mandate * The 1987 amendments to the Nuclear Waste Policy Act (NWPA) established the U S Nuclear Waste Technical Review Board established the U.S. Nuclear Waste Technical Review Board. * The Board evaluates the technical and scientific validity of DOE activities related to: - transportation, packaging and storage of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) - site characterization, design, development, and operations of facilities for

356

Development of an NDA system for high-level waste from the Chernobyl new safe confinement construction site  

SciTech Connect

In early 2009, preliminary excavation work has begun in preparation for the construction of the New Safe Confinement (NSC) at the Chernobyl Nuclear Power Plant (ChNPP) in Ukraine. The NSC is the structure that will replace the present containment structure and will confine the radioactive remains of the ChNPP Unit-4 reactor for the next 100 years. It is expected that special nuclear material (SNM) that was ejected from the Unit-4 reactor during the accident in 1986 could be uncovered and would therefore need to be safeguarded. ChNPP requested the assistance of the United States Department of Energy/National Nuclear Security Administration (NNSA) with developing a new non-destructive assay (NDA) system that is capable of assaying radioactive debris stored in 55-gallon drums. The design of the system has to be tailored to the unique circumstances and work processes at the NSC construction site and the ChNPP. This paper describes the Chernobyl Drum Assay System (CDAS), the solution devised by Los Alamos National Laboratory, Sonalysts Inc., and the ChNPP, under NNSA's International Safeguards and Engagement Program (INSEP). The neutron counter measures the spontaneous fission neutrons from the {sup 238}U, {sup 240}Pu, {sup 244}Cm in a waste drum and estimates the mass contents of the SNMs in the drum by using of isotopic compositions determined by fuel burnup. The preliminary evaluation on overall measurement uncertainty shows that the system meets design performance requirements imposed by the facility.

Lee, Sang-yoon [Los Alamos National Laboratory; Browne, Michael C [Los Alamos National Laboratory; Rael, Carlos D [Los Alamos National Laboratory; Carroll, Colin J [SONALYST INC.; Sunshine, Alexander [NA-243; Novikov, Alexander [CHNPP; Lebedev, Evgeny [CHNPP

2010-01-01T23:59:59.000Z

357

Nuclear waste/nuclear power: their futures are linked  

SciTech Connect

This paper briefly reviews current aspects of radioactive waste disposal techniques and transportation. Addressed are high-level and low-level radioactive wastes, interim spent fuel storage and transportation. The waste options being explored by DOE are listed. Problems of public acceptance will be more difficult to overcome than technical problems. (DMC)

Skoblar, L.T.

1981-01-01T23:59:59.000Z

358

Nuclear Waste Policy Act Signed | National Nuclear Security Administra...  

NLE Websites -- All DOE Office Websites (Extended Search)

Waste Policy Act Signed | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response...

359

Statement of work for conceptual design of solidified high-level waste interim storage system project (phase I)  

SciTech Connect

The U.S. Department of Energy (DOE) has embarked upon a course to acquire Hanford Site tank waste treatment and immobilization services using privatized facilities. This plan contains a two phased approach. Phase I is a ``proof-of-principle/commercial demonstration- scale`` effort and Phase II is a full-scale production effort. In accordance with the planned approach, interim storage (IS) and disposal of various products from privatized facilities are to be DOE furnished. The path forward adopted for Phase I solidification HLW IS entails use of Vaults 2 and 3 in the Spent Nuclear Fuel Canister Storage Building, to be located in the Hanford Site 200 East Area. This Statement of Work describes the work scope to be performed by the Architect-Engineer to prepare a conceptual design for the solidified HLW IS System.

Calmus, R.B., Westinghouse Hanford

1996-12-17T23:59:59.000Z

360

Nuclear waste management. Quarterly progress report, January-March, 1981  

SciTech Connect

Reports and summaries are provided for the following programs: high-level waste process development; alternative waste forms; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton solidification; thermal outgassing; iodine-129 fixation; NWVP off-gas analysis; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; verification instrument development; mobility of organic complexes of radionuclide in soils; low-level waste generation reduction handbook; waste management system studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology program; high-level waste form preparation; development of backfill materials; development of structural engineered barriers; disposal charge analysis; analysis of spent fuel policy implementation; spent fuel and pool component integrity program; analysis of postulated criticality events in a storage array of spent LWR fuel; asphalt emulsion sealing of uranium mill tailings; liner evaluation for uranium mill tailings; multilayer barriers for sealing of uranium tailings; application of long-term chemical biobarriers for uranium tailings; and revegetation of inactive uranium tailings sites.

Chikalla, T.D.; Powell, J.A. (comp.)

1981-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


361

NUCLEAR WASTE CONSULTANTS, INC.  

E-Print Network (OSTI)

Attached please find the Subtask 1.2 Management, prepared by Water, Waste bibliographical information relevant on NNWSI, now comprising 356 titles. scheduled, semi-annual update report contract. Update Report: Data Inventory and and Land (WWL). The report presents the to the all documents in the WWL library Please note that this is the last for Subtask 1.2 under the current The WWL report has received a managerial review by M. Logsdon (NWC), and the report was prepared under WWL's QA procedures, consistent with the NWC QA-manual. 009-1.2-- NNWSI DATABASE- Aucust IS- 19M-2 Auou~~~~ ~ _ _ _ 5._,1_88

Technical Review Branch

1988-01-01T23:59:59.000Z

362

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

SciTech Connect

High-level waste (HLW) throughput (i.e., the amount of waste processed per unit time) is a function of two critical parameters: waste loading (WL) and melt rate. For the Waste Treatment and Immobilization Plant (WTP) at the Hanford Site and the Defense Waste Processing Facility (DWPF) at the Savannah River Site (SRS), increasing HLW throughput would significantly reduce the overall mission life cycle costs for the Department of Energy (DOE). The objective of this study was to generate supplemental validation data that could be used to determine the applicability of the current liquidus temperature (TL) model to expanded DWPF glass composition regions of interest based on higher WLs. Two specific flowsheets were used in this study to provide such insight: (1) Higher WL glasses (45 and 50%) based on future sludge batches that have (and have not) undergone the Al-dissolution process. (2) Coupled operations supported by the Salt Waste Processing Facility (SWPF), which increase the TiO{sub 2} concentration in glass to greater than 2 wt%. Glasses were also selected to address technical issues associated with Al{sub 2}O{sub 3} solubility, nepheline formation, and homogeneity issues for coupled operations. A test matrix of 28 glass compositions was developed to provide insight into these issues. The glasses were fabricated and characterized using chemical composition analysis, X-ray Diffraction (XRD), TL measurement and the Product Consistency Test (PCT). The results of this study are summarized below: (1) TiO{sub 2} concentrations up to {approx} 3.5 wt% were retained in DWPF type glasses, where retention is defined as the absence of crystalline TiO{sub 2} (i.e., unreacted or undissolved) in the as-fabricated glasses. Although this TiO{sub 2} content does not bound the projected SWPF high output flowsheet (up to 6 wt% TiO{sub 2} may be required in glass), these data demonstrate the potential for increasing the TiO{sub 2} limit in glass above the current limit of 2 wt% (based strictly on retention or solubility). (2) For those study glasses that had very close compositional overlap with the model development and/or model validation ranges of the current DWPF TL model (except TiO{sub 2} and MgO concentrations), there was very little difference in the predicted and measured TL values. Even though the TiO{sub 2} concentrations were above the 2 wt% upper limit, the results indicate that the current T{sub L} model is applicable in this compositional region with TiO{sub 2} contents up to approximately 3.5 wt%. (3) As the target glass compositions diverge from the model development and validation ranges, the T{sub L} data suggest that the model under-predicted the measured values. These discrepancies imply that there are individual oxides or oxide combinations that need to be accounted for in the model. These oxides include B{sub 2}O{sub 3}, SiO{sub 2}, MnO, TiO{sub 2} and/or their combinations. More data would be required to fill in these anticipated DWPF compositional regions for higher WL glasses so that the model coefficients could be refit to account for these differences. (4) Based on PCT response of HWL-21 and HWL-22 (two glasses that were prone to nepheline formation) it appears that increasing the B{sub 2}O{sub 3} concentration in glass does not consistently suppress the formation of nepheline in glasses with higher Al{sub 2}O{sub 3} and/or Na{sub 2}O content. Although the chemical durabilities of the quenched versions of these glasses were very acceptable, the canister centerline cooled (ccc) glasses exhibited a considerable decrease in durability and were found to contain nepheline via XRD. In fact, one of the glasses had a release that was 5 times greater than that of the Environmental Assessment (EA) benchmark glass. These results suggest a need for a more fundamental understanding of the compositional and kinetic effects of nepheline formation in high WL glasses. (5) Data have been generated in support of the replacement of the homogeneity constraint with the Al{sub 2}O{sub 3} and/or sum of alkali constraints for coupled o

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

2008-10-23T23:59:59.000Z

363

Property/composition relationships for Hanford high-level waste glasses melting at 115{degrees}C volume 1: Chapters 1-11  

Science Conference Proceedings (OSTI)

A Composition Variation study (CVS) is being performed within the Pacific Northwest Laboratory Vitrification Technology Development (PVTD) project in support of a future high-level nuclear waste vitrification plant at the Hanford site in Washington. From 1989 to 1994, over 120 nonradioactive glasses were melted and properties measured in five statistically-designed experimental phases. Glass composition is represented by the 10 components SiO{sub 2}, B{sub 2}O{sub 3}, Al{sub 2}O{sub 3}, Fe{sub 2}O{sub 3}, ZrO{sub 2}, Na{sub 2}O, Li{sub 2}O, CaO, MgO, and Others (all remaining components). The properties measured include viscosity ({eta}), electrical conductivity ({epsilon}), glass transition temperature (T{sub g} ), thermal expansion of solid glass ({alpha}{sub s}) and molten glass ({alpha}{sub m}), crystallinity (quenched and canister centerline cooled glasses), liquidus temperature (T{sub L}), durability based on normalized elemental releases from the Materials Characterization Center-1 28-day dissolution test (MCC-1, r{sub mi}) and the 7-day Product Consistency Test (PCT, r{sub pi}), and solution pHs from MCC-1 and PCT. Amorphous phase separation was also evaluated. Empirical first- and second-order mixture models were fit using the CVS data to relate the various properties to glass composition. Equations for calculating the uncertainty associated with property values predicted by the models were also developed. The models were validated using both internal and external data. Other modeling approaches (e.g., non-bridging oxygen, free energy of hydration, phase-equilibria T{sub L}) were investigated for specific properties. A preliminary Qualified Composition Region was developed to identify glass compositions with high confidence of being processable in a melter and meeting waste form acceptance criteria.

Hrma, P.R.; Piepel, G.F.

1994-12-01T23:59:59.000Z

364

The Social and Ethical Aspects of Nuclear Waste  

E-Print Network (OSTI)

2002). Current issues in nuclear energy: Radioactive waste.trans/trfact03.htm Nuclear Energy Agency. (1995). TheSweden: Komentus. Nuclear Energy Agency, Radioactive Waste

Marshall, Alan

2005-01-01T23:59:59.000Z

365

MEASUREMENT AND CALCULATION OF RADIONUCLIDE ACTIVITIES IN SAVANNAH RIVER SITE HIGH LEVEL WASTE SLUDGE FOR ACCEPTANCE OF DEFENSE WASTE PROCESSING FACILITY GLASS IN A FEDERAL REPOSITORY  

SciTech Connect

This paper describes the results of the analyses of High Level Waste (HLW) sludge slurry samples and of the calculations necessary to decay the radionuclides to meet the reporting requirement in the Waste Acceptance Product Specifications (WAPS) [1]. The concentrations of 45 radionuclides were measured. The results of these analyses provide input for radioactive decay calculations used to project the radionuclide inventory at the specified index years, 2015 and 3115. This information is necessary to complete the Production Records at Savannah River Site's Defense Waste Processing Facility (DWPF) so that the final glass product resulting from Macrobatch 5 (MB5) can eventually be submitted to a Federal Repository. Five of the necessary input radionuclides for the decay calculations could not be measured directly due to their low concentrations and/or analytical interferences. These isotopes are Nb-93m, Pd-107, Cd-113m, Cs-135, and Cm-248. Methods for calculating these species from concentrations of appropriate other radionuclides will be discussed. Also the average age of the MB5 HLW had to be calculated from decay of Sr-90 in order to predict the initial concentration of Nb-93m. As a result of the measurements and calculations, thirty-one WAPS reportable radioactive isotopes were identified for MB5. The total activity of MB5 sludge solids will decrease from 1.6E+04 {micro}Ci (1 {micro}Ci = 3.7E+04 Bq) per gram of total solids in 2008 to 2.3E+01 {micro}Ci per gram of total solids in 3115, a decrease of approximately 700 fold. Finally, evidence will be given for the low observed concentrations of the radionuclides Tc-99, I-129, and Sm-151 in the HLW sludges. These radionuclides were reduced in the MB5 sludge slurry to a fraction of their expected production levels due to SRS processing conditions.

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

2008-12-31T23:59:59.000Z

366

Nuclear Utility Mixed Waste Stream Characterization Study  

Science Conference Proceedings (OSTI)

This report presents industry experience at nuclear utilities in characterizing the hazardous component of potential mixed waste streams. It identifies key considerations for characterizing mixed waste; provides background information, including actual sample results, on the majority of plant processes with a potential to generate mixed waste; and presents a methodology for characterizing mixed waste.

1994-12-31T23:59:59.000Z

367

DOE/EIS-0287 Idaho High-Level Waste & Facilities Disposition Draft Environmental Impact Statement (December 1999)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

HLW & FD EIS HLW & FD EIS 3-13 DOE/EIS-0287D Calcine storag e i n b i n s ets Calcine storag e i n b i n s et s Cesium ion exchange & grouting Cesium ion exchange & grouting NWCF* NWCF* Calcine Mixed transuranic waste/SBW Mixed transuranic waste/NGLW Low-level waste disposa l*** disposa l*** Tank heels Transuranic waste (from tank heels) * * * * Mixed transuranic waste/ NGLW Mixed transuranic waste/ NGLW M i x e d t r a nsuran ic w a s t e / M i x e d t r a nsuran ic w a s t e / S B W s t o rage in Ta n k F a r m S B W s t o rage in Ta n k F a r m Low-leve l waste Low-leve l waste FIGURE 3-2. Continued Current Operations Alternative. LEGEND * Including high-temperature and maximum achievable control technology upgrades. Mixed transuranic waste/ newly generated liquid waste New Waste Calcining Facility ** Calcine would be transferred from bin set #1 to bin set #6 or #7.

368

Approach for enhancing nuclear materials tracking and reporting in waste  

SciTech Connect

Recent policy from the Department of Energy/Office of Safeguards and Security (DOE/OSS) has identified the need to report nuclear materials in waste in a manner that is consistent with the Department of Energy's Nuclear Materials Information System (NMIS), which uses Form 471 as its official record. NMIS is used to track nuclear material inventories while they are subject to safeguards. This requirement necessitates the reevaluation of existing business practices that are used to track and report these nuclear materials. This paper provides a methodology for applying a systems approach to the evaluation of the flow of nuclear waste materials from a generating facility through to permanent disposal. This methodology can be used to integrate existing systems and leverage data already gathered that support both the waste reporting requirements and the NMIS requirements. In order to consider an active waste reporting system that covers waste management through to final disposal, the requirements for characterization, certification, and transportation for disposal at the Waste Isolation Pilot Plant (WIPP) are used as an example. These requirements are found in the WIPP Waste Acceptance Criteria (WIPP/WAC) and associated requirement documents. This approach will prevent inconsistencies in reported data and address current and future needs. For example, spent fuel (which the U.S. intends to dispose of as high-level waste) has not been viewed as particularly attractive in terms of proliferation in comparison to materials associated with other parts of the nuclear fuel cycle. However, collecting high-level waste (or some types of defense waste) in one location where it will be left for hundreds or thousands of years presents proliferation and safeguards issues that need to be considered as part of a systems evaluation. This paper brings together information on domestic and international safeguards practices and considers the current system of documentation used by the U.S. Department of Energy for radioactive waste disposal. The information presented represents current practices, and we recognize that the practices were designed to address different goals. After providing an overview of these areas, some steps that may help develop safeguards systems for geologic repositories in the U.S. context are discussed.

Longmire, V. L. (Victoria L.); Seitz, S. L. (Sharon L.); Sinkule, B. J. (Barbara J.)

2001-06-01T23:59:59.000Z

369

Uranium immobilization and nuclear waste  

SciTech Connect

Considerable information useful in nuclear waste storage can be gained by studying the conditions of uranium ore deposit formation. Further information can be gained by comparing the chemistry of uranium to nuclear fission products and other radionuclides of concern to nuclear waste disposal. Redox state appears to be the most important variable in controlling uranium solubility, especially at near neutral pH, which is characteristic of most ground water. This is probably also true of neptunium, plutonium, and technetium. Further, redox conditions that immobilize uranium should immobilize these elements. The mechanisms that have produced uranium ore bodies in the Earth's crust are somewhat less clear. At the temperatures of hydrothermal uranium deposits, equilibrium models are probably adequate, aqueous uranium (VI) being reduced and precipitated by interaction with ferrous-iron-bearing oxides and silicates. In lower temperature roll-type uranium deposits, overall equilibrium may not have been achieved. The involvement of sulfate-reducing bacteria in ore-body formation has been postulated, but is uncertain. Reduced sulfur species do, however, appear to be involved in much of the low temperature uranium precipitation. Assessment of the possibility of uranium transport in natural ground water is complicated because the system is generally not in overall equilibrium. For this reason, Eh measurements are of limited value. If a ground water is to be capable of reducing uranium, it must contain ions capable of reducing uranium both thermodynamically and kinetically. At present, the best candidates are reduced sulfur species.

Duffy, C.J.; Ogard, A.E.

1982-02-01T23:59:59.000Z

370

Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (DOE/EIS-0287) (11/28/06)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

811 Federal Register 811 Federal Register / Vol. 71, No. 228 / Tuesday, November 28, 2006 / Notices Information Relay Service (FIRS) at 1-800-877-8339. [FR Doc. E6-20124 Filed 11-27-06; 8:45 am] BILLING CODE 4000-01-P DEPARTMENT OF ENERGY Amended Record of Decision: Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement AGENCY: Department of Energy. ACTION: Amended Record of Decision. SUMMARY: The U.S. Department of Energy (DOE) is amending its Record of Decision (ROD) published December 19, 2005 (70 Federal Register [FR] 75165), pursuant to the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (Final EIS) (DOE/EIS-0287, September 2002). The Final EIS analyzed two sets of alternatives for accomplishing DOE's

371

DOE/EIS-0287-SA-01: Supplement Analysis for the Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement (June 2005)  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

7 7 -SA-Ol SUPPLEMENT ANALYSIS For The Idaho High-Level Waste and Facilities Disposition Final Environmental Impact Statement June 2005 United States Department of Energy Idaho Operations Office 1.0 2.0 3.0 4.0 5.0 6.0 DOEÆIS-0287 -SA-O 1 TABLE OF CONTENTS Introduction......................................................................................................................... 4 Background......................................................................................................................... 4 Areas of Review.................................................................................................................. 6 3.1 3.2 3.3 3.4 Proposed Waste Treatment Technology.......... .......................................................

372

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

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.

Croff, A.G.; Liberman, M.S.; Morrison, G.W.

1982-01-01T23:59:59.000Z

373

STATUS OF THE DEVELOPMENT OF IN-TANK/AT-TANK SEPARATIONS TECHNOLOGIES FOR FOR HIGH-LEVEL WASTE PROCESSING FOR THE U.S. DEPARTMENT OF ENERGY  

SciTech Connect

Within the U.S. Department of Energy's (DOE) Office of Technology Innovation and Development, the Office of Waste Processing manages a research and development program related to the treatment and disposition of radioactive waste. At the Savannah River (South Carolina) and Hanford (Washington) Sites, approximately 90 million gallons of waste are distributed among 226 storage tanks (grouped or collocated in 'tank farms'). This waste may be considered to contain mixed and stratified high activity and low activity constituent waste liquids, salts and sludges that are collectively managed as high level waste (HLW). A large majority of these wastes and associated facilities are unique to the DOE, meaning many of the programs to treat these materials are 'first-of-a-kind' and unprecedented in scope and complexity. As a result, the technologies required to disposition these wastes must be developed from basic principles, or require significant re-engineering to adapt to DOE's specific applications. Of particular interest recently, the development of In-tank or At-Tank separation processes have the potential to treat waste with high returns on financial investment. The primary objective associated with In-Tank or At-Tank separation processes is to accelerate waste processing. Insertion of the technologies will (1) maximize available tank space to efficiently support permanent waste disposition including vitrification; (2) treat problematic waste prior to transfer to the primary processing facilities at either site (i.e., Hanford's Waste Treatment and Immobilization Plant (WTP) or Savannah River's Salt Waste Processing Facility (SWPF)); and (3) create a parallel treatment process to shorten the overall treatment duration. This paper will review the status of several of the R&D projects being developed by the U.S. DOE including insertion of the ion exchange (IX) technologies, such as Small Column Ion Exchange (SCIX) at Savannah River. This has the potential to align the salt and sludge processing life cycle, thereby reducing the Defense Waste Processing Facility (DWPF) mission by 7 years. Additionally at the Hanford site, problematic waste streams, such as high boehmite and phosphate wastes, could be treated prior to receipt by WTP and thus dramatically improve the capacity of the facility to process HLW. Treatment of boehmite by continuous sludge leaching (CSL) before receipt by WTP will dramatically reduce the process cycle time for the WTP pretreatment facility, while treatment of phosphate will significantly reduce the number of HLW borosilicate glass canisters produced at the WTP. These and other promising technologies will be discussed.

Aaron, G.; Wilmarth, B.

2011-09-19T23:59:59.000Z

374

Action plan for response to abnormal conditions in Hanford high level radioactive liquid waste storage tanks containing flammable gases. Revision 1  

DOE Green Energy (OSTI)

Radioactive liquid waste tends to produce hydrogen as a result of the interaction of gamma radiation and water. In tanks containing organic chelating agents, additional hydrogen gas as well as nitrous oxide and ammonia can be produced by thermal and radiolytic decomposition of these organics. Several high-level radioactive liquid waste storage tanks, located underground at the Hanford Site, contain waste that retains the gases produced in them until large quantities are released rapidly to the tank vapor space. Tanks filled to near capacity have relatively little vapor space; therefore, if the waste suddenly releases a large amount of hydrogen and nitrous oxide, a flammable gas mixture may result. The most notable waste tank with a flammable gas problem is tank 241-SY-101. Waste in this tank has occasionally released enough flammable gas to burn if an ignition source had been present inside of the tank. Several other waste tanks exhibit similar behavior to a lesser magnitude. Administrative controls have been developed to assure that these Flammable Gas Watch List tanks are safely maintained. Responses have also been developed for off-normal conditions which might develop in these tanks. In addition, scientific and engineering studies are underway to further understand and mitigate the behavior of the Flammable Gas Watch List tanks.

Sherwood, D.J.

1994-03-01T23:59:59.000Z

375

Basic Research for Evaluating Nuclear Waste Form Performance  

Science Conference Proceedings (OSTI)

Technical Paper / Argonne National Laboratory Specialists’ Workshop on Basic Research Needs for Nuclear Waste Management / Radioactive Waste

Don J. Bradley

376

The Nuclear Waste Policy Act, as amended, with appropriations acts appended. Revision 1  

SciTech Connect

This act provides for the development of repositories for the disposal of high-level radioactive wastes, low-level radioactive wastes, and spent nuclear fuels. In addition, it establishes research and development programs, as well as demonstration programs regarding the disposal of these wastes. This Act consists of the Act of Jan. 7, 1983 (Public Law 97-425; 96 Stat. 2201), as amended by Public Law 100-203 and Public Law 102-486.

NONE

1995-02-01T23:59:59.000Z

377

Nuclear Waste and the Distant Future Nuclear Waste and the Distant Future  

E-Print Network (OSTI)

Nuclear Waste and the Distant Future 1 Nuclear Waste and the Distant Future PER F. PETERSON WILLIAM://www.issues.org/22.4/peterson.html Regulation of nuclear hazards must be consistent with rules governing other of the radioactive material generated by nuclear energy decays away over short times ranging from minutes to several

Kammen, Daniel M.

378

Nuclear waste vitrification efficiency: cold cap reactions  

SciTech Connect

The cost and schedule of nuclear waste treatment and immobilization are greatly affected by the rate of glass production. Various factors influence the performance of a waste-glass melter. One of the most significant, and also one of the least understood, is the process of batch melting. Studies are being conducted to gain fundamental understanding of the batch reactions, particularly those that influence the rate of melting, and models are being developed to link batch makeup and melter operation to the melting rate. Batch melting takes place within the cold cap, i.e., a batch layer floating on the surface of molten glass. The conversion of batch to glass consists of various chemical reactions, phase transitions, and diffusion-controlled processes. These include water evaporation (slurry feed contains as high as 60% water), gas evolution, the melting of salts, the formation of borate melt, reactions of borate melt with molten salts and with amorphous oxides (Fe2O3 and Al2O3), the formation of intermediate crystalline phases, the formation of a continuous glass-forming melt, the growth and collapse of primary foam, and the dissolution of residual solids. To this list we also need to add the formation of secondary foam that originates from molten glass but accumulates on the bottom of the cold cap. This study presents relevant data obtained for a high-level-waste melter feed and introduces a one-dimensional (1D) mathematical model of the cold cap as a step toward an advanced three-dimensional (3D) version for a complete model of the waste glass melter. The 1D model describes the batch-to-glass conversion within the cold cap as it progresses in a vertical direction. With constitutive equations and key parameters based on measured data, and simplified boundary conditions on the cold-cap interfaces with the glass melt and the plenum space of the melter, the model provides sensitivity analysis of the response of the cold cap to the batch makeup and melter conditions. The model demonstrates that batch foaming has a decisive influence on the rate of melting. Understanding the dynamics of the foam layer at the bottom of the cold cap and the heat transfer through it appears crucial for a reliable prediction of the rate of melting as a function of the melter-feed makeup and melter operation parameters. Although the study is focused on a batch for waste vitrification, the authors expect that the outcome will also be relevant for commercial glass melting.

Hrma, Pavel R.; Kruger, Albert A.; Pokorny, Richard

2012-12-15T23:59:59.000Z

379

Report on the EH special review of occupational safety and health programs for the Hanford high-level waste tanks  

DOE Green Energy (OSTI)

This report documents a review of the occupational safety and health programs currently in effect at the tank farms located at the Hanford Reservation in Washington State. The review was undertaken by the Office of Environment, Safety and Health (EH) at the request of the Office of Environmental Restoration and Waste Management (EM) to determine whether issues important to worker safety and health have been properly identified, prioritized, and resolved. To carry out this task, four categories of safety and health activities were reviewed. The first category involved examination of management issues associated with the toxicity, flammability, and explosivity of gases and vapors generated by wastes. The second focused on waste tank characterization and technology programs that have been initiated at Hanford to identify, characterize and mitigate hazards associated with tanks generating hydrogen or containing ferrocyanides or organic compounds. The third evaluated worker safety and health programs necessary to support hazardous waste site operations it the tank farms. The fourth dealt with interim corrective actions currently in place, pending full implementation of site-specific safety and health programs to address waste vapor toxicity, flammability, and explosivity hazards.

Not Available

1992-11-01T23:59:59.000Z

380

DOWNSTREAM IMPACTS OF SLUDGE MASS REDUCTION VIA ALUMINUM DISSOLUTION ON DWPF PROCESSING OF SAVANNAH RIVER SITE HIGH LEVEL WASTE - 9382  

Science Conference Proceedings (OSTI)

The SRS sludge that was to become a major fraction of Sludge Batch 5 (SB5) for the Defense Waste Processing Facility (DWPF) contained a large fraction of H-Modified PUREX (HM) sludge, containing a large fraction of aluminum compounds that could adversely impact the processing and increase the vitrified waste volume. It is beneficial to reduce the non-radioactive fraction of the sludge to minimize the number of glass waste canisters that must be sent to a Federal Repository. Removal of aluminum compounds, such as boehmite and gibbsite, from sludge can be performed with the addition of NaOH solution and heating the sludge for several days. Preparation of SB5 involved adding sodium hydroxide directly to the waste tank and heating the contents to a moderate temperature through slurry pump operation to remove a fraction of this aluminum. The Savannah River National Laboratory (SRNL) was tasked with demonstrating this process on actual tank waste sludge in our Shielded Cells Facility. This paper evaluates some of the impacts of aluminum dissolution on sludge washing and DWPF processing by comparing sludge processing with and without aluminum dissolution. It was necessary to demonstrate these steps to ensure that the aluminum removal process would not adversely impact the chemical and physical properties of the sludge which could result in slower processing or process upsets in the DWPF.

Pareizs, J; Cj Bannochie, C; Michael Hay, M; Daniel McCabe, D

2009-01-14T23:59:59.000Z

Note: This page contains sample records for the topic "high-level nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


381

Nuclear Waste Management. Semiannual progress report, October 1984-March 1985  

Science Conference Proceedings (OSTI)

Progress reports are presented for the following studies on radioactive waste management: defense waste technology; nuclear waste materials characterization center; and supporting studies. 19 figs., 29 tabs.

McElroy, J.L.; Powell, J.A. (comps.)

1985-06-01T23:59:59.000Z

382

NuclearNuclear ""BurningBurning"" of Nuclearof Nuclear ""WasteWaste"" Constantine P. Tzanos  

E-Print Network (OSTI)

1. INTRODUCTION 1.1. Yucca Mountain Project The Yucca Mountain site in Nevada has been designated as United States choice for nuclear waste repository. Yucca Mountain is in a remote dry area, on federal has been made to characterize the nature of the discontinuities of the Yucca Mountain proposed nuclear

383

Salt disposal of heat-generating nuclear waste.  

SciTech Connect

This report summarizes the state of salt repository science, reviews many of the technical issues pertaining to disposal of heat-generating nuclear waste in salt, and proposes several avenues for future science-based activities to further the technical basis for disposal in salt. There are extensive salt formations in the forty-eight contiguous states, and many of them may be worthy of consideration for nuclear waste disposal. The United States has extensive experience in salt repository sciences, including an operating facility for disposal of transuranic wastes. The scientific background for salt disposal including laboratory and field tests at ambient and elevated temperature, principles of salt behavior, potential for fracture damage and its mitigation, seal systems, chemical conditions, advanced modeling capabilities and near-future developments, performance assessment processes, and international collaboration are all discussed. The discussion of salt disposal issues is brought current, including a summary of recent international workshops dedicated to high-level waste disposal in salt. Lessons learned from Sandia National Laboratories' experience on the Waste Isolation Pilot Plant and the Yucca Mountain Project as well as related salt experience with the Strategic Petroleum Reserve are applied in this assessment. Disposal of heat-generating nuclear waste in a suitable salt formation is attractive because the material is essentially impermeable, self-sealing, and thermally conductive. Conditions are chemically beneficial, and a significant experience base exists in understanding this environment. Within the period of institutional control, overburden pressure will seal fractures and provide a repository setting that limits radionuclide movement. A salt repository could potentially achieve total containment, with no releases to the environment in undisturbed scenarios for as long as the region is geologically stable. Much of the experience gained from United States repository development, such as seal system design, coupled process simulation, and application of performance assessment methodology, helps define a clear strategy for a heat-generating nuclear waste repository in salt.

Leigh, Christi D. (Sandia National Laboratories, Carlsbad, NM); Hansen, Francis D.

2011-01-01T23:59:59.000Z

384

Plasma Mass Filters For Nuclear Waste Reprocessing  

SciTech Connect

Practical disposal of nuclear waste requires high-throughput separation techniques. The most dangerous part of nuclear waste is the fission product, which contains the most active and mobile radioisotopes and produces most of the heat. We suggest that the fission products could be separated as a group from nuclear waste using plasma mass filters. Plasmabased processes are well suited to separating nuclear waste, because mass rather than chemical properties are used for separation. A single plasma stage can replace several stages of chemical separation, producing separate streams of bulk elements, fission products, and actinoids. The plasma mass filters may have lower cost and produce less auxiliary waste than chemical processing plants. Three rotating plasma configurations are considered that act as mass filters: the plasma centrifuge, the Ohkawa filter, and the asymmetric centrifugal trap.

Abraham J. Fetterman and Nathaniel J. Fisch

2011-05-26T23:59:59.000Z

385

APPLICATION OF A THIN FILM EVAPORATOR SYSTEM FOR MANAGEMENT OF LIQUID HIGH-LEVEL WASTES AT HANFORD  

Science Conference Proceedings (OSTI)

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.

TEDESCHI AR; WILSON RA

2010-01-14T23:59:59.000Z

386

Design and performance of a full-scale spray calciner for nonradioactive high-level-waste-vitrification studies  

SciTech Connect

In the spray calcination process, liquid waste is spray-dried in a heated-wall spray dryer (termed a spray calciner), and then it may be combined in solid form with a glass-forming frit. This mixture is then melted in a continuous ceramic melter or in an in-can melter. Several sizes of spray calciners have been tested at PNL- laboratory scale, pilot scale and full scale. Summarized here is the experience gained during the operation of PNL's full-scale spray calciner, which has solidified approx. 38,000 L of simulated acid wastes and approx. 352,000 L of simulated neutralized wastes in 1830 h of processing time. Operating principles, operating experience, design aspects, and system descriptions of a full-scale spray calciner are discussed. Individual test run summaries are given in Appendix A. Appendices B and C are studies made by Bechtel Inc., under contract by PNL. These studies concern, respectively, feed systems for the spray calciner process and a spray calciner vibration analysis. Appendix D is a detailed structural analysis made at PNL of the spray calciner. These appendices are included in the report to provide a complete description of the spray calciner and to include all major studies made concerning PNL's full-scale spray calciner.

Miller, F.A.

1981-06-01T23:59:59.000Z

387

Expected near-field thermal environments in a sequentially loaded spent-fuel or high-level waste repository in salt  

SciTech Connect

This report describes the effect of realistic waste emplacement schedules on repository thermal environments. Virtually all estimates to date have been based on instantaneous loading of wastes having uniform properties throughout the repository. However, more realistic scenarios involving sequential emplacement of wastes reflect the gradual filling of the repository over its lifetime. These cases provide temperatures that can be less extreme than with the simple approximation. At isolated locations in the repository, the temperatures approach the instantaneous-loading limit. However, for most of the repository, temperature rises in the near-field are 10 to 40 years behind the conservative estimates depending on the waste type and the location in the repository. Results are presented for both spent-fuel and high-level reprocessing waste repositories in salt, for a regional repository concept, and for a single national repository concept. The national repository is filled sooner and therefore more closely approximates the instantaneously loaded repository. However, temperatures in the near-field are still 20/sup 0/C or more below the values in the simple model for 40 years after startup of repository emplacement operations. The results suggest that current repository design concepts based on the instantaneous-loading predictions are very conservative. Therefore, experiments to monitor temperatures in a test and evaluation facility, for example, will need to take into account the reduced temperatures in order to provide data used in predicting repository performance.

Rickertsen, L.D.; Arbital, J.G.; Claiborne, H.C.

1982-01-01T23:59:59.000Z

388

USED NUCLEAR MATERIALS AT SAVANNAH RIVER SITE: ASSET OR WASTE?  

SciTech Connect

The nuclear industry, both in the commercial and the government sectors, has generated large quantities of material that span the spectrum of usefulness, from highly valuable (“assets”) to worthless (“wastes”). In many cases, the decision parameters are clear. Transuranic waste and high level waste, for example, have no value, and is either in a final disposition path today, or – in the case of high level waste – awaiting a policy decision about final disposition. Other materials, though discardable, have intrinsic scientific or market value that may be hidden by the complexity, hazard, or cost of recovery. An informed decision process should acknowledge the asset value, or lack of value, of the complete inventory of materials, and the structure necessary to implement the range of possible options. It is important that informed decisions are made about the asset value for the variety of nuclear materials available. For example, there is a significant quantity of spent fuel available for recycle (an estimated $4 billion value in the Savannah River Site’s (SRS) L area alone); in fact, SRS has already blended down more than 300 metric tons of uranium for commercial reactor use. Over 34 metric tons of surplus plutonium is also on a path to be used as commercial fuel. There are other radiological materials that are routinely handled at the site in large quantities that should be viewed as strategically important and / or commercially viable. In some cases, these materials are irreplaceable domestically, and failure to consider their recovery could jeopardize our technological leadership or national defense. The inventories of nuclear materials at SRS that have been characterized as “waste” include isotopes of plutonium, uranium, americium, and helium. Although planning has been performed to establish the technical and regulatory bases for their discard and disposal, recovery of these materials is both economically attractive and in the national interest.

Magoulas, V.

2013-06-03T23:59:59.000Z

389

Accelerated Chemical Aging of Crystalline Nuclear Waste Forms  

Science Conference Proceedings (OSTI)

Symposium, Materials Science of Nuclear Waste Management ... thereof) will ultimately determine whether nuclear energy is deemed environmentally friendly.

390

An Investigation into the Oxidation State of Molybdenum in Simplified High Level Nuclear Waste Glass Compositions  

E-Print Network (OSTI)

a full simulated HLW stream based upon 4:1 ratio of high burn up UO2/mixed oxide (HBU/MOX) fuel. EXPERIMENTAL A series of simplified simulated HLW glasses (based on the 4:1 HBU/MOX composition) were melted

Sheffield, University of

391

Nitrous oxide production from radiolysis of simulted high-level nuclear waste solutions  

DOE Green Energy (OSTI)

Nitrous oxide gas (N{sub 2}O) is produced by the radiolysis of aqueous nitrate or nitrite solutions in the presence of organic compounds. When ethylenediaminetetraacetic acid (EDTA) or N- (2-hydroxyethyl)-ethylenediaminetriacetic acid (HEDTA) is present, the G-value for hydrogen increases and N{sub 2}O become the major gaseous product (G=0.54). A survey of organic compounds indicates the amount of N{sub 2}O formed depends on the structure of the organic. With highly oxidized organics (carbonate, formate, acetate and oxalate), little or no N{sub 2}O is formed. Aromatic and aliphatic organics (sodium tetraphenylborate, benzene, phenol, n-paraffin, and tributylphosphate) produce small amounts of N{sub 2}O. Water soluble, easily oxidized organics (methanol, ethanol, isopropanol, n-butanol, acetone, and ethylene glycol) produce large amounts of N{sub 2}O relative to the previous two categories. Nitrous oxide production is not greatly affected by pH between neutral and pH=13, but increases significantly in acid solution. The G-value for N{sub 2}O production in 10 wt% potassium tetraphenylborate slurries has been measured under process conditions important at the Savannah River Site.

Walker, D.D.; Hobbs, D.T.; Tiffany, J.B.; Bibler, N.E. [Westinghouse Savannah River Co., Aiken, SC (United States); Meisel, D. [Argonne National Lab., IL (United States)

1992-07-01T23:59:59.000Z

392

Nitrous oxide production from radiolysis of simulted high-level nuclear waste solutions  

DOE Green Energy (OSTI)

Nitrous oxide gas (N{sub 2}O) is produced by the radiolysis of aqueous nitrate or nitrite solutions in the presence of organic compounds. When ethylenediaminetetraacetic acid (EDTA) or N- (2-hydroxyethyl)-ethylenediaminetriacetic acid (HEDTA) is present, the G-value for hydrogen increases and N{sub 2}O become the major gaseous product (G=0.54). A survey of organic compounds indicates the amount of N{sub 2}O formed depends on the structure of the organic. With highly oxidized organics (carbonate, formate, acetate and oxalate), little or no N{sub 2}O is formed. Aromatic and aliphatic organics (sodium tetraphenylborate, benzene, phenol, n-paraffin, and tributylphosphate) produce small amounts of N{sub 2}O. Water soluble, easily oxidized organics (methanol, ethanol, isopropanol, n-butanol, acetone, and ethylene glycol) produce large amounts of N{sub 2}O relative to the previous two categories. Nitrous oxide production is not greatly affected by pH between neutral and pH=13, but increases significantly in acid solution. The G-value for N{sub 2}O production in 10 wt% potassium tetraphenylborate slurries has been measured under process conditions important at the Savannah River Site.

Walker, D.D.; Hobbs, D.T.; Tiffany, J.B.; Bibler, N.E. (Westinghouse Savannah River Co., Aiken, SC (United States)); Meisel, D. (Argonne National Lab., IL (United States))

1992-01-01T23:59:59.000Z

393

Evaluation of high-level nuclear waste tanks having a potential flammable gas hazard  

DOE Green Energy (OSTI)

In 1990 the U.S. Department of Energy declared an unreviewed safety question as a result of the behavior of tank 241-SY-101. This tank exhibited episodic releases of flammable gases that on a couple of occasions exceeded the lower flammability limit of hydrogen in air. Over the past six years a considerable amount of knowledge has been gained about the chemical and physical processes that govern the behavior of tank 241-SY-101 and the other tanks associated with a potential flammable gas hazard. This paper presents an overview of the current understanding of gas generation, retention, and release and covers the results of direct sampling of the tanks to determine the gas composition and the amount of stored gas.

Johnson, G.D.; Barton, W.B.; Hill, R.C.; et al, Fluor Daniel Hanford

1997-02-14T23:59:59.000Z

394

Roadmapping Process Improvements by Experience at the Idaho National Engineering and Environmental Laboratory High Level Waste Program and Synergistic Interfaces with Decision-Making  

SciTech Connect

Six technology roadmaps were developed for various technologies under consideration for the treatment of sodium bearing liquid and calcine wastes. In the process of creating these roadmaps, a number of process improvements were identified for each of the formal roadmapping phases as described in the Department of Energy’s draft roadmapping guidance. The lessons learned, presented as beneficial improvements to the Idaho National Engineering and Environmental Laboratory (INEEL) High Level Waste Program, are proposed to be added to the draft guidance. Additionally, synergistic interfaces between the roadmapping and decision-making processes were observed and reported on. With these improvements, technology roadmapping has become an effective integration tool at the INEEL for planning technology development.

Murphy, James Anthony; Olson, Arlin Leland

2001-02-01T23:59:59.000Z

395

Control of high level radioactive waste-glass melters. Part 6, Noble metal catalyzed formic acid decomposition, and formic acid/denitration  

DOE Green Energy (OSTI)

A necessary step in Defense Waste Processing Facility (DWPF) melter feed preparation for the immobilization of High Level Radioactive Waste (HLW) is reduction of Hg(II) to Hg(0), permitting steam stripping of the Hg. Denitrition and associated NOx evolution is a secondary effect of the use of formic acid as the mercury-reducing agent. Under certain conditions the presence of transition or noble metals can result in significant formic acid decomposition, with associated CO{sub 2} and H{sub 2} evolution. These processes can result in varying redox properties of melter feed, and varying sequential gaseous evolution of oxidants and hydrogen. Electrochemical methods for monitoring the competing processes are discussed. Laboratory scale techniques have been developed for simulating the large-scale reactions, investigating the relative effectiveness of the catalysts, and the effectiveness of catalytic poisons. The reversible nitrite poisoning of formic acid catalysts is discussed.

Bickford, D.F.; Coleman, C.J.; Hsu, C.L.W.; Eibling, R.E.

1990-12-31T23:59:59.000Z

396

Case histories of EA documents for nuclear waste  

SciTech Connect

Nuclear power programs and policies in the United States have been subject to environmental assessment under the National Environmental Policy Act (NEPA) since 1971. NEPA documentation prepared for programmatic policy decision-making within the nuclear fuel cycle and concurrent federal policy are examined as they relate to radioactive waste management in this paper. Key programmatic environmental impact statements that address radioactive waste management include: the Atomic Energy Commission document on management of commercial high-level and transuranium-contaminated radioactive waste, which focussed on development of engineered retrievable surface storage facilities (RSSF); the Nuclear Regulatory Commission (NRC) document on use of recycled plutonium in mixed oxide fuel in light water cooled reactors, which focussed on plutonium recycle and RSSF; the NRC statement on handling of spent light water power reactor fuel, which focussed on spent fuel storage; and the Department of Energy (DOE) statement on management of commercially generated radioactive wastes, which focussed on development of deep geologic repositories. DOE is currently pursuing the deep geologic repository option, with monitored retrievable storage as a secondary option.

Vocke, R.W.

1985-01-01T23:59:59.000Z