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1

Chapter 19 - Nuclear Waste Fund  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |Energyon ArmedWaste andAccessCO2 Injection Begins8: Variable Frequency Drive Evaluation19.0

2

Nuclear Waste Fund Activities Management Team | Department of Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014NuclearCommission,ScienceWaste Fund

3

GC Commits to Transparency on Nuclear Waste Fund Fee Adequacy Decisions  

Broader source: Energy.gov [DOE]

Consistent with the Administration's commitment to transparency, DOE General Counsel Scott Blake Harris has decided that all future determinations as to the adequacy of the Nuclear Waste Fund fee...

4

Department of Energy's Nuclear Waste Fund's Fiscal Year 2014 Financial Statement Audit  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in Review: TopEnergy DOEDealing WithDevelopmentReport and7DepartmentNuclear Waste Fund's

5

Nuclear Waste Reduction  

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

Nuclear Waste Reduction Pyroprocessing is a promising technology for recycling used nuclear fuel and improving the associated waste management options. The process...

6

Bubblers Speed Nuclear Waste Processing at SRS  

SciTech Connect (OSTI)

At the Department of Energy's Savannah River Site, American Recovery and Reinvestment Act funding has supported installation of bubbler technology and related enhancements in the Defense Waste Processing Facility (DWPF). The improvements will accelerate the processing of radioactive waste into a safe, stable form for storage and permit expedited closure of underground waste tanks holding 37 million gallons of liquid nuclear waste.

None

2010-11-14T23:59:59.000Z

7

Bubblers Speed Nuclear Waste Processing at SRS  

ScienceCinema (OSTI)

At the Department of Energy's Savannah River Site, American Recovery and Reinvestment Act funding has supported installation of bubbler technology and related enhancements in the Defense Waste Processing Facility (DWPF). The improvements will accelerate the processing of radioactive waste into a safe, stable form for storage and permit expedited closure of underground waste tanks holding 37 million gallons of liquid nuclear waste.

None

2014-08-06T23:59:59.000Z

8

Animal Waste Technology Fund (Maryland)  

Broader source: Energy.gov [DOE]

A bill passed in 2012 transferred responsibility for animal waste management technology projects to the Maryland Department of Agriculture. The Department will maintain the Animal Waste Technology...

9

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

10

Swedish nuclear waste efforts  

SciTech Connect (OSTI)

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

11

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

E-Print Network [OSTI]

as a geologic repository for disposal of spent nuclear fuel and high level radioactive waste. #12;The YuccaNuclearNuclear ""BurningBurning"" of Nuclearof Nuclear ""WasteWaste"" Constantine P. Tzanos Argonne-level radioactive waste that has accumulated at 72 commercial and 4 DOE sites. s U.S. Congress adopted the Nuclear

12

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington is intended to update Congress and the Secretary of Energy on the U.S. Nuclear Waste Technical Review Board-level radioactive waste (HLW) is evolving. The letter is issued in accordance with provisions of the Nuclear Waste

13

United States Nuclear Waste Technical Review Board  

E-Print Network [OSTI]

United States Nuclear Waste Technical Review Board Experience Gained From Programs to Manage High-Level Radioactive Waste and Spent Nuclear Fuel in the United States and Other Countries A Report to Congress and the Secretary of Energy April 2011 #12;#12;U.S. Nuclear Waste Technical Review Board Experience Gained From

14

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

jlc029va UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300, D.C. 20585 Dear Speaker Gingrich, Senator Thurmond, and Secretary Peña: The Nuclear Waste Technical Review Board (the Board) herewith submits this second report of 1997 as required by the Nuclear Waste

15

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

con202vf UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300, the Nuclear Waste Technical Review Board (Board) submits its second report of 2003 in accordance with provisions of the Nuclear Waste Policy Amendments Act of 1987, Public Law 100-203. The Act requires the Board

16

Appendix A Nuclear Waste Technical Review Board  

E-Print Network [OSTI]

39 Appendices Appendices #12;Appendix A Nuclear Waste Technical Review Board Members: Curricula to the Nuclear Waste Technical Review Board. President Clinton appointed Dr. Cohon chairman on January 17, 1997, and Asia and on energy facility siting, including nuclear waste ship- ping and storage. In addition to his

17

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

JEC187V3 UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 of Energy 1000 Independence Avenue, SW Washington, DC 20585 Dear Secretary O'Leary: At the Nuclear Waste UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington, VA

18

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington are pleased to transmit a technical report prepared by the Nuclear Waste Technical Review Board (Board. Based on its review of data gathered by the DOE and the Center for Nuclear Waste Regulatory Analyses

19

Appendix A Nuclear Waste Technical Review Board  

E-Print Network [OSTI]

59 Appendices Appendices #12;Appendix A Nuclear Waste Technical Review Board Members: Curricula Cohon to serve on the Nuclear Waste Technical Review Board. President Clinton appointed Dr. Cohon, and Asia and on energy-facility siting, including nuclear waste shipping and storage. In addition to his

20

Recovery Act Funding Leads to Record Year for Transuranic Waste Shipments  

Broader source: Energy.gov [DOE]

With the help of American Recovery and Reinvestment Act funding, the Waste Isolation Pilot Plant (WIPP) received the most transuranic waste shipments in a single year since waste operations began...

Note: This page contains sample records for the topic "nuclear waste fund" 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

Nuclear waste management. Semiannual progress report, October 1983-March 1984  

SciTech Connect (OSTI)

Progress in the following studies on radioactive waste management is reported: defense waste technology; Nuclear Waste Materials Characterization Center; waste isolation; and supporting studies. 58 figures, 22 tables.

McElroy, J.L.; Powell, J.A.

1984-06-01T23:59:59.000Z

22

Nuclear Waste Technical Review Board Performance Plan  

E-Print Network [OSTI]

to disposing of spent nuclear fuel and high-level radioactive waste were set forth by Congress in the NWPA. The goals are to develop a repository or repositories for disposing of high-level radioactive waste spent nuclear fuel and high- level radioactive waste. The Board's general goals and strategic objectives

23

Uranium immobilization and nuclear waste  

SciTech Connect (OSTI)

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

24

Global Nuclear Energy Partnership Waste Treatment Baseline  

SciTech Connect (OSTI)

The Global Nuclear Energy Partnership program (GNEP) is designed to demonstrate a proliferation-resistant and sustainable integrated nuclear fuel cycle that can be commercialized and used internationally. Alternative stabilization concepts for byproducts and waste streams generated by fuel recycling processes were evaluated and a baseline of waste forms was recommended for the safe disposition of waste streams. Waste forms are recommended based on the demonstrated or expected commercial practicability and technical maturity of the processes needed to make the waste forms, and performance of the waste form materials when disposed. Significant issues remain in developing technologies to process some of the wastes into the recommended waste forms, and a detailed analysis of technology readiness and availability may lead to the choice of a different waste form than what is recommended herein. Evolving regulations could also affect the selection of waste forms.

Dirk Gombert; William Ebert; James Marra; Robert Jubin; John Vienna

2008-05-01T23:59:59.000Z

25

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

the Department of Energy's (DOE) work related to the packaging and transport of such waste. Consistent with its and waste package performance. In light of continuing technical challenges, the Board believes that the DOEcon144vf UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300

26

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.

27

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

SciTech Connect (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

28

Plasma Mass Filters For Nuclear Waste Reprocessing  

SciTech Connect (OSTI)

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

2011-05-25T23:59:59.000Z

29

Plasma Mass Filters For Nuclear Waste Reprocessing  

SciTech Connect (OSTI)

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

30

Removing nuclear waste, one shipment at a time  

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

Removing nuclear waste, one shipment at a time Removing nuclear waste, one shipment at a time The Lab's 1,000th shipment of transuranic waste recently left Los Alamos, on its way...

31

Deep Geologic Nuclear Waste Disposal - No New Taxes - 12469  

SciTech Connect (OSTI)

To some, the perceived inability of the United States to dispose of high-level nuclear waste justifies a moratorium on expansion of nuclear power in this country. Instead, it is more an example of how science yields to social pressure, even on a subject as technical as nuclear waste. Most of the problems, however, stem from confusion on the part of the public and their elected officials, not from a lack of scientific knowledge. We know where to put nuclear waste, how to put it there, how much it will cost, and how well it will work. And it's all about the geology. The President's Blue Ribbon Commission on America's Nuclear Future has drafted a number of recommendations addressing nuclear energy and waste issues (BRC 2011) and three recommendations, in particular, have set the stage for a new strategy to dispose of high-level nuclear waste and to manage spent nuclear fuel in the United States: 1) interim storage for spent nuclear fuel, 2) resumption of the site selection process for a second repository, and 3) a quasi-government entity to execute the program and take control of the Nuclear Waste Fund in order to do so. The first two recommendations allow removal and storage of spent fuel from reactor sites to be used in the future, and allows permanent disposal of actual waste, while the third controls cost and administration. The Nuclear Waste Policy Act of 1982 (NPWA 1982) provides the second repository different waste criteria, retrievability, and schedule, so massive salt returns as the candidate formation of choice. The cost (in 2007 dollars) of disposing of 83,000 metric tons of heavy metal (MTHM) high-level waste (HLW) is about $ 83 billion (b) in volcanic tuff, $ 29 b in massive salt, and $ 77 b in crystalline rock. Only in salt is the annual revenue stream from the Nuclear Waste Fund more than sufficient to accomplish this program without additional taxes or rate hikes. The cost is determined primarily by the suitability of the geologic formation, i.e., how well it performs on its own for millions of years with little engineering assistance from humans. It is critical that the states most affected by this issue (WA, SC, ID, TN, NM and perhaps others) develop an independent multi-state agreement in order for a successful program to move forward. Federal approval would follow. Unknown to most, the United States has a successful operating deep permanent geologic nuclear repository for high and low activity waste, called the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico. Its success results from several factors, including an optimal geologic and physio-graphic setting, a strong scientific basis, early regional community support, frequent interactions among stakeholders at all stages of the process, long-term commitment from the upper management of the U.S. Department of Energy (DOE) over several administrations, strong New Mexico State involvement and oversight, and constant environmental monitoring from before nuclear waste was first emplaced in the WIPP underground (in 1999) to the present. WIPP is located in the massive bedded salts of the Salado Formation, whose geological, physical, chemical, redox, thermal, and creep-closure properties make it an ideal formation for long-term disposal, long-term in this case being greater than 200 million years. These properties also mean minimal engineering requirements as the rock does most of the work of isolating the waste. WIPP has been operating for twelve years, and as of this writing, has disposed of over 80,000 m{sup 3} of nuclear weapons waste, called transuranic or TRU waste (>100 nCurie/g but <23 Curie/1000 cm{sup 3}) including some high activity waste from reprocessing of spent fuel from old weapons reactors. All nuclear waste of any type from any source can be disposed in this formation better, safer and cheaper than in any other geologic formation. At the same time, it is critical that we complete the Yucca Mountain license application review so as not to undermine the credibility of the Nuclear Regulatory Commission and the scientific commun

Conca, James [RJLee Group, Inc., Pasco WA 509.205.7541 (United States); Wright, Judith [UFA Ventures, Inc., Richland, WA (United States)

2012-07-01T23:59:59.000Z

32

Doing the impossible: Recycling nuclear waste  

ScienceCinema (OSTI)

A Science Channel feature explores how Argonne techniques could be used to safely reduce the amount of radioactive waste generated by nuclear power?the most plentiful carbon-neutral energy source. Read more at http://www.anl.gov/Media_Center/ArgonneNow/Fall_2009/nuclear.html

None

2013-04-19T23:59:59.000Z

33

Nuclear waste incineration technology status  

SciTech Connect (OSTI)

The incinerators developed and/or used for radioactive waste combustion are discussed and suggestions are made for uses of incineration in radioactive waste management programs and for incinerators best suited for specific applications. Information on the amounts and types of radioactive wastes are included to indicate the scope of combustible wastes being generated and in existence. An analysis of recently developed radwaste incinerators is given to help those interested in choosing incinerators for specific applications. Operating information on US and foreign incinerators is also included to provide additional background information. Development needs are identified for extending incinerator applications and for establishing commercial acceptance.

Ziegler, D.L.; Lehmkuhl, G.D.; Meile, L.J.

1981-07-15T23:59:59.000Z

34

Appendix A U.S. Nuclear Waste Technical Review  

E-Print Network [OSTI]

Appendices Appendices 31 #12;#12;Appendix A Appendix A U.S. Nuclear Waste Technical Review Board.S. Nuclear Waste Technical Review Board as Chairman on September 10, 2004, by President George W. Bush. Dr­2004), 4 years as chair, on the U.S. Nuclear Regulatory Commission's Advisory Committee on Nuclear Waste

35

Appendix A U.S. Nuclear Waste Technical Review  

E-Print Network [OSTI]

Appendices Appendices 37 #12;#12;Appendix A Appendix A U.S. Nuclear Waste Technical Review Board as chair, on the U.S. Nuclear Regulatory Commission's Advisory Commit tee on Nuclear Waste. His areas to the Nuclear Waste Technical Review Board on June 26, 2002, by President George W. Bush. Dr. Abkowitz

36

Nuclear Waste Technical Review Board Members Appendix A 53  

E-Print Network [OSTI]

51 Appendix A Nuclear Waste Technical Review Board Members #12;#12;Appendix A 53 B. John Garrick, Ph.D., P.E. Chairman Dr. B. John Garrick was appointed to the U.S. Nuclear Waste Technical Review, on the U.S. Nuclear Regula- tory Commission's Advisory Committee on Nuclear Waste. His areas of expertise

37

Nuclear waste management. Quarterly progress report, January-March 1980  

SciTech Connect (OSTI)

Reported are: high-level waste immobilization, alternative waste forms, nuclear waste materials characterization, TRU waste immobilization, TRU waste decontamination, krypton solidification, thermal outgassing, iodine-129 fixation, 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, engineered barriers, criteria for defining waste isolation, and spent fuel and pool component integrity. (DLC)

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

1980-06-01T23:59:59.000Z

38

Recovery of fissile materials from nuclear wastes  

DOE Patents [OSTI]

A process for recovering fissile materials such as uranium, and plutonium, and rare earth elements, from complex waste feed material, and converting the remaining wastes into a waste glass suitable for storage or disposal. The waste feed is mixed with a dissolution glass formed of lead oxide and boron oxide resulting in oxidation, dehalogenation, and dissolution of metal oxides. Carbon is added to remove lead oxide, and a boron oxide fusion melt is produced. The fusion melt is essentially devoid of organic materials and halogens, and is easily and rapidly dissolved in nitric acid. After dissolution, uranium, plutonium and rare earth elements are separated from the acid and recovered by processes such as PUREX or ion exchange. The remaining acid waste stream is vitrified to produce a waste glass suitable for storage or disposal. Potential waste feed materials include plutonium scrap and residue, miscellaneous spent nuclear fuel, and uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, organic material and other carbon-containing material.

Forsberg, Charles W. (Oak Ridge, TN)

1999-01-01T23:59:59.000Z

39

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

in 1992 by the National Academy Press in a report titled Ground Water at Yucca Mountain--How High Can of affiliation with the Yucca Mountain Project, and their lack of previous involvement in evaluating Mr General's office on possible future upwelling of water into the proposed nuclear waste repository at Yucca

40

Scientific Solutions to Nuclear Waste Environmental Challenges  

SciTech Connect (OSTI)

The Hidden Cost of Nuclear Weapons The Cold War arms race drove an intense plutonium production program in the U.S. This campaign produced approximately 100 tons of plutonium over 40 years. The epicenter of plutonium production in the United States was the Hanford site, a 586 square mile reservation owned by the Department of Energy and located on the Colombia River in Southeastern Washington. Plutonium synthesis relied on nuclear reactors to convert uranium to plutonium within the reactor fuel rods. After a sufficient amount of conversion occurred, the rods were removed from the reactor and allowed to cool. They were then dissolved in an acid bath and chemically processed to separate and purify plutonium from the rest of the constituents in the used reactor fuel. The acidic waste was then neutralized using sodium hydroxide and the resulting mixture of liquids and precipitates (small insoluble particles) was stored in huge underground waste tanks. The byproducts of the U.S. plutonium production campaign include over 53 million gallons of high-level radioactive waste stored in 177 large underground tanks at Hanford and another 34 million gallons stored at the Savannah River Site in South Carolina. This legacy nuclear waste represents one of the largest environmental clean-up challenges facing the world today. The nuclear waste in the Hanford tanks is a mixture of liquids and precipitates that have settled into sludge. Some of these tanks are now over 60 years old and a small number of them are leaking radioactive waste into the ground and contaminating the environment. The solution to this nuclear waste challenge is to convert the mixture of solids and liquids into a durable material that won't disperse into the environment and create hazards to the biosphere. What makes this difficult is the fact that the radioactive half-lives of some of the radionuclides in the waste are thousands to millions of years long. (The half-life of a radioactive substance is the amount of time it takes for one-half of the material to undergo radioactive decay.) In general, the ideal material would need to be durable for approximately 10 half-lives to allow the activity to decay to negligible levels. However, the potential health effects of each radionuclide vary depending on what type of radiation is emitted, the energy of that emission, and the susceptibility for the human body to accumulate and concentrate that particular element. Consequently, actual standards tend to be based on limiting the dose (energy deposited per unit mass) that is introduced into the environment. The Environmental Protection Agency (EPA) has the responsibility to establish standards for nuclear waste disposal to protect the health and safety of the public. For example, the Energy Policy Act of 1992 directed the EPA to establish radiation protection standards for the Yucca Mountain geologic repository for nuclear wastes. The standards for Yucca Mountain were promulgated in 2008, and limit the dose to 15 millirem per year for the first 10,000 years, and 100 milirem per year between 10,000 years and 1 million years (40 CFR Part 197; http://www.epa.gov/radiation/yucca/2008factsheet.html). So, the challenge is two-fold: (1) develop a material (a waste form) that is capable of immobilizing the waste over geologic time scales, and (2) develop a process to convert the radioactive sludge in the tanks into this durable waste form material. Glass: Hard, durable, inert, and with infinite chemical versatility Molten glass is a powerful solvent liquid, which can be designed to dissolve almost anything. When solidified, it can be one of the most chemically inert substances known to man. Nature's most famous analogue to glass is obsidian, a vitreous product of volcanic activity; formations over 17 million years old have been found. Archaeologists have found man-made glass specimens that are five thousand years old.

Johnson, Bradley R.

2014-01-30T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

Nuclear Waste Assessment System for Technical Evaluation (NUWASTE)  

E-Print Network [OSTI]

Nuclear Waste Assessment System for Technical Evaluation (NUWASTE): Status and Initial Results A Report to the U.S. Congress and the Secretary of Energy U.S. Nuclear Waste Technical Review Board June 2011 Topical Report #2 #12;ii U.S.U.S. Nuclear Waste Technical Review Board Report Availability

42

U.S. Nuclear Waste Technical Review Board Members  

E-Print Network [OSTI]

Appendix A Appendix A U.S. Nuclear Waste Technical Review Board Members Jared L. Cohon, Ph.D.; Chairman On June 29, 1995, President Bill Clinton appointed Jared Cohon to the Nuclear Waste Technical, and Asia and on energy facil ity siting, including nuclear waste shipping and storage. In addition to his

43

Appendix A U.S. Nuclear Waste Technical Review  

E-Print Network [OSTI]

Appendices Appendices 25 #12;#12;Appendix A Appendix A U.S. Nuclear Waste Technical Review Board Members Michael L. Corradini, Ph.D.; Chairman Dr. Michael L. Corradini was appointed to the Nuclear Waste and The Secretary of Energy Mark D. Abkowitz, Ph.D. Dr. Mark D. Abkowitz was appointed to the Nuclear Waste

44

Appendix A U.S. Nuclear Waste Technical Review Board  

E-Print Network [OSTI]

7 Appendices Appendices #12;Appendix A U.S. Nuclear Waste Technical Review Board Members: Curricula to the Nuclear Waste Technical Review Board. President Clinton appointed Dr. Cohon chairman on January 17, 1997, and Asia and on energy facility siting, including nuclear waste shipping and storage. In addition to his

45

U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD Report to January to December 2000 The U.S. Congress are available at www.nwtrb.gov, the NWTRB Web site. #12;#12;#12;NUCLEAR WASTE TECHNICAL REVIEW BOARD Dr. Jared L . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 vii Table of Contents Appendices Appendix A Nuclear Waste Technical Review Board Members

46

U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD Report to January 1, 2001, to January 31, 2002 The U All NWTRB reports are available at www.nwtrb.gov, the NWTRB Web site. #12;#12;#12;NUCLEAR WASTE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 Appendices Appendix A U.S. Nuclear Waste Technical Review Board Members

47

U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD Report to The U.S. Congress And The Secretary of Energy January 1, 2003, to December 31, 2003 #12;U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD Report to The U.nwtrb.gov, the NWTRB Web site. #12;#12;#12;NUCLEAR WASTE TECHNICAL REVIEW BOARD 2003 Dr. Michael L. Corradini

48

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

SciTech Connect (OSTI)

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

49

Nuclear Materials: Reconsidering Wastes and Assets - 13193  

SciTech Connect (OSTI)

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. (authors)

Michalske, T.A. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States)

2013-07-01T23:59:59.000Z

50

International nuclear waste management fact book  

SciTech Connect (OSTI)

The International Nuclear Waste Management Fact Book has been compiled to provide current data on fuel cycle and waste management facilities, R and D programs, and key personnel in 24 countries, including the US; four multinational agencies; and 20 nuclear societies. This document, which is in its second year of publication supersedes the previously issued International Nuclear Fuel Cycle Fact Book (PNL-3594), which appeared annually for 12 years. The content has been updated to reflect current information. The Fact Book is organized as follows: National summaries--a section for each country that summarizes nuclear policy, describes organizational relationships, and provides addresses and names of key personnel and information on facilities. International agencies--a section for each of the international agencies that has significant fuel cycle involvement and a list of nuclear societies. Glossary--a list of abbreviations/acronyms of organizations, facilities, and technical and other terms. The national summaries, in addition to the data described above, feature a small map for each country and some general information that is presented from the perspective of the Fact Book user in the US.

Abrahms, C W; Patridge, M D; Widrig, J E

1995-11-01T23:59:59.000Z

51

Seal welded cast iron nuclear waste container  

DOE Patents [OSTI]

This invention identifies methods and articles designed to circumvent metallurgical problems associated with hermetically closing an all cast iron nuclear waste package by welding. It involves welding nickel-carbon alloy inserts which are bonded to the mating plug and main body components of the package. The welding inserts might be bonded in place during casting of the package components. When the waste package closure weld is made, the most severe thermal effects of the process are restricted to the nickel-carbon insert material which is far better able to accommodate them than is cast iron. Use of nickel-carbon weld inserts should eliminate any need for pre-weld and post-weld heat treatments which are a problem to apply to nuclear waste packages. Although the waste package closure weld approach described results in a dissimilar metal combination, the relative surface area of nickel-to-iron, their electrochemical relationship, and the presence of graphite in both materials will act to prevent any galvanic corrosion problem.

Filippi, Arthur M. (Pittsburgh, PA); Sprecace, Richard P. (Murrysville, PA)

1987-01-01T23:59:59.000Z

52

Nuclear waste: our radioactive hot potato  

SciTech Connect (OSTI)

Nuclear industry inevitably produces nuclear waste, whose prudent, prompt and economic disposal is important to the national welfare. Technological problems of containment and isolation have apparently been solved. Underground or geologic disposal sites have the potential form permanent isolation, with salt, basalt, granite, shale, and tuff currently receiving principal attention as repository host rocks. Bedded salt deposits may offer the principal mechanical advantages, but in the northwestern United States the abundance of basalt at existing test sites has made it the subject of experimentation. However, psychological, political, and allegedly environmental obstructionism have stalled the process and virtually immobilized current construction. A program is suggested with the purpose of satisfying technical requirements for public protection while allaying the exaggerated fears of anti-nuclear factions.

Conselman, F.B.

1984-01-01T23:59:59.000Z

53

Other U.S. Nuclear Waste Technical Review Board Correspondence  

E-Print Network [OSTI]

Appendix F Appendix F Other U.S. Nuclear Waste Technical Review Board Correspondence · Letter UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington, VA: Acknowledgement of letter · Letter from Robert R. Loux, Executive Director, Nevada Agency for Nuclear Projects

54

The necessity for permanence : making a nuclear waste storage facility  

E-Print Network [OSTI]

The United States Department of Energy is proposing to build a nuclear waste storage facility in southern Nevada. This facility will be designed to last 10,000 years. It must prevent the waste from contaminating the ...

Stupay, Robert Irving

1991-01-01T23:59:59.000Z

55

Extraction of cesium and strontium from nuclear waste  

DOE Patents [OSTI]

Cesium is extracted from acidified nuclear waste by contacting the waste with a bis 4,4[prime](5) [1-hydroxy-2-ethylhexyl]benzo 18-crown-6 compound and a cation exchanger in a matrix solution. Strontium is extracted from acidified nuclear waste by contacting the waste with a bis 4,4[prime](5[prime]) [1-hydroxyheptyl]cyclohexo 18-crown-6 compound, and a cation exchanger in a matrix solution. 3 figs.

Davis, M.W. Jr.; Bowers, C.B. Jr.

1988-06-07T23:59:59.000Z

56

Extraction of cesium and strontium from nuclear waste  

DOE Patents [OSTI]

Cesium is extracted from acidified nuclear waste by contacting the waste with a bis 4,4'(5) [1-hydroxy-2-ethylhexyl]benzo 18-crown-6 compound and a cation exchanger in a matrix solution. Strontium is extracted from acidified nuclear waste by contacting the waste with a bis 4,4'(5') [1-hydroxyheptyl]cyclohexo 18-crown-6 compound, and a cation exchanger in a matrix solution.

Davis, Jr., Milton W. (Lexington, SC); Bowers, Jr., Charles B. (Columbia, SC)

1988-01-01T23:59:59.000Z

57

Nuclear Waste Policy Act.doc  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in3.pdfEnergyDepartment of Energy Advanced1, 2014NuclearCommission,ScienceWaste

58

Nye County, Nevada 1992 nuclear waste repository program: Program overview. Final report  

SciTech Connect (OSTI)

The purpose of this document is to provide an overview of the Nye County FY92 Nuclear Waste Repository Program (Program). Funds to pay for Program costs will come from the Federal Nuclear Waste Fund, which was established under the Nuclear Waste Policy Act of 1982 (NWPA). In early 1983, the Yucca Mountain was identified as a potentially suitable site for the nation`s first geologic repository for spent reactor fuel and high-level radioactive waste. Later that year, the Nye County Board of County Commissioners (Board) established the capability to monitor the Federal effort to implement the NWPA and evaluate the potential impacts of repository-related activities on Nye County. Over the last eight years, the County`s program has grown in complexity and cost in order to address DOE`s evolving site characterization studies, and prepare for the potential for facility construction and operation. Changes were necessary as well, in response to Congress`s redirection of the repository program specified in the amendments, to the NWPA approved in 1987. In early FY 1991, the County formally established a project office to plan and implement its program of work. The Repository Project Office`s (RPO) mission and functions are provided in Section 2.0. The RPO organization structure is described in Section 3.0.

NONE

1998-07-01T23:59:59.000Z

59

An Underwater Robotic Network for Monitoring Nuclear Waste Storage Pools  

E-Print Network [OSTI]

An Underwater Robotic Network for Monitoring Nuclear Waste Storage Pools Sarfraz Nawaz1 , Muzammil Manchester M60 1QD simon.watson@postgrad.manchester.ac.uk peter.n.green@manchester.ac.uk Abstract. Nuclear to build maps of their internal structure which can then be used for waste removal and pool decommissioning

Jeavons, Peter

60

Nuclear waste management. Quarterly progress report, April-June 1981  

SciTech Connect (OSTI)

Reports and summaries are presented for the following: high-level waste process development; alternative waste forms; TMI zeolite vitrification demonstration program; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton implantation; 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 radionuclides in soils; handbook of methods to decrease the generation of low-level waste; waste management system studies; waste management safety 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; and analysis of spent fuel policy implementation.

Chikalla, T.D.; Powell, J.A.

1981-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

Dismantlement and Radioactive Waste Management of DPRK Nuclear Facilities  

SciTech Connect (OSTI)

One critical aspect of any denuclearization of the Democratic Peoples Republic of Korea (DPRK) involves dismantlement of its nuclear facilities and management of their associated radioactive wastes. The decommissioning problem for its two principal operational plutonium facilities at Yongbyun, the 5MWe nuclear reactor and the Radiochemical Laboratory reprocessing facility, alone present a formidable challenge. Dismantling those facilities will create radioactive waste in addition to existing inventories of spent fuel and reprocessing wastes. Negotiations with the DPRK, such as the Six Party Talks, need to appreciate the enormous scale of the radioactive waste management problem resulting from dismantlement. The two operating plutonium facilities, along with their legacy wastes, will result in anywhere from 50 to 100 metric tons of uranium spent fuel, as much as 500,000 liters of liquid high-level waste, as well as miscellaneous high-level waste sources from the Radiochemical Laboratory. A substantial quantity of intermediate-level waste will result from disposing 600 metric tons of graphite from the reactor, an undetermined quantity of chemical decladding liquid waste from reprocessing, and hundreds of tons of contaminated concrete and metal from facility dismantlement. Various facilities for dismantlement, decontamination, waste treatment and packaging, and storage will be needed. The shipment of spent fuel and liquid high level waste out of the DPRK is also likely to be required. Nuclear facility dismantlement and radioactive waste management in the DPRK are all the more difficult because of nuclear nonproliferation constraints, including the call by the United States for complete, verifiable and irreversible dismantlement, or CVID. It is desirable to accomplish dismantlement quickly, but many aspects of the radioactive waste management cannot be achieved without careful assessment, planning and preparation, sustained commitment, and long completion times. The radioactive waste management problem in fact offers a prospect for international participation to engage the DPRK constructively. DPRK nuclear dismantlement, when accompanied with a concerted effort for effective radioactive waste management, can be a mutually beneficial goal.

Jooho, W.; Baldwin, G. T.

2005-04-01T23:59:59.000Z

62

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

SciTech Connect (OSTI)

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

63

NUCLEAR WASTE VITRIFICATION EFFICIENCY COLD CAP REACTIONS  

SciTech Connect (OSTI)

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 (Fe{sub 2}O{sub 3} and Al{sub 2}O{sub 3}), 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.

KRUGER AA; HRMA PR; POKORNY R

2011-07-29T23:59:59.000Z

64

Review of Nuclear Safety Culture at the Hanford Site Waste Treatment...  

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

Review of Nuclear Safety Culture at the Hanford Site Waste Treatment and Immobilization Plant Project, October 2010 Review of Nuclear Safety Culture at the Hanford Site Waste...

65

Method of preparing nuclear wastes for tansportation and interim storage  

DOE Patents [OSTI]

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

66

WIPP Uses Recovery Act Funding to Reduce Nuclear Waste Footprint |  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartment ofDepartment of Energy While dry storageMarch| DepartmentofCarlsbadDepartment of

67

Nuclear waste management. Quarterly progress report, April-June 1980  

SciTech Connect (OSTI)

The status of the following programs is reported: high-level waste immobilization; alternative waste forms; Nuclear Waste Materials Characterization Center; TRU waste immobilization; TRU waste decontamination; krypton solidification; thermal outgassing; iodine-129 fixation; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; mobility of organic complexes of fission products in soils; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology; systems study on engineered barriers; criteria for defining waste isolation; spent fuel and fuel pool component integrity program; analysis of spent fuel policy implementation; asphalt emulsion sealing of uranium tailings; application of long-term chemical biobarriers for uranium tailings; and development of backfill material.

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

1980-09-01T23:59:59.000Z

68

Nuclear waste management. Quarterly progress report, October through December 1980  

SciTech Connect (OSTI)

Progress reports and summaries are presented under the following headings: 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; monitoring and physical characterization of unsaturated zone transport; well-logging instrumentation development; mobility of organic complexes of radionuclides in soils; waste management system studies; waste management safety studies; assessment of effectiveness of geologic isolation systems; waste/rock interactions technology; high level waste form preparation; development of backfill material; development of structural engineered barriers; ONWI disposal charge analysis; spent fuel and fuel component integrity program; analysis of spent fuel policy implementation; analysis of postulated criticality events in a storage array of spent LWR fuel; asphalt emulsion sealing of uranium tailings; liner evaluation for uranium mill tailings; multilayer barriers for sealing of uranium tailings; application of long-term chemical biobarriers for uranium tailings; revegetation of inactive uranium tailing sites; verification instrument development.

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

1981-03-01T23:59:59.000Z

69

RADIOACTIVE WASTE MANAGEMENT IN THE CHERNOBYL EXCLUSION ZONE - 25 YEARS SINCE THE CHERNOBYL NUCLEAR POWER PLANT ACCIDENT  

SciTech Connect (OSTI)

Radioactive waste management is an important component of the Chernobyl Nuclear Power Plant accident mitigation and remediation activities of the so-called Chernobyl Exclusion Zone. This article describes the localization and characteristics of the radioactive waste present in the Chernobyl Exclusion Zone and summarizes the pathways and strategy for handling the radioactive waste related problems in Ukraine and the Chernobyl Exclusion Zone, and in particular, the pathways and strategies stipulated by the National Radioactive Waste Management Program. The brief overview of the radioactive waste issues in the ChEZ presented in this article demonstrates that management of radioactive waste resulting from a beyond-designbasis accident at a nuclear power plant becomes the most challenging and the costliest effort during the mitigation and remediation activities. The costs of these activities are so high that the provision of radioactive waste final disposal facilities compliant with existing radiation safety requirements becomes an intolerable burden for the current generation of a single country, Ukraine. The nuclear accident at the Fukushima-1 NPP strongly indicates that accidents at nuclear sites may occur in any, even in a most technologically advanced country, and the Chernobyl experience shows that the scope of the radioactive waste management activities associated with the mitigation of such accidents may exceed the capabilities of a single country. Development of a special international program for broad international cooperation in accident related radioactive waste management activities is required to handle these issues. It would also be reasonable to consider establishment of a dedicated international fund for mitigation of accidents at nuclear sites, specifically, for handling radioactive waste problems in the ChEZ. The experience of handling Chernobyl radioactive waste management issues, including large volumes of radioactive soils and complex structures of fuel containing materials can be fairly useful for the entire world's nuclear community and can help make nuclear energy safer.

Farfan, E.; Jannik, T.

2011-10-01T23:59:59.000Z

70

Nuclear waste treatment program. Annual report for FY 1985  

SciTech Connect (OSTI)

Two of the US Department of Energy's (DOE) nuclear waste management-related goals are: (1) to ensure that waste management is not an obstacle to the further deployment of light-water reactors (LWR) and the closure of the nuclear fuel cycle and (2) to fulfill its institutional responsibility for providing safe storage and disposal of existing and future nuclear wastes. As part of its approach to achieving these goals, the Office of Terminal Waste Disposal and Remedial Action of DOE established what is now called the Nuclear Waste Treatment Program (NWTP) at the Pacific Northwest Laboratory (PNL) during the second half of FY 1982. To support DOE's attainment of its goals, the NWTP is to provide (1) documented technology necessary for the design and operation of nuclear waste treatment facilities by commercial enterprises as part of a licensed waste management system and (2) problem-specific treatment approaches, waste form and treatment process adaptations, equipment designs, and trouble-shooting assistance, as required, to treat existing wastes. This annual report describes progress during FY 1985 toward meeting these two objectives. The detailed presentation is organized according to the task structure of the program.

Powell, J.A. (ed.)

1986-04-01T23:59:59.000Z

71

Nuclear waste treatment program: Annual report for FY 1987  

SciTech Connect (OSTI)

Two of the US Department of Energy's (DOE) nuclear waste management-related goals are to ensure that waste management is not an obstacle to the further development of light-water reactors and the closure of the nuclear fuel cycle and to fulfill its institutional responsibility for providing safe storage and disposal of existing and future nuclear wastes. As part of its approach to achieving these goals, the Office of Remedial Action and Waste Technology of DOE established what is now called the Nuclear Waste Treatment Program (NWTP) at the Pacific Northwest Laboratory during the second half of FY 1982. To support DOE's attainment of its goals, the NWTP is to provide technology necessary for the design and operation of nuclear waste treatment facilities by commercial enterprises as part of a licensed waste management system and problem-specific treatment approaches, waste form and treatment process adaptations, equipment designs, and trouble-shooting assistance, as required to treat existing wastes. This annual report describes progress during FY 1987 towards meeting these two objectives. 24 refs., 59 figs., 24 tabs.

Brouns, R.A.; Powell, J.A. (comps.)

1988-09-01T23:59:59.000Z

72

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents [OSTI]

A method for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package.

Forsberg, Charles W. (Oak Ridge, TN)

1998-01-01T23:59:59.000Z

73

Depleted uranium as a backfill for nuclear fuel waste package  

DOE Patents [OSTI]

A method is described for packaging spent nuclear fuel for long-term disposal in a geological repository. At least one spent nuclear fuel assembly is first placed in an unsealed waste package and a depleted uranium fill material is added to the waste package. The depleted uranium fill material comprises flowable particles having a size sufficient to substantially fill any voids in and around the assembly and contains isotopically-depleted uranium in the +4 valence state in an amount sufficient to inhibit dissolution of the spent nuclear fuel from the assembly into a surrounding medium and to lessen the potential for nuclear criticality inside the repository in the event of failure of the waste package. Last, the waste package is sealed, thereby substantially reducing the release of radionuclides into the surrounding medium, while simultaneously providing radiation shielding and increased structural integrity of the waste package. 6 figs.

Forsberg, C.W.

1998-11-03T23:59:59.000Z

74

acidic nuclear wastes: Topics by E-print Network  

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

acidic nuclear wastes First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 NuclearNuclear ""BurningBurning""...

75

automated nuclear waste: Topics by E-print Network  

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

automated nuclear waste First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 NuclearNuclear...

76

acidic nuclear waste: Topics by E-print Network  

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

acidic nuclear waste First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 NuclearNuclear ""BurningBurning""...

77

Design of the Prototypical Cryomodule for the EUROTRANS Superconducting Linac for Nuclear Waste Transmutation  

E-Print Network [OSTI]

Design of the Prototypical Cryomodule for the EUROTRANS Superconducting Linac for Nuclear Waste Transmutation

Barbanotti, S; Blache, P; Commeaux, C; Duthil, P; Panzeri, N; Pierini, P; Rampnoux, E; Souli, M

2008-01-01T23:59:59.000Z

78

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

79

Process to separate transuranic elements from nuclear waste  

DOE Patents [OSTI]

A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR).

Johnson, Terry R. (Wheaton, IL); Ackerman, John P. (Downers Grove, IL); Tomczuk, Zygmunt (Orland Park, IL); Fischer, Donald F. (Glen Ellyn, IL)

1989-01-01T23:59:59.000Z

80

Method for forming microspheres for encapsulation of nuclear waste  

DOE Patents [OSTI]

Microspheres for nuclear waste storage are formed by gelling droplets containing the waste in a gelation fluid, transferring the gelled droplets to a furnace without the washing step previously used, and heating the unwashed gelled droplets in the furnace under temperature or humidity conditions that result in a substantially linear rate of removal of volatile components therefrom.

Angelini, Peter (Oak Ridge, TN); Caputo, Anthony J. (Knoxville, TN); Hutchens, Richard E. (Knoxville, TN); Lackey, Walter J. (Oak Ridge, TN); Stinton, David P. (Knoxville, TN)

1984-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

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

82

State of Nevada, Agency for Nuclear Projects/Nuclear Waste Project Office narrative report, January 1992  

SciTech Connect (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

1992-12-31T23:59:59.000Z

83

Department of Energy plan for recovery and utilization of nuclear byproducts from defense wastes. Volume 2  

SciTech Connect (OSTI)

Nuclear wastes from the defense production cycle contain many uniquely useful, intrinsically valuable, and strategically important materials. These materials have a wide range of known and potential applications in food technology, agriculture, energy, public health, medicine, industrial technology, and national security. Furthermore, their removal from the nuclear waste stream can facilitate waste management and yield economic, safety, and environmental advantages in the management and disposal of the residual nuclear wastes that have no redemptive value. This document is the program plan for implementing the recovery and beneficial use of these valuable materials. An Executive Summary of this document, DOE/DP-0013, Vol. 1, January 1983, is available. Program policy, goals and strategy are stated in Section 2. Implementation tasks, schedule and funding are detailed in Section 3. The remaining five sections and the appendixes provide necessary background information to support these two sections. Section 4 reviews some of the unique properties of the individual byproduct materials and describes both demonstrated and potential applications. The amounts of byproduct materials that are available now for research and demonstration purposes, and the amounts that could be recovered in the future for expanded applications are detailed in Section 5. Section 6 describes the effects byproduct recovery and utilization have on the management and final disposal of nuclear wastes. The institutional issues that affect the recovery, processing and utilization of nuclear byproducts are discussed in Section 7. Finally, Section 8 presents a generalized mathematical process by which applications can be evaluated and prioritized (rank-ordered) to provide planning data for program management.

Not Available

1983-08-01T23:59:59.000Z

84

Expected brine movement at potential nuclear waste repository salt sites  

SciTech Connect (OSTI)

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

85

alumina nuclear waste: Topics by E-print Network  

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

alumina nuclear waste First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Nuclear magnetic resonance-based...

86

Nuclear physics information needed for accelerator driven transmutation of nuclear waste  

SciTech Connect (OSTI)

There is renewed interest in using accelerator driven neutron sources to address the problem of high-level long-lived nuclear waste. Several laboratories have developed systems that may have a significant impact on the future use of nuclear power, adding options for dealing with long-lived actinide wastes and fission products, and for power production. This paper describes a new Los Alamos concept using thermal neutrons and examines the nuclear data requirements. 7 refs., 3 figs., 1 tab.

Lisowski, P.W.; Bowman, C.D.; Arthur, E.D.; Young, P.G.

1991-01-01T23:59:59.000Z

87

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect (OSTI)

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

A. Alsaed

2005-07-28T23:59:59.000Z

88

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect (OSTI)

The purpose of this calculation is to estimate the probability of misloading a commercial spent nuclear fuel waste package with a fuel assembly(s) that has a reactivity (i.e., enrichment and/or burnup) outside the waste package design. The waste package designs are based on the expected commercial spent nuclear fuel assemblies and previous analyses (Macheret, P. 2001, Section 4.1 and Table 1). For this calculation, a misloaded waste package is defined as a waste package that has a fuel assembly(s) loaded into it with an enrichment and/or burnup outside the waste package design. An example of this type of misload is a fuel assembly designated for the 21-PWR Control Rod waste package being incorrectly loaded into a 21-PWR Absorber Plate waste package. This constitutes a misloaded 21-PWR Absorber Plate waste package, because the reactivity (i.e., enrichment and/or burnup) of a 21-PWR Control Rod waste package fuel assembly is outside the design of a 21-PWR Absorber Plate waste package. These types of misloads (i.e., fuel assembly with enrichment and/or burnup outside waste package design) are the only types that are evaluated in this calculation. This calculation utilizes information from ''Frequency of SNF Misload for Uncanistered Fuel Waste Package'' (CRWMS M&O 1998) as the starting point. The scope of this calculation is limited to the information available. The information is based on the whole population of fuel assemblies and the whole population of waste packages, because there is no information about the arrival of the waste stream at this time. The scope of this calculation deviates from that specified in ''Technical Work Plan for: Risk and Criticality Department'' (BSC 2002a, Section 2.1.30) in that only waste package misload is evaluated. The remaining issues identified (i.e., flooding and geometry reconfiguration) will be addressed elsewhere. The intended use of the calculation is to provide information and inputs to the Preclosure Safety Analysis Department. Before using the results of this calculation, the reader is cautioned to verify that the assumptions made in this calculation regarding the waste stream, the loading process, and the staging of the spent nuclear fuel assemblies are applicable.

J.K. Knudson

2003-10-02T23:59:59.000Z

89

Nuclear waste policy and public acceptance in France  

SciTech Connect (OSTI)

In France, the development of an extensive nuclear program has traditionally met relative support from the public and the politicals. Yet, facing some unusual opposition in 1991, the government declared a moratorium on the selection process of a geological repository. The authors review in this paper the successive steps of the nuclear waste storage program over the last 12 years, from the successful siting of two LLW storage facilities by ANDRA, the national agency for the storage of nuclear waste, to the more difficult years of the search for a suitable site to host the HLW repository which led to a new approach of the issue.

Guais, J.C. [NUSYS, Paris (France)

1993-12-31T23:59:59.000Z

90

RH-TRU Waste Shipments from Battelle Columbus Laboratories to the Hanford Nuclear Facility for Interim Storage  

SciTech Connect (OSTI)

Battelle Columbus Laboratories (BCL), located in Columbus, Ohio, must complete decontamination and decommissioning (D&D) activities for nuclear research buildings and grounds by 2006, as directed by Congress. Most of the resulting waste (approximately 27 cubic meters [m3]) is remote-handled (RH) transuranic (TRU) waste destined for disposal at the Waste Isolation Pilot Plant (WIPP). The BCL, under a contract to the U.S. Department of Energy (DOE) Ohio Field Office, has initiated a plan to ship the TRU waste to the DOE Hanford Nuclear Facility (Hanford) for interim storage pending the authorization of WIPP for the permanent disposal of RH-TRU waste. The first of the BCL RH-TRU waste shipments was successfully completed on December 18, 2002. This BCL shipment of one fully loaded 10-160B Cask was the first shipment of RH-TRU waste in several years. Its successful completion required a complex effort entailing coordination between different contractors and federal agencies to establish necessary supporting agreements. This paper discusses the agreements and funding mechanisms used in support of the BCL shipments of TRU waste to Hanford for interim storage. In addition, this paper presents a summary of the efforts completed to demonstrate the effectiveness of the 10-160B Cask system. Lessons learned during this process are discussed and may be applicable to other TRU waste site shipment plans.

Eide, J.; Baillieul, T. A.; Biedscheid, J.; Forrester, T,; McMillan, B.; Shrader, T.; Richterich, L.

2003-02-26T23:59:59.000Z

91

U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 20042009  

E-Print Network [OSTI]

Appendix G Appendix G U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 2004­2009 (Revised March 2004) Statement of the Board The Nuclear Waste Policy Amendments Act of 1987 directed the U-level radioactive waste. The Act also established the U.S. Nuclear Waste Technical Review Board as an indepen dent

92

Introduction This paper provides the perspective of the members of the Nuclear Waste Tech-  

E-Print Network [OSTI]

Introduction This paper provides the perspective of the members of the Nuclear Waste Tech- nical.S. program for managing spent nuclear fuel and high-level radioactive waste. It discusses the Board's opinion Waste Management Program In 1982, the U.S. Congress enacted the Nuclear Waste Policy Act (Public Law 97

93

U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 20042009  

E-Print Network [OSTI]

Appendix G Appendix G U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 2004­2009 (Revised March 2004) Statement of the Board The Nuclear Waste Policy Amendments Act of 1987 directed the U-level radioactive waste. The Act also established the U.S. Nuclear Waste Technical Review Board as an independent

94

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

SciTech Connect (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

95

Nuclear Waste Disposal: An Independent View of the Big Picture and a Proposal for CARD  

E-Print Network [OSTI]

1 Nuclear Waste Disposal: An Independent View of the Big Picture and a Proposal for CARD Presented to isolate nuclear waste successfully from the biosphere for the long term can be developed if our society to this impasse? In the 1940's at the beginning of the nuclear age, nuclear waste was seen as a "problem" only

California at Santa Cruz, University of

96

ULTRASONIC ARRAY TECHNIQUE FOR THE INSPECTION OF COPPER LINED CANISTERS FOR NUCLEAR WASTE FUEL  

E-Print Network [OSTI]

ULTRASONIC ARRAY TECHNIQUE FOR THE INSPECTION OF COPPER LINED CANISTERS FOR NUCLEAR WASTE FUEL and Waste Management Co.) for encapsulation of nuclear waste. Due to the radiation emitted by the nuclear, and characterization. The applicability of linear array technique for inspection of copper lined canisters for nuclear

97

An Investigation into the Oxidation State of Molybdenum in Simplified High Level Nuclear Waste Glass Compositions  

E-Print Network [OSTI]

An Investigation into the Oxidation State of Molybdenum in Simplified High Level Nuclear Waste of Mo in glasses containing simplified simulated high level nuclear waste (HLW) streams has been originating from the reprocessing of spent nuclear fuel. Experiments using simulated nuclear waste streams

Sheffield, University of

98

Reproductive Life Events in the Population Living in the Vicinity of a Nuclear Waste Reprocessing Plant  

E-Print Network [OSTI]

: There is concern about the health of populations living close to nuclear waste reprocessing plants. We conducted the health of the population living in the vicinity of nuclear waste reprocessing plants was raised the Dounreay nuclear waste reprocessing plant (United Kingdom).[1] Similar studies around the French nuclear

Paris-Sud XI, Universit de

99

Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography  

E-Print Network [OSTI]

This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry storage containers. We show that the tracking of individual muons, as they enter and exit a structure, can potentially improve the accuracy and availability of data on nuclear waste and the contents of Dry Storage Containers (DSC) used for spent fuel storage at CANDU plants. This could be achieved in near real time, with the potential for unattended and remotely monitored operations. We show that the expected sensitivity, in the case of the DSC, exceeds the IAEA detection target for nuclear material accountancy.

Jonkmans, G; Jewett, C; Thompson, M

2012-01-01T23:59:59.000Z

100

The First Recovery Act Funded Waste Shipment depart from the Advanced Mixed  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear SecurityTensile Strain Switched Ferromagnetism in Layered NbS2 and NbSe2Different ImpactsTheRollingCompetitionWaste

Note: This page contains sample records for the topic "nuclear waste fund" 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

NUCLEAR WASTE GLASSES CONTINUOUS MELTING AND BULK VITRIFICAITON  

SciTech Connect (OSTI)

This contribution addresses various aspects of nuclear waste vitrification. Nuclear wastes have a variety of components and composition ranges. For each waste composition, the glass must be formulated to possess acceptable processing and product behavior defined in terms of physical and chemical properties that guarantee the glass can be easily made and resist environmental degradation. Glass formulation is facilitated by developing property-composition models, and the strategy of model development and application is reviewed. However, the large variability of waste compositions presents numerous additional challenges: insoluble solids and molten salts may segregate; foam may hinder heat transfer and slow down the process; molten salts may accumulate in container refractory walls; the glass on cooling may precipitate crystalline phases. These problems need targeted exploratory research. Examples of specific problems and their possible solutions are discussed.

KRUGER AA; HRMA PR

2008-03-24T23:59:59.000Z

102

National briefing summaries: Nuclear fuel cycle and waste management  

SciTech Connect (OSTI)

Since 1976, the International Program Support Office (IPSO) at the Pacific Northwest Laboratory (PNL) has collected and compiled publicly available information concerning foreign and international radioactive waste management programs. This National Briefing Summaries is a printout of an electronic database that has been compiled and is maintained by the IPSO staff. The database contains current information concerning the radioactive waste management programs (with supporting information on nuclear power and the nuclear fuel cycle) of most of the nations (except eastern European countries) that now have or are contemplating nuclear power, and of the multinational agencies that are active in radioactive waste management. Information in this document is included for three additional countries (China, Mexico, and USSR) compared to the prior issue. The database and this document were developed in response to needs of the US Department of Energy.

Schneider, K.J.; Bradley, D.J.; Fletcher, J.F.; Konzek, G.J.; Lakey, L.T.; Mitchell, S.J.; Molton, P.M.; Nightingale, R.E.

1991-04-01T23:59:59.000Z

103

Public meetings on nuclear waste management: their function and organization  

SciTech Connect (OSTI)

This report focuses on public meetings as a vehicle for public participation in nuclear waste management. The nature of public meetings is reviewed and the functions served by meetings highlighted. The range of participants and their concerns are addressed, including a review of the participants from past nuclear waste management meetings. A sound understanding of the expected participants allows DOE to tailor elements of the meeting, such as notification, format, and agenda to accommodate the attendees. Finally, the report discusses the organization of public meetings on nuclear waste management in order to enhance the DOE's functions for such meetings. Possible structures are suggested for a variety of elements that are relevant prior to, during and after the public meeting. These suggestions are intended to supplement the DOE Public Participation Manual.

Duvernoy, E.G.; Marcus, A.A.; Overcast, T.; Schilling, A.H.

1981-05-01T23:59:59.000Z

104

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

SciTech Connect (OSTI)

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

105

Disposal of radioactive waste from nuclear research facilities  

E-Print Network [OSTI]

Swiss radioactive wastes originate from nuclear power plants (NPP) and from medicine (e.g. radiation sources), industry (e.g. fire detectors) and research (e.g. CERN, PSI). Their conditioning, characterisation and documentation has to meet the demands given by the Swiss regulatory authorities including all information needed for a safe disposal in future repositories. For NPP wastes, arisings as well as the processes responsible for the buildup of short and long lived radionuclides are well known, and the conditioning procedures are established. The radiological inventories are determined on a routinely basis using a combined system of measurements and calculational programs. For waste from research, the situation is more complicated. The wide spectrum of different installations combined with a poorly known history of primary and secondary radiation results in heterogeneous waste sorts with radiological inventories quite different from NPP waste and difficult to measure long lived radionuclides. In order to c...

Maxeiner, H; Kolbe, E

2003-01-01T23:59:59.000Z

106

Process to separate transuranic elements from nuclear waste  

DOE Patents [OSTI]

A process is described for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

1989-03-21T23:59:59.000Z

107

Process to separate transuranic elements from nuclear waste  

DOE Patents [OSTI]

A process for removing transuranic elements from a waste chloride electrolytic salt containing transuranic elements in addition to rare earth and other fission product elements so the salt waste may be disposed of more easily and the valuable transuranic elements may be recovered for reuse. The salt is contacted with a cadmium-uranium alloy which selectively extracts the transuranic elements from the salt. The waste salt is generated during the reprocessing of nuclear fuel associated with the Integral Fast Reactor (IFR). 2 figs.

Johnson, T.R.; Ackerman, J.P.; Tomczuk, Z.; Fischer, D.F.

1988-07-12T23:59:59.000Z

108

Salt disposal of heat-generating nuclear waste.  

SciTech Connect (OSTI)

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

109

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network [OSTI]

to evaluate the development of the transportation, aging, and disposal (TAD) canister system. The of Energy's (DOE) activities related to disposing of, packaging, and transporting high-level radioactive to develop a proposed repository for the permanent disposal of such waste at Yucca Mountain in Nevada

110

DESIGN OF THE PROTOTYPICAL CRYOMODULE FOR THE EUROTRANS SUPERCONDUCTING LINAC FOR NUCLEAR WASTE  

E-Print Network [OSTI]

DESIGN OF THE PROTOTYPICAL CRYOMODULE FOR THE EUROTRANS SUPERCONDUCTING LINAC FOR NUCLEAR WASTE of the accelerator workpackage of the EUROTRANS program for the design of a nuclear waste transmutation system

Boyer, Edmond

111

Waste Minimization Policy at the Romanian Nuclear Power Plant  

SciTech Connect (OSTI)

The radioactive waste management system at Cernavoda Nuclear Power Plant (NPP) in Romania was designed to maintain acceptable levels of safety for workers and to protect human health and the environment from exposure to unacceptable levels of radiation. In accordance with terminology of the International Atomic Energy Agency (IAEA), this system consists of the ''pretreatment'' of solid and organic liquid radioactive waste, which may include part or all of the following activities: collection, handling, volume reduction (by an in-drum compactor, if appropriate), and storage. Gaseous and aqueous liquid wastes are managed according to the ''dilute and discharge'' strategy. Taking into account the fact that treatment/conditioning and disposal technologies are still not established, waste minimization at the source is a priority environmental management objective, while waste minimization at the disposal stage is presently just a theoretical requirement for future adopted technologies . The necessary operational and maintenance procedures are in place at Cernavoda to minimize the production and contamination of waste. Administrative and technical measures are established to minimize waste volumes. Thus, an annual environmental target of a maximum 30 m3 of radioactive waste volume arising from operation and maintenance has been established. Within the first five years of operations at Cernavoda NPP, this target has been met. The successful implementation of the waste minimization policy has been accompanied by a cost reduction while the occupational doses for plant workers have been maintained at as low as reasonably practicable levels. This paper will describe key features of the waste management system along with the actual experience that has been realized with respect to minimizing the waste volumes at the Cernavoda NPP.

Andrei, V.; Daian, I.

2002-02-26T23:59:59.000Z

112

GEOHYDROLOGICAL STUDIES FOR NUCLEAR WASTE ISOLATION AT THE HANFORD RESERVATION -- Vol. I: Executive Summary; Vol. II: Final Report  

E-Print Network [OSTI]

NUCLEAR WASTE ISOLATION AT THE HANFORD RESERVATION Volume I:of Washington state." Rockwell Hanford Operations Topicalmodel evaluation at the Hanford nuclear waste facility."

Apps, J.

2010-01-01T23:59:59.000Z

113

Introduction to Nuclear Waste Management Nuclear Waste is a type of radioactive waste that is usually the by-product of  

E-Print Network [OSTI]

-Difficulty storing radioactive material -Waste disposal (heavy water, space jettison, underground) -Boat Transport -Impact water supply, nuclear fallout -6% world energy from nuclear energy, 14% world that is run through a turbine to produce energy. -Water or liquid metal are used to cool reactors. What

Auerbach, Scott M.

114

Nuclear Waste Technical Review Board Strategic Plan FY 20082013  

E-Print Network [OSTI]

on compliance activities, in conduct- ing its evaluation, the Board will encourage DOE through its science its review of DOE activities into three technical areas: preclosure operations, including surface-facility design and operations and the transport of spent nuclear fuel and high-level radioactive waste from

115

aqueous nuclear wastes: Topics by E-print Network  

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

aqueous nuclear wastes First Page Previous Page 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 Next Page Last Page Topic Index 1 Standard practice for analysis...

116

Backfill composition for secondary barriers in nuclear waste repositories  

DOE Patents [OSTI]

A backfill composition for sorbing and retaining hazardous elements of nuclear wastes comprises 50 to 70% by weight of quartz, 10 to 30% by weight of montmorillonite, 1 to 10% by weight of phosphate mineral, 1 to 10% by weight of ferrous mineral, 1 to 10% by weight of sulfate mineral and 1 to 10% by weight of attapulgite.

Beall, G.W.; Allard, B.M.

1980-05-30T23:59:59.000Z

117

U.S. Nuclear Waste Technical Review Board  

E-Print Network [OSTI]

U.S. Nuclear Waste Technical Review Board Report to The U.S. Congressand The Secretary of Energy.S. Congress and The Secretary of Energy March 1, 2006­December 31, 2007 #12;NWTRB reports are available online at www.nwtrb.gov. The Board appreciates the assistance of DOE in providing many of the graphics

118

A ThreeDimensional Finite Element Simulation for Transport of Nuclear Waste Contamination in Porous Media  

E-Print Network [OSTI]

A Three­Dimensional Finite Element Simulation for Transport of Nuclear Waste Contamination of South Carolina, Columbia, South Carolina ABSTRACT: Model equations for transport of nuclear­waste based up on the inherent physics. A three­dimensional finite element method for nuclear waste

Ewing, Richard E.

119

U.S. NUclear WaSte techNical revieW Board  

E-Print Network [OSTI]

U.S. NUclear WaSte techNical revieW Board Report to The U.S. Congress and The Secretary STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD 2300 Clarendon Boulevard, Suite 1300 Arlington, VA 22201 June Speaker Hastert, Senator Stevens, and Secretary Bodman: The U.S. Nuclear Waste Technical Review Board

120

Nuclear Waste Technical Review Board Strategic Plan for FY 2001-2006  

E-Print Network [OSTI]

Appendix G Nuclear Waste Technical Review Board Strategic Plan for FY 2001-2006 (Revised March 2001) Statement of the Chairman The U.S. Nuclear Waste Technical Review Board was established as an independent agency of the United States Government on December 22, 1987, in the Nuclear Waste Policy Amendments Act

Note: This page contains sample records for the topic "nuclear waste fund" 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

U.S. Nuclear Waste Technical Review Board Jared L. Cohon, Ph.D.; Chairman  

E-Print Network [OSTI]

Appendix A Appendix A U.S. Nuclear Waste Technical Review Board Members Jared L. Cohon, Ph.D.; Chairman On June 29, 1995, President Bill Clinton appointed Jared Cohon to the Nuclear Waste Technical siting, including nuclear waste shipping and storage. In addition to his academic experience, he served

122

U.S. Nuclear Waste Technical Review Board Fiscal Year 2002-2007 Strategic Plan  

E-Print Network [OSTI]

Appendix G Appendix G U.S. Nuclear Waste Technical Review Board Fiscal Year 2002-2007 Strategic Plan Statement of the Chairman The U.S. Nuclear Waste Technical Review Board was established of Energy Mission The Board's mission, established in the Nuclear Waste Policy Amendments Act (NWPAA

123

Numerical methods for the simulation of a corrosion model in a nuclear waste deep repository $  

E-Print Network [OSTI]

Numerical methods for the simulation of a corrosion model in a nuclear waste deep repository $ C of the French nuclear waste management agency ANDRA, investigations are conducted to optimize and finalize by the Nuclear Waste Management Agency ANDRA Corresponding author. Phone: +49 30 20372 560, Fax: +49 30 2044975

Paris-Sud XI, Université de

124

Communication Between the U.S. Nuclear Waste Technical Review Board  

E-Print Network [OSTI]

Appendix F Appendix F Communication Between the U.S. Nuclear Waste Technical Review Board on Energy and Air Quality on March 25, 2004 143 #12;#12;Appendix F UNITED STATES NUCLEAR WASTE TECHNICAL much for your written questions related to my testimony on behalf of the Nuclear Waste Technical Review

125

Seismic modeling and analysis of a prototype heated nuclear waste storage tunnel, Yucca Mountain, Nevada  

E-Print Network [OSTI]

Seismic modeling and analysis of a prototype heated nuclear waste storage tunnel, Yucca Mountain was heated to replicate the effects of long-term storage of decaying nuclear waste and to study the effects for the long- term storage of high-level nuclear waste from reactors and decom- missioned atomic weapons

Snieder, Roel

126

Abbreviations and Acronyms Board U. S. Nuclear Waste Technical Review Board  

E-Print Network [OSTI]

Abbreviations and Acronyms Board U. S. Nuclear Waste Technical Review Board CFR Code of Federal Commission NWTRB U.S. Nuclear Waste Technical Review Board PCI pellet-cladding interaction PTn Paintbrush, Conceptual Model, and Results of the TSPA-VA." Presentation to Nuclear Waste Technical Review Board's Panel

127

Numerical Zoom for Multiscale Problems with an Application to Nuclear Waste Disposal  

E-Print Network [OSTI]

Numerical Zoom for Multiscale Problems with an Application to Nuclear Waste Disposal Jean of a nuclear waste repository site. Key words: Multiscale, Finite Element, Domain Decomposition, Chimera, Numerical Zoom, Nuclear Waste. PACS: 02.30.Jr, 47.11.Fg, 28.41.Kw, 47.55.P- 1 Introduction The present paper

128

U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 20032008  

E-Print Network [OSTI]

Appendix G Appendix G U.S. Nuclear Waste Technical Review Board Strategic Plan: Fiscal Years 2003­2008 Statement of the Chairman The Nuclear Waste Policy Amendments Act of 1987 directed the U.S. Department as the location of a permanent repository for dis posing of spent nuclear fuel and high-level radioactive waste

129

Response to West Cumbria MRWS consultation: Why a deep nuclear waste repository should not be  

E-Print Network [OSTI]

Response to West Cumbria MRWS consultation: Why a deep nuclear waste repository should not be sited geological nuclear waste repository. There a suspicion of predetermination because the only district that has. National and international guidance on how best to select potential sites for deep geological nuclear waste

130

SWAMI: An Autonomous Mobile Robot for Inspection of Nuclear Waste Storage Facilities  

E-Print Network [OSTI]

SWAMI: An Autonomous Mobile Robot for Inspection of Nuclear Waste Storage Facilities Ron Fulbright Inspector (SWAMI) is a prototype mobile robot designed to perform autonomous inspection of nuclear waste user interface building tool called UIM/X. Introduction Safe disposal of nuclear waste is a difficult

Stephens, Larry M.

131

Radioactive Waste Management in Non-Nuclear Countries - 13070  

SciTech Connect (OSTI)

This paper challenges internationally accepted concepts of dissemination of responsibilities between all stakeholders involved in national radioactive waste management infrastructure in the countries without nuclear power program. Mainly it concerns countries classified as class A and potentially B countries according to International Atomic Energy Agency. It will be shown that in such countries long term sustainability of national radioactive waste management infrastructure is very sensitive issue that can be addressed by involving regulatory body in more active way in the infrastructure. In that way countries can mitigate possible consequences on the very sensitive open market of radioactive waste management services, comprised mainly of radioactive waste generators, operators of end-life management facilities and regulatory body. (authors)

Kubelka, Dragan; Trifunovic, Dejan [SORNS, Frankopanska 11, HR-10000 Zagreb (Croatia)] [SORNS, Frankopanska 11, HR-10000 Zagreb (Croatia)

2013-07-01T23:59:59.000Z

132

Nuclear Waste Removal Using Particle Beams Incineration with Fast Neutrons  

E-Print Network [OSTI]

The management of nuclear waste is one of the major obstacles to the acceptability of nuclear power as a main source of energy for the future. TARC, a new experiment at CERN, is testing the practicality of Carlo Rubbia's idea to make use of Adiabatic Resonance Crossing to transmute long-lived fission fragments into short-lived or stable nuclides. Spallation neutrons produced in a large Lead assembly have a high probability to be captured at the energies of cross-section resonances in elements such as 99Tc, 129I, etc. An accelerator-driven sub-critical device using Thorium (Energy Amplifier) would be very effective in eliminating TRansUranic elements which constitute the most dangerous part of nuclear waste while producing from it large amounts of energy. In addition, such a system could transform, at a high rate and little energetic cost, long-lived fission fragments into short-lived elements.

Revol, Jean Pierre Charles

1997-01-01T23:59:59.000Z

133

Malonamides as new extractants for nuclear waste solutions  

SciTech Connect (OSTI)

A new class of extractants has been investigated: pentaalkylpropane diamides. Due to their chelating effect on metallic cations, they extract trivalent actinides such as Am{sup 3+} and Cm{sup 3+} from acidic nitrate nuclear waste solutions. These solvents are completely incinerable and do not generate large amounts of waste. A review is provided of their chemical properties, leading to the choice of the proper molecule. The results of a bench-scale experiment performed in a mixer-settler battery are presented.

Cuillerdier, C.; Musikas, C.; Hoel, P.; Nigond, L.; Vitart, X. (Commissariat a L'Energie Atomique, Fontenay-aux-Roses (France))

1991-09-01T23:59:59.000Z

134

Process for recovery of palladium from nuclear fuel reprocessing wastes  

DOE Patents [OSTI]

Palladium is selectively removed from spent nuclear fuel reprocessing waste by adding sugar to a strong nitric acid solution of the waste to partially denitrate the solution and cause formation of an insoluble palladium compound. The process includes the steps of: (a) adjusting the nitric acid content of the starting solution to about 10 M; (b) adding 50% sucrose solution in an amount sufficient to effect the precipitation of the palladium compound; (c) heating the solution at reflux temperature until precipitation is complete; and (d) centrifuging the solution to separate the precipitated palladium compound from the supernatant liquid.

Campbell, D.O.; Buxton, S.R.

1980-06-16T23:59:59.000Z

135

Coincidence counter design for the assay of vitrified nuclear waste  

SciTech Connect (OSTI)

For the termination of nuclear safeguards and transfers to waste storage, the verification of the plutonium content in vitrified nuclear waste is required by international safeguards agreements. A novel design has been used to develop a coincidence counter for measuring vitrified nuclear waste. The authors have devised a method to measure the {sup 244}Cm content and to calculate the plutonium content from the curium-to-plutonium ratio. In order to provide unattended inspection, the counter is designed for continuous operation in the presence of highly radioactive samples: 3.0 {times} 10{sup 7} Rad/h gamma and 9.0 {times} 10{sup 7}/s neutron fluence. Operability under these conditions has been obtained by designing a heavily shielded detector with radiation hard components subtending a limited solid angle. A counting technique, Localized Source Term Coincidence Counting, has been developed to allow neutron assay of this type of sample. The system will be installed at the Power Nuclear Corporation Tokai Vitrification Facility in the later part of 1998.

Beddingfield, D.H.; Menlove, H.O. [Los Alamos National Lab., NM (United States); Iwamoto, T.; Tomikawa, H. [Power Nuclear Corp. (Japan)

1998-12-31T23:59:59.000Z

136

Nuclear Waste Policy Act Signed | National Nuclear Security Administration  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the ContributionsArms Control R&D ConsortiumNuclearSafeguardsResearch

137

"Hanford: A Conversation About Nuclear Waste and Cleanup"  

SciTech Connect (OSTI)

In ''Hanford: A Conversation about Nuclear Waste and Cleanup'', Roy Gephart takes us on a journey through a world of facts, values, conflicts, and choices facing the most complex environmental cleanup project in the United States, the U.S. Department of Energy's Hanford Site. Starting with the top-secret Manhattan Project, Hanford was used to create tons of plutonium for nuclear weapons. Hundreds of tons of waste remain. In an easy-to-read, illustrated text, Gephart crafts the story of Hanford becoming the world's first nuclear weapons site to release large amounts of contaminants into the environment. This was at a time when radiation biology was in its infancy, industry practiced unbridled waste dumping, and the public trusted what it was told. The plutonium market stalled with the end of the Cold War. Public accountability and environmental compliance ushered in a new cleanup mission. Today, Hanford is driven by remediation choices whose outcomes remain uncertain. It's a story whose epilogue will be written by future generations. This book is an information resource, written for the general reader as well as the technically trained person wanting an overview of Hanford and cleanup issues facing the nuclear weapons complex. Each chapter is a topical mini-series. It's an idea guide that encourages readers to be informed consumers of Hanford news, to recognize that knowledge, high ethical standards, and social values are at the heart of coping with Hanford's past and charting its future. Hanford history is a window into many environmental conflicts facing our nation; it's about building upon success and learning from failure. And therein lies a key lesson, when powerful interests are involved, no generation is above pretense. Roy E. Gephart is a geohydrologist and senior program manager at the Pacific Northwest National Laboratory, Richland, Washington. He has 30 years experience in environmental studies and the nuclear waste industry.

Gephart, Roy E.

2003-05-10T23:59:59.000Z

138

Nuclear Waste Policy Act | Department of Energy  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAXBalanced ScorecardReactor Technology Subcommittee of NEACSummary NucleariNuclear

139

Nuclear waste vitrification efficiency: Cold cap reactions  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas Conchas recoveryLaboratory | NationalJohnSecurityControls |NavyNuclearLife

140

OFFICE OF RIVER PROTECTION NUCLEAR WASTE PROGRAM  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the ContributionsArms Control R&DNuclear fuel recycling inNewsTimSF-428S Supplemental

Note: This page contains sample records for the topic "nuclear waste fund" 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

Small businesses selected for nuclear waste services  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administrationcontroller systemsBiSite Cultural ResourcestepidumProjectsMore than 140Small

142

EIS-0203: Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs  

Broader source: Energy.gov [DOE]

Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs

143

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

SciTech Connect (OSTI)

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 Sites (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

144

Alcohol-free alkoxide process for containing nuclear waste  

DOE Patents [OSTI]

Disclosed is a method of containing nuclear waste. A composition is first prepared of about 25 to about 80%, calculated as SiO.sub.2, of a partially hydrolyzed silicon compound, up to about 30%, calculated as metal oxide, of a partially hydrolyzed aluminum or calcium compound, about 5 to about 20%, calculated as metal oxide, of a partially hydrolyzed boron or calcium compound, about 3 to about 25%, calculated as metal oxide, of a partially hydrolyzed sodium, potassium or lithium compound, an alcohol in a weight ratio to hydrolyzed alkoxide of about 1.5 to about 3% and sufficient water to remove at least 99% of the alcohol as an azeotrope. The azeotrope is boiled off and up to about 40%, based on solids in the product, of the nuclear waste, is mixed into the composition. The mixture is evaporated to about 25 to about 45% solids and is melted and cooled.

Pope, James M. (Monroeville, PA); Lahoda, Edward J. (Edgewood, PA)

1984-01-01T23:59:59.000Z

145

Transmutation of nuclear waste in accelerator-driven systems  

E-Print Network [OSTI]

Today more than ever energy is not only a cornerstone of human development, but also a key to the environmental sustainability of economic activity. In this context, the role of nuclear power may be emphasized in the years to come. Nevertheless, the problems of nuclear waste, safety and proliferation still remain to be solved. It is believed that the use of accelerator-driven systems (ADSs) for nuclear waste transmutation and energy production would address these problems in a simple, clean and economically viable, and therefore sustainable, manner. This thesis covers the major nuclear physics aspects of ADSs, in particular the spallation process and the core neutronics specific to this type of systems. The need for accurate nuclear data is described, together with a detailed analysis of the specific isotopes and energy ranges in which this data needs to be improved and the impact of their uncertainty. Preliminary experimental results for some of these isotopes, produced by the Neutron Time-of-Flight (n_TOF) ...

Herrera-Martnez, A

2004-01-01T23:59:59.000Z

146

Status of the Nevada Nuclear Waste Storage investigations  

SciTech Connect (OSTI)

The Nevada Nuclear Waste Storage Investigations (NNWSI) are part of the National Waste Terminal Storage (NWTS) program being conducted by the Department of Energy. Within the NWTS program, the NNWSI is the component that focuses on siting evaluations on and near the Nevada Test Site (NTS). The objectives of the Nevada project include evaluating the suitability of a Test and Evaluation Facility (TEF) site on or near the NTS, evaluating the suitability of a commercial nuclear waste repository site on or near the NTS, and supporting the NWTS program with research that is uniquely possible at NTS. Current engineering studies suggest that TEF and repository surface facilities would need to be located on gently sloping alluvium east of Yucca Mountain. Access from surface facilities to underground waste emplacement areas would be by vertical shafts and horizontal drifts, or possibly by inclined adits. The current NNWSI schedule includes an exploratory shaft location and horizon recommendation in 12/82, with a start of exploratory shaft drilling in 9/83. Because of the complexities of horizon selection, it is possible that the exploratory shaft depth or horizon recommendation may involve the exploration of more than one horizon. Phase I of the exploratory shaft, determination of TEF site suitability, is currently scheduled for 7/85. Phase II of the exploratory shaft, determination of repository site suitability, is currently scheduled for 3/87. This schedule is consistent with the current NWTS TEF and repository site selection schedules.

Lincoln, R. C.

1982-01-01T23:59:59.000Z

147

Potential applications of nanostructured materials in nuclear waste management.  

SciTech Connect (OSTI)

This report summarizes the results obtained from a Laboratory Directed Research & Development (LDRD) project entitled 'Investigation of Potential Applications of Self-Assembled Nanostructured Materials in Nuclear Waste Management'. The objectives of this project are to (1) provide a mechanistic understanding of the control of nanometer-scale structures on the ion sorption capability of materials and (2) develop appropriate engineering approaches to improving material properties based on such an understanding.

Braterman, Paul S. (The University of North Texas, Denton, TX); Phol, Phillip Isabio; Xu, Zhi-Ping (The University of North Texas, Denton, TX); Brinker, C. Jeffrey; Yang, Yi (University of New Mexico, Albuquerque, NM); Bryan, Charles R.; Yu, Kui; Xu, Huifang (University of New Mexico, Albuquerque, NM); Wang, Yifeng; Gao, Huizhen

2003-09-01T23:59:59.000Z

148

Method of determining a content of a nuclear waste container  

DOE Patents [OSTI]

A method and apparatus are provided for identifying contents of a nuclear waste container. The method includes the steps of forming an image of the contents of the container using digital radiography, visually comparing contents of the image with expected contents of the container and performing computer tomography on the container when the visual inspection reveals an inconsistency between the contents of the image and the expected contents of the container.

Bernardi, Richard T. (Prospect Heights, IL); Entwistle, David (Buffalo Grove, IL)

2003-04-22T23:59:59.000Z

149

Consideration of nuclear criticality when disposing of transuranic waste at the Waste Isolation Pilot Plant  

SciTech Connect (OSTI)

Based on general arguments presented in this report, nuclear criticality was eliminated from performance assessment calculations for the Waste Isolation Pilot Plant (WIPP), a repository for waste contaminated with transuranic (TRU) radioisotopes, located in southeastern New Mexico. At the WIPP, the probability of criticality within the repository is low because mechanisms to concentrate the fissile radioisotopes dispersed throughout the waste are absent. In addition, following an inadvertent human intrusion into the repository (an event that must be considered because of safety regulations), the probability of nuclear criticality away from the repository is low because (1) the amount of fissile mass transported over 10,000 yr is predicted to be small, (2) often there are insufficient spaces in the advective pore space (e.g., macroscopic fractures) to provide sufficient thickness for precipitation of fissile material, and (3) there is no credible mechanism to counteract the natural tendency of the material to disperse during transport and instead concentrate fissile material in a small enough volume for it to form a critical concentration. Furthermore, before a criticality would have the potential to affect human health after closure of the repository--assuming that a criticality could occur--it would have to either (1) degrade the ability of the disposal system to contain nuclear waste or (2) produce significantly more radioisotopes than originally present. Neither of these situations can occur at the WIPP; thus, the consequences of a criticality are also low.

RECHARD,ROBERT P.; SANCHEZ,LAWRENCE C.; STOCKMAN,CHRISTINE T.; TRELLUE,HOLLY R.

2000-04-01T23:59:59.000Z

150

Nuclear Waste Technical Review Board Correspondence with the Department of Energy  

E-Print Network [OSTI]

85 Appendix E Nuclear Waste Technical Review Board Correspondence with the Department of Energy #12;#12;Appendix E 87 Nuclear Waste Technical Review Board Correspondence with the Department of Energy I n of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The letters typically provide

151

Review of Concrete Biodeterioration in Relation to Buried Nuclear Waste  

SciTech Connect (OSTI)

Long-term storage of low level radioactive material in below ground concrete disposal units (DUs) (Saltstone Disposal Facility) is a means of depositing wastes generated from nuclear operations of the U.S. Department of Energy. Based on the currently modeled degradation mechanisms, possible microbial induced effects on the structural integrity of buried low level wastes must be addressed. Previous international efforts related to microbial impacts on concrete structures that house low level radioactive waste showed that microbial activity can play a significant role in the process of concrete degradation and ultimately structural deterioration. This literature review examines the recent research in this field and is focused on specific parameters that are applicable to modeling and prediction of the fate of concrete vaults housing stored wastes and the wastes themselves. Rates of concrete biodegradation vary with the environmental conditions, illustrating a need to understand the bioavailability of key compounds involved in microbial activity. Specific parameters require pH and osmotic pressure to be within a certain range to allow for microbial growth as well as the availability and abundance of energy sources like components involved in sulfur, iron and nitrogen oxidation. Carbon flow and availability are also factors to consider in predicting concrete biodegradation. The results of this review suggest that microbial activity in Saltstone, (grouted low level radioactive waste) is unlikely due to very high pH and osmotic pressure. Biodegradation of the concrete vaults housing the radioactive waste however, is a possibility. The rate and degree of concrete biodegradation is dependent on numerous physical, chemical and biological parameters. Results from this review point to parameters to focus on for modeling activities and also, possible options for mitigation that would minimize concrete biodegradation. In addition, key chemical components that drive microbial activity on concrete surfaces are discussed.

Turick, C; Berry, C.

2012-10-15T23:59:59.000Z

152

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC).  

SciTech Connect (OSTI)

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. Achieving the objective of modeling the performance of a disposal scenario requires describing processes involved in waste form degradation and radionuclide release at the subcontinuum scale, beginning with mechanistic descriptions of chemical reactions and chemical kinetics at the atomic scale, and upscaling into effective, validated constitutive models for input to high-fidelity continuum scale codes for coupled multiphysics simulations of release and transport. Verification and validation (V&V) is required throughout the system to establish evidence-based metrics for the level of confidence in M&S codes and capabilities, including at the subcontiunuum scale and the constitutive models they inform or generate. This Report outlines the nature of the V&V challenge at the subcontinuum scale, an approach to incorporate V&V concepts into subcontinuum scale modeling and simulation (M&S), and a plan to incrementally incorporate effective V&V into subcontinuum scale M&S destined for use in the NEAMS Waste IPSC work flow to meet requirements of quantitative confidence in the constitutive models informed by subcontinuum scale phenomena.

Schultz, Peter Andrew

2011-12-01T23:59:59.000Z

153

Nuclear waste repository transparency technology test bed demonstrations at WIPP  

SciTech Connect (OSTI)

Secretary of Energy, Bill Richardson, has stated that one of the nuclear waste legacy issues is ``The challenge of managing the fuel cycle's back end and assuring the safe use of nuclear power.'' Waste management (i.e., the back end) is a domestic and international issue that must be addressed. A key tool in gaining acceptance of nuclear waste repository technologies is transparency. Transparency provides information to outside parties for independent assessment of safety, security, and legitimate use of materials. Transparency is a combination of technologies and processes that apply to all elements of the development, operation, and closure of a repository system. A test bed for nuclear repository transparency technologies has been proposed to develop a broad-based set of concepts and strategies for transparency monitoring of nuclear materials at the back end of the fuel/weapons cycle. WIPP is the world's first complete geologic repository system for nuclear materials at the back end of the cycle. While it is understood that WIPP does not currently require this type of transparency, this repository has been proposed as realistic demonstration site to generate and test ideas, methods, and technologies about what transparency may entail at the back end of the nuclear materials cycle, and which could be applicable to other international repository developments. An integrated set of transparency demonstrations was developed and deployed during the summer, and fall of 1999 as a proof-of-concept of the repository transparency technology concept. These demonstrations also provided valuable experience and insight into the implementation of future transparency technology development and application. These demonstrations included: Container Monitoring Rocky Flats to WIPP; Underground Container Monitoring; Real-Time Radiation and Environmental Monitoring; Integrated level of confidence in the system and information provided. As the world's only operating deep geologic repository, the Waste Isolation Pilot Plant (WIPP) offers a unique opportunity to serve as an international cooperative test bed for developing and demonstrating technologies and processes in a fully operational repository system setting. To address the substantial national security implications for the US resulting from the lack of integrated, transparent management and disposition of nuclear materials at the back-end of the nuclear fuel and weapons cycles, it is proposed that WIPP be used as a test bed to develop and demonstrate technologies that will enable the transparent and proliferation-resistant geologic isolation of nuclear materials. The objectives of this initiative are to: (1) enhance public confidence in safe, secure geologic isolation of nuclear materials; (2) develop, test, and demonstrate transparency measures and technologies for the back-end of nuclear fuel cycle; and (3) foster international collaborations leading to workable, effective, globally-accepted standards for the transparent monitoring of geological repositories for nuclear materials. Test-bed activities include: development and testing of monitoring measures and technologies; international demonstration experiments; transparency workshops; visiting scientist exchanges; and educational outreach. These activities are proposed to be managed by the Department of Energy/Carlsbad Area Office (DOE/CAO) as part of The Center for Applied Repository and Underground Studies (CARUS).

BETSILL,J. DAVID; ELKINS,NED Z.; WU,CHUAN-FU; MEWHINNEY,JAMES D.; AAMODT,PAUL

2000-01-27T23:59:59.000Z

154

Removal of Pertechnetate from Simulated Nuclear Waste Streams Using Supported Zerovalent Iron  

E-Print Network [OSTI]

Removal of Pertechnetate from Simulated Nuclear Waste Streams Using Supported Zerovalent Iron John and reduction of pertechnetate anions (TcO4 -) from complex waste mixtures was investigated as an alternative approach to current waste-processing schemes. Although applicable to pertechnetate-containing waste streams

155

Conceivable new recycling of nuclear waste by nuclear power companies in their plants  

E-Print Network [OSTI]

We outline the basic principles and the needed experiments for a conceivable new recycling of nuclear waste by the power plants themselves to avoid its transportation and storage to a (yet unknown) dumping area. Details are provided in an adjoining paper and in patents pending.

Ruggero Maria Santilli

1997-04-09T23:59:59.000Z

156

National briefing summaries: Nuclear fuel cycle and waste management  

SciTech Connect (OSTI)

The National Briefing Summaries is a compilation of publicly available information concerning the nuclear fuel cycle and radioactive waste management strategies and programs of 21 nations, including the United States and three international agencies that have publicized their activities in this field. It presents available highlight information with references that may be used by the reader for additional information. The information in this document is compiled primarily for use by the US Department of Energy and other US federal agencies and their contractors to provide summary information on radioactive waste management activities in other countries. This document provides an awareness to managers and technical staff of what is occurring in other countries with regard to strategies, activities, and facilities. The information may be useful in program planning to improve and benefit United States' programs through foreign information exchange. Benefits to foreign exchange may be derived through a number of exchange activities.

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

1988-12-01T23:59:59.000Z

157

Cesium and Strontium Specific Exchangers for Nuclear Waste Effluent Remediation  

SciTech Connect (OSTI)

During the past 50 years, nuclear defense activities have produced large quantities of nuclear waste that now require safe and permanent disposal. The general procedure to be implemented involves the removal of cesium and strontium from the waste solutions for disposal in permanently vitrified media. This requires highly selective sorbents or ion exchangers. Further, at the high radiation doses present in the solution, organic exchangers or sequestrants are likely to decompose over time. Inorganic ion exchangers are resistant to radiation damage and can exhibit remarkably high selectivities. We have synthesized three families of tunnel-type ion exchangers. The crystal structures of these compounds as well as their protonated phases, coupled with ion exchange titrations, were determined and this information was used to develop an understanding of their ion exchange behavior. The ion exchange selectivities of these phases could be regulated by isomorphous replacement of the framework metals by larger or smaller radius metals. In the realm of layered compounds, we prepared alumina, silica, and zirconia pillared clays and sodium micas. The pillared clays yielded very high Kd values for Cs+ and were very effective in removing Cs+ from groundwaters. The sodium micas also had a high affinity for Cs+ but an even greater attraction for S42+. They also possess the property of trapping these ions permanently as the layers slowly decrease their interlayer distance as loading occurs. Sodium nonatitanate exhibited extremely high Kd values for Sr2+ in alkaline tank wastes and should be considered for removal of Sr2+ in such cases. For tank wastes containing complexing agents, we have found that adding Ca2+ to the solution releases the complexed Sr2+ which may then be removed with the CST exchanger.

A. Clearfield; A. I. Bortun; L. A. Bortun; E. A. Bhlume; P. Sylvester; G. M. Graziano

2000-09-01T23:59:59.000Z

158

Flammability Control In A Nuclear Waste Vitrification System  

SciTech Connect (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 in pretreatment has been studied to eliminate the production of hydrogen in the pretreatment systems, which requires nuclear grade monitoring equipment. An alternative reductant, glycolic acid, has been studied as a substitute for formic acid. However, in the melter, the potential for greater formation of flammable gases exists with glycolic acid. Melter flammability is difficult to control because flammable mixtures can be formed during surges in offgases that both increase the amount of flammable species and decrease the temperature in the vapor space of the melter. A flammable surge can exceed the 60% of the LFL with no way to mitigate it. Therefore, careful control of the melter feed composition based on scaled melter surge testing is required. The results of engineering scale melter tests with the formic-nitric flowsheet and the use of these data in the melter flammability model are presented.

Zamecnik, John R.; Choi, Alexander S.; Johnson, Fabienne C.; Miller, Donald H.; Lambert, Daniel P.; Stone, Michael E.; Daniel, William E. Jr.

2013-07-25T23:59:59.000Z

159

E-Print Network 3.0 - active nuclear wastes Sample Search Results  

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

disposal site for transuranic (TRU) radio- active waste created during... , americium, curium, and neptunium are created during the produc- tion of nuclear weapons. Transuranic...

160

E-Print Network 3.0 - assessing nuclear waste Sample Search Results  

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

University - Waste-to-Energy Research and Technology Council (WTERT) Collection: Renewable Energy 14 NRE 2110 Introduction to Nuclear and Radiological Engineering (Required)...

Note: This page contains sample records for the topic "nuclear waste fund" 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.


161

Corrosion-induced gas generation in a nuclear waste repository: Reactive geochemistry and multiphase flow effect  

E-Print Network [OSTI]

Mallants, D. , 2002. Gas generation and migration in Boomof Post-Disposal Gas Generation in a Repository for SpentCorrosion-Induced Gas Generation in a Nuclear Waste

Xu, T.

2009-01-01T23:59:59.000Z

162

THERMODYNAMIC TABLES FOR NUCLEAR WASTE ISOLATION, V.1: AQUEOUSSOLUTIONS DATABASE  

SciTech Connect (OSTI)

Tables of consistent thermodynamic property values for nuclear waste isolation are given. The tables include critically assessed values for Gibbs energy of formation. enthalpy of formation, entropy and heat capacity for minerals; solids; aqueous ions; ion pairs and complex ions of selected actinide and fission decay products at 25{sup o}C and zero ionic strength. These intrinsic data are used to calculate equilibrium constants and standard potentials which are compared with typical experimental measurements and other work. Recommendations for additional research are given.

Phillips, S.L.; Hale, F.V.; Silvester, L.F.

1988-05-01T23:59:59.000Z

163

Removing nuclear waste, one shipment at a time  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas ConchasPassive Solar HomePromisingStories » Removing nuclear waste, one shipment at a time

164

State of Nevada, Agency for Nuclear Projects/Nuclear Waste Project Office narrative report, October 1--December 31, 1991  

SciTech Connect (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

165

Nuclear waste management. Semiannual progress report, April 1983-September 1983  

SciTech Connect (OSTI)

The status of the following programs is reported: waste stabilization; waste isolation; low-level waste management; remedial action; and supporting studies. 58 figures, 39 tables.

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

1984-01-01T23:59:59.000Z

166

Characterising encapsulated nuclear waste using cosmic-ray muon tomography  

E-Print Network [OSTI]

Tomographic imaging techniques using the Coulomb scattering of cosmic-ray muons have been shown previously to successfully identify and characterise low- and high-Z materials within an air matrix using a prototype scintillating-fibre tracker system. Those studies were performed as the first in a series to assess the feasibility of this technology and image reconstruction techniques in characterising the potential high-Z contents of legacy nuclear waste containers for the UK Nuclear Industry. The present work continues the feasibility study and presents the first images reconstructed from experimental data collected using this small-scale prototype system of low- and high-Z materials encapsulated within a concrete-filled stainless-steel container. Clear discrimination is observed between the thick steel casing, the concrete matrix and the sample materials assayed. These reconstructed objects are presented and discussed in detail alongside the implications for future industrial scenarios.

Anthony Clarkson; David J. Hamilton; Matthias Hoek; David G. Ireland; John R. Johnstone; Ralf Kaiser; Tibor Keri; Scott Lumsden; David F. Mahon; Bryan McKinnon; Morgan Murray; Sin Nutbeam-Tuffs; Craig Shearer; Guangliang Yang; Colin Zimmerman

2014-10-27T23:59:59.000Z

167

Characterising encapsulated nuclear waste using cosmic-ray muon tomography  

E-Print Network [OSTI]

Tomographic imaging techniques using the Coulomb scattering of cosmic-ray muons have been shown previously to successfully identify and characterise low- and high-Z materials within an air matrix using a prototype scintillating-fibre tracker system. Those studies were performed as the first in a series to assess the feasibility of this technology and image reconstruction techniques in characterising the potential high-Z contents of legacy nuclear waste containers for the UK Nuclear Industry. The present work continues the feasibility study and presents the first images reconstructed from experimental data collected using this small-scale prototype system of low- and high-Z materials encapsulated within a concrete-filled stainless-steel container. Clear discrimination is observed between the thick steel casing, the concrete matrix and the sample materials assayed. These reconstructed objects are presented and discussed in detail alongside the implications for future industrial scenarios.

Clarkson, Anthony; Hoek, Matthias; Ireland, David G; Johnstone, John R; Kaiser, Ralf; Keri, Tibor; Lumsden, Scott; Mahon, David F; McKinnon, Bryan; Murray, Morgan; Nutbeam-Tuffs, Sin; Shearer, Craig; Yang, Guangliang; Zimmerman, Colin

2014-01-01T23:59:59.000Z

168

CHARACTERIZATION OF DEFENSE NUCLEAR WASTE USING HAZARDOUS WASTE GUIDANCE. APPLICATIONS TO HANFORD SITE ACCELERATED HIGH-LEVEL WASTE TREATMENT AND DISPOSAL MISSION0  

SciTech Connect (OSTI)

Federal hazardous waste regulations were developed for management of industrial waste. These same regulations are also applicable for much of the nation's defense nuclear wastes. At the U.S. Department of Energy's (DOE) Hanford Site in southeast Washington State, one of the nation's largest inventories of nuclear waste remains in storage in large underground tanks. The waste's regulatory designation and its composition and form constrain acceptable treatment and disposal options. Obtaining detailed knowledge of the tank waste composition presents a significant portion of the many challenges in meeting the regulatory-driven treatment and disposal requirements for this waste. Key in applying the hazardous waste regulations to defense nuclear wastes is defining the appropriate and achievable quality for waste feed characterization data and the supporting evidence demonstrating that applicable requirements have been met at the time of disposal. Application of a performance-based approach to demonstrating achievable quality standards will be discussed in the context of the accelerated high-level waste treatment and disposal mission at the Hanford Site.

Hamel, William; Huffman, Lori; Lerchen, Megan; Wiemers, Karyn

2003-02-27T23:59:59.000Z

169

SRNL CRP progress report [Development of Melt Processed Ceramics for Nuclear Waste Immobilization  

SciTech Connect (OSTI)

A multi-phase ceramic waste form is being developed at the Savannah River National Laboratory (SRNL) for treatment of secondary waste streams generated by reprocessing commercial spent nuclear. The envisioned waste stream contains a mixture of transition, alkali, alkaline earth, and lanthanide metals. Ceramic waste forms are tailored (engineered) to incorporate waste components as part of their crystal structure based on knowledge from naturally found minerals containing radioactive and non-radioactive species similar to the radionuclides of concern in wastes from fuel reprocessing. The ability to tailor ceramics to mimic naturally occurring crystals substantiates the long term stability of such crystals (ceramics) over geologic timescales of interest for nuclear waste immobilization [1]. A durable multiphase ceramic waste form tailored to incorporate all the waste components has the potential to broaden the available disposal options and thus minimize the storage and disposal costs associated with aqueous reprocessing.

Amoroso, J.; Marra, J.

2014-10-02T23:59:59.000Z

170

Separation of technetium from nuclear waste stream simulants. Final report  

SciTech Connect (OSTI)

The author studied liquid anion exchangers, such as Aliquat-336 nitrate, various pyridinium nitrates, and related salts, so that they may be applied toward a specific process for extracting (partitioning) and recovering {sup 99}TcO{sub 4}{sup {minus}} from nuclear waste streams. Many of the waste streams are caustic and contain a variety of other ions. For this reason, the author studied waste stream simulants that are caustic and contain appropriate concentrations of selected, relevant ions. Methods of measuring the performance of the exchangers and extractant systems included contact experiments. Batch contact experiments were used to determine the forward and reverse extraction parameters as a function of temperature, contact time, phase ratio, concentration, solvent (diluent), and other physical properties. They were also used for stability and competition studies. Specifically, the author investigated the solvent extraction behavior of salts of perrhenate (ReO{sub 4}{sup {minus}}), a stable (non-radioactive) chemical surrogate for {sup 99}TcO{sub 4}{sup {minus}}. Results are discussed for alternate organic solvents; metalloporphyrins, ferrocenes, and N-cetyl pyridium nitrate as alternate extractant salts; electroactive polymers; and recovery of ReO{sub 4}{sup {minus}} and TcO{sub 4}{sup {minus}}.

Strauss, S.H. [Colorado State Univ., Fort Collins, CO (United States). Dept. of Chemistry

1995-09-11T23:59:59.000Z

171

Transmutation of Nuclear Waste and the future MYRRHA Demonstrator  

E-Print Network [OSTI]

While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, there are also, in particular in the "old world" major objections. These are both concerns about safety, in particular in the wake of the Fukushima nuclear accident and concerns about the long-term burden that is constituted by the radiotoxic waste from the spent fuel. With regard to the second topic, the present contribution will outline the concept of Partitioning & Transmutation (P&T), as scientific and technological answer. Deployment of P&T may use dedicated "Transmuter" or "Burner" reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a most attractive solution. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven system. The present paper aims at a short introduction into the field that has been characterized by a high collaborative activity during the last decade in Europe, in order to focus, in its later part, on the MYRRHA project as the European ADS technology demonstrator.

Alex C. Mueller

2012-10-16T23:59:59.000Z

172

Transmutation of Nuclear Waste and the future MYRRHA Demonstrator  

E-Print Network [OSTI]

While a considerable and world-wide growth of the nuclear share in the global energy mix is desirable for many reasons, there are also, in particular in the "old world" major objections. These are both concerns about safety, in particular in the wake of the Fukushima nuclear accident and concerns about the long-term burden that is constituted by the radiotoxic waste from the spent fuel. With regard to the second topic, the present contribution will outline the concept of Partitioning & Transmutation (P&T), as scientific and technological answer. Deployment of P&T may use dedicated "Transmuter" or "Burner" reactors, using a fast neutron spectrum. For the transmutation of waste with a large content (up to 50%) of (very long-lived) Minor Actinides, a sub-critical reactor, using an external neutron source is a most attractive solution. It is constituted by coupling a proton accelerator, a spallation target and a subcritical core. This promising new technology is named ADS, for accelerator-driven syste...

Mueller, Alex C

2012-01-01T23:59:59.000Z

173

Disposition of nuclear waste using subcritical accelerator-driven systems  

SciTech Connect (OSTI)

Spent fuel from nuclear power plants contains large quantities of Pu, other actinides, and fission products (FP). This creates challenges for permanent disposal because of the long half-lives of some isotopes and the potential for diversion of the fissile material. Two issues of concern for the US repository concept are: (1) long-term radiological risk peaking tens-of-thousands of years in the future; and (2) short-term thermal loading (decay heat) that limits capacity. An accelerator-driven neutron source can destroy actinides through fission, and can convert long-lived fission products to shorter-lived or stable isotopes. Studies over the past decade have established that accelerator transmutation of waste (ATW) can have a major beneficial impact on the nuclear waste problem. Specifically, the ATW concept the authors are evaluating: (1) destroys over 99.9% of the actinides; (2) destroys over 99.9% of the Tc and I; (3) separates Sr-90 and Cs-137; (4) separates uranium from the spent fuel; (5) produces electric power.

Venneri, F.; Li, N.; Williamson, M.; Houts, M.; Lawrence, G.

1998-12-31T23:59:59.000Z

174

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

175

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

176

WASTE PROCESSING ANNUAL NUCLEAR SAFETY RELATED R AND D REPORT FOR CY2008  

SciTech Connect (OSTI)

The Engineering and Technology Office of Waste Processing identifies and reduces engineering and technical risks associated with key waste processing project decisions. The risks, and actions taken to mitigate those risks, are determined through technology readiness assessments, program reviews, technology information exchanges, external technical reviews, technical assistance, and targeted technology development and deployment (TDD). The Office of Waste Processing TDD program prioritizes and approves research and development scopes of work that address nuclear safety related to processing of highly radioactive nuclear wastes. Thirteen of the thirty-five R&D approved work scopes in FY2009 relate directly to nuclear safety, and are presented in this report.

Fellinger, A.

2009-10-15T23:59:59.000Z

177

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center  

SciTech Connect (OSTI)

This report provides a quantitative inventory and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. From December 1963 through May 2000, liquid radioactive wastes generated by spent nuclear fuel reprocessing were converted into a solid, granular form called calcine. This report also contains a description of the calcine storage bins.

M. D. Staiger

2007-06-01T23:59:59.000Z

178

Nuclear Waste Technical Review Board Thermal-Response Evaluation of Yucca Mountain  

E-Print Network [OSTI]

Nuclear Waste Technical Review Board Thermal-Response Evaluation of Yucca Mountain During of the thermal response of the proposed Yucca Mountain repository for various thermal loadings. The U. S. Nuclear Waste Technical Review Board (NWTRB) staff has developed calculation tools that allow performing

179

Standard Guide for Preparing Waste Management Plans for Decommissioning Nuclear Facilities  

E-Print Network [OSTI]

1.1 This guide addresses the development of waste management plans for potential waste streams resulting from decommissioning activities at nuclear facilities, including identifying, categorizing, and handling the waste from generation to final disposal. 1.2 This guide is applicable to potential waste streams anticipated from decommissioning activities of nuclear facilities whose operations were governed by the Nuclear Regulatory Commission (NRC) or Agreement State license, under Department of Energy (DOE) Orders, or Department of Defense (DoD) regulations. 1.3 This guide provides a description of the key elements of waste management plans that if followed will successfully allow for the characterization, packaging, transportation, and off-site treatment or disposal, or both, of conventional, hazardous, and radioactive waste streams. 1.4 This guide does not address the on-site treatment, long term storage, or on-site disposal of these potential waste streams. 1.5 This standard does not purport to address ...

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

180

EXPERIMENTAL METHODS TO ESTIMATE ACCUMULATED SOLIDS IN NUCLEAR WASTE TANKS  

SciTech Connect (OSTI)

The Department of Energy has a large number of nuclear waste tanks. It is important to know if fissionable materials can concentrate when waste is transferred from staging tanks prior to feeding waste treatment plants. Specifically, there is a concern that large, dense particles, e.g., plutonium containing, could accumulate in poorly mixed regions of a blend tank heel for tanks that employ mixing jet pumps. At the request of the DOE Hanford Tank Operations Contractor, Washington River Protection Solutions, the Engineering Development Laboratory of the Savannah River National Laboratory performed a scouting study in a 1/22-scale model of a waste tank to investigate this concern and to develop measurement techniques that could be applied in a more extensive study at a larger scale. Simulated waste tank solids and supernatant were charged to the test tank and rotating liquid jets were used to remove most of the solids. Then the volume and shape of the residual solids and the spatial concentration profiles for the surrogate for plutonium were measured. This paper discusses the overall test results, which indicated heavy solids only accumulate during the first few transfer cycles, along with the techniques and equipment designed and employed in the test. Those techniques include: Magnetic particle separator to remove stainless steel solids, the plutonium surrogate from a flowing stream; Magnetic wand used to manually remove stainless steel solids from samples and the tank heel; Photographs were used to determine the volume and shape of the solids mounds by developing a composite of topographical areas; Laser rangefinders to determine the volume and shape of the solids mounds; Core sampler to determine the stainless steel solids distribution within the solids mounds; Computer driven positioner that placed the laser rangefinders and the core sampler over solids mounds that accumulated on the bottom of a scaled staging tank in locations where jet velocities were low. These devices and techniques were very effective to estimate the movement, location, and concentrations of the solids representing plutonium and are expected to perform well at a larger scale. The operation of the techniques and their measurement accuracies will be discussed as well as the overall results of the accumulated solids test.

Duignan, M.; Steeper, T.; Steimke, J.

2012-12-10T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

SORPTION OF URANIUM, PLUTONIUM AND NEPTUNIUM ONTO SOLIDS PRESENT IN HIGH CAUSTIC NUCLEAR WASTE STORAGE TANKS  

SciTech Connect (OSTI)

Solids such as granular activated carbon, hematite and sodium phosphates, if present as sludge components in nuclear waste storage tanks, have been found to be capable of precipitating/sorbing actinides like plutonium, neptunium and uranium from nuclear waste storage tank supernatant liqueur. Thus, the potential may exists for the accumulation of fissile materials in such nuclear waste storage tanks during lengthy nuclear waste storage and processing. To evaluate the nuclear criticality safety in a typical nuclear waste storage tank, a study was initiated to measure the affinity of granular activated carbon, hematite and anhydrous sodium phosphate to sorb plutonium, neptunium and uranium from alkaline salt solutions. Tests with simulated and actual nuclear waste solutions established the affinity of the solids for plutonium, neptunium and uranium upon contact of the solutions with each of the solids. The removal of plutonium and neptunium from the synthetic salt solution by nuclear waste storage tank solids may be due largely to the presence of the granular activated carbon and transition metal oxides in these storage tank solids or sludge. Granular activated carbon and hematite also showed measurable affinity for both plutonium and neptunium. Sodium phosphate, used here as a reference sorbent for uranium, as expected, exhibited high affinity for uranium and neptunium, but did not show any measurable affinity for plutonium.

Oji, L; Bill Wilmarth, B; David Hobbs, D

2008-05-30T23:59:59.000Z

182

A ThreeDimensional Finite Element Simulation for Transport of Nuclear Waste Contamination in Porous Media \\Lambda  

E-Print Network [OSTI]

A Three­Dimensional Finite Element Simulation for Transport of Nuclear Waste Contamination for transport of nuclear­waste contamination in three­dimensional porous media are presented with a description of contamination of groundwater by high­level nuclear waste and a wide variety of other sources makes a proper

Ewing, Richard E.

183

Commercial nuclear reactors and waste: the current status  

SciTech Connect (OSTI)

During the last five years, the declared size of the commercial light water reactor (LWR) nuclear power industry in the US has steadily decreased. As of January 1980, the total number of power plants had dropped to 191 from the 226 in December 31, 1974. At least another nine were cancelled in the last few months. This report was developed as the first of a series to track implications to waste management due to such changes in the declared size of the industry. For the presently declared size, key conclusions are: the declared reactors will peak at a capacity of 162 GWe and consume about 10/sup 6/ MTU as enrichment feed. As few as two repositories of about 100,000 MTHM capacity each would hold the waste. Predisposal storage (reactor basins and AFRs) would peak at less than 100,000 MTHM (in the year 2020) with one repository opening in the year 1997 and the other in the year 2020. Most of the 100,000 MTHM would have to be in AFR storage unless current practice regarding reactor basin size was radically changed.

Platt, A.M.; Robinson, J.V.

1980-04-01T23:59:59.000Z

184

Redox reaction and foaming in nuclear waste glass melting  

SciTech Connect (OSTI)

This document was prepared by Pacific Northwest Laboratory (PNL) and is an attempt to analyze and estimate the effects of feed composition variables and reducing agent variables on the expected chemistry of reactions occurring in the cold cap and in the glass melt in the nuclear waste glass Slurry-fed, joule-heated melters as they might affect foaming during the glass-making process. Numerous redox reactions of waste glass components and potential feed additives, and the effects of other feed variables on these reactions are reviewed with regard to their potential effect on glass foaming. A major emphasis of this report is to examine the potential positive or negative aspects of adjusting feed with formic acid as opposed to other feed modification techniques including but not limited to use of other reducing agents. Feed modification techniques other than the use of reductants that should influence foaming behavior include control of glass melter feed pH through use of nitric acid. They also include partial replacement of sodium salts by lithium salts. This latter action (b) apparently lowers glass viscosity and raises surface tension. This replacement should decrease foaming by decreasing foam stability.

Ryan, J.L.

1995-08-01T23:59:59.000Z

185

Ceramic package fabrication for YMP nuclear waste disposal  

SciTech Connect (OSTI)

The purpose of this work is to develop alternate materials/design concepts to metal barriers for the Nevada Nuclear Waste Storage Investigations Project. There is some potential that site conditions may prove to be too aggressive for successful employment of the metal alloys under current consideration or that performance assessment models will predict metal container degradation rates that are inconsistent with the goal of substantially complete containment included in the NRC regulations. In the event that the anticipated lifetimes of metal containers are considered inadequate, alternate materials (i.e. ceramics or ceramic/metal composites) will be chosen due to superior corrosion resistance. This document was prepared using information taken from the open literature, conversations and correspondence with vendors, news releases and data presented at conferences to determine what form such a package might take. This discussion presents some ceramic material selection criteria, alternatives for the materials which might be used and alternatives for potential fabrication routes. This includes {open_quotes}stand alone{close_quotes} ceramic components and ceramic coatings/linings for metallic structures. A list of companies providing verbal or written information concerning the production of ceramic or ceramic lined waste containers appears at the end of this discussion.

Wilfinger, K.

1994-08-01T23:59:59.000Z

186

Identifying suitable "piercement" salt domes for nuclear waste storage sites  

SciTech Connect (OSTI)

Piercement salt domes of the northern interior salt basins of the Gulf of Mexico are being considered as permanent storage sites for both nuclear and chemically toxic wastes. The suitable domes are stable and inactive, having reached their final evolutionary configuration at least 30 million years ago. They are buried to depths far below the level to which erosion will penetrate during the prescribed storage period and are not subject to possible future reactivation. The salt cores of these domes are themselves impermeable, permitting neither the entry nor exit of ground water or other unwanted materials. In part, a stable dome may be recognized by its present geometric configuration, but conclusive proof depends on establishing its evolutionary state. The evolutionary state of a dome is obtained by reconstructing the growth history of the dome as revealed by the configuration of sedimentary strata in a large area (commonly 3,000 square miles or more) surrounding the dome. A high quality, multifold CDP reflection seismic profile across a candidate dome will provide much of the necessary information when integrated with available subsurface control. Additional seismic profiles may be required to confirm an apparent configuration of the surrounding strata and an interpreted evolutionary history. High frequency seismic data collected in the near vicinity of a dome are also needed as a supplement to the CDP data to permit accurate depiction of the configuration of shallow strata. Such data must be tied to shallow drill hole control to confirm the geologic age at which dome growth ceased. If it is determined that a dome reached a terminal configuration many millions of years ago, such a dome is incapable of reactivation and thus constitutes a stable storage site for nuclear wastes.

Kehle, R.

1980-08-01T23:59:59.000Z

187

The Encapsulated Nuclear Heat Source for Proliferation-Resistant Low-Waste Nuclear Energy  

SciTech Connect (OSTI)

Encapsulated Nuclear Heat Source (ENHS) is a small innovative reactor suitable for use in developing countries. The reference design is a SOMWe lead-bismuth eutectic (Pb-Bi) cooled fast reactor. It is designed so that the fuel is installed and sealed into the reactor module at the factory. The nuclear controls, a major portion of the instrumentation and the Pb-Bi covering the core are also installed at the factory. At the site of operations the reactor module is inserted into a pool of Pb-Bi that contains the steam generators. Major components, such as the pool vessel and steam generators, are permanent and remain in place while the reactor module is replaced every 15 years. At the end of life the sealed reactor module is removed and returned to an internationally controlled recycling center. Thus, the ENHS provides a unique capability for ensuring the security of the nuclear fuel throughout its life. The design also can minimize the user country investment in nuclear technology and staff. Following operation and return of the module to the recycling facility, the useable components, including the fuel, are refurbished and available for reuse. A fuel cycle compatible with this approach has been identified that reduces the amount of nuclear waste.

Brown, N; Carelli, M; Conway, L; Dzodzo, M; Greenspan, E; Hossain, Q; Saphier, D; Shimada, H; Sienicki, J; Wade, D

2001-04-01T23:59:59.000Z

188

Nuclear waste management. Semiannual progress report, October 1982-March 1983  

SciTech Connect (OSTI)

This document is one of a series of technical progress reports designed to report radioactive waste management programs at the Pacific Northwest Laboratory. Accomplishments in the following programs are reported: waste stabilization; Materials Characterization Center; waste isolation; low-level waste management; remedial action; and supporting studies.

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

1983-06-01T23:59:59.000Z

189

Integrated data base report - 1994: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect (OSTI)

The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel and commercial and U.S. government-owned radioactive wastes. Except for transuranic wastes, inventories of these materials are reported as of December 31, 1994. Transuranic waste inventories are reported as of December 31, 1993. All spent nuclear fuel and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

NONE

1995-09-01T23:59:59.000Z

190

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

191

Technical Aspects Regarding the Management of Radioactive Waste from Decommissioning of Nuclear Facilities  

SciTech Connect (OSTI)

The proper application of the nuclear techniques and technologies in Romania started in 1957, once with the commissioning of the Research Reactor VVR-S from IFIN-HH-Magurele. During the last 45 years, appear thousands of nuclear application units with extremely diverse profiles (research, biology, medicine, education, agriculture, transport, all types of industry) which used different nuclear facilities containing radioactive sources and generating a great variety of radioactive waste during the decommissioning after the operation lifetime is accomplished. A new aspect appears by the planning of VVR-S Research Reactor decommissioning which will be a new source of radioactive waste generated by decontamination, disassembling and demolition activities. By construction and exploitation of the Radioactive Waste Treatment Plant (STDR)--Magurele and the National Repository for Low and Intermediate Radioactive Waste (DNDR)--Baita, Bihor county, in Romania was solved the management of radioactive wastes arising from operation and decommissioning of small nuclear facilities, being assured the protection of the people and environment. The present paper makes a review of the present technical status of the Romanian waste management facilities, especially raising on treatment capabilities of ''problem'' wastes such as Ra-266, Pu-238, Am-241 Co-60, Co-57, Sr-90, Cs-137 sealed sources from industrial, research and medical applications. Also, contain a preliminary estimation of quantities and types of wastes, which would result during the decommissioning project of the VVR-S Research Reactor from IFIN-HH giving attention to some special category of wastes like aluminum, graphite and equipment, components and structures that became radioactive through neutron activation. After analyzing the technical and scientific potential of STDR and DNDR to handle big amounts of wastes resulting from the decommissioning of VVR-S Research Reactor and small nuclear facilities, the necessity of up-gradation of these nuclear objectives before starting the decommissioning plan is revealed. A short presentation of the up-grading needs is also presented.

Dragolici, F.; Turcanu, C. N.; Rotarescu, G.; Paunica, I.

2003-02-25T23:59:59.000Z

192

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

193

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

194

Feasibility of very deep borehole disposal of US nuclear defense wastes  

E-Print Network [OSTI]

This thesis analyzes the feasibility of emplacing DOE-owned defense nuclear waste from weapons production into a permanent borehole repository drilled ~4 km into granite basement rock. Two canister options were analyzed ...

Dozier, Frances Elizabeth

2011-01-01T23:59:59.000Z

195

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

196

Department of Energy Issues Funding Opportunity Announcements...  

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

Issues Funding Opportunity Announcements to Enhance Nuclear Energy Education Department of Energy Issues Funding Opportunity Announcements to Enhance Nuclear Energy Education March...

197

THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH GEOLOGIC DISPOSAL OF NUCLEAR WASTE  

E-Print Network [OSTI]

released by the buried wastes and heat remain ing in theOF 10-YEAR-OLD WASTES Waste Heat Source C h a r a c t e r ia t e r s e c t i o n s . WASTE HEAT SOURCE CHARACTERIZATION

Wang, J.S.Y.

2010-01-01T23:59:59.000Z

198

Low-Level Radioactive Waste Disposal Regional Facility Act (Pennsylvania)  

Broader source: Energy.gov [DOE]

This act establishes a low-level radioactive waste disposal regional facility siting fund that requires nuclear power reactor constructors and operators to pay to the Department of Environmental...

199

Engineered barrier development for a nuclear waste repository in basalt: an integration of current knowledge  

SciTech Connect (OSTI)

This document represents a compilation of data and interpretive studies conducted as part of the engineered barriers program of the Basalt Waste Isolation Project. The overall objective of these studies is to provide information on barrier system designs, emplacement and isolation techniques, and chemical reactions expected in a nuclear waste repository located in the basalts underlying the Hanford Site within the state of Washington. Backfills, waste-basalt interactions, sorption, borehole plugging, etc., are among the topics discussed.

Smith, M.J.

1980-05-01T23:59:59.000Z

200

Helium solubility in SON68 nuclear waste glass  

SciTech Connect (OSTI)

Helium behavior in a sodium borosilicate glass (SON68) dedicated to the immobilization of high-level nuclear waste is examined. Two experimental approaches on nonradioactive glass specimens are implemented: pressurized helium infusion experiments and {sup 3}He ion implantation experiments. The temperature variation of helium solubility in SON68 glass was determined and analyzed with the harmonic oscillator model to determine values of the energy of interaction E(0) at the host sites (about -4000 J/mol), the vibration frequency (about 1.7 x 10{sup 11} s{sup -1}), and the density of solubility sites (2.2 x 10{sup 21} sites cm{sup -3}). The implantation experiments show that a non diffusive transport phenomenon (i.e., athermal diffusion) is involved in the material when the helium concentration exceeds 2.3 x 10{sup 21} He cm{sup -3}, and thus probably as soon as it exceeds the density of solubility sites accessible to helium in the glass. We propose that this transport mechanism could be associated with the relaxation of the stress gradient induced by the implanted helium profile, which is favored by the glass damage. Microstructural characterization by TEM and ESEM of glass specimens implanted with high helium concentrations showed a homogeneous microstructure free of bubbles, pores, or cracking at a scale of 10 nm. (authors)

Fares, Toby; Peuget, Sylvain; Bouty, Olivier; Broudic, Veronique; Maugeri, Emilio; Bes, Rene; Jegou, Christophe [CEA, DEN, DTCD SECM LMPA, F-30207 Marcoule, Bagnols Sur Cez, (France); Chamssedine, Fadel; Sauvage, Thierry [CNRS, CEMHTI, F-245071 Orleans, (France); Deschanels, Xavier [LNAR, Marcoule Inst Separat Chem, F-30207 Bagnols Sur Ceze, (France)

2012-12-15T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

Hydrogeologic effects of natural disruptive events on nuclear waste repositories  

SciTech Connect (OSTI)

Some possible hydrogeologic effects of disruptive events that may affect repositories for nuclear wastte are described. A very large number of combinations of natural events can be imagined, but only those events which are judged to be most probable are covered. Waste-induced effects are not considered. The disruptive events discussed above are placed into four geologic settings. Although the geology is not specific to given repository sites that have been considered by other agencies, the geology has been generalized from actual field data and is, therefore, considered to be physically reasonable. The geologic settings considered are: (1) interior salt domes of the Gulf Coast, (2) bedded salt of southeastern New Mexico, (3) argillaceous rocks of southern Nevanda, and (4) granitic stocks of the Basin and Range Province. Log-normal distributions of permeabilities of rock units are given for each region. Chapters are devoted to: poresity and permeability of natural materials, regional flow patterns, disruptive events (faulting, dissolution of rock forming minerals, fracturing from various causes, rapid changes of hydraulic regimen); possible hydrologic effects of disruptive events; and hydraulic fracturing.

Davis, S.N.

1980-06-01T23:59:59.000Z

202

State-of-the-art review of materials properties of nuclear waste forms.  

SciTech Connect (OSTI)

The Materials Characterization Center (MCC) was established at the Pacific Northwest Laboratory to assemble a standardized nuclear waste materials data base for use in research, systems and facility design, safety analyses, and waste management decisions. This centralized data base will be provided through the means of a Nuclear Waste Materials Handbook. The first issue of the Handbook will be published in the fall of 1981 in looseleaf format so that it can be updated as additional information becomes available. To ensure utmost reliability, all materials data appearing in the Handbook will be obtained by standard procedures defined in the Handbook and approved by an independent Materials Review Board (MRB) comprised of materials experts from Department of Energy laboratories and from universities and industry. In the interim before publication of the Handbook there is need for a report summarizing the existing materials data on nuclear waste forms. This review summarizes materials property data for the nuclear waste forms that are being developed for immobilization of high-level radioactive waste. It is intended to be a good representation of the knowledge concerning the properties of HLW forms as of March 1981. The table of contents lists the following topics: introduction which covers waste-form categories, and important waste-form materials properties; physical properties; mechanical properties; chemical durability; vaporization; radiation effects; and thermal phase stability.

Mendel, J. E.; Nelson, R. D.; Turcotte, R. P.; Gray, W. J.; Merz, M. D.; Roberts, F. P.; Weber, W. J.; Westsik, Jr., J. H.; Clark, D. E.

1981-04-01T23:59:59.000Z

203

Precipitation process for the removal of technetium values from nuclear waste solutions  

DOE Patents [OSTI]

High efficiency removal of techetium 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, D.D.; Ebra, M.A.

1985-11-21T23:59:59.000Z

204

Na, Mg, Ni and Cs distribution and speciation after long-term alteration of a simulated nuclear waste glass  

E-Print Network [OSTI]

distribution and speciation of Na, Mg, Ni and Cs in a simulated (inactive) nuclear waste glass were studied and Cs represent dose determining long-lived radionuclides (59 Ni, 135 Cs) in vitrified nuclear wasteNa, Mg, Ni and Cs distribution and speciation after long-term alteration of a simulated nuclear

205

Two Approaches to the Geologic Disposal of Long-Lived Nuclear Waste: Yucca Mountain, Nevada and the Waste Isolation Pilot Plant, Carlsbad, New Mexico  

SciTech Connect (OSTI)

A key component of the US energy program is to provide for the safe and permanent isolation of spent nuclear fuel and long-lived radioactive waste produced through programs related to national defense and the generation of electric power by nuclear utilities. To meet this challenge, the US Department of Energy (DOE) has developed a multi-faceted approach to the geologic disposal of long-lived nuclear wastes. Two sites are being developed or studied as current or potential deep geologic repositories for long lived radioactive wastes, the Waste Isolation Pilot Plant (WIPP) near Carlsbad, New Mexico and Yucca Mountain, Nevada.

Levich, R. A.; Patterson, R. L.; Linden, R. M.

2002-02-26T23:59:59.000Z

206

Processing constraints on high-level nuclear waste glasses for Hanford Waste Vitrification Plant  

SciTech Connect (OSTI)

The work presented in this paper is a part of a major technology program supported by the US Department of Energy (DOE) in preparation for the planned operation of the Hanford Waste Vitrification Plant (HWVP). Because composition of Hanford waste varies greatly, processability is a major concern for successful vitrification. This paper briefly surveys general aspects of waste glass processability and then discusses their ramifications for specific examples of Hanford waste streams.

Hrma, P. [Pacific Northwest Lab., Richland, WA (United States)

1993-12-31T23:59:59.000Z

207

Systems engineering programs for geologic nuclear waste disposal  

SciTech Connect (OSTI)

The design sequence and system programs presented begin with general approximate solutions that permit inexpensive analysis of a multitude of possible wastes, disposal media, and disposal process properties and configurations. It then continues through progressively more precise solutions as parts of the design become fixed, and ends with repository and waste form optimization studies. The programs cover both solid and gaseous waste forms. The analytical development, a program listing, a users guide, and examples are presented for each program. Sensitivity studies showing the effects of disposal media and waste form thermophysical properties and repository layouts are presented as examples.

Klett, R. D.; Hertel, Jr., E. S.; Ellis, M. A.

1980-06-01T23:59:59.000Z

208

U.S. Nuclear Waste Technical Review Board  

E-Print Network [OSTI]

Preclosure Operations of the Waste Management System . . . . . . . . . . . . . . . . .2 Transportation-Aging-Disposal . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .2 Transportation and Radionuclide Transport. . . . . . . . . . . . . . . . . . .5 Realistic Performance Assessments

209

alkaline nuclear wastes: Topics by E-print Network  

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

Geosciences Websites Summary: -solid waste for CO2 mitigation and reduction of greenhouse effect gases into the atmosphere. ? 2008 ElsevierCarbonation of alkaline paper mill...

210

alkaline nuclear waste: Topics by E-print Network  

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

Geosciences Websites Summary: -solid waste for CO2 mitigation and reduction of greenhouse effect gases into the atmosphere. ? 2008 ElsevierCarbonation of alkaline paper mill...

211

Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report  

SciTech Connect (OSTI)

The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Argyle, Mark Don; Lauerhass, Lance; Bendixsen, Carl Lee; Hinckley, Steve Harold

2000-11-01T23:59:59.000Z

212

Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-2000 Status Report  

SciTech Connect (OSTI)

The Low-Activity Waste Process Technology Program anticipated that grouting will be used for disposal of low-level and transuranic wastes generated at the Idaho Nuclear Technology Engineering Center (INTEC). During fiscal year 2000, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed using silica gel and other absorbents to solidify sodium-bearing wastes. A feasibility study and conceptual design were completed for the construction of a grout pilot plant for simulated wastes and demonstration facility for actual wastes.

Herbst, A.K.; McCray, J.A.; Kirkham, R.J.; Pao, J.; Argyle, M.D.; Lauerhass, L.; Bendixsen, C.L.; Hinckley, S.H.

2000-10-31T23:59:59.000Z

213

Can we talk? Communications management for the Waste Isolation Pilot Plant, a complex nuclear waste management project  

SciTech Connect (OSTI)

Sandia Nuclear Waste Management Program is pursuing for DOE an option for permanently disposing radioactive waste in deep geologic repositories. Included in the Program are the Waste Isolation Pilot Plant (WIPP) Project for US defense program mixed waste the Yucca Mountain Project (YMP) for spent power reactor fuel and vitrified high-level waste, projects for other waste types, and development efforts in environmental decision support technologies. WIPP and YMP are in the public arena, of a controversial nature, and provide significant management challenges. Both projects have large project teams, multiple organization participants, large budgets, long durations, are very complex, have a high degree of programmatic risk, and operate in an extremely regulated environment requiring legal defensibility. For environmental projects like these to succeed, SNL`s Program is utilizing nearly all areas in PMI`s Project Management Body of Knowledge (PMBOK) to manage along multiple project dimensions such as the physical sciences (e.g., geophysics and geochemistry; performance assessment; decision analysis) management sciences (controlling the triple constraint of performance, cost and schedule), and social sciences (belief systems; public participation; institutional politics). This discussion focuses primarily on communication challenges active on WIPP. How is the WIPP team meeting the challenges of managing communications?`` and ``How are you approaching similar challenges?`` will be questions for a dialog with the audience.

Goldstein, S.A.; Pullen, G.M.; Brewer, D.R.

1995-07-01T23:59:59.000Z

214

Computational Efficient Upscaling Methodology for Predicting Thermal Conductivity of Nuclear Waste forms  

SciTech Connect (OSTI)

This study evaluated different upscaling methods to predict thermal conductivity in loaded nuclear waste form, a heterogeneous material system. The efficiency and accuracy of these methods were compared. Thermal conductivity in loaded nuclear waste form is an important property specific to scientific researchers, in waste form Integrated performance and safety code (IPSC). The effective thermal conductivity obtained from microstructure information and local thermal conductivity of different components is critical in predicting the life and performance of waste form during storage. How the heat generated during storage is directly related to thermal conductivity, which in turn determining the mechanical deformation behavior, corrosion resistance and aging performance. Several methods, including the Taylor model, Sachs model, self-consistent model, and statistical upscaling models were developed and implemented. Due to the absence of experimental data, prediction results from finite element method (FEM) were used as reference to determine the accuracy of different upscaling models. Micrographs from different loading of nuclear waste were used in the prediction of thermal conductivity. Prediction results demonstrated that in term of efficiency, boundary models (Taylor and Sachs model) are better than self consistent model, statistical upscaling method and FEM. Balancing the computation resource and accuracy, statistical upscaling is a computational efficient method in predicting effective thermal conductivity for nuclear waste form.

Li, Dongsheng; Sun, Xin; Khaleel, Mohammad A.

2011-09-28T23:59:59.000Z

215

DOE Seeks to Invest up to $15 Million in Funding for Nuclear...  

Energy Savers [EERE]

following areas: Used Fuel Separations Technology, Advanced Nuclear Fuel Development, Fast Burner Reactors and Advanced Transmutation Systems, Advanced Fuel Cycle Systems...

216

Nuclear-waste-management. Quarterly progress report, July-September 1981  

SciTech Connect (OSTI)

Progress reports and summaries are presented for the following: high-level waste process development, alternate waste forms; TMI zeolite vitrification demonstration program; nuclear waste materials characterization center; TRU waste immobilization; TRU waste decontamination; krypton implantation; 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 radionuclides in soils; handbook of methods to decrease the generation of low-level waste; waste management system studies; waste management safety 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 fuel 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 uranium tailings; application of long-term chemical biobarriers for uranium tailings; and revegetation of inactive uranium tailings sites.

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

1981-12-01T23:59:59.000Z

217

Nuclear waste management technical support in the development of nuclear waste form criteria for the NRC. Task 1. Waste package overview  

SciTech Connect (OSTI)

In this report the current state of waste package development for high level waste, transuranic waste, and spent fuel in the US and abroad has been assessed. Specifically, reviewed are recent and on-going research on various waste forms, container materials and backfills and tentatively identified those which are likely to perform most satisfactorily in the repository environment. Radiation effects on the waste package components have been reviewed and the magnitude of these effects has been identified. Areas requiring further research have been identified. The important variables affecting radionuclide release from the waste package have been described and an evaluation of regulatory criteria for high level waste and spent fuel is presented. Finally, for spent fuel, high level, and TRU waste, components which could be used to construct a waste package having potential to meet NRC performance requirements have been described and identified.

Dayal, R.; Lee, B.S.; Wilke, R.J.; Swyler, K.J.; Soo, P.; Ahn, T.M.; McIntyre, N.S.; Veakis, E.

1982-02-01T23:59:59.000Z

218

THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH GEOLOGIC DISPOSAL OF NUCLEAR WASTE  

E-Print Network [OSTI]

waste repository design AERE-R--9343 Atomic Energy Researchof the thermal s t r e s s field. AERE R-8999, Atomic Energy

Wang, J.S.Y.

2010-01-01T23:59:59.000Z

219

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 2, Part B  

SciTech Connect (OSTI)

Two types of projects in the spent nuclear fuel and environmental restoration and waste management activities at the Idaho National Engineering Laboratory (INEL) are described. These are: foreseeable proposed projects where some funding for preliminary planning and/or conceptual design may already be authorized, but detailed design or planning will not begin until the Department of Energy (DOE) has determined that the requirements of the National Environmental Policy Act process for the project have been completed; planned or ongoing projects not yet completed but whose National Environmental Policy Act documentation is already completed or is expected to be completed before the Record of Decision for this Envirorunental Impact Statement (EIS) is issued. The section on project summaries describe the projects (both foreseeable proposed and ongoing).They provide specific information necessary to analyze the environmental impacts of these projects. Chapter 3 describes which alternative(s) each project supports. Summaries are included for (a) spent nuclear fuel projects, (b) environmental remediation projects, (c) the decontamination and decommissioning of surplus INEL facilities, (d) the construction, upgrade, or replacement of existing waste management facilities, (e) infrastructure projects supporting waste management activities, and (f) research and development projects supporting waste management activities.

Not Available

1994-06-01T23:59:59.000Z

220

The Italian Activities in the Field of Nuclear Waste Management - 12439  

SciTech Connect (OSTI)

The Italian situation in the field of nuclear waste management is characterized by a relative small quantity of wastes, as a consequence of the giving up of energy production by nuclear generation in 1986. Notwithstanding this situation, Italy is a unique case study since the country needs to undertake the final decommissioning of four shut-down NPPs (size 100-200 MWe), each one different from the others. Therefore all the regulatory, technical, and financial actions are needed in the same way as if there was actual nuclear generation. Furthermore, the various non-power generating applications of nuclear energy still require management, a legal framework, a regulatory body, an industrial structure, and technical know-how. Notwithstanding the absence of energy production from nuclear sources, the country has the burden of radioactive waste management from the previous nuclear operations, which obliges it to implement at first a robust legislative framework, then to explore all the complex procedures to achieve the localization of the national interim storage facility, not excluding the chance to have a European regional facility for geologic disposal, under the clauses of the Council Directive of 19 July 2011 'Establishing a Community Framework for the Responsible and Safe Management of Radioactive Waste'. Then, as far as industrial, medical and R and D aspects, the improvement of the legislative picture, the creation of a regulatory body, is a good start for the future, to achieve the best efficiency of the Italian system. (authors)

Giorgiantoni, Giorgio; Marzo, Giuseppe A.; Sepielli, Massimo [ENEA, C. R. Casaccia, Roma (Italy)

2012-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

State of the art review of radioactive waste volume reduction techniques for commercial nuclear power plants  

SciTech Connect (OSTI)

A review is made of the state of the art of volume reduction techniques for low level liquid and solid radioactive wastes produced as a result of: (1) operation of commercial nuclear power plants, (2) storage of spent fuel in away-from-reactor facilities, and (3) decontamination/decommissioning of commercial nuclear power plants. The types of wastes and their chemical, physical, and radiological characteristics are identified. Methods used by industry for processing radioactive wastes are reviewed and compared to the new techniques for processing and reducing the volume of radioactive wastes. A detailed system description and report on operating experiences follow for each of the new volume reduction techniques. In addition, descriptions of volume reduction methods presently under development are provided. The Appendix records data collected during site surveys of vendor facilities and operating power plants. A Bibliography is provided for each of the various volume reduction techniques discussed in the report.

Not Available

1980-04-01T23:59:59.000Z

222

Recovery of fissile materials from plutonium residues, miscellaneous spent nuclear fuel, and uranium fissile wastes  

SciTech Connect (OSTI)

A new process is proposed that converts complex feeds containing fissile materials into a chemical form that allows the use of existing technologies (such as PUREX and ion exchange) to recover the fissile materials and convert the resultant wastes to glass. Potential feed materials include (1) plutonium scrap and residue, (2) miscellaneous spent nuclear fuel, and (3) uranium fissile wastes. The initial feed materials may contain mixtures of metals, ceramics, amorphous solids, halides, and organics. 14 refs., 4 figs.

Forsberg, C.W.

1997-03-01T23:59:59.000Z

223

The dilemma of siting a high-level nuclear waste repository  

SciTech Connect (OSTI)

This books presents a siting process that the authors believe will prove successful within the adversarial world that characterizes most attempts to build waste-disposal facilities. They come to the following conclusions: a volunatary siting process stands the best chance of breaking the `not-in-my-backyard` problem; and without public acknowledgement that a facility is needed, any proposal to build a high-level nuclear waste storage facility will meet with opposition.

Easterline, D.; Kunreuther, H.

1995-12-31T23:59:59.000Z

224

2011 SECRETARIAL DETERMINATION OF THE ADEQUACY OF THE NUCLEAR WASTE FUND  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustionImprovement Awardflash2007-42attachment1.pdfmodule(EE) |the Fuel Cell Technologies ProgramFEE |

225

Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee |  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Office of Inspector GeneralDepartment of Energy fromCommentsRevolvingPerformance | DepartmentSecretarial

226

The Department of Energy's Nuclear Waste Fund's Fiscal Year 2011 Financial Statements  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustion |Energy Usage »of Energy StrainClientDesignOfficeThe Allegations Regarding ProhibitedThe

227

Department of Energy's Nuclear Waste Fund's Fiscal Year 2012 Financial Statement Audit  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:Year in Review: TopEnergy DOEDealing WithDevelopmentReport and7Department

228

Use of tangential filtration unit for processing liquid waste from nuclear laundries  

SciTech Connect (OSTI)

Nuclear facilities produce large quantities of weakly contaminated effluents charged with insoluble and soluble products. In collaboration with CEA, TECHNICATOME has developed an ultrafiltration process for liquid waste from nuclear facilities associated with prior insolubilization of radiochemical activity. This process, seeded ultrafiltration, is based on the use of a decloggable mineral filter media and combines very high separation efficiency with long membrane life. The efficiency of the tangential filtration unit which has been processing effluents from the Cadarache Nuclear Research Center (CEA-France) nuclear facility since mid-1988, has been confirmed on several sites.

Augustin, X.; Buzonniere, A. de [Technicatome, Gif-sur-Yvette (France); Barnier, H. [CEA Cadarache, St. Paul-lez-Durance (France)

1993-12-31T23:59:59.000Z

229

Materials characterization center workshop on compositional and microstructural analysis of nuclear waste materials. Summary report  

SciTech Connect (OSTI)

The purpose of the Workshop on Compositional and Microstructural Analysis of Nuclear Waste Materials, conducted November 11 and 12, 1980, was to critically examine and evaluate the various methods currently used to study non-radioactive, simulated, nuclear waste-form performance. Workshop participants recognized that most of the Materials Characterization Center (MCC) test data for inclusion in the Nuclear Waste Materials Handbook will result from application of appropriate analytical procedures to waste-package materials or to the products of performance tests. Therefore, the analytical methods must be reliable and of known accuracy and precision, and results must be directly comparable with those from other laboratories and from other nuclear waste materials. The 41 participants representing 18 laboratories in the United States and Canada were organized into three working groups: Analysis of Liquids and Solutions, Quantitative Analysis of Solids, and Phase and Microstructure Analysis. Each group identified the analytical methods favored by their respective laboratories, discussed areas needing attention, listed standards and reference materials currently used, and recommended means of verifying interlaboratory comparability of data. The major conclusions from this workshop are presented.

Daniel, J.L.; Strachan, D.M.; Shade, J.W.; Thomas, M.T.

1981-06-01T23:59:59.000Z

230

Radioactive Waste Management and Nuclear Facility Decommissioning Progress in Iraq - 13216  

SciTech Connect (OSTI)

Management of Iraq's radioactive wastes and decommissioning of Iraq's former nuclear facilities are the responsibility of Iraq's Ministry of Science and Technology (MoST). The majority of Iraq's former nuclear facilities are in the Al-Tuwaitha Nuclear Research Center located a few kilometers from the edge of Baghdad. These facilities include bombed and partially destroyed research reactors, a fuel fabrication facility and radioisotope production facilities. Within these facilities are large numbers of silos, approximately 30 process or waste storage tanks and thousands of drums of uncharacterised radioactive waste. There are also former nuclear facilities/sites that are outside of Al-Tuwaitha and these include the former uranium processing and waste storage facility at Jesira, the dump site near Adaya, the former centrifuge facility at Rashdiya and the former enrichment plant at Tarmiya. In 2005, Iraq lacked the infrastructure needed to decommission its nuclear facilities and manage its radioactive wastes. The lack of infrastructure included: (1) the lack of an organization responsible for decommissioning and radioactive waste management, (2) the lack of a storage facility for radioactive wastes, (3) the lack of professionals with experience in decommissioning and modern waste management practices, (4) the lack of laws and regulations governing decommissioning or radioactive waste management, (5) ongoing security concerns, and (6) limited availability of electricity and internet. Since its creation eight years ago, the MoST has worked with the international community and developed an organizational structure, trained staff, and made great progress in managing radioactive wastes and decommissioning Iraq's former nuclear facilities. This progress has been made, despite the very difficult implementing conditions in Iraq. Within MoST, the Radioactive Waste Treatment and Management Directorate (RWTMD) is responsible for waste management and the Iraqi Decommissioning Directorate (IDD) is responsible for decommissioning activities. The IDD and the RWTMD work together on decommissioning projects. The IDD has developed plans and has completed decommissioning of the GeoPilot Facility in Baghdad and the Active Metallurgical Testing Laboratory (LAMA) in Al-Tuwaitha. Given this experience, the IDD has initiated work on more dangerous facilities. Plans are being developed to characterize, decontaminate and decommission the Tamuz II Research Reactor. The Tammuz Reactor was destroyed by an Israeli air-strike in 1981 and the Tammuz II Reactor was destroyed during the First Gulf War in 1991. In addition to being responsible for managing the decommissioning wastes, the RWTMD is responsible for more than 950 disused sealed radioactive sources, contaminated debris from the first Gulf War and (approximately 900 tons) of naturally-occurring radioactive materials wastes from oil production in Iraq. The RWTMD has trained staff, rehabilitated the Building 39 Radioactive Waste Storage building, rehabilitated portions of the French-built Radioactive Waste Treatment Station, organized and secured thousands of drums of radioactive waste organized and secured the stores of disused sealed radioactive sources. Currently, the IDD and the RWTMD are finalizing plans for the decommissioning of the Tammuz II Research Reactor. (authors)

Al-Musawi, Fouad; Shamsaldin, Emad S.; Jasim, Hadi [Ministry of Science and Technology (MoST), Al-Jadraya, P.O. Box 0765, Baghdad (Iraq)] [Ministry of Science and Technology (MoST), Al-Jadraya, P.O. Box 0765, Baghdad (Iraq); Cochran, John R. [Sandia National Laboratories1, New Mexico, Albuquerque New Mexico 87185 (United States)] [Sandia National Laboratories1, New Mexico, Albuquerque New Mexico 87185 (United States)

2013-07-01T23:59:59.000Z

231

Glass former composition and method for immobilizing nuclear waste using the same  

DOE Patents [OSTI]

An alkoxide glass former composition has silica-containing constituents present as solid particulates of a particle size of 0.1 to 0.7 micrometers in diameter in a liquid carrier phase substantially free of dissolved silica. The glass former slurry is resistant to coagulation and may contain other glass former metal constituents. The immobilization of nuclear waste employs the described glass former by heating the same to reduce the volume, mixing the same with the waste, and melting the resultant mixture to encapsulate the waste in the resultant glass.

Cadoff, Laurence H. (Wilkins Township, Allegheny County, PA); Smith-Magowan, David B. (Washington, DC)

1988-01-01T23:59:59.000Z

232

Precipitate hydrolysis process for the removal of organic compounds from nuclear waste slurries  

DOE Patents [OSTI]

A process for removing organic compounds from a nuclear waste slurry comprising reacting a mixture of radioactive waste precipitate slurry and an acid in the presence of a catalytically effective amount of a copper(II) catalyst whereby the organic compounds in the precipitate slurry are hydrolyzed to form volatile organic compounds which are separated from the reacting mixture. The resulting waste slurry, containing less than 10 percent of the original organic compounds, is subsequently blended with high level radioactive sludge land transferred to a vitrification facility for processing into borosilicate glass for long-term storage. 2 figs., 3 tabs.

Doherty, J.P.; Marek, J.C.

1987-02-25T23:59:59.000Z

233

Precipitate hydrolysis process for the removal of organic compounds from nuclear waste slurries  

DOE Patents [OSTI]

A process for removing organic compounds from a nuclear waste slurry comprising reacting a mixture of radioactive waste precipitate slurry and an acid in the presence of a catalytically effective amount of a copper (II) catalyst whereby the organic compounds in the precipitate slurry are hydrolyzed to form volatile organic compounds which are separated from the reacting mixture. The resulting waste slurry, containing less than 10 percent of the orginal organic compounds, is subsequently blended with high level radioactive sludge and transferred to a virtrification facility for processing into borosilicate glass for long-term storage.

Doherty, Joseph P. (Elkton, MD); Marek, James C. (Augusta, GA)

1989-01-01T23:59:59.000Z

234

Renewable Development Fund (RDF)  

Broader source: Energy.gov [DOE]

Xcel Energy's Renewable Development Fund (RDF) was created in 1999 pursuant to the 1994 Radioactive Waste Management Facility Authorization Law (Minn. Stat. 116C.779). Originally, Xcel Energy was...

235

Radioactive waste shipments to Hanford Retrievable Storage from the General Electric Vallecitos Nuclear Center, Pleasanton, California  

SciTech Connect (OSTI)

During the next two decades the transuranic (TRU) wastes now stored in the burial trenches and storage facilities at the Hanford Site are to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant near Carlsbad, New Mexico for final disposal. Approximately 3.8% of the TRU waste to be retrieved for shipment to WIPP was generated at the General Electric (GE) Vallecitos Nuclear Center (VNC) in Pleasanton, California and shipped to the Hanford Site for storage. The purpose of this report is to characterize these radioactive solid wastes using process knowledge, existing records, and oral history interviews. The waste was generated almost exclusively from the activities, of the Plutonium Fuels Development Laboratory and the Plutonium Analytical Laboratory. Section 2.0 provides further details of the VNC physical plant, facility operations, facility history, and current status. The solid radioactive wastes were associated with two US Atomic Energy Commission/US Department of Energy reactor programs -- the Fast Ceramic Reactor (FCR) program, and the Fast Flux Test Reactor (FFTR) program. These programs involved the fabrication and testing of fuel assemblies that utilized plutonium in an oxide form. The types and estimated quantities of waste resulting from these programs are discussed in detail in Section 3.0. A detailed discussion of the packaging and handling procedures used for the VNC radioactive wastes shipped to the Hanford Site is provided in Section 4.0. Section 5.0 provides an in-depth look at this waste including the following: weight and volume of the waste, container types and numbers, physical description of the waste, radiological components, hazardous constituents, and current storage/disposal locations.

Vejvoda, E.J.; Pottmeyer, J.A.; DeLorenzo, D.S.; Weyns-Rollosson, M.I. [Los Alamos Technical Associates, Inc., NM (United States); Duncan, D.R. [Westinghouse Hanford Co., Richland, WA (United States)

1993-10-01T23:59:59.000Z

236

Burying democracy along with UK nuclear waste Stuart Haszeldine 3 May 2012 s.haszeldine@ed.ac.uk  

E-Print Network [OSTI]

Burying democracy along with UK nuclear waste Stuart Haszeldine 3 May 2012 s, in addition to the existing costs of decommissioning cleanup. So this is a problem about which a national up to accept. The UK was first to deploy civil nuclear power in 1956, but waste disposal has

237

The U.S. Nuclear Waste Technical Review Board evaluates the technical and scientific validity of ac-  

E-Print Network [OSTI]

Chapter 1 Overview The U.S. Nuclear Waste Technical Review Board evaluates the technical in Nevada for its suit- ability as a location for a repository for high-level ra- dioactive waste (HLW) and spent nuclear fuel (SNF). The U.S. Department of Energy (DOE) began studying Yucca Mountain

238

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

239

A compilation of reports of the Advisory Committee on nuclear waste, July 1996--June 1997  

SciTech Connect (OSTI)

This compilation contains 11 reports issued by the Advisory Committee on Nuclear Waste (ACNW) during the ninth year of its operation. The reports were submitted to the Chairman and Commissioners of the U.S. Nuclear Regulatory Commission. All reports prepared by the Committee have been made available to the public through the NRC Public Document Room, the U.S. Library of Congress, and the internet at http://www.nrc.gov/ACRSACNW.

NONE

1997-08-01T23:59:59.000Z

240

Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel  

SciTech Connect (OSTI)

Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electrorefiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form.

Michael F. Simpson; Michael N. Patterson; Joon Lee; Yifeng Wang; Joshua Versey; Ammon Williams; Supathorn Phongikaroon; James Allensworth; Man-Sung Yim

2013-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

Management of salt waste from electrochemical processing of used nuclear fuel  

SciTech Connect (OSTI)

Electrochemical processing of used nuclear fuel involves operation of one or more cells containing molten salt electrolyte. Processing of the fuel results in contamination of the salt via accumulation of fission products and transuranic (TRU) actinides. Upon reaching contamination limits, the salt must be removed and either disposed or treated to remove the contaminants and recycled back to the process. During development of the Experimental Breeder Reactor-II spent fuel treatment process, waste salt from the electro-refiner was to be stabilized in a ceramic waste form and disposed of in a high-level waste repository. With the cancellation of the Yucca Mountain high-level waste repository, other options are now being considered. One approach that involves direct disposal of the salt in a geologic salt formation has been evaluated. While waste forms such as the ceramic provide near-term resistance to corrosion, they may not be necessary to ensure adequate performance of the repository. To improve the feasibility of direct disposal, recycling a substantial fraction of the useful salt back to the process equipment could minimize the volume of the waste. Experiments have been run in which a cold finger is used for this purpose to crystallize LiCl from LiCl/CsCl. If it is found to be unsuitable for transportation, the salt waste could also be immobilized in zeolite without conversion to the ceramic waste form. (authors)

Simpson, M.F.; Patterson, M.N. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415 (United States); Lee, J.; Wang, Y. [Sandia National Laboratory, Albuquerque, NM (United States); Versey, J.; Phongikaroon, S. [University of Idaho, Idaho Falls, ID (United States)

2013-07-01T23:59:59.000Z

242

A review of the environmental survivability of telerobotic control sensor systems for use in nuclear waste tanks  

SciTech Connect (OSTI)

This report was prepared by the Oak Ridge National Laboratory (ORNL) and funded by the Department of Energy (DOE) Office of Technology Development (OTD) Robotics Technology Development Program (RTDP). During the next few years field deployment of remotely operated systems in nuclear waste cleanup operations will increase dramatically as DOE strives to efficiently and safely remediate the many waste storage sites. Typically, the most fragile components in remote systems are the sensors that provide feedback to the operators or to computer control algorithms. The purpose of this review is to determine the availability of environmentally hardened sensors to support control of a manipulator or vehicle system in a waste tank environment. The emphasis of the report is on the environmental ruggedness of currently available sensors. For the purpose of this review a set of nominal requirements for survivability were adopted conditions in the single-shell tanks at Hanford. This report is designed to be a practical guide to the state of the art in commercially available environmentally tolerant sensors for use with robotic systems. It is neither intended to be an exhaustive review of the technical literature on potential measurement techniques nor a complete physical review of the functioning of particular sensor systems. This report is intended to be a living document. As additional, corrected, or updated information is received from sensor manufacturers, it will be incorporated into the report database. The physical report will then be periodically revised and released in updated format. The authors wish to apologize to any sources of environmentally hardened sensors that were omitted during this review and encourage submission of new or updated data.

Holcomb, D.E.; Burks, B.L.

1994-05-01T23:59:59.000Z

243

Terminating Safeguards on Excess Special Nuclear Material: Defense TRU Waste Clean-up and Nonproliferation - 12426  

SciTech Connect (OSTI)

The Department of Energy (DOE) and the National Nuclear Security Administration (NNSA) manages defense nuclear material that has been determined to be excess to programmatic needs and declared waste. When these wastes contain plutonium, they almost always meet the definition of defense transuranic (TRU) waste and are thus eligible for disposal at the Waste Isolation Pilot Plant (WIPP). The DOE operates the WIPP in a manner that physical protections for attractiveness level D or higher special nuclear material (SNM) are not the normal operating condition. Therefore, there is currently a requirement to terminate safeguards before disposal of these wastes at the WIPP. Presented are the processes used to terminate safeguards, lessons learned during the termination process, and how these approaches might be useful for future defense TRU waste needing safeguards termination prior to shipment and disposal at the WIPP. Also described is a new criticality control container, which will increase the amount of fissile material that can be loaded per container, and how it will save significant taxpayer dollars. Retrieval, compliant packaging and shipment of retrievably stored legacy TRU waste has dominated disposal operations at WIPP since it began operations 12 years ago. But because most of this legacy waste has successfully been emplaced in WIPP, the TRU waste clean-up focus is turning to newly-generated TRU materials. A major component will be transuranic SNM, currently managed in safeguards-protected vaults around the weapons complex. As DOE and NNSA continue to consolidate and shrink the weapons complex footprint, it is expected that significant quantities of transuranic SNM will be declared surplus to the nation's needs. Safeguards termination of SNM varies due to the wide range of attractiveness level of the potential material that may be directly discarded as waste. To enhance the efficiency of shipping waste with high TRU fissile content to WIPP, DOE designed an over-pack container, similar to the pipe component, called the criticality control over-pack, which will significantly enhance the efficiency of disposal. Hundreds of shipments of transuranic SNM, suitably packaged to meet WIPP waste acceptance criteria and with safeguards terminated have been successfully emplaced at WIPP (primarily from the Rocky Flats site clean-up) since WIPP opened. DOE expects that thousands more may eventually result from SNM consolidation efforts throughout the weapons complex. (authors)

Hayes, Timothy [Los Alamos National Laboratory, Carlsbad Operations Group (United States); Nelson, Roger [Department Of Energy, Carlsbad Operations Office (United States)

2012-07-01T23:59:59.000Z

244

Application of oil exploration techniques toward finding a nuclear waste repository site  

SciTech Connect (OSTI)

Through its National Waste Terminal Storage (NWTS) Program, the U.S. Department of Energy is responsible for providing facilities to permanently dispose of high-level nuclear waste in a manner that will ensure public health and safety and that will be environmentally acceptable. Three separately coordinated projects, which have placed emphasis on deep, underground repositories are described. They are the Basalt Waste Isolation Project at the Hanford Reservation in Wahington, the Nevada Nuclear Waste Storage investigations at the Nevada Test Site, and a project focusing on suitable rock types in the remainder of the conterminous 48 states, managed by the Office of Nuclear Waste Isolation in Columbus, Ohio. The last-mentioned project is presently focused on rock salt and crystalline formations as suitable host rocks for a repository. The geologic evaluation considers stratigraphy, structure, hydrogeology, seismicity, tectonic history, Quaternary features, physiography, energy/mineral resources, and geotechnical factors. Extensive use has been made of drill stem test data, a mainstay of petroleum exploration since the late 1920`s, to obtain hydrologic information. (JMT)

Laughon, R.B.

1982-06-01T23:59:59.000Z

246

Integrated data base report--1996: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect (OSTI)

The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and U.S. Department of Energy (DOE) spent nuclear fuel (SNF) and commercial and U.S. government-owned radioactive wastes. Inventories of most of these materials are reported as of the end of fiscal year (FY) 1996, which is September 30, 1996. Commercial SNF and commercial uranium mill tailings inventories are reported on an end-of-calendar year (CY) basis. All SNF and radioactive waste data reported are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest DOE/Energy Information Administration (EIA) projections of U.S. commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are SNF, high-level waste, transuranic waste, low-level waste, uranium mill tailings, DOE Environmental Restoration Program contaminated environmental media, naturally occurring and accelerator-produced radioactive material, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given through FY 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions.

NONE

1997-12-01T23:59:59.000Z

247

Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report  

SciTech Connect (OSTI)

The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

A. K. Herbst; J. A. McCray; R. J. Kirkham; J. Pao; S. H. Hinckley

1999-09-30T23:59:59.000Z

248

Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-99 Status Report  

SciTech Connect (OSTI)

The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1999, grout formulations were studied for transuranic waste derived from INTEC liquid sodium-bearing waste and for projected newly generated low-level liquid waste. Additional studies were completed on radionuclide leaching, microbial degradation, waste neutralization, and a small mockup for grouting the INTEC underground storage tank residual heels.

Herbst, Alan Keith; Mc Cray, John Alan; Kirkham, Robert John; Pao, Jenn Hai; Hinckley, Steve Harold

1999-10-01T23:59:59.000Z

249

CORROSION OF NUCLEAR WASTE GLASSES IN NON-SATURATED CONDITIONS: TIME-TEMPERATURE BEHAVIOUR  

E-Print Network [OSTI]

CORROSION OF NUCLEAR WASTE GLASSES IN NON-SATURATED CONDITIONS: TIME-TEMPERATURE BEHAVIOUR Michael Rostovsky Lane, 2/14, Moscow, 119121, Russia This paper examines corrosion behaviour of radioactive term natural tests. These demonstrated diminishing of release rates of radionuclides by time. Corrosion

Sheffield, University of

250

Functionalized ultra-porous titania nanofiber membranes as nuclear waste separation and sequestration scaffolds for nuclear fuels recycle.  

SciTech Connect (OSTI)

Advanced nuclear fuel cycle concept is interested in reducing separations to a simplified, one-step process if possible. This will benefit from the development of a one-step universal getter and sequestration material so as a simplified, universal waste form was proposed in this project. We have developed a technique combining a modified sol-gel chemistry and electrospinning for producing ultra-porous ceramic nanofiber membranes with controllable diameters and porous structures as the separation/sequestration materials. These ceramic nanofiber materials have been determined to have high porosity, permeability, loading capacity, and stability in extreme conditions. These porous fiber membranes were functionalized with silver nanoparticles and nanocrystal metal organic frameworks (MOFs) to introduce specific sites to capture gas species that are released during spent nuclear fuel reprocessing. Encapsulation into a durable waste form of ceramic composition was also demonstrated.

Liu, Haiqing; Bell, Nelson Simmons; Cipiti, Benjamin B.; Lewis, Tom Goslee,; Sava, Dorina Florentina; Nenoff, Tina Maria

2012-09-01T23:59:59.000Z

251

An underground characterization program for a nuclear fuel waste disposal vault in plutonic rock  

SciTech Connect (OSTI)

The Canadian Nuclear Fuel Waste Management Program (CNFWMP) is developing a concept for disposing of nuclear fuel waste that involves placing and sealing it in a disposal vault excavated 500 to 1,000 m deep in the stable plutonic rock of the Canadian Shield. In this concept, engineered and natural barriers serve to isolate the waste from the biosphere. Since 1983, underground characterization and testing in support of the CNFWMP has been ongoing at the Underground Research Laboratory (URL) in southeastern Manitoba. This paper draws on experience gained at the URL to recommend an approach to underground characterization that would provide the necessary information to make design decisions for a disposal vault in plutonic rock.

Thompson, P.M.; Everitt, R.A. [AECL Research, Pinawa, Manitoba (Canada). Whiteshell Labs.

1993-12-31T23:59:59.000Z

252

Process for the recovery of curium-244 from nuclear waste  

SciTech Connect (OSTI)

A process has been designed for the recovery of curium from purex waste. Curium and americium are separated from the lanthanides by a TALSPEAK extraction process using differential extraction. Equations were derived for the estimation of the economically optimum conditions for the extraction using laboratory batch extraction data. The preparation of feed for the extraction involves the removal of nitric acid from the Purex waste by vaporization under reduced pressure, the leaching of soluble nitrates from the resulting cake, and the oxalate precipitation of a pure lanthanide-actinide fraction. Final separation of the curium from americium is done by ion-exchange. The steps of the process, except ion-exchange, were tested on a laboratory scale and workable conditions were determined.

Posey, J.C.

1980-10-01T23:59:59.000Z

253

THE STRUCTURAL CHEMISTRY OF MOLYBDENUM IN MODEL HIGH LEVEL NUCLEAR WASTE GLASSES, INVESTIGATED BY MO K-EDGE X-RAY ABSORPTION  

E-Print Network [OSTI]

THE STRUCTURAL CHEMISTRY OF MOLYBDENUM IN MODEL HIGH LEVEL NUCLEAR WASTE GLASSES, INVESTIGATED of molybdenum in model UK high level nuclear waste glasses was investigated by X-ray Absorption Spectroscopy (XAS). Molybdenum K-edge XAS data were acquired from several inactive simulant high level nuclear waste

Sheffield, University of

254

A Plan for Nuclear Waste http://www.washingtonpost.com/wp-dyn/content/article/2006/01/29/AR2006012900719_pf.html  

E-Print Network [OSTI]

A Plan for Nuclear Waste http://www.washingtonpost.com/wp-dyn/content/article/2006/01/29/AR2006012900719_pf.html A Plan for Nuclear Waste By John Deutch and Ernest J. Moniz U.S. policy for managing of radioactive waste and heighten rather than reduce public concerns about expanded nuclear power. We agree

Deutch, John

255

Molybdenum in Nuclear Waste Glasses -Incorporation and Redox state R.J. Short, R.J. Hand, N.C. Hyatt,  

E-Print Network [OSTI]

Molybdenum in Nuclear Waste Glasses - Incorporation and Redox state R.J. Short, R.J. Hand, N form in simulated high level nuclear waste (HLW) glass melts have been studied. It was found less attention has been paid to the effects of redox on nuclear waste glasses. One particular element

Sheffield, University of

256

A new DOE standard for transuranic waste nuclear safety analysis  

SciTech Connect (OSTI)

The DOE Office of Environmental Management (EM) observed through onsite assessments and a review of site-specific lessons learned that transuranic (TRU) waste operations could benefit from standardization of assumptions and approaches used to analyze hazards and select controls. EM collected and compared safety analysis information from DOE sites, including a comparison of the type of TRU waste accidents evaluated and controls selected, as well as specific Airborne Release Fractions (ARFs), Respirable Fractions (RFs), and Damage Ratios (DRs) assumed in accident analyses. This paper recounts the efforts by the DOE and its contractors to bring consistency to the safety analysis process supporting TRU waste operations through an integrated re-engineering effort. EM embarked on a process to re-engineer and standardize TRU safety analysis activities complex-wide. The effort involved DOE headquarters, field offices, and contractors. Five teams were formed to analyze and develop the necessary technical basis for a DOE Technical Standard. The teams looked at general issues including Safety Basis (SB), drum integrity and inspection criteria, hazard controls and analysis, safety analysis review and approval process, and implementation of hazard controls. (authors)

Triay, I.; Chung, D. [U.S. Department of Energy, Washington, D.C. (United States); Woody, J. [Atlas Consulting, Knoxville, TN (United States); Foppe, T. [Carlsbad Technical Assistance Contractor, Carlsbad, NM (United States); Mewhinney, C. [Sandia National Laboratories, Carlsbad, NM (United States); Jennings, S. [Los Alamos National Laboratories, Carlsbad, NM (United States)

2007-07-01T23:59:59.000Z

257

Materials characterization center workshop on the irradiation effects in nuclear waste forms  

SciTech Connect (OSTI)

The Workshop on Irradiation Effects in Nuclear Waste Forms sponsored by the Materials Characterization Center (MCC) brought together experts in radiation damage in materials and waste-management technology to review the problems associated with irradiation effects on waste-form integrity and to evaluate standard methods for generating data to be included in the Nuclear Waste Materials Handbook. The workshop reached the following conclusions: the concept of Standard Test for the Effects of Alpha-Decay in Nuclear Waste Solids, (MCC-6) for evaluating the effects of alpha decay is valid and useful, and as a result of the workshop, modifications to the proposed procedure will be incorpoated in a revised version of MCC-6; the MCC-6 test is not applicable to the evaluation of radiation damage in spent fuel; plutonium-238 is recommended as the dopant for transuranic and defense high-level waste forms, and when high doses are required, as in the case of commercial high-level waste forms, /sup 244/Cm can be used; among the important property changes caused by irradiation are those that lead to greater leachability, and additionally, radiolysis of the leachant may increase leach rates; research is needed in this area; ionization-induced changes in physical properties can be as important as displacement damage in some materials, and a synergism is also likely to exist from the combined effects of ionization and displacement damage; and the effect of changing the temperature and dose rates on property changes induced by radiation damage needs to be determined.

Roberts, F.P.; Turcotte, R.P.; Weber, W.J.

1981-01-01T23:59:59.000Z

258

Calcine Waste Storage at the Idaho Nuclear Technology and Engineering Center  

SciTech Connect (OSTI)

A potential option in the program for long-term management of high-level wastes at the Idaho Nuclear Technology and Engineering Center (INTEC), at the Idaho National Engineering and Environmental Laboratory, calls for retrieving calcine waste and converting it to a more stable and less dispersible form. An inventory of calcine produced during the period December 1963 to May 1999 has been prepared based on calciner run, solids storage facilities operating, and miscellaneous operational information, which gives the range of chemical compositions of calcine waste stored at INTEC. Information researched includes calciner startup data, waste solution analyses and volumes calcined, calciner operating schedules, solids storage bin capacities, calcine storage bin distributor systems, and solids storage bin design and temperature monitoring records. Unique information on calcine solids storage facilities design of potential interest to remote retrieval operators is given.

M. D. Staiger

1999-06-01T23:59:59.000Z

259

A Novel Fuel/Reactor Cycle to Implement the 300 Years Nuclear Waste Policy Approach - 12377  

SciTech Connect (OSTI)

A thorium-based fuel cycle system can effectively burn the currently accumulated commercial used nuclear fuel and move to a sustainable equilibrium where the actinide levels in the high level waste are low enough to yield a radiotoxicity after 300 years lower than that of the equivalent uranium ore. The second step of the Westinghouse approach to solving the waste 'problem' has been completed. The thorium fuel cycle has indeed the potential of burning the legacy TRU and achieve the waste objective proposed. Initial evaluations have been started for the third step, development and selection of appropriate reactors. Indications are that the probability of show-stoppers is rather remote. It is, therefore, believed that development of the thorium cycle and associated technologies will provide a permanent solution to the waste management. Westinghouse is open to the widest collaboration to make this a reality. (authors)

Carelli, M.D.; Franceschini, F.; Lahoda, E.J. [Westinghouse Electric Company LLC., Cranberry Township, PA (United States); Petrovic, B. [Georgia Institute of Technology, Atlanta, GA (United States)

2012-07-01T23:59:59.000Z

260

Progress and Status of the Ignalina Nuclear Power Plant's New Solid Waste Management and Storage Facilities  

SciTech Connect (OSTI)

A considerable amount of dry radioactive waste from former NPP operation has accumulated up to date and is presently stored at the Ignalina NPP site, Lithuania. Current storage capacities are nearly exhausted and more waste is to come from future decommissioning of the two RMBKtype reactors. Additionally, the existing storage facilities does not comply to the state-of-the-art technology for handling and storage of radioactive waste. In 2005, INPP faced this situation of a need for waste processing and subsequent interim storage of these wastes by contracting NUKEM with the design, construction, installation and commissioning of new waste management and storage facilities. The subject of this paper is to describe the scope and the status of the new solid waste management and storage facilities at the Ignalina Nuclear Power Plant. In summary: The turnkey contract for the design, supply and commission of the SWMSF was awarded in December 2005. The realisation of the project was initially planned within 48 month. The basic design was finished in August 2007 and the Technical Design Documentation and Preliminary Safety Analyses Report was provided to Authorities in October 2007. The construction license is expected in July 2008. The procurement phase was started in August 2007, start of onsite activities is expected in November 2007. The start of operation of the SWMSF is scheduled for end of 2009. (authors)

Rausch, J.; Henderson, R.W. [NUKEM Technologies GmbH, Alzenau (Germany); Penkov, V. [State Enterprise Ignalina Nuclear Power Plant, Visaginas (Lithuania)

2008-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

Secretary Chu Announces Funding for 71 University-Led Nuclear Research and  

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

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262

Secretary Chu Announces Funding for 71 University-Led Nuclear Research and  

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

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263

Secretary Chu Announces Funding for 71 University-Led Nuclear Research and  

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

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264

Final Recovery Act-Funded Demolition Underway at Y-12 | National Nuclear  

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNational NuclearhasAdministration77Nuclear SecurityFAPAC-NM | NationalFYSecurity

265

THERMAL IMPACT OF WASTE EMPLACEMENT AND SURFACE COOLING ASSOCIATED WITH GEOLOGIC DISPOSAL OF NUCLEAR WASTE  

E-Print Network [OSTI]

Cnalk R i v e r , Ontar o, AECL-b308, 46 p . Tnomas, R. K. ,with a radioactive waste vault. AECL--6308, Atomic Energys t u d y . P a r t 2. AECL-6188-2, Atomic Energy of Canada,

Wang, J.S.Y.

2010-01-01T23:59:59.000Z

266

Volume reduction and vitrification of nuclear waste with thermal plasma  

SciTech Connect (OSTI)

A process for efficient and safe destruction of organics and vitrification of low/medium level radioactive waste is presented. A transferred arc plasma torch is employed as the heat source. The process handles several types of feed: combustibles, inorganic materials and metals. A non-leaching glassy solid which can be stored without further treatment is obtained as the final product. High volume-reduction factors can be achieved with this process. A wet gas cleaning system leads to extremely clean off-gas.

Hoffelner, W. [Moser-Glaser and Co., Muttenz (Switzerland); Chrubasik, A. [NUKEM GmbH, Alzenau (Germany); Eschenbach, R.C. [RETECH Inc., Ukiah, CA (United States)

1993-12-31T23:59:59.000Z

267

Disposition of Nuclear Waste Using Subcritical Accelerator-Driven Systems  

SciTech Connect (OSTI)

ATW destroys virtually all the plutonium and higher actinides without reprocessing the spent fuel in a way that could lead to weapons material diversion. An ATW facility consists of three major elements: (1) a high-power proton linear accelerator; (2) a pyrochemical spent fuel treatment i waste cleanup system; (3) a liquid lead-bismuth cooled burner that produces and utilizes an intense source-driven neutron flux for transmutation in a heterogeneous (solid fuel) core. The concept is the result of many years of development at LANL as well as other major international research centers. Once demonstrated and developed, ATW could be an essential part of a global non-proliferation strategy for countries that could build up large quantities of plutonium from their commercial reactor waste. ATW technology, initially proposed in the US, has received wide and rapidly increasing attention abroad, especially in Europe and the Far East with major programs now being planned, organized and tided. Substantial convergence presently exists on the technology choices among the programs, opening the possibility of a strong and effective international collaboration on the phased development of the ATW technology.

Doolen, G.D.; Venneri, F.; Li, N.; Williamson, M.A.; Houts, M.; Lawrence, G.

1998-06-27T23:59:59.000Z

268

Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors  

DOE Patents [OSTI]

A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify /sup 233/U, /sup 235/U and /sup 239/Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as /sup 240/Pu, /sup 244/Cm and /sup 252/Cf, and the spontaneous alpha particle emitter /sup 241/Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether permanent low-level burial is appropriate for the waste sample.

Caldwell, J.T.; Kunz, W.E.; Atencio, J.D.

1982-03-31T23:59:59.000Z

269

Nevada Nuclear-Waste-Storage Investigations. Quarterly report, April-June 1982  

SciTech Connect (OSTI)

The Nevada Nuclear Waste Storage Investigations (NNWSI) are studying the Nevada Test Site (NTS) area to establish whether it would qualify as a licensable location for a commercial nuclear waste repository; determining whether specific underground rock masses in the NTS area are technically acceptable for permanently disposing of highly radioactive solid wastes; and developing and demonstrating the capability to safely handle and store commercial spent reactor fuel and high-level waste. Progress reports for the following eight tasks are presented: systems; waste package; site; repository; regulatory and institutional; test facilities; land acquisition; and program management. Some of the highlights are: A code library was established to provide a central location for documentation of repository performance assessment codes. A two-dimensional finite element code, SAGUARO, was developed for modeling saturated/unsaturated groundwater flow. The results of an initial experiment to determine canister penetration rates due to corrosion indicate the expected strong effect of toxic environmental conditions on the corrosion rate of carbon steel in tuff-conditioned water. Wells USW-H3 and USW-H4 at Yucca Mountain have been sampled for groundwater analysis. A summary characterizing and relating the mineralogy and petrology of Yucca Mountain tuffs was compiled from the findings of studies of core samples from five drill holes.

None

1982-09-01T23:59:59.000Z

270

Apparatus and method for quantitative assay of generic transuranic wastes from nuclear reactors  

DOE Patents [OSTI]

A combination of passive and active neutron measurements which yields quantitative information about the isotopic composition of transuranic wastes from nuclear power or weapons material manufacture reactors is described. From the measurement of prompt and delayed neutron emission and the incidence of two coincidentally emitted neutrons from induced fission of fissile material in the sample, one can quantify .sup.233 U, .sup.235 U and .sup.239 Pu isotopes in waste samples. Passive coincidence counting, including neutron multiplicity measurement and determination of the overall passive neutron flux additionally enables the separate quantitative evaluation of spontaneous fission isotopes such as .sup.240 Pu, .sup.244 Cm and .sup.252 Cf, and the spontaneous alpha particle emitter .sup.241 Am. These seven isotopes are the most important constituents of wastes from nuclear power reactors and once the mass of each isotope present is determined by the apparatus and method of the instant invention, the overall alpha particle activity can be determined to better than 1 nCi/g from known radioactivity data. Therefore, in addition to the quantitative analysis of the waste sample useful for later reclamation purposes, the alpha particle activity can be determined to decide whether "permanent" low-level burial is appropriate for the waste sample.

Caldwell, John T. (Los Alamos, NM); Kunz, Walter E. (Santa Fe, NM); Atencio, James D. (Los Alamos, NM)

1984-01-01T23:59:59.000Z

271

Waste component recycle, treatment, and disposal integrated demonstration (WeDID) nuclear weapon dismantlement activities  

SciTech Connect (OSTI)

One of the drivers in the dismantlement and disposal of nuclear weapon components is Envirorunental Protection Agency (EPA) guidelines. The primary regulatory driver for these components is the Resource Conservation Recovery Act (RCRA). Nuclear weapon components are heterogeneous and contain a number of hazardous materials including heavy metals, PCB`S, selfcontained explosives, radioactive materials, gas-filled tubes, etc. The Waste Component Recycle, Treatment, Disposal and Integrated Demonstration (WeDID) is a Department of Energy (DOE) Environmental Restoration and Waste Management (ERWM) sponsored program. It also supports DOE Defense Program (DP) dismantlement activities. The goal of WeDID is to demonstrate the end-to-end disposal process for Sandia National Laboratories designed nuclear weapon components. One of the primary objectives of WeDID is to develop and demonstrate advanced system treatment technologies that will allow DOE to continue dismantlement and disposal unhindered even as environmental regulations become more stringent. WeDID is also demonstrating waste minimization techniques by recycling a significant weight percentage of the bulk/precious metals found in weapon components and by destroying the organic materials typically found in these components. WeDID is concentrating on demonstrating technologies that are regulatory compliant, are cost effective, technologically robust, and are near-term to ensure the support of DOE dismantlement time lines. The waste minimization technologies being demonstrated by WeDID are cross cutting and should be able to support a number of ERWM programs.

Wheelis, W.T.

1993-04-12T23:59:59.000Z

272

Waste component recycle, treatment, and disposal integrated demonstration (WeDID) nuclear weapon dismantlement activities  

SciTech Connect (OSTI)

One of the drivers in the dismantlement and disposal of nuclear weapon components is Envirorunental Protection Agency (EPA) guidelines. The primary regulatory driver for these components is the Resource Conservation Recovery Act (RCRA). Nuclear weapon components are heterogeneous and contain a number of hazardous materials including heavy metals, PCB'S, selfcontained explosives, radioactive materials, gas-filled tubes, etc. The Waste Component Recycle, Treatment, Disposal and Integrated Demonstration (WeDID) is a Department of Energy (DOE) Environmental Restoration and Waste Management (ERWM) sponsored program. It also supports DOE Defense Program (DP) dismantlement activities. The goal of WeDID is to demonstrate the end-to-end disposal process for Sandia National Laboratories designed nuclear weapon components. One of the primary objectives of WeDID is to develop and demonstrate advanced system treatment technologies that will allow DOE to continue dismantlement and disposal unhindered even as environmental regulations become more stringent. WeDID is also demonstrating waste minimization techniques by recycling a significant weight percentage of the bulk/precious metals found in weapon components and by destroying the organic materials typically found in these components. WeDID is concentrating on demonstrating technologies that are regulatory compliant, are cost effective, technologically robust, and are near-term to ensure the support of DOE dismantlement time lines. The waste minimization technologies being demonstrated by WeDID are cross cutting and should be able to support a number of ERWM programs.

Wheelis, W.T.

1993-04-12T23:59:59.000Z

273

Integrated data base report--1995: US spent nuclear fuel and radioactive waste inventories, projections, and characteristics  

SciTech Connect (OSTI)

The information in this report summarizes the U.S. Department of Energy (DOE) data base for inventories, projections, and characteristics of domestic spent nuclear fuel and radioactive waste. This report is updated annually to keep abreast of continual waste inventory and projection changes in both the government and commercial sectors. Baseline information is provided for DOE program planning purposes and to support DOE program decisions. Although the primary purpose of this document is to provide background information for program planning within the DOE community, it has also been found useful by state and local governments, the academic community, and some private citizens.

NONE

1996-12-01T23:59:59.000Z

274

Sandia National Laboratories: Study Could Help Improve Nuclear Waste  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -theErikGroundbreakingStandards Solar

275

Nuclear-waste encapsulation by metal-matrix casting  

SciTech Connect (OSTI)

Several encapsulation casting processes are described that were developed or used at the Pacific Northwest Laboratory to embed simulated high-level wastes of two different forms (glass marbles and ceramic pellets) in metal matrices. Preliminary evaluations of these casting processes and the products are presented. Demonstrations have shown that 5- to 10-mm-dia glass marbles can be encapsulated on an engineering scale with lead or lead alloys by gravity or vacuum processes. Marbles approx. 12 mm in dia were successfully encapsulated in a lead alloy on a production scale. Also, 4- to 9-mm-dia ceramic pellets in containers of various sizes were completely penetrated and the individual pellets encased with aluminum-12 wt % silicon alloy by vacuum processes. Indications are that of the casting processes tested, aluminum 12 wt % silicon alloy vacuum-cast around ceramic pellets had the highest degree of infiltration or coverage of pellet surfaces.

Nelson, R.G.; Nesbitt, J.F.; Slate, S.C.

1981-05-01T23:59:59.000Z

276

Truck and rail charges for shipping spent fuel and nuclear waste  

SciTech Connect (OSTI)

The Pacific Northwest Laboratory developed techniques for calculating estimates of nuclear-waste shipping costs and compiled a listing of representative data that facilitate incorporation of reference shipping costs into varius logistics analyses. The formulas that were developed can be used to estimate costs that will be incurred for shipping spent fuel or nuclear waste by either legal-weight truck or general-freight rail. The basic data for this study were obtained from tariffs of a truck carrier licensed to serve the 48 contiguous states and from various rail freight tariff guides. Also, current transportation regulations as issued by the US Department of Transportation and the Nuclear Regulatory Commission were investigated. The costs that will be incurred for shipping spent fuel and/or nuclear waste, as addressed by the tariff guides, are based on a complex set of conditions involving the shipment origin, route, destination, weight, size, and volume and the frequency of shipments, existing competition, and the length of contracts. While the complexity of these conditions is an important factor in arriving at a ''correct'' cost, deregulation of the transportation industry means that costs are much more subject to negotiation and, thus, the actual fee that will be charged will not be determined until a shipping contract is actually signed. This study is designed to provide the baseline data necessary for making comparisons of the estimated costs of shipping spent fuel and/or nuclear wastes by truck and rail transportation modes. The scope of the work presented in this document is limited to the costs incurred for shipping, and does not include packaging, cask purchase/lease costs, or local fees placed on shipments of radioactive materials.

McNair, G.W.; Cole, B.M.; Cross, R.E.; Votaw, E.F.

1986-06-01T23:59:59.000Z

277

Decontamination of Nuclear Liquid Wastes Status of CEA and AREVA R and D: Application to Fukushima Waste Waters - 12312  

SciTech Connect (OSTI)

Liquid wastes decontamination processes are mainly based on two techniques: Bulk processes and the so called Cartridges processes. The first technique has been developed for the French nuclear fuel reprocessing industry since the 60's in Marcoule and La Hague. It is a proven and mature technology which has been successfully and quickly implemented by AREVA at Fukushima site for the processing of contaminated waters. The second technique, involving cartridges processes, offers new opportunities for the use of innovative adsorbents. The AREVA process developed for Fukushima and some results obtained on site will be presented as well as laboratory scale results obtained in CEA laboratories. Examples of new adsorbents development for liquid wastes decontamination are also given. A chemical process unit based on co-precipitation technique has been successfully and quickly implemented by AREVA at Fukushima site for the processing of contaminated waters. The asset of this technique is its ability to process large volumes in a continuous mode. Several chemical products can be used to address specific radioelements such as: Cs, Sr, Ru. Its drawback is the production of sludge (about 1% in volume of initial liquid volume). CEA developed strategies to model the co-precipitation phenomena in order to firstly minimize the quantity of added chemical reactants and secondly, minimize the size of co-precipitation units. We are on the way to design compact units that could be mobilized very quickly and efficiently in case of an accidental situation. Addressing the problem of sludge conditioning, cementation appears to be a very attractive solution. Fukushima accident has focused attention on optimizations that should be taken into account in future studies: - To better take account for non-typical aqueous matrixes like seawater; - To enlarge the spectrum of radioelements that can be efficiently processed and especially short lives radioelements that are usually less present in standard effluents resulting from nuclear activities; - To develop reversible solid adsorbents for cartridge-type applications in order to minimize wastes. (authors)

Fournel, B.; Barre, Y.; Lepeytre, C.; Peycelon, H. [CEA Marcoule, DTCD, BP17171, 30207 Bagnols sur Ceze (France); Grandjean, A. [Institut de Chimie Separative de Marcoule, UMR5257 CEA-CNRS-UM2-ENSCM, BP17171, 30207 Bagnols sur Ceze (France); Prevost, T.; Valery, J.F. [AREVA NC, Paris La Defense (France); Shilova, E.; Viel, P. [CEA Saclay, DSM/IRAMIS/SPCSI, 91191 Gif sur Yvette (France)

2012-07-01T23:59:59.000Z

278

Epsilon Metal Waste Form for Immobilization of Noble Metals from Used Nuclear Fuel  

SciTech Connect (OSTI)

Epsilon metal (?-metal), an alloy of Mo, Pd, Rh, Ru, and Tc, is being developed as a waste form to treat and immobilize the undissolved solids and dissolved noble metals from aqueous reprocessing of commercial used nuclear fuel. Epsilon metal is an attractive waste form for several reasons: increased durability relative to borosilicate glass, it can be fabricated without additives (100% waste loading), and in addition it also benefits borosilicate glass waste loading by eliminating noble metals from the glass and thus the processing problems related there insolubility in glass. This work focused on the processing aspects of the epsilon metal waste form development. Epsilon metal is comprised of refractory metals resulting in high reaction temperatures to form the alloy, expected to be 1500 - 2000C making it a non-trivial phase to fabricate by traditional methods. Three commercially available advanced technologies were identified: spark-plasma sintering, microwave sintering, and hot isostatic pressing, and investigated as potential methods to fabricate this waste form. Results of these investigations are reported and compared in terms of bulk density, phase assemblage (X-ray diffraction and elemental analysis), and microstructure (scanning electron microscopy).

Crum, Jarrod V.; Strachan, Denis M.; Rohatgi, Aashish; Zumhoff, Mac R.

2013-10-01T23:59:59.000Z

279

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center  

SciTech Connect (OSTI)

This comprehensive report provides definitive volume, mass, and composition (chemical and radioactivity) of calcined waste stored at the Idaho Nuclear Technology and Engineering Center. Calcine composition data are required for regulatory compliance (such as permitting and waste disposal), future treatment of the caline, and shipping the calcine to an off-Site-facility (such as a geologic repository). This report also contains a description of the calcine storage bins. The Calcined Solids Storage Facilities (CSSFs) were designed by different architectural engineering firms and built at different times. Each CSSF has a unique design, reflecting varying design criteria and lessons learned from historical CSSF operation. The varying CSSF design will affect future calcine retrieval processes and equipment. Revision 4 of this report presents refinements and enhancements of calculations concerning the composition, volume, mass, chemical content, and radioactivity of calcined waste produced and stored within the CSSFs. The historical calcine samples are insufficient in number and scope of analysis to fully characterize the entire inventory of calcine in the CSSFs. Sample data exist for all the liquid wastes that were calcined. This report provides calcine composition data based on liquid waste sample analyses, volume of liquid waste calcined, calciner operating data, and CSSF operating data using several large Microsoft Excel (Microsoft 2003) databases and spreadsheets that are collectively called the Historical Processing Model. The calcine composition determined by this method compares favorably with historical calcine sample data.

Staiger, M. Daniel, Swenson, Michael C.

2011-09-01T23:59:59.000Z

280

The NOx system in nuclear waste. 1997 annual progress report  

SciTech Connect (OSTI)

'The authors highlight their results from the title project. The project is a coordinated effort of the three Co-PIs to assist the Safety Programs at the Hanford and other DOE Environmental Management Sites. The authors present in the report their observations and interactively discuss their implications for safety concerns. They focus on three issues: (1) Reducing radicals in the NOx system The authors show that the only reducing radical that lasts longer than a few ns in typical waste solutions, and is capable of generating hydrogen, is NO{sub 3}{sup 2-}. The authors measured the lifetime of this species across the whole pH range (3 {le} pH {le} 14) and found it to be shorter than -15 \\265s, before it dissociates to give the strongly oxidizing NO, radicals. They found that it reacts with many proton donors (H{sup +}, phosphate, borate, NH{prime}, amines) in a reaction that is not merely an acid-base equilibrium reaction but is probably a dissociative proton transfer. They estimate the redox potential from theoretical considerations and obtain an experimental verification. They conclude that it is highly unlikely, although thermodynamically possible, that this radi-cal will generate hydrogen in waste solutions. (2) Aging of organic chelators and their degradation products by NO, Methodologies to study the degradation of organic substrates (including the important waste components, formate and oxalate) to CO;, or carbonate, by NO, were developed. This radical dimerizes and disproportionates to nitrate and nitrite. Therefore, mineraliza-tion of the organic substrates competes with the disproportionation of NO,. Among the organic substrates, formate and oxalate are also mineralized but because they are of low fuel value their mineralization is not very helpful, yet it consumes NO,. (3) Interfacial processes in aqueous suspensions Yields of charge transfer from solid silica particles to water and other liquids were meas-ured. If the particles are small enough, essentially all of the charge that is originally depos-ited in the solid escapes into the liquid. This implies that the solid/liquid interface does not provide a significant barrier to the transfer of charges into the solution when the particles are very small (I 20 nm). Electrons may reach the liquid and generate hydrogen, for example. On the other hand, the same mechanism may also provide a pathway for oxidative aging of organics by holes even when the organic is dissolved in the liquid or adsorbed on the solid surface. The authors have started to study reactions of NO,. Methodology and instrumentation to measure reactions of relevant organic radicals with NO, and with its parent NO, were developed. Because of low extinction coefficients, conductivity will be the method of choice.'

Meisel, D. [Argonne National Lab., IL (US). Chemistry Div.; Camaioni, D.; Orlando, T. [Pacific Northwest National Lab., Richland, WA (US)

1997-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

Chemistry, radiation, and interfaces in suspensions of nuclear waste simulants  

SciTech Connect (OSTI)

We focus in this report on three issues that are of central importance in the management of radioactive high-level liquid waste (HLLW). We show that the only reducing radical that lasts longer than a few ps in typical HLLW, and is capable of generating hydrogen, is NO{sub 3}{sup 2-}. We measured the lifetime of this species across the whole pH range (3 {le} pH {le} 14) and found it to be shorter than {approx}15 {mu}s, before is dissociates to give strongly oxidizing NO{sub 2} radicals. We found that it reacts with many proton donors (H{sup +} phosphate, borate, NH{sub 4}{sup +}, amines) in a reaction that is not merely an acid-base equilibrium reaction. Using high-level ab initio calculations we estimate its redox potential and pK{sub a}. We have developed methodologies to study the degradation of organic additives to the HLLW (to CO{sub 2} or carbonate) by NO{sub 2}. Relative rates of degradation of several complexants were determined using competition kinetics and {sup 13}C NMR and proton NMR detection techniques. Direct absorption of the radiation (low-energy electrons as well as photons above the ionization threshold) in NaNO{sub 3} single crystal at the solid/vacuum interface led to production of NO, O, and O- as the major products.

Meisel, D.; Cook, A.; Camaioni, D.; Orlando, T.

1997-08-01T23:59:59.000Z

282

Politics and promises of nuclear waste disposal: the view from Nevada  

SciTech Connect (OSTI)

DOE`s betrayal of the principles and standards of the Nuclear Waste Policy Act (NWPA) has distorted the agency`s repository-siting decisions. Leadership is needed to make midcourse corrections and to return to the promise of state-federal cooperation on which the act was built. NWPA managed to incorporate the interests of diverse factions into a decision-making process that was viewed as an equitable and workable solution to the nation`s nuclear waste disposal dilemma. The House of Representatives subcommittees report documents conclusively a substantial and pervasive bias in favor of the selection of sites at Yucca Mountain and Hanford and a politization of the siting process.

Bryan, R.H.

1987-10-01T23:59:59.000Z

283

Assessing the Feasibility of Interrogating Nuclear Waste Storage Silos using Cosmic-ray Muons  

E-Print Network [OSTI]

Muon radiography is a fast growing field in applied scientific research. In recent years, many detector technologies and imaging techniques using the Coulomb scattering and absorption properties of cosmic-ray muons have been developed for the non-destructive assay of various structures across a wide range of applications. This work presents the first results that assess the feasibility of using muons to interrogate waste silos within the UK Nuclear Industry. Two such approaches, using different techniques that exploit each of these properties, have previously been published, and show promising results from both simulation and experimental data for the detection of shielded high-Z materials and density variations from volcanic assay. Both detector systems are based on scintillator and photomultiplier technologies. Results from dedicated simulation studies using both these technologies and image reconstruction techniques are presented for an intermediate-sized nuclear waste storage facility filled with concrete...

Ambrosino, F; Cimmino, L; D'Alessandro, R; Ireland, D G; Kaiser, R; Mahon, D F; Mori, N; Noli, P; Saracino, G; Shearer, C; Viliani, L; Yang, G

2014-01-01T23:59:59.000Z

284

Nuclear Waste Disposal and Strategies for Predicting Long-Term Performance of Material  

SciTech Connect (OSTI)

Ceramics have been an important part of the nuclear community for many years. On December 2, 1942, an historic event occurred under the West Stands of Stagg Field, at the University of Chicago. Man initiated his first self-sustaining nuclear chain reaction and controlled it. The impact of this event on civilization is considered by many as monumental and compared by some to other significant events in history, such as the invention of the steam engine and the manufacturing of the first automobile. Making this event possible and the successful operation of this first man-made nuclear reactor, was the use of forty tons of UO2. The use of natural or enriched UO2 is still used today as a nuclear fuel in many nuclear power plants operating world-wide. Other ceramic materials, such as 238Pu, are used for other important purposes, such as ceramic fuels for space exploration to provide electrical power to operate instruments on board spacecrafts. Radioisotopic Thermoelectric Generators (RTGs) are used to supply electrical power and consist of a nuclear heat source and converter to transform heat energy from radioactive decay into electrical power, thus providing reliable and relatively uniform power over the very long lifetime of a mission. These sources have been used in the Galileo spacecraft orbiting Jupiter and for scientific investigations of Saturn with the Cassini spacecraft. Still another very important series of applications using the unique properties of ceramics in the nuclear field, are as immobilization matrices for management of some of the most hazardous wastes known to man. For example, in long-term management of radioactive and hazardous wastes, glass matrices are currently in production immobilizing high-level radioactive materials, and cementious forms have also been produced to incorporate low level wastes. Also, as part of nuclear disarmament activities, assemblages of crystalline phases are being developed for immobilizing weapons grade plutonium, to not only produce environmentally friendly products, but also forms that are proliferation resistant. All of these waste forms as well as others, are designed to take advantage of the unique properties of the ceramic systems.

Wicks, G.G.

2001-03-28T23:59:59.000Z

285

Sub-Seabed Repository for Nuclear Waste - a Strategic Alternative - 13102  

SciTech Connect (OSTI)

It was recognized at the outset of nuclear power generation in the 1950's that the waste products would require isolation away from humans for periods in excess of 10,000 years. After years studying alternatives, the DOE recommended pursuing the development of a SNF/HLW disposal facility within Yucca Mountain in the desert of Nevada. That recommendation became law with passage of the NWPAA, effectively stopping development of other approaches to the waste problem. In the face of political resistance from the state of Nevada, the 2010 decision to withdraw the license application for the geologic repository at Yucca Mountain has delayed further the most mature option for safe, long-term disposal of SNF and HLW. It is time to revisit an alternative option, sub-seabed disposal within the US Exclusive Economic Zone (EEZ), which would permanently sequester waste out of the biosphere, and out of the reach of saboteurs or terrorists. A proposal is made for a full scale pilot project to demonstrate burying radioactive waste in stable, deep ocean sediments. While much of the scientific work on pelagic clays has been done to develop a sub-seabed waste sequestration capability, this proposal introduces technology from non-traditional sources such as riser-less ocean drilling and the Navy's Sound Surveillance System. The political decisions affecting the issue will come down to site selection and a thorough understanding of comparative risks. The sub-seabed sequestration of nuclear waste has the potential to provide a robust solution to a critical problem for this clean and reliable energy source. (authors)

McAllister, Keith R. [Department of the Navy, 15 Turkey Foot Court, Darnestown, MD 20878 (United States)] [Department of the Navy, 15 Turkey Foot Court, Darnestown, MD 20878 (United States)

2013-07-01T23:59:59.000Z

286

MODELING SOLIDIFICATION-INDUCED STRESSES IN CERAMIC WASTE FORMS CONTAINING NUCLEAR WASTES  

SciTech Connect (OSTI)

The goal of this work is to produce a ceramic waste form (CWF) that permanently occludes radioactive waste. This is accomplished by absorbing radioactive salts into zeolite, mixing with glass frit, heating to a molten state 915 C to form a sodalite glass matrix, and solidifying for long-term storage. Less long term leaching is expected if the solidifying cooling rate doesnt cause cracking. In addition to thermal stress, this paper proposes that a stress is formed during solidification which is very large for fast cooling rates during solidification and can cause severe cracking. A solidifying glass or ceramic cylinder forms a dome on the cylinder top end. The temperature distribution at the time of solidification causes the stress and the dome. The dome height, the length deficit, produces an axial stress when the solid returns to room temperature with the inherent outer region in compression, the inner in tension. Large tensions will cause cracking of the specimen. The temperature deficit, derived by dividing the length deficit by the coefficient of thermal expansion, allows solidification stress theory to be extended to the circumferential stress. This paper derives the solidification stress theory, gives examples, explains how to induce beneficial stresses, and compares theory to experimental data.

Charles W. Solbrig; Kenneth J. Bateman

2010-11-01T23:59:59.000Z

287

Long-Term Environmental Monitoring of an Operating Deep Geologic Nuclear Waste Repository  

SciTech Connect (OSTI)

In the present energy dilemma in which we find ourselves, the magnitude of humanity's energy needs requires that we embrace a multitude of various energy sources and applications. Nuclear energy must be a major portion of the distribution. One often-cited strategic hurdle to the commercial production of nuclear energy is the apparent lack of an acceptable nuclear waste repository. This issue has been quietly addressed at the U. S. Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP; see http://www.wipp.energy.gov), the closest population center of significant size being Carlsbad, New Mexico. WIPP has been operating for about nine years, disposing of over 250,000 drum-equivalents of nuclear waste. From the standpoint of addressing operational and environmental risk, as well as public fear, WIPP has had extensive human health and environmental monitoring. The Carlsbad Environmental Monitoring and Research Center is in the Institute for Energy and the Environment, in the College of Engineering at New Mexico State University. Located in Carlsbad, NM, CEMRC has been the independent monitoring facility for the area around WIPP from 1993 to the present, i.e., from six years before disposal operations began to nine years of waste disposal operations (www.cemcr.org). Based on the radiological analyses of monitoring samples completed to date for area residents and site workers, and for selected aerosols, soils, sediments, drinking water and surface waters, there is no evidence of increases in radiological contaminants in the region of WIPP that could be attributed to releases from WIPP. Levels of radiological and non-radiological analytes measured since operations began in 1999 have been within the range of baseline levels measured previously, and are within the ranges measured by other entities at the State and local levels since well before disposal phase operations began in 1999. (authors)

Conca, J.; Kirchner, Th.; Monk, J.; Sage, S. [Carlsbad Environmental Monitoring and Research Center, IEE NMSU, 1400 University Drive, Carlsbad, NM (United States)

2008-07-01T23:59:59.000Z

288

A Characteristics-Based Approach to Radioactive Waste Classification in Advanced Nuclear Fuel Cycles  

E-Print Network [OSTI]

on the impact of waste heat load on waste involve coupling waste heat load with metrics radionuclides in the waste, heat generated by

Djokic, Denia

2013-01-01T23:59:59.000Z

289

Application of Direct Assessment Approaches and Methodologies to Cathodically Protected Nuclear Waste Transfer Lines  

SciTech Connect (OSTI)

The U.S. Department of Energy's (DOE) Hanford Site is responsible for the safe storage, retrieval, treatment, and disposal of approximately 54 million gallons (204 million liters) of radioactive waste generated since the site's inception in 1943. Today, the major structures involved in waste management at Hanford include 149 carbon steel single-shell tanks, 28 carbon-steel double-shell tanks, plus a network of buried metallic transfer lines and ancillary systems (pits, vaults, catch tanks, etc.) required to store, retrieve, and transfer waste within the tank farm system. Many of the waste management systems at Hanford are still in use today. In response to uncertainties regarding the structural integrity of these systems,' an independent, comprehensive integrity assessment of the Hanford Site piping system was performed. It was found that regulators do not require the cathodically protected pipelines located within the Hanford Site to be assessed by External Corrosion Direct Assessment (ECDA) or any other method used to ensure integrity. However, a case study is presented discussing the application of the direct assessment process on pipelines in such a nuclear environment. Assessment methodology and assessment results are contained herein. An approach is described for the monitoring, integration of outside data, and analysis of this information in order to identify whether coating deterioration accompanied by external corrosion is a threat for these waste transfer lines.

Dahl, Megan M. [ARES Corporation, Richland, WA (United States); Pikas, Joseph [Schiff Associates, Sugar Land TX (United States); Edgemon, Glenn L. [ARES Corporation, Richland, WA (United States); Philo, Sarah [ARES Corporation, Richland, WA (United States)

2013-01-22T23:59:59.000Z

290

Development of an Integrated Raman and Turbidity Fiber Optic Sensor for the In-Situ Analysis of High Level Nuclear Waste - 13532  

SciTech Connect (OSTI)

Stored nuclear waste must be retrieved from storage, treated, separated into low- and high-level waste streams, and finally put into a disposal form that effectively encapsulates the waste and isolates it from the environment for a long period of time. Before waste retrieval can be done, waste composition needs to be characterized so that proper safety precautions can be implemented during the retrieval process. In addition, there is a need for active monitoring of the dynamic chemistry of the waste during storage since the waste composition can become highly corrosive. This work describes the development of a novel, integrated fiber optic Raman and light scattering probe for in situ use in nuclear waste solutions. The dual Raman and turbidity sensor provides simultaneous chemical identification of nuclear waste as well as information concerning the suspended particles in the waste using a common laser excitation source. (authors)

Gasbarro, Christina; Bello, Job [EIC Laboratories, Inc., 111 Downey St., Norwood, MA, 02062 (United States)] [EIC Laboratories, Inc., 111 Downey St., Norwood, MA, 02062 (United States); Bryan, Samuel; Lines, Amanda; Levitskaia, Tatiana [Pacific Northwest National Laboratory, PO Box 999, Richland, WA, 99352 (United States)] [Pacific Northwest National Laboratory, PO Box 999, Richland, WA, 99352 (United States)

2013-07-01T23:59:59.000Z

291

Field testing of waste forms using lysimeters  

SciTech Connect (OSTI)

The Low-Level Waste Data Base Development - EPICOR-II Resin/Liner Investigation Program funded by the US Nuclear Regulatory Commission is obtaining information on performance of radioactive waste in a disposal environment. Waste forms manufactured from ion exchange resins used to clean up water from the accident at Three Mile Island Nuclear Power Station are being examined in field tests. This paper presents a description of the field testing and results from the first year of operation. 8 refs., 8 figs., 4 tabs.

McConnell, J.W. Jr.; Rogers, R.D.

1987-01-01T23:59:59.000Z

292

MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY  

SciTech Connect (OSTI)

The Idaho National Engineering Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy is in part due to the history of the INL as the National Reactor Testing Station, in part to its mission to recover highly enriched uranium from SNF and in part to its mission to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facility, some dating back 50 years in the site history. The success of the INL SNF program is measured by its ability to: 1) achieve safe existing storage, 2) continue to receive SNF from other locations, both foreign and domestic, 3) repackage SNF from wet storage to interim dry storage, and 4) prepare the SNF for dispositioning in a federal repository. Because of the diversity in the SNF and the facilities at the INL, the INL is addressing almost very condition that may exist in the SNF world. Many of solutions developed by the INL are applicable to other SNF storage sites as they develop their management strategy. The SNF being managed by the INL are in a variety of conditions, from intact assemblies to individual rods or plates to powders, rubble, and metallurgical mounts. Some of the fuel has been in wet storage for over forty years. The fuel is stored bare, or in metal cans and either wet under water or dry in vaults, caissons or casks. Inspections have shown varying degrees of corrosion and degradation of the fuel and the storage cans. Some of the fuel has been recanned under water, and the conditions of the fuel inside the second or third can are unknown. The fuel has been stored in one of 10 different facilities: five wet pools and one casks storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The wet pools range from forty years old to the most modern pool in the US Department of Energy (DOE) complex. The near-term objective is moving the fuel in the older wet storage facilities to interim dry storage facilities, thus permitting the shutdown and decommission of the older facilities. Two wet pool facilities, one at the Idaho Nuclear Technology and Engineering Center and the other at Test Area North, were targeted for initial SNF movements since these were some of the oldest at the INL. Because of the difference in the SNF materials different types of drying processes had to be developed. Passive drying, as is done with typical commercial SNF was not an option because on the condition of some of the fuel, the materials to be dried, and the low heat generation of some of the SNF. There were also size limitations in the existing facility. Active dry stations were designed to address the specific needs of the SNF and the facilities.

Hill, Thomas J

2005-09-01T23:59:59.000Z

293

Proceedings of the 1993 international conference on nuclear waste management and environmental remediation. Volume 3: Environmental remediation and environmental management issues  

SciTech Connect (OSTI)

This conference was held in 1993 in Prague, Czech Republic to provide a forum for exchange of state-of-the-art information on radioactive waste management. Papers are divided into the following sections: Low/Intermediate level waste disposal from an international viewpoint; Solid waste volume reduction, treatment and packaging experience; Design of integrated systems for management of nuclear wastes; Mixed waste (hazardous and radioactive) treatment and disposal; Advanced low/intermediate level waste conditioning technologies including incineration; National programs for low/intermediate waste management; Low/Intermediate waste characterization, assay, and tracking systems; Disposal site characterization and performance assessment; Radioactive waste management and practices in developing countries; Waste management from unconventional (e.g. VVER) nuclear power reactors; Waste minimization, avoidance and recycling in nuclear power plants; Liquid waste treatment processes and experience; Low/Intermediate waste storage facilities--design and experience; Low/Intermediate waste forms and acceptance criteria for disposal; Management of non-standard or accident waste; and Quality assurance and control in nuclear waste management. Individual papers have been processed separately for inclusion in the appropriate data bases.

Baschwitz, R.; Kohout, R.; Marek, J.; Richter, P.I.; Slate, S.C. [eds.

1993-12-31T23:59:59.000Z

294

Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program FY-98 Status Report  

SciTech Connect (OSTI)

The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

Herbst, Alan Keith; Mc Cray, John Alan; Rogers, Adam Zachary; Simmons, R. F.; Palethorpe, S. J.

1999-03-01T23:59:59.000Z

295

Idaho Nuclear Technology and Engineering Center Low-Activity Waste Process Technology Program, FY-98 Status Report  

SciTech Connect (OSTI)

The Low-Activity Waste Process Technology Program at the Idaho Nuclear Technology and Engineering Center (INTEC) anticipates that large volumes of low-level/low-activity wastes will need to be grouted prior to near-surface disposal. During fiscal year 1998, three grout formulations were studied for low-activity wastes derived from INTEC liquid sodium-bearing waste. Compressive strength and leach results are presented for phosphate bonding cement, acidic grout, and alkaline grout formulations. In an additional study, grout formulations are recommended for stabilization of the INTEC underground storage tank residual heels.

Herbst, A.K.; Rogers, A.Z.; McCray, J.A.; Simmons, R.F.; Palethorpe, S.J.

1999-03-01T23:59:59.000Z

296

Yucca Mountain: How Do Global and Federal Initiatives Impact Clark County's Nuclear Waste Program?  

SciTech Connect (OSTI)

Since 1987, Clark County has been designated by the U.S. Department of Energy (DOE) as an 'Affected Unit of Local Government' (AULG). The AULG designation is an acknowledgement by the federal government that activities associated with the Yucca Mountain proposal could result in considerable impacts on Clark County residents and the community as a whole. As an AULG, Clark County is authorized to identify 'any potential economic, social, public health and safety, and environmental impacts of a repository', 42 U.S.C. Section 10135(c)(1)(B)(i) under provisions of the Nuclear Waste Policy Act Amendments (NWPAA). Clark County's oversight program contains key elements of (1) technical and scientific analysis (2) transportation analysis (3) impact assessment and monitoring (4) policy and legislative analysis and monitoring, and (5) public outreach. Clark County has conducted numerous studies of potential impacts, many of which are summarized in Clark County's Impact Assessment Report that was submitted DOE and the President of the United States in February 2002. Given the unprecedented magnitude and duration of DOE's proposal, as well as the many unanswered questions about the transportation routes, number of shipments, and the modal mix that will ultimately be used, impacts to public health and safety and security, as well as socioeconomic impacts, can only be estimated. In order to refine these estimates, Clark County Comprehensive Planning Department's Nuclear Waste Division updates, assesses, and monitors impacts on a regular basis. Clark County's Impact Assessment program covers not only unincorporated Clark County but all five jurisdictions of Las Vegas, North Las Vegas, Henderson, Mesquite, and Boulder City as well as tribal jurisdictions that fall within Clark County's geographic boundary. National and global focus on nuclear power and nuclear waste could have significant impact on the Yucca Mountain Program, and therefore, Clark County's oversight of that program. (authors)

Navis, I.; McGehee, B. [Clark County Department of Comprehensive Planning - Nuclear Waste Division, Las Vegas, NV (United States)

2008-07-01T23:59:59.000Z

297

Scale-up considerations relevant to experimental studies of nuclear waste-package behavior  

SciTech Connect (OSTI)

Results from a study that investigated whether testing large-scale nuclear waste-package assemblages was technically warranted are reported. It was recognized that the majority of the investigations for predicting waste-package performance to date have relied primarily on laboratory-scale experimentation. However, methods for the successful extrapolation of the results from such experiments, both geometrically and over time, to actual repository conditions have not been well defined. Because a well-developed scaling technology exists in the chemical-engineering discipline, it was presupposed that much of this technology could be applicable to the prediction of waste-package performance. A review of existing literature documented numerous examples where a consideration of scaling technology was important. It was concluded that much of the existing scale-up technology is applicable to the prediction of waste-package performance for both size and time extrapolations and that conducting scale-up studies may be technically merited. However, the applicability for investigating the complex chemical interactions needs further development. It was recognized that the complexity of the system, and the long time periods involved, renders a completely theoretical approach to performance prediction almost hopeless. However, a theoretical and experimental study was defined for investigating heat and fluid flow. It was concluded that conducting scale-up modeling and experimentation for waste-package performance predictions is possible using existing technology. A sequential series of scaling studies, both theoretical and experimental, will be required to formulate size and time extrapolations of waste-package performance.

Coles, D.G.; Peters, R.D.

1986-04-01T23:59:59.000Z

298

Solid Waste Management Program (South Dakota)  

Broader source: Energy.gov [DOE]

South Dakota's Solid Waste Management Program offers loans and grants for solid waste disposal, recycling, and waste tire projects. Funds are available for private or public projects, and...

299

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410  

SciTech Connect (OSTI)

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for use as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)

Phillips, Chris; Willis, William; Carter, Robert [EnergySolutions Federal EPC., 2345 Stevens Drive, Richland, WA, 99354 (United States)] [EnergySolutions Federal EPC., 2345 Stevens Drive, Richland, WA, 99354 (United States); Baker, Stephen [UK National Nuclear Laboratory, Warrington, Cheshire (United Kingdom)] [UK National Nuclear Laboratory, Warrington, Cheshire (United Kingdom)

2013-07-01T23:59:59.000Z

300

Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model - 13413  

SciTech Connect (OSTI)

This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system, and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity. (authors)

Djokic, Denia [Department of Nuclear Engineering, University of California - Berkeley, 4149 Etcheverry Hall, Berkeley, CA 94720-1730 (United States)] [Department of Nuclear Engineering, University of California - Berkeley, 4149 Etcheverry Hall, Berkeley, CA 94720-1730 (United States); Piet, Steven J.; Pincock, Layne F.; Soelberg, Nick R. [Idaho National Laboratory - INL, 2525 North Fremont Avenue, Idaho Falls, ID 83415 (United States)] [Idaho National Laboratory - INL, 2525 North Fremont Avenue, Idaho Falls, ID 83415 (United States)

2013-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" from the National Library of EnergyBeta (NLEBeta).
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301

Waste Classification based on Waste Form Heat Generation in Advanced Nuclear Fuel Cycles Using the Fuel-Cycle Integration and Tradeoffs (FIT) Model  

SciTech Connect (OSTI)

This study explores the impact of wastes generated from potential future fuel cycles and the issues presented by classifying these under current classification criteria, and discusses the possibility of a comprehensive and consistent characteristics-based classification framework based on new waste streams created from advanced fuel cycles. A static mass flow model, Fuel-Cycle Integration and Tradeoffs (FIT), was used to calculate the composition of waste streams resulting from different nuclear fuel cycle choices. This analysis focuses on the impact of waste form heat load on waste classification practices, although classifying by metrics of radiotoxicity, mass, and volume is also possible. The value of separation of heat-generating fission products and actinides in different fuel cycles is discussed. It was shown that the benefits of reducing the short-term fission-product heat load of waste destined for geologic disposal are neglected under the current source-based radioactive waste classification system , and that it is useful to classify waste streams based on how favorable the impact of interim storage is in increasing repository capacity.

Denia Djokic; Steven J. Piet; Layne F. Pincock; Nick R. Soelberg

2013-02-01T23:59:59.000Z

302

Expected near-field thermal performance for nuclear waste repositories at potential salt sites: Technical report  

SciTech Connect (OSTI)

Thermal analyses were made for the environmental assessments of seven potential salt sites for a nuclear waste repository. These analyses predicted that potential repository sites in domal salts located in the Gulf Coast will experience higher temperature than those in bedded salts of Paradox and Palo Duro Basins, mainly because of higher ambient temperatures at depth. The TEMPV5 code, a semi-analytical heat transfer code for finite line sources, calculated temperatures for commercial high-level waste (CHLW) and spent fuel from pressurized-water reactors (SFPWR). Benchmarks with HEATING6, THAC-SIP-3D, STEALTH, and SPECTROM-41 showed that TEMPV5 agreed closely in the very near field around the waste package and approximately in the near-field and far-field regions of the repository. The analyses used site-specific thermal conductivities that were increased by 40% to compensate for reductions caused by testing technique, salt impurities, and other heterogeneities, and sampling disturbance. Analyses showed peak salt temperatures of 236/sup 0/C (CHLW) and 134/sup 0/C (SFPWR) for the bedded salt and 296/sup 0/C (CHLW) and 180/sup 0/C (SFPWR) for the domal salt. Analyses with uncorrected laboratory thermal conductivities would increase peak salt temperatures by about 120/sup 0/C for CHLW and about 60/sup 0/C for SFPWR. These temperature increases would increase the thermally induced flow of brine and accelerate corrosion of the waste package. 30 refs., 35 figs., 48 tabs.

McNulty, E.G.

1987-08-01T23:59:59.000Z

303

Simulation of Self-Irradiation of High-Sodium Content Nuclear Waste Glasses  

SciTech Connect (OSTI)

Alkali-borosilicate glasses are widely used in nuclear industry as a matrix for immobilisation of hazardous radioactive wastes. Durability or corrosion resistance of these glasses is one of key parameters in waste storage and disposal safety. It is influenced by many factors such as composition of glass and surrounding media, temperature, time and so on. As these glasses contain radioactive elements most of their properties including corrosion resistance are also impacted by self-irradiation. The effect of external gamma-irradiation on the short-term (up to 27 days) dissolution of waste borosilicate glasses at moderate temperatures (30 deg. to 60 deg. C) was studied. The glasses studied were Magnox Waste glass used for immobilisation of HLW in UK, and K-26 glass used in Russia for ILW immobilisation. Glass samples were irradiated under {gamma}-source (Co-60) up to doses 1 and 11 MGy. Normalised rates of elemental release and activation energy of release were measured for Na, Li, Ca, Mg, B, Si and Mo before and after irradiation. Irradiation up to 1 MGy results in increase of leaching rate of almost all elements from both MW and K-26 with the exception of Na release from MW glass. Further irradiation up to a dose of 11 MGy leads to the decrease of elemental release rates to nearly initial value. Another effect of irradiation is increase of activation energies of elemental release. (authors)

Pankov, Alexey S.; Ojovan, Michael I. [Immobilisation Science Laboratory, Department of Engineering Materials, University of Sheffield, Sir Robert Hadfield Building, Mappin Street, Sheffield, S1 3JD (United Kingdom); Batyukhnova, Olga G. [International Education Training Centre, SUE SIA 'Radon', The 7-th Rostovsky Lane 2/14, Moscow, 119121 (Russian Federation); Lee, William E. [Department of Materials, Imperial College London, South Kensington Campus, Exhibition Road, London, SW7 2AZ (United Kingdom)

2007-07-01T23:59:59.000Z

304

Evaluation of nuclear facility decommissioning projects: Summary status report: Three Mile Island Unit 2. Radioactive waste and laundry shipments  

SciTech Connect (OSTI)

This document summarizes information concerning radioactive waste and laundry shipments from the Three Mile Island Nuclear Station Unit 2 to radioactive waste disposal sites and to protective clothing decontamination facilities (laundries) since the loss of coolant accident experienced on March 28, 1979. Data were collected from radioactive shipment records, summarized, and placed in a computerized data information retrieval/manipulation system which permits extraction of specific information. This report covers the period of April 9, 1979 through April 19, 1987. Included in this report are: waste disposal site locations, dose rates, curie content, waste description, container type and number, volumes and weights. This information is presented in two major categories: protective clothing (laundry) and radioactive waste. Each of the waste shipment reports is in chronological order.

Doerge, D. H.; Haffner, D. R.

1988-06-01T23:59:59.000Z

305

Status of development of actinide blanket processing flowsheets for accelerator transmutation of nuclear waste  

SciTech Connect (OSTI)

An accelerator-driven subcritical nuclear system is briefly described that transmutes actinides and selected long-lived fission products. An application of this accelerator transmutation of nuclear waste (ATW) concept to spent fuel from a commercial nuclear power plant is presented as an example. The emphasis here is on a possible aqueous processing flowsheet to separate the actinides and selected long-lived fission products from the remaining fission products within the transmutation system. In the proposed system the actinides circulate through the thermal neutron flux as a slurry of oxide particles in heavy water in two loops with different average residence times: one loop for neptunium and plutonium and one for americium and curium. Material from the Np/Pu loop is processed with a short cooling time (5-10 days) because of the need to keep the total actinide inventory, low for this particular ATW application. The high radiation and thermal load from the irradiated material places severe constraints on the separation processes that can be used. The oxide particles are dissolved in nitric acid and a quarternary, ammonium anion exchanger is used to extract neptunium, plutonium, technetium, and palladium. After further cooling (about 90 days), the Am, Cm and higher actinides are extracted using a TALSPEAK-type process. The proposed operations were chosen because they have been successfully tested for processing high-level radioactive fuels or wastes in gram to kilogram quantities.

Dewey, H.J.; Jarvinen, G.D.; Marsh, S.F.; Schroeder, N.C.; Smith, B.F.; Villarreal, R.; Walker, R.B.; Yarbro, S.L.; Yates, M.A.

1993-09-01T23:59:59.000Z

306

Supercritical Fluid Extraction of Radionuclides - A Green Technology for Nuclear Waste Management  

SciTech Connect (OSTI)

Supercritical fluid carbon dioxide (SF-CO2) is capable of extracting radionuclides including cesium, strontium, uranium, plutonium and lanthanides directly from liquid and solid samples with proper complexing agents. Of particular interest is the ability of SF-CO2 to dissolve uranium dioxide directly using a CO2-soluble tri-nbutylphosphate- nitric acid (TBP-HNO3) extractant to form a highly soluble UO2(NO3)2(TBP)2 complex that can be transported and separated from Cs, Sr, and other transition metals. This method can also dissolve plutonium dioxide in SF-CO2. The SF-CO2 extraction technology offers several advantages over conventional solvent-based methods including ability to extract radionuclides directly from solids, easy separation of solutes from CO2, and minimization of liquid waste generation. Potential applications of the SF-CO2 extraction technology for nuclear waste treatment and for reprocessing of spent nuclear fuels will be discussed. Information on current demonstrations of the SF-CO2 technology by nuclear companies and research organizations in different countries will be reviewed.

Wai, Chien M.

2003-09-10T23:59:59.000Z

307

Nevada Nuclear Waste Storage Investigations: Exploratory Shaft Facility fluids and materials evaluation  

SciTech Connect (OSTI)

The objective of this study was to determine if any fluids or materials used in the Exploratory Shaft Facility (ESF) of Yucca Mountain will make the mountain unsuitable for future construction of a nuclear waste repository. Yucca Mountain, an area on and adjacent to the Nevada Test Site in southern Nevada, USA, is a candidate site for permanent disposal of high-level radioactive waste from commercial nuclear power and defense nuclear activities. To properly characterize Yucca Mountain, it will be necessary to construct an underground test facility, in which in situ site characterization tests can be conducted. The candidate repository horizon at Yucca Mountain, however, could potentially be compromised by fluids and materials used in the site characterization tests. To minimize this possibility, Los Alamos National Laboratory was directed to evaluate the kinds of fluids and materials that will be used and their potential impacts on the site. A secondary objective was to identify fluids and materials, if any, that should be prohibited from, or controlled in, the underground. 56 refs., 19 figs., 11 tabs.

West, K.A.

1988-11-01T23:59:59.000Z

308

Emptying of the Storage for Solid Radioactive Waste in the Greifswald Nuclear Power Plant  

SciTech Connect (OSTI)

On the Greifswald site, 8 WWER 440 reactor units are located and also several facilities to handle fuel and radwaste. After the reunification of Germany, the final decision was taken to decommission all these Russian designed reactors. Thus, EWN is faced with a major decommissioning project in the field of nuclear power stations. One of the major tasks before the dismantling of the plant is the complete disposal of the operational waste. Among other facilities, a store for solid radioactive waste is located on the site, which has been filled over 17 years of operation of units 1 to 4. The paper presents the disposal technology development and results achieved. This activity is the first project in the operational history of the Russian type serial reactor line WWER-440.

Hartmann, B.; Fischer, J.

2002-02-26T23:59:59.000Z

309

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

SciTech Connect (OSTI)

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

Troyer, G.L.

1997-03-17T23:59:59.000Z

310

Information basis for developing comprehensive waste management system-US-Japan joint nuclear energy action plan waste management working group phase I report.  

SciTech Connect (OSTI)

The activity of Phase I of the Waste Management Working Group under the United States - Japan Joint Nuclear Energy Action Plan started in 2007. The US-Japan JNEAP is a bilateral collaborative framework to support the global implementation of safe, secure, and sustainable, nuclear fuel cycles (referred to in this document as fuel cycles). The Waste Management Working Group was established by strong interest of both parties, which arise from the recognition that development and optimization of waste management and disposal system(s) are central issues of the present and future nuclear fuel cycles. This report summarizes the activity of the Waste Management Working Group that focused on consolidation of the existing technical basis between the U.S. and Japan and the joint development of a plan for future collaborative activities. Firstly, the political/regulatory frameworks related to nuclear fuel cycles in both countries were reviewed. The various advanced fuel cycle scenarios that have been considered in both countries were then surveyed and summarized. The working group established the working reference scenario for the future cooperative activity that corresponds to a fuel cycle scenario being considered both in Japan and the U.S. This working scenario involves transitioning from a once-through fuel cycle utilizing light water reactors to a one-pass uranium-plutonium fuel recycle in light water reactors to a combination of light water reactors and fast reactors with plutonium, uranium, and minor actinide recycle, ultimately concluding with multiple recycle passes primarily using fast reactors. Considering the scenario, current and future expected waste streams, treatment and inventory were discussed, and the relevant information was summarized. Second, the waste management/disposal system optimization was discussed. Repository system concepts were reviewed, repository design concepts for the various classifications of nuclear waste were summarized, and the factors to consider in repository design and optimization were then discussed. Japan is considering various alternatives and options for the geologic disposal facility and the framework for future analysis of repository concepts was discussed. Regarding the advanced waste and storage form development, waste form technologies developed in both countries were surveyed and compared. Potential collaboration areas and activities were next identified. Disposal system optimization processes and techniques were reviewed, and factors to consider in future repository design optimization activities were also discussed. Then the potential collaboration areas and activities related to the optimization problem were extracted.

Nutt, M.; Nuclear Engineering Division

2010-05-25T23:59:59.000Z

311

Recommendation by the Secretary of Energy Regarding the Suitability of the Yucca Mountain Site for a Repository Under the Nuclear Waste Policy Act of 1982  

Broader source: Energy.gov [DOE]

Recommendation by the Secretary of Energy Regarding the Suitability of the Yucca Mountain Site for a Repository Under the Nuclear Waste Policy Act of 1982

312

Integrated Data Base report--1993: U.S. spent nuclear fuel and radioactive waste inventories, projections, and characteristics. Revision 10  

SciTech Connect (OSTI)

The Integrated Data Base Program has compiled historic data on inventories and characteristics of both commercial and DOE spent nuclear fuel; also, commercial and US government-owned radioactive wastes through December 31, 1993. These data are based on the most reliable information available from government sources, the open literature, technical reports, and direct contacts. The information forecasted is consistent with the latest US Department of Energy/Energy Information Administration projections of US commercial nuclear power growth and the expected DOE-related and private industrial and institutional activities. The radioactive materials considered, on a chapter-by-chapter basis, are spent nuclear fuel, high-level waste, transuranic waste, low-level waste, commercial uranium mill tailings, DOE Environmental Restoration Program wastes, commercial reactor and fuel-cycle facility decommissioning wastes, and mixed (hazardous and radioactive) low-level waste. For most of these categories, current and projected inventories are given the calendar-year 2030, and the radioactivity and thermal power are calculated based on reported or estimated isotopic compositions. In addition, characteristics and current inventories are reported for miscellaneous radioactive materials that may require geologic disposal. 256 refs., 38 figs., 141 tabs.

Not Available

1994-12-01T23:59:59.000Z

313

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Waste Integrated Performance and Safety Codes (IPSC) : FY10 development and integration.  

SciTech Connect (OSTI)

This report describes the progress in fiscal year 2010 in developing the Waste Integrated Performance and Safety Codes (IPSC) in support of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The goal of the Waste IPSC is to develop an integrated suite of computational modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. Waste IPSC activities in fiscal year 2010 focused on specifying a challenge problem to demonstrate proof of concept, developing a verification and validation plan, and performing an initial gap analyses to identify candidate codes and tools to support the development and integration of the Waste IPSC. The current Waste IPSC strategy is to acquire and integrate the necessary Waste IPSC capabilities wherever feasible, and develop only those capabilities that cannot be acquired or suitably integrated, verified, or validated. This year-end progress report documents the FY10 status of acquisition, development, and integration of thermal-hydrologic-chemical-mechanical (THCM) code capabilities, frameworks, and enabling tools and infrastructure.

Criscenti, Louise Jacqueline; Sassani, David Carl; Arguello, Jose Guadalupe, Jr.; Dewers, Thomas A.; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Wang, Yifeng; Schultz, Peter Andrew

2011-02-01T23:59:59.000Z

314

Nuclear waste processing based on FOOF and KrF sub 2  

SciTech Connect (OSTI)

A class of oxygen fluoride compounds and krypton difluoride show great promise in recovering and purifying plutonium and other actinides from nuclear waste and residues. Since 1983, a significant effort has been expended in three main areas of research; one area is the characterization of the reactive species and the fluorination reaction chemistry with various actinides. The second area is to develop efficient production methods for the active chemicals. The third area is actually in demonstrating application of these fluorinating agents to actinides recovery and purification. Substantial progress has been made in all three areas and some of the highlights in their research effort is discussed. 11 refs., 2 figs., 2 tabs.

Kim, K.C.; Blum, T.W.

1991-01-01T23:59:59.000Z

315

Significance of the microstructure of Pacific red clays to nuclear waste disposal  

E-Print Network [OSTI]

. Tieh (member) Richard H. Bennett (member) Rober t O. Reid (Head oi' Department) Richard Rezak (member) August 1987 ABSTRACT The Significance of' the Micr ostr uctur e of Pacif'ic Red Clays to Nuclear Waste Disposal. (August 1987) Patti Jo..., 1966) B and C:"Stepped (FF) face-to -face" (Smalley and Cabrera, 1969). V' Chain of stepped face-to- face and (EE) edge-to-edge particles (O' Brien, 1971). "Turbostratic" structure (Aylmore and Quirked 1960). /7 Figure 2, Particle r le ar...

Burkett, Patti Jo

1987-01-01T23:59:59.000Z

316

A Characteristics-Based Approach to Radioactive Waste Classification in Advanced Nuclear Fuel Cycles  

E-Print Network [OSTI]

Nuclear Fuel, Nuclear Engineering and Technology, in Engineering -? Nuclear Engineering and the in Engineering -? Nuclear Engineering and the

Djokic, Denia

2013-01-01T23:59:59.000Z

317

Progress toward bridging from atomistic to continuum modeling to predict nuclear waste glass dissolution.  

SciTech Connect (OSTI)

This report summarizes research performed for the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Subcontinuum and Upscaling Task. The work conducted focused on developing a roadmap to include molecular scale, mechanistic information in continuum-scale models of nuclear waste glass dissolution. This information is derived from molecular-scale modeling efforts that are validated through comparison with experimental data. In addition to developing a master plan to incorporate a subcontinuum mechanistic understanding of glass dissolution into continuum models, methods were developed to generate constitutive dissolution rate expressions from quantum calculations, force field models were selected to generate multicomponent glass structures and gel layers, classical molecular modeling was used to study diffusion through nanopores analogous to those in the interfacial gel layer, and a micro-continuum model (K{mu}C) was developed to study coupled diffusion and reaction at the glass-gel-solution interface.

Zapol, Peter (Argonne National Laboratory, Argonne, IL); Bourg, Ian (Lawrence Berkeley National Laboratories, Berkeley, CA); Criscenti, Louise Jacqueline; Steefel, Carl I. (Lawrence Berkeley National Laboratories, Berkeley, CA); Schultz, Peter Andrew

2011-10-01T23:59:59.000Z

318

The National Nuclear Laboratory's Approach to Processing Mixed Wastes and Residues - 13080  

SciTech Connect (OSTI)

The National Nuclear Laboratory (NNL) treats a wide variety of materials produced as by-products of the nuclear fuel cycle, mostly from uranium purification and fuel manufacture but also including materials from uranium enrichment and from the decommissioning of obsolete plants. In the context of this paper, treatment is defined as recovery of uranium or other activity from residues, the recycle of uranium to the fuel cycle or preparation for long term storage and the final disposal or discharge to the environment of the remainder of the material. NNL's systematic but flexible approach to residue assessment and treatment is described in this paper. The approach typically comprises up to five main phases. The benefits of a systematic approach to waste and residue assessments and processing are described in this paper with examples used to illustrate each phase of work. Benefits include early identification of processing routes or processing issues and the avoidance of investment in inappropriate and costly plant or processes. (authors)

Greenwood, Howard; Docrat, Tahera; Allinson, Sarah J.; Coppersthwaite, Duncan P.; Sultan, Ruqayyah; May, Sarah [National Nuclear Laboratory, Springfields, Preston, UK, PR4 0XJ (United Kingdom)] [National Nuclear Laboratory, Springfields, Preston, UK, PR4 0XJ (United Kingdom)

2013-07-01T23:59:59.000Z

319

International Nuclear Energy Research Initiative Development of Computational Models for Pyrochemical Electrorefiners of Nuclear Waste Transmutation Systems  

SciTech Connect (OSTI)

In support of closing the nuclear fuel cycle using non-aqueous separations technology, this project aims to develop computational models of electrorefiners based on fundamental chemical and physical processes. Spent driver fuel from Experimental Breeder Reactor-II (EBR-II) is currently being electrorefined in the Fuel Conditioning Facility (FCF) at Idaho National Laboratory (INL). And Korea Atomic Energy Research Institute (KAERI) is developing electrorefining technology for future application to spent fuel treatment and management in the Republic of Korea (ROK). Electrorefining is a critical component of pyroprocessing, a non-aqueous chemical process which separates spent fuel into four streams: (1) uranium metal, (2) U/TRU metal, (3) metallic high-level waste containing cladding hulls and noble metal fission products, and (4) ceramic high-level waste containing sodium and active metal fission products. Having rigorous yet flexible electrorefiner models will facilitate process optimization and assist in trouble-shooting as necessary. To attain such models, INL/UI has focused on approaches to develop a computationally-light and portable two-dimensional (2D) model, while KAERI/SNU has investigated approaches to develop a computationally intensive three-dimensional (3D) model for detailed and fine-tuned simulation.

M.F. Simpson; K.-R. Kim

2010-12-01T23:59:59.000Z

320

Selected, annotated bibliography of studies relevant to the isolation of nuclear wastes. [705 references  

SciTech Connect (OSTI)

This annotated bibliography of 705 references represents the first in a series to be published by the Ecological Sciences Information Center containing scientific, technical, economic, and regulatory information relevant to nuclear waste isolation. Most references discuss deep geologic disposal, with fewer studies of deep seabed disposal; space disposal is also included. The publication covers both domestic and foreign literature for the period 1954 to 1980. Major chapters selected are Chemical and Physical Aspects; Container Design and Performance; Disposal Site; Envirnmental Transport; General Studies and Reviews; Geology, Hydrology and Site Resources; Regulatory and Economic Aspects; Repository Design and Engineering; Transportation Technology; Waste Production; and Waste Treatment. Specialized data fields have been incorporated to improve the ease and accuracy of locating pertinent references. Specific radionuclides for which data are presented are listed in the Measured Radionuclides field, and specific parameters which affect the migration of these radionuclides are presented in the Measured Parameters field. The references within each chapter are arranged alphabetically by leading author, corporate affiliation, or title of the document. When the author is not given, the corporate affiliation appears first. If these two levels of authorship are not given, the title of the document is used as the identifying level. Indexes are provided for author(s), keywords, subject category, title, geographic location, measured parameters, measured radionuclides, and publication description.

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

1980-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

Summary of four release consequence analyses for hypothetical nuclear waste repositories in salt and granite  

SciTech Connect (OSTI)

Release consequence methology developed under the Assessment of Effectiveness of Geologic Isolation Systems (AEGIS) program has now been applied to four hypothetical repository sites. This paper summarizes the results of these four studies in order to demonstrate that the far-field methodology developed under the AEGIS program offers a practical approach to the post-closure safety assessment of nuclear waste repositories sited in deep continental geologic formations. The four studies are briefly described and compared according to the following general categories: physical description of the repository (size, inventory, emplacement depth); geologic and hydrologic description of the site and the conceptual hydrologic model for the site; description of release scenario; hydrologic model implementation and results; engineered barriers and leach rate modeling; transport model implementation and results; and dose model implementation and results. These studies indicate the following: numerical modeling is a practical approach to post-closure safety assessment analysis for nuclear waste repositories; near-field modeling capability needs improvement to permit assessment of the consequences of human intrusion and pumping well scenarios; engineered barrier systems can be useful in mitigating consequences for postulated release scenarios that short-circuit the geohydrologic system; geohydrologic systems separating a repository from the natural biosphere discharge sites act to mitigate the consequences of postulated breaches in containment; and engineered barriers of types other than the containment or absorptive type may be useful.

Cole, C.R.; Bond, F.W.

1980-12-01T23:59:59.000Z

322

Geologic and hydrologic investigations of a potential nuclear waste disposal site at Yucca Mountain, southern Nevada  

SciTech Connect (OSTI)

Yucca Mountain in southern Nye County, Nevada, has been selected by the United States Department of Energy as one of three potential sites for the nation`s first high-level nuclear waste repository. Its deep water table, closed-basin ground-water flow, potentially favorable host rock, and sparse population have made the Yucca Mountain area a viable candidate during the search for a nuclear waste disposal site. Yucca Mountain, however, lies within the southern Great Basin, a region of known contemporary tectonism and young volcanic activity, and the characterization of tectonism and volcanism remains as a fundamental problem for the Yucca Mountain site. The United States Geological Survey has been conducting extensive studies to evaluate the geologic setting of Yucca Mountain, as well as the timing and rates of tectonic and volcanic activity in the region. A workshop was convened by the Geologic Survey in Denver, Colorado, on August 19, 20, and 21, 1985, to review the scientific progress and direction of these studies. Considerable debate resulted. This collection of papers represents the results of some of the studies presented at the workshop, but by no means covers all of the scientific results and viewpoints presented. Rather, the volume is meant to serve as a progress report on some of the studies within the Geological Survey`s continuing research program toward characterizing the tectonic framework of Yucca Mountain. Individual papers were processed separately for the data base.

Carr, M.D.; Yount, J.C. (eds.)

1988-12-31T23:59:59.000Z

323

Commercial nuclear power 1990  

SciTech Connect (OSTI)

This report presents the status at the end of 1989 and the outlook for commercial nuclear capacity and generation for all countries in the world with free market economies (FME). The report provides documentation of the US nuclear capacity and generation projections through 2030. The long-term projections of US nuclear capacity and generation are provided to the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM) for use in estimating nuclear waste fund revenues and to aid in planning the disposal of nuclear waste. These projections also support the Energy Information Administration's annual report, Domestic Uranium Mining and Milling Industry: Viability Assessment, and are provided to the Organization for Economic Cooperation and Development. The foreign nuclear capacity projections are used by the DOE uranium enrichment program in assessing potential markets for future enrichment contracts. The two major sections of this report discuss US and foreign commercial nuclear power. The US section (Chapters 2 and 3) deals with (1) the status of nuclear power as of the end of 1989; (2) projections of nuclear capacity and generation at 5-year intervals from 1990 through 2030; and (3) a discussion of institutional and technical issues that affect nuclear power. The nuclear capacity projections are discussed in terms of two projection periods: the intermediate term through 2010 and the long term through 2030. A No New Orders case is presented for each of the projection periods, as well as Lower Reference and Upper Reference cases. 5 figs., 30 tabs.

Not Available

1990-09-28T23:59:59.000Z

324

Preliminary technique assessment for nondestructive evaluation certification of the NNWSI [Nevada Nuclear Waste Storage Investigations] disposal container closure  

SciTech Connect (OSTI)

Under the direction of the Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM) program, the Nevada Nuclear Waste Storage Investigations (NNWSI) project is evaluating a candidate repository site at Yucca Mountain, Nevada, for permanent disposal of high-level nuclear waste. The Lawrence Livermore National Laboratory (LLNL), a participant in the NNWSI project, is developing waste package designs to meet the NRC requirements. One aspect of this waste package is the nondestructive testing of the final closure of the waste container. The container closure weld can best be nondestructively examined (NDE) by a combination of ultrasonics and liquid penetrants. This combination can be applied remotely and can meet stringent quality control requirements common to nuclear applications. Further development in remote systems and inspection will be required to meet anticipated requirements for flaw detection reliability and sensitivity. New research is not required but might reduce cost or inspection time. Ultrasonic and liquid penetrant methods can examine all closure methods currently being considered, which include fusion welding and inertial welding, among others. These NDE methods also have a history of application in high radiation environments and a well developed technology base for remote operation that can be used to reduce development and design costs. 43 refs., 23 figs., 3 tabs.

Day, R.A.

1988-12-31T23:59:59.000Z

325

Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario Hydro`s Bruce Nuclear Generating Station `A`  

SciTech Connect (OSTI)

Ontario Hydro at the Bruce Nuclear Generating Station `A` has undertaken a program to render the station`s liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants to satisfy regulatory requirements for emissions. The system will remove radionuclide and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology.

Day, J.E.; Baker, R.L.

1995-05-01T23:59:59.000Z

326

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

SciTech Connect (OSTI)

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

Hasan, M.A.; Selim, Y.T.; Lasheen, Y.F. [Hot Labs and Waste Management Center, Atomic Energy Authority, 3 Ahmed El-Zomor St., El-Zohour District, Naser City, 11787, Cairo (Egypt)] [Hot Labs and Waste Management Center, Atomic Energy Authority, 3 Ahmed El-Zomor St., El-Zohour District, Naser City, 11787, Cairo (Egypt)

2013-07-01T23:59:59.000Z

327

NWTS program criteria for mined geologic disposal of nuclear waste: repository performance and development criteria. Public draft  

SciTech Connect (OSTI)

This document, DOE/NWTS-33(3) is one of a series of documents to establish the National Waste Terminal Storage (NWTS) program criteria for mined geologic disposal of high-level radioactive waste. For both repository performance and repository development it delineates the criteria for design performance, radiological safety, mining safety, long-term containment and isolation, operations, and decommissioning. The US Department of Energy will use these criteria to guide the development of repositories to assist in achieving performance and will reevaluate their use when the US Nuclear Regulatory Commission issues radioactive waste repository rules.

none,

1982-07-01T23:59:59.000Z

328

Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain  

SciTech Connect (OSTI)

Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices.

Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr. [Lawrence Livermore National Lab., CA (United States)] [Lawrence Livermore National Lab., CA (United States); Gdowski, G.E. [KMI, Inc., Albuquerque, NM (United States)] [KMI, Inc., Albuquerque, NM (United States)

1993-02-01T23:59:59.000Z

329

Site-specific EIS ordered but injunctive relief deined in nuclear waste storage case  

SciTech Connect (OSTI)

The Energy Research and Development Administration (ERDA) received appropriations in 1976-77 to construct 22 tanks for storage of high level radioactive wastes generated by its nuclear weapons materials production program. The tanks were to replace older, leaking tanks at the Hanford Reservation in Richland, Washington and the Savannah River Plant in Aiken, South Carolina. The Natural Resources Defense Council (NRDC) had unsuccessfully requested that ERDA obtain a construction permit from the Nuclear Regulatory Commission (NRC). NRDC also petitioned NRC to exercise its licensing authority over the tanks under Section 202(4) of the Energy Reorganization Act of 1974. In response to the NRDC request, ERDA claimed the tanks were only for short-term storage and therefore a license was unnecessary. NRC claimed it lacked jurisdiction over the tanks. NRDC filed suit in United States District Court for the District of Columbia, alleging that ERDA had violated Section 102(2)(C) of the National Environmental Policy Act, and that both ERDA and NRC had violated Section 202(4) of the Energy Reorganization Act. NRDC requested an injunction against further construction of the tanks. Although ERDA did not have to obtain an NRC construction permit for the nuclear waste storage tanks at Hanford Reservation and Savannah River Plant, the programmatic Environmental Impact Statement submitted was insufficient and site-specific statements must be prepared. Injunctive relief pending the statements was denied for the social and economic costs of delaying the tanks project. NRC decisions even remotely connected to its licensing power should be contested in federal courts of appeals, not district courts. The court gave NRDC a hollow victory by ordering a more specific EIS, but denying an injunction.

Barnhart y Chavez, S.

1980-01-01T23:59:59.000Z

330

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 (OSTI)

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

1992-05-01T23:59:59.000Z

331

Money Related Decommissioning and Funding Decision Making  

SciTech Connect (OSTI)

'Money makes the world go round', as the song says. It definitely influences decommissioning decision-making and financial assurance for future decommissioning. This paper will address two money-related decommissioning topics. The first is the evaluation of whether to continue or to halt decommissioning activities at Fermi 1. The second is maintaining adequacy of financial assurance for future decommissioning of operating plants. Decommissioning costs considerable money and costs are often higher than originally estimated. If costs increase significantly and decommissioning is not well funded, decommissioning activities may be deferred. Several decommissioning projects have been deferred when decision-makers determined future spending is preferable than current spending, or when costs have risen significantly. Decommissioning activity timing is being reevaluated for the Fermi 1 project. Assumptions for waste cost-escalation significantly impact the decision being made this year on the Fermi 1 decommissioning project. They also have a major impact on the estimated costs for decommissioning currently operating plants. Adequately funding full decommissioning during plant operation will ensure that the users who receive the benefit pay the full price of the nuclear-generated electricity. Funding throughout operation also will better ensure that money is available following shutdown to allow decommissioning to be conducted without need for additional funds.

Goodman, Lynne S. [Detroit Edison Company, 6400 N. Dixie Highway, Newport, Michigan 48162 (United States)

2008-01-15T23:59:59.000Z

332

Development of high-waste loaded high-level nuclear waste glasses for high-temperature melter  

SciTech Connect (OSTI)

This paper describes the approach taken in formulating glasses that can be processed at 1150 to 1500{degrees}C by applying glass property/composition models developed at Pacific Northwest Laboratory. Compositions and melting temperatures for glasses with high waste loading that are acceptable and able to be processed were determined for two different Hanford waste types. The glasses meet high-level waste glass acceptability criteria and are suitable for processing in a continuous Joule-heated melter.

Kim, D.S.; Hrma, P.; Lamar, D.A.; Elliott, M.L. [Pacific Northwest Lab., Richland, WA (United States)

1994-12-31T23:59:59.000Z

333

Development of high-waste loaded high-level nuclear waste glasses for high-temperature melter  

SciTech Connect (OSTI)

This paper describes the approach taken in formulating glasses that can be processed at 1150 to 1500{degrees}C by applying glass property/composition models developed at Pacific Northwest Laboratory. Compositions and melting temperatures for glasses with high waste loading that are acceptable and able to be processed were determined for two different Hanford waste types. The glasses meet high-level waste glass acceptability criteria and are suitable for processing in a continuous Joule-heated melter.

Kim, D.S.; Hrma, P.R.; Lamar, D.A.; Elliott, M.L.

1994-04-01T23:59:59.000Z

334

Challenge problem and milestones for : Nuclear Energy Advanced Modeling and Simulation (NEAMS) waste Integrated Performance and Safety Codes (IPSC).  

SciTech Connect (OSTI)

This report describes the specification of a challenge problem and associated challenge milestones for the Waste Integrated Performance and Safety Codes (IPSC) supporting the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign. The NEAMS challenge problems are designed to demonstrate proof of concept and progress towards IPSC goals. The goal of the Waste IPSC is to develop an integrated suite of modeling and simulation capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive waste storage or disposal system. The Waste IPSC will provide this simulation capability (1) for a range of disposal concepts, waste form types, engineered repository designs, and geologic settings, (2) for a range of time scales and distances, (3) with appropriate consideration of the inherent uncertainties, and (4) in accordance with robust verification, validation, and software quality requirements. To demonstrate proof of concept and progress towards these goals and requirements, a Waste IPSC challenge problem is specified that includes coupled thermal-hydrologic-chemical-mechanical (THCM) processes that describe (1) the degradation of a borosilicate glass waste form and the corresponding mobilization of radionuclides (i.e., the processes that produce the radionuclide source term), (2) the associated near-field physical and chemical environment for waste emplacement within a salt formation, and (3) radionuclide transport in the near field (i.e., through the engineered components - waste form, waste package, and backfill - and the immediately adjacent salt). The initial details of a set of challenge milestones that collectively comprise the full challenge problem are also specified.

Freeze, Geoffrey A.; Wang, Yifeng; Howard, Robert; McNeish, Jerry A.; Schultz, Peter Andrew; Arguello, Jose Guadalupe, Jr.

2010-09-01T23:59:59.000Z

335

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

SciTech Connect (OSTI)

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

336

Overview on backfill materials and permeable reactive barriers for nuclear waste disposal facilities.  

SciTech Connect (OSTI)

A great deal of money and effort has been spent on environmental restoration during the past several decades. Significant progress has been made on improving air quality, cleaning up and preventing leaching from dumps and landfills, and improving surface water quality. However, significant challenges still exist in all of these areas. Among the more difficult and expensive environmental problems, and often the primary factor limiting closure of contaminated sites following surface restoration, is contamination of ground water. The most common technology used for remediating ground water is surface treatment where the water is pumped to the surface, treated and pumped back into the ground or released at a nearby river or lake. Although still useful for certain remediation scenarios, the limitations of pump-and-treat technologies have recently been recognized, along with the need for innovative solutions to ground-water contamination. Even with the current challenges we face there is a strong need to create geological repository systems for dispose of radioactive wastes containing long-lived radionuclides. The potential contamination of groundwater is a major factor in selection of a radioactive waste disposal site, design of the facility, future scenarios such as human intrusion into the repository and possible need for retrieving the radioactive material, and the use of backfills designed to keep the radionuclides immobile. One of the most promising technologies for remediation of contaminated sites and design of radioactive waste repositories is the use of permeable reactive barriers (PRBs). PRBs are constructed of reactive material(s) to intercept and remove the radionuclides from the water and decontaminate the plumes in situ. The concept of PRBs is relatively simple. The reactive material(s) is placed in the subsurface between the waste or contaminated area and the groundwater. Reactive materials used thus far in practice and research include zero valent iron, hydroxyapatite, magnesium oxide, and others. As the contaminant moves through the reactive material, the contaminant is either sorbed by the reactive material or chemically reacts with the material to form a less harmful substance. Because of the high risk associated with failure of a geological repository for nuclear waste, most nations favor a near-field multibarrier engineered system using backfill materials to prevent release of radionuclides into the surrounding groundwater.

Moore, Robert Charles; Hasan, Ahmed Ali Mohamed; Holt, Kathleen Caroline; Hasan, Mahmoud A. (Egyptian Atomic Energy Authority, Cairo, Egypt)

2003-10-01T23:59:59.000Z

337

Experimental Methods to Estimate Accumulated Solids in Nuclear Waste Tanks - 13313  

SciTech Connect (OSTI)

The Department of Energy has a large number of nuclear waste tanks. It is important to know if fissionable materials can concentrate when waste is transferred from staging tanks prior to feeding waste treatment plants. Specifically, there is a concern that large, dense particles, e.g., plutonium containing, could accumulate in poorly mixed regions of a blend tank heel for tanks that employ mixing jet pumps. At the request of the DOE Hanford Tank Operations Contractor, Washington River Protection Solutions, the Engineering Development Laboratory of the Savannah River National Laboratory performed a scouting study in a 1/22-scale model of a waste tank to investigate this concern and to develop measurement techniques that could be applied in a more extensive study at a larger scale. Simulated waste tank solids and supernatant were charged to the test tank and rotating liquid jets were used to remove most of the solids. Then the volume and shape of the residual solids and the spatial concentration profiles for the surrogate for plutonium were measured. This paper discusses the overall test results, which indicated heavy solids only accumulate during the first few transfer cycles, along with the techniques and equipment designed and employed in the test. Those techniques include: - Magnetic particle separator to remove stainless steel solids, the plutonium surrogate from a flowing stream. - Magnetic wand used to manually remove stainless steel solids from samples and the tank heel. - Photographs were used to determine the volume and shape of the solids mounds by developing a composite of topographical areas. - Laser range finders to determine the volume and shape of the solids mounds. - Core sampler to determine the stainless steel solids distribution within the solids mounds. - Computer driven positioner that placed the laser range finders and the core sampler over solids mounds that accumulated on the bottom of a scaled staging tank in locations where jet velocities were low. These devices and techniques were very effective to estimate the movement, location, and concentrations of the solids representing plutonium and are expected to perform well at a larger scale. The operation of the techniques and their measurement accuracies will be discussed as well as the overall results of the accumulated solids test. (authors)

Duignan, Mark R.; Steeper, Timothy J.; Steimke, John L. [Savannah River Nuclear Solutions, Savannah River National Laboratory, Aiken, SC 29808 (United States)] [Savannah River Nuclear Solutions, Savannah River National Laboratory, Aiken, SC 29808 (United States)

2013-07-01T23:59:59.000Z

338

Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act  

SciTech Connect (OSTI)

Chapter six describes the basis for facility design, the completed facility conceptual design, the completed analytical work relating to the resolution of design issues, and future design-related work. The basis for design and the conceptual design information presented in this chapter meet the requirements of the Nuclear Waste Policy Act of 1982, for a conceptual repository design that takes into account site-specific requirements. This information is presented to permit a critical evaluation of planned site characterization activities. Chapter seven describes waste package components, emplacement environment, design, and status of research and development that support the Nevada Nuclear Waste Storage Investigation (NNWSI) Project. The site characterization plan (SCP) discussion of waste package components is contained entirely within this chapter. The discussion of emplacement environment in this chapter is limited to considerations of the environment that influence, or which may influence, if perturbed, the waste packages and their performance (particularly hydrogeology, geochemistry, and borehole stability). The basis for conceptual waste package design as well as a description of the design is included in this chapter. The complete design will be reported in the advanced conceptual design (ACD) report and is not duplicated in the SCP. 367 refs., 173 figs., 68 tabs.

NONE

1988-01-01T23:59:59.000Z

339

Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source  

E-Print Network [OSTI]

1 Sodium Cooled Fast Reactors and the Pyro-Process: Conversion of Nuclear Waste into a Fuel Source. Belanger Chair, Department of Physics #12;2 Abstract A review of the sodium cooled fast reactor........................................................................................23 1.3.5 Reactor Startup

Belanger, David P.

340

Assessment of alternatives for management of ORNL retrievable transuranic waste. Nuclear Waste Program: transuranic waste (Activity No. AR 05 15 15 0; ONL-WT04)  

SciTech Connect (OSTI)

Since 1970, solid waste with TRU or U-233 contamination in excess of 10 ..mu..Ci per kilogram of waste has been stored in a retrievable fashion at ORNL, such as in ss drums, concrete casks, and ss-lined wells. This report describes the results of a study performed to identify and evaluate alternatives for management of this waste and of the additional waste projected to be stored through 1995. The study was limited to consideration of the following basic strategies: Strategy 1: Leave waste in place as is; Strategy 2: Improve waste confinement; and Strategy 3: Retrieve waste and process for shipment to a Federal repository. Seven alternatives were identified and evaluated, one each for Strategies 1 and 2 and five for Strategy 3. Each alternative was evaluated from the standpoint of technical feasibility, cost, radiological risk and impact, regulatory factors and nonradiological environmental impact.

Not Available

1980-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" from the National Library of EnergyBeta (NLEBeta).
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341

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

SciTech Connect (OSTI)

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

342

Rhenium Solubility in Borosilicate Nuclear Waste Glass: Implications for the Processing and Immobilization of Technetium-99  

SciTech Connect (OSTI)

The immobilization of 99Tc in a suitable host matrix has proved to be an arduous task for the researchers in nuclear waste community around the world. At the Hanford site in Washington State, the total amount of 99Tc in low-activity waste (LAW) is ~1300 kg and the current strategy is to immobilize the 99Tc in borosilicate glass with vitrification. In this context, the present article reports on the solubility/retention of rhenium, a nonradioactive surrogate for 99Tc, in a LAW borosilicate glass. Due to the radioactive nature of technetium, rhenium was chosen as a simulant because of the similarity between their ionic radii and other chemical aspects. The glasses containing Re (0 10,000 ppm by mass) were synthesized in vacuum-sealed quartz ampoules in order to minimize the loss of Re by volatilization during melting at 1000 C. The rhenium was found to predominantly exist as Re (VII) in all the glasses as observed by X-ray absorption near-edge structure (XANES). The solubility of Re in borosilicate glasses was determined to be ~3000 ppm (by mass) with inductively coupled plasma-optical emission spectroscopy (ICP-OES). At higher rhenium concentrations, some additional material was retained in the glasses in the form of crystalline inclusions that were detected by X-ray diffraction (XRD) and laser ablation-ICP mass spectrometry (LA-ICP-MS). The implications of these results on the immobilization of 99Tc from radioactive wastes in borosilicate glasses have been discussed.

McCloy, John S.; Riley, Brian J.; Goel, Ashutosh; Liezers, Martin; Schweiger, Michael J.; Rodriguez, Carmen P.; Hrma, Pavel R.; Kim, Dong-Sang; Lukens, Wayne W.; Kruger, Albert A.

2012-10-26T23:59:59.000Z

343

Toward Understanding the Effect of Nuclear Waste Glass Composition of Sulfur Solubility  

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

The concentration of sulfur in nuclear waste glass melter feed must be maintained below the point where salt accumulates on the melt surface. The allowable concentrations may range from 0.37 to over 2.05 weight percent (of SO3 on a calcined oxide basis) depending on the composition of the melter feed and processing conditions. If the amount of sulfur exceeds the melt tolerance level, a molten salt will accumulate, which may upset melter operations and potentially shorten the useful life of the melter. At the Hanford site, relatively conservative limits have been placed on sulfur loading in melter feed, which in turn significantly increases the amount of glass that will be produced. Crucible-scale sulfur solubility data and scaled melter sulfur tolerance data have been collected on simulated Hanford waste glasses over the last 15 years. These data were compiled and analyzed. A model was developed to predict the solubility of SO3 in glass based on 252 simulated Hanford low-activity waste (LAW) glass compositions. This model represents the data well, accounting for over 85% of the variation in data, and was well validated. The model was also found to accurately predict the tolerance for sulfur in melter feed for 13 scaled melter tests of simulated LAW glasses. The model can be used to help estimate glass volumes and make informed decisions on process options. The model also gives quantitative estimates of component concentration effects on sulfur solubility. The components that most increase sulfur solubility are Li2O > V2O5> CaO ? P2O5 > Na2O ? B2O3 > K2O. The components that most decrease sulfur solubility are Cl > Cr2O3 > Al2O3 > ZrO2 ? SnO2 > Others ? SiO2. The order of component effects is similar to previous literature data, in most cases.

Vienna, John D.; Kim, Dong-Sang; Muller, I. S.; Kruger, Albert A.; Piepel, Gregory F.

2014-02-13T23:59:59.000Z

344

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

SciTech Connect (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

345

Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste repository--BMT1 of the DECOVALEX III project. Part 1: Conceptualization  

E-Print Network [OSTI]

Numerical study of the THM effects on the near-field safety of a hypothetical nuclear waste on the safety of a hypothetical nuclear waste repository at the near-field and are presented in three on the safety of nuclear waste repositories. To achieve the second objective, hypothetical benchmark test

Paris-Sud XI, Université de

346

INTERNATIONAL STUDIES OF ENHANCED WASTE LOADING AND IMPROVED MELT RATE FOR HIGH ALUMINA CONCENTRATION NUCLEAR WASTE GLASSES  

SciTech Connect (OSTI)

The goal of this study was to determine the impacts of glass compositions with high aluminum concentrations on melter performance, crystallization and chemical durability for Savannah River Site (SRS) and Hanford waste streams. Glass compositions for Hanford targeted both high aluminum concentrations in waste sludge and a high waste loading in the glass. Compositions for SRS targeted Sludge Batch 5, the next sludge batch to be processed in the Defense Waste Processing Facility (DWPF), which also has a relatively high aluminum concentration. Three frits were selected for combination with the SRS waste to evaluate their impact on melt rate. The glasses were melted in two small-scale test melters at the V. G. Khlopin Radium Institute. The results showed varying degrees of spinel formation in each of the glasses. Some improvements in melt rate were made by tailoring the frit composition for the SRS feeds. All of the Hanford and SRS compositions had acceptable chemical durability.

Fox, K; David Peeler, D; James Marra, J

2008-09-11T23:59:59.000Z

347

TEMP: A finite line heat transfer code for geologic repositories for nuclear waste  

SciTech Connect (OSTI)

TEMP is a FORTRAN computer code for calculating temperatures in a geologic repository for nuclear waste. It will calculate the incremental temperature contributed by a single heat source, by an infinite array of heat sources, or by heat sources geometrically arranged in a finite array. In the finite array geometry, different types of heat sources can be placed in different regions at different times to more closely approximate the emplacement of waste in a repository. TEMP uses a semi-analytical technique for solving the equation for a heat producing finite length line source in an infinite and isotropic medium. Temperature contributions from individual heat sources are superimposed to determine the temperature at a specific location and time in a repository of multiple heat sources. Thermal conductivity of the geologic medium can be a function of temperature, and, when it is, an approximation is made for the temperature dependence of thermal diffusivity. This report derives the equations solved by TEMP and documents its accuracy by comparing its results to known analytical solutions and to the finite-difference and finite-element heat transfer codes HEATING5, HEATING6, THAC-SIP-3D, SPECTROM-41, and STEALTH-2D. The temperature results from TEMP are shown to be very accurate when compared to the analytical solutions and to the results from the finite-difference and finite-element codes. 8 refs., 97 figs., 39 tabs.

Wurm, K.J.; Bloom, S.G.; Atterbury, W.G.; Hetteberg, J.R.

1987-10-01T23:59:59.000Z

348

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

SciTech Connect (OSTI)

Following completion of the preliminary risk assessment of the potential Yucca Mountain Site by Pacific Northwest Laboratory (PNL) in 1988, the Office of Civilian Radioactive Waste Management (OCRWM) of the US Department of Energy (DOE) requested the Performance Assessment Scientific Support (PASS) Program at PNL to develop an integrated system model and computer code that provides performance and risk assessment analysis capabilities for a potential high-level nuclear waste repository. The system model that has been developed addresses the cumulative radionuclide release criteria established by the US Environmental Protection Agency (EPA) and estimates population risks in terms of dose to humans. The system model embodied in the SUMO (System Unsaturated Model) code will also allow benchmarking of other models being developed for the Yucca Mountain Project. The system model has three natural divisions: (1) source term, (2) far-field transport, and (3) dose to humans. This document gives a detailed description of the mathematics of each of these three divisions. Each of the governing equations employed is based on modeling assumptions that are widely accepted within the scientific community.

Eslinger, P.W.; Miley, T.B.; Engel, D.W.; Chamberlain, P.J. II

1992-09-01T23:59:59.000Z

349

Low-temperature lithium diffusion in simulated high-level boroaluminosilicate nuclear waste glasses  

SciTech Connect (OSTI)

Ion exchange is recognized as an integral, if underrepresented, mechanism influencing glass corrosion. However, due to the formation of various alteration layers in the presence of water, it is difficult to conclusively deconvolute the mechanisms of ion exchange from other processes occurring simultaneously during corrosion. In this work, an operationally inert non-aqueous solution was used as an alkali source material to isolate ion exchange and study the solid-state diffusion of lithium. Specifically, the experiments involved contacting glass coupons relevant to the immobilization of high-level nuclear waste, SON68 and CJ-6, which contained Li in natural isotope abundance, with a non-aqueous solution of 6LiCl dissolved in dimethyl sulfoxide at 90 C for various time periods. The depth profiles of major elements in the glass coupons were measured using time-of-flight secondary ion mass spectrometry (ToF-SIMS). Lithium interdiffusion coefficients, DLi, were then calculated based on the measured depth profiles. The results indicate that the penetration of 6Li is rapid in both glasses with the simplified CJ-6 glass (D6Li ? 4.0-8.0 10-21 m2/s) exhibiting faster exchange than the more complex SON68 glass (DLi ? 2.0-4.0 10-21 m2/s). Additionally, sodium ions present in the glass were observed to participate in ion exchange reactions; however, different diffusion coefficients were necessary to fit the diffusion profiles of the two alkali ions. Implications of the diffusion coefficients obtained in the absence of alteration layers to the long-term performance of nuclear waste glasses in a geological repository system are also discussed.

Neeway, James J.; Kerisit, Sebastien N.; Gin, Stephane; Wang, Zhaoying; Zhu, Zihua; Ryan, Joseph V.

2014-12-01T23:59:59.000Z

350

Crystallization of Sodium Phosphate Dodecahydrate and Re-Crystallization to Natrophosphate in Simulated Hanford Nuclear Waste  

SciTech Connect (OSTI)

The nuclear waste at the Hanford site near Richland, WA, has large concentrations of phosphate in the form of the phosphate ion, sodium phosphate dodecahydrate (Na3PO4.12H2O.(1/4)NaOH) and natrophosphate (Na7F(PO4)3.19H2O). Sodium phosphate dodecahydrate can form a gel and natrophosphate can crystallize large particles, complicating the processing of slurries of both salts. The gel is regarded as more problematic, so natrophosphate has historically been crystallized to prevent phosphate gelling. This study determined that natrophosphate crystals can grow to large size in short time periods (a few weeks), time periods relevant to short process shutdowns. Solutions of NaOH and NaAl(OH)4 were blended at different ratios with stock solutions containing NaOH, NaF, and Na3PO4 at 50 C. The mixtures were allowed to cool to 22 C, and the crystal growth was monitored by Polarized Light Microscopy (PLM) for 17 days. Four of the six blends investigated gelled rapidly due to Na3PO4.12H2O.(1/4)NaOH precipitation. The gel slowly dissipated over time as the solids recrystallized into natrophosphate. In one case, the natrophosphate reached sizes of greater than 1000 microns in diameter in just 4 days. This rapid gelling and crystallization kinetics is important to engineers trying to manage nuclear wastes high in phosphates. Hanford tank farm engineers are developing sampling plans to support temperature and process control strategies for preventing the formation of solid phosphates. They are also developing methods of suspending large natrophosphate crystals.

Reynolds, J. G.

2014-12-22T23:59:59.000Z

351

The Transmutation of Nuclear Waste in the Two-Zone Subcritical System Driven by High- Intensity Neutron Generator - 12098  

SciTech Connect (OSTI)

The main problems of transmutation of high-level radioactive waste (minor actinides and long-lived fission products) are considered in our work. The range of radioactive waste of nuclear power is analyzed. The conditions under which the transmutation of radioactive waste will be most effective are analyzed too. The modeling results of a transmutation of the main radioactive isotopes are presented and discussed. The transmutation of minor actinides and long-lived fission products are modeled in our work (minor actinides - Np-237, Am-241, Am-242, Am-243, Cm-244, Cm-245; long-lived fission products - I-129, Tc-99). The two-zone subcritical system is calculated with help of different neutron-physical codes (MCNP, Scale, Montebarn, Origen). The ENDF/B-VI nuclear data library used in above calculations. Thus, radioactive wastes can be divided into two main groups that need to be transmuted. The minor actinides form the first group and the long-lived fission products form the second one. For the purpose of effective transmutation these isotopes must be extracted from the spent nuclear fuel with the help of either PUREX technology or pyrometallurgical technology. The two-zone reactor system with fast and thermal regions is more effective for nuclear waste transmutation than the one-zone reactor. Modeling results show that nearly all radioactive wastes can be transmuted in the two-zone subcritical system driven by a high-intensity neutron generator with the external neutron source strength of 1.10{sup 13} n/sec. Obviously, transmutation rate will increase with a rise of the external neutron source strength. From the results above we can also see that the initial loading of radioactive isotopes into the reactor system should exceed by mass those isotopes that are finally produced. (authors)

Babenko, V.O. [Bogolyubov Institute for Theoretical Physics, Metrolohichna str. 14-b, Kiev, 03680 (Ukraine); Gulik, V.I.; Pavlovych, V.M. [Institute for Nuclear Research, pr. Nauky 47, Kyiv, 03680 (Ukraine)

2012-07-01T23:59:59.000Z

352

Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) verification and validation plan. version 1.  

SciTech Connect (OSTI)

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Waste Integrated Performance and Safety Codes (NEAMS Waste IPSC) is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to quantitatively assess the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. To meet this objective, NEAMS Waste IPSC M&S capabilities will be applied to challenging spatial domains, temporal domains, multiphysics couplings, and multiscale couplings. A strategic verification and validation (V&V) goal is to establish evidence-based metrics for the level of confidence in M&S codes and capabilities. Because it is economically impractical to apply the maximum V&V rigor to each and every M&S capability, M&S capabilities will be ranked for their impact on the performance assessments of various components of the repository systems. Those M&S capabilities with greater impact will require a greater level of confidence and a correspondingly greater investment in V&V. This report includes five major components: (1) a background summary of the NEAMS Waste IPSC to emphasize M&S challenges; (2) the conceptual foundation for verification, validation, and confidence assessment of NEAMS Waste IPSC M&S capabilities; (3) specifications for the planned verification, validation, and confidence-assessment practices; (4) specifications for the planned evidence information management system; and (5) a path forward for the incremental implementation of this V&V plan.

Bartlett, Roscoe Ainsworth; Arguello, Jose Guadalupe, Jr.; Urbina, Angel; Bouchard, Julie F.; Edwards, Harold Carter; Freeze, Geoffrey A.; Knupp, Patrick Michael; Wang, Yifeng; Schultz, Peter Andrew; Howard, Robert (Oak Ridge National Laboratory, Oak Ridge, TN); McCornack, Marjorie Turner

2011-01-01T23:59:59.000Z

353

FLUIDIZED BED STEAM REFORMING MINERALIZATION FOR HIGH ORGANIC AND NITRATE WASTE STREAMS FOR THE GLOBAL NUCLEAR ENERGY PARTNERSHIP  

SciTech Connect (OSTI)

Waste streams that may be generated by the Global Nuclear Energy Partnership (GNEP) Advanced Energy Initiative may contain significant quantities of organics (0-53 wt%) and/or nitrates (0-56 wt%). Decomposition of high nitrate streams requires reducing conditions, e.g. organic additives such as sugar or coal, to reduce the NO{sub x} in the off-gas to N{sub 2} to meet the Clean Air Act (CAA) standards during processing. Thus, organics will be present during waste form stabilization regardless of which GNEP processes are chosen, e.g. organics in the feed or organics for nitrate destruction. High organic containing wastes cannot be stabilized with the existing HLW Best Developed Available Technology (BDAT) which is HLW vitrification (HLVIT) unless the organics are removed by preprocessing. Alternative waste stabilization processes such as Fluidized Bed Steam Reforming (FBSR) operate at moderate temperatures (650-750 C) compared to vitrification (1150-1300 C). FBSR converts organics to CAA compliant gases, creates no secondary liquid waste streams, and creates a stable mineral waste form that is as durable as glass. For application to the high Cs-137 and Sr-90 containing GNEP waste streams a single phase mineralized Cs-mica phase was made by co-reacting illite clay and GNEP simulated waste. The Cs-mica accommodates up to 30% wt% Cs{sub 2}O and all the GNEP waste species, Ba, Sr, Rb including the Cs-137 transmutation to Ba-137. For reference, the cesium mineral pollucite (CsAlSi{sub 2}O{sub 6}), currently being studied for GNEP applications, can only be fabricated at {ge} 1000 C. Pollucite mineralization creates secondary aqueous waste streams and NO{sub x}. Pollucite is not tolerant of high concentrations of Ba, Sr or Rb and forces the divalent species into different mineral host phases. The pollucite can accommodate up to 33% wt% Cs{sub 2}O.

Jantzen, C; Michael Williams, M

2008-01-11T23:59:59.000Z

354

Sandia National Laboratories performance assessment methodology for long-term environmental programs : the history of nuclear waste management.  

SciTech Connect (OSTI)

Sandia National Laboratories (SNL) is the world leader in the development of the detailed science underpinning the application of a probabilistic risk assessment methodology, referred to in this report as performance assessment (PA), for (1) understanding and forecasting the long-term behavior of a radioactive waste disposal system, (2) estimating the ability of the disposal system and its various components to isolate the waste, (3) developing regulations, (4) implementing programs to estimate the safety that the system can afford to individuals and to the environment, and (5) demonstrating compliance with the attendant regulatory requirements. This report documents the evolution of the SNL PA methodology from inception in the mid-1970s, summarizing major SNL PA applications including: the Subseabed Disposal Project PAs for high-level radioactive waste; the Waste Isolation Pilot Plant PAs for disposal of defense transuranic waste; the Yucca Mountain Project total system PAs for deep geologic disposal of spent nuclear fuel and high-level radioactive waste; PAs for the Greater Confinement Borehole Disposal boreholes at the Nevada National Security Site; and PA evaluations for disposal of high-level wastes and Department of Energy spent nuclear fuels stored at Idaho National Laboratory. In addition, the report summarizes smaller PA programs for long-term cover systems implemented for the Monticello, Utah, mill-tailings repository; a PA for the SNL Mixed Waste Landfill in support of environmental restoration; PA support for radioactive waste management efforts in Egypt, Iraq, and Taiwan; and, most recently, PAs for analysis of alternative high-level radioactive waste disposal strategies including repositories deep borehole disposal and geologic repositories in shale and granite. Finally, this report summarizes the extension of the PA methodology for radioactive waste disposal toward development of an enhanced PA system for carbon sequestration and storage systems. These efforts have produced a generic PA methodology for the evaluation of waste management systems that has gained wide acceptance within the international community. This report documents how this methodology has been used as an effective management tool to evaluate different disposal designs and sites; inform development of regulatory requirements; identify, prioritize, and guide research aimed at reducing uncertainties for objective estimations of risk; and support safety assessments.

Marietta, Melvin Gary; Anderson, D. Richard; Bonano, Evaristo J.; Meacham, Paul Gregory (Raytheon Ktech, Albuquerque, NM)

2011-11-01T23:59:59.000Z

355

Calculations of Induced Activity in the ATLAS Experiment for Nuclear Waste Zoning.  

E-Print Network [OSTI]

Extensive calculations were performed with the general activation formula using the fluxes of high-energy hadrons and low-energy neutrons previously obtained from simulations with the GCALOR code of the ATLAS detector. Three sets of proton cross-sections were used for hadrons energy above 20 MeV: (a) one set calculated with the YIELDX code (i.e., the Silberberg-Tsao formula of partial proton spallation cross-sections), (b) one set calculated with the Rudstam formula, and (c) the ??best-estimate' dataset which was a compilation of the available experimental and calculated data. In the energy region below 20 MeV, neutron activation cross-sections were taken from evaluated nuclear data files. The activity of each nuclide for a predefined operation scenario (i.e., number and duration of irradiation and shutdown cycles) was normalized to reference values taken from the European or Swiss legislations, to obtain an aggregate estimate of the radiological hazard comparable with a nuclear waste zoning definition cr...

Morev, M N

2007-01-01T23:59:59.000Z

356

RADIOACTIVE WASTE STORAGE IN MINED CAVERNS IN CRYSTALLINE ROCK-RESULTS OF FIELD INVESTIGATIONS AT STRIPA, SWEDEN  

E-Print Network [OSTI]

for Nuclear Waste Management, Materials Research Society.for Nuclear Waste Management, Materials Research Society.

Witherspoon, P.A.

2010-01-01T23:59:59.000Z

357

Comparison of Different Upscaling Methods for Predicting Thermal Conductivity of Complex Heterogeneous Materials System: Application on Nuclear Waste Forms  

SciTech Connect (OSTI)

To develop a strategy in thermal conductivity prediction of a complex heterogeneous materials system, loaded nuclear waste forms, the computational efficiency and accuracy of different upscaling methods have been evaluated. The effective thermal conductivity, obtained from microstructure information and local thermal conductivity of different components, is critical in predicting the life and performance of waste form during storage. Several methods, including the Taylor model, Sachs model, self-consistent model, and statistical upscaling method, were developed and implemented. Microstructure based finite element method (FEM) prediction results were used to as benchmark to determine the accuracy of the different upscaling methods. Micrographs from waste forms with varying waste loadings were used in the prediction of thermal conductivity in FEM and homogenization methods. Prediction results demonstrated that in term of efficiency, boundary models (e.g., Taylor model and Sachs model) are stronger than the self-consistent model, statistical upscaling method, and finite element method. However, when balancing computational efficiency and accuracy, statistical upscaling is a useful method in predicting effective thermal conductivity for nuclear waste forms.

Li, Dongsheng; Sun, Xin; Khaleel, Mohammad A.

2012-06-16T23:59:59.000Z

358

Proceedings of the 1993 international conference on nuclear waste management and environmental remediation. Volume 2: High level radioactive waste and spent fuel management  

SciTech Connect (OSTI)

This conference was held in 1993 in Prague, Czech Republic to provide a forum for exchange of state-of-the-art information on radioactive waste management. Volume 2 contains 109 papers divided into the following sections: recent developments in environmental remediation technologies; decommissioning of nuclear power reactors; environmental restoration site characterization and monitoring; decontamination and decommissioning of other nuclear facilities; prediction of contaminant migration and related doses; treatment of wastes from decontamination and decommissioning operations; management of complex environmental cleanup projects; experiences in actual cleanup actions; decontamination and decommissioning demolition technologies; remediation of obsolete sites from uranium mining and milling; ecological impacts from radioactive environmental contamination; national environmental management regulations--issues and assessments; significant issues and strategies in environmental management; acceptance criteria for very low-level radioactive wastes; processes for public involvement in environmental activities and decisions; recent experiences in public participation activities; established and emerging environmental management organizations; and economic considerations in environmental management. Individual papers have been processed separately for inclusion in the appropriate data bases.

Ahlstroem, P.E.; Chapman, C.C.; Kohout, R.; Marek, J. [eds.

1993-12-31T23:59:59.000Z

359

Performance assessment of the direct disposal in unsaturated tuff or spent nuclear fuel and high-level waste owned by USDOE: Volume 2, Methodology and results  

SciTech Connect (OSTI)

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

360

Radioactive waste isolation in salt: special advisory report on the status of the Office of Nuclear Waste Isolation's plans for repository performance assessment  

SciTech Connect (OSTI)

Repository performance assessment is analysis that identifies events and processes that might affect a repository system for isolation of radioactive waste, examines their effects on barriers to waste migration, and estimates the probabilities of their occurrence and their consequences. In 1983 Battelle Memorial Institute's Office of Nuclear Waste Isolation (ONWI) prepared two plans - one for performance assessment for a waste repository in salt and one for verification and validation of performance assessment technology. At the request of the US Department of Energy's Salt Repository Project Office (SRPO), Argonne National Laboratory reviewed those plans and prepared this report to advise SRPO of specific areas where ONWI's plans for performance assessment might be improved. This report presents a framework for repository performance assessment that clearly identifies the relationships among the disposal problems, the processes underlying the problems, the tools for assessment (computer codes), and the data. In particular, the relationships among important processes and 26 model codes available to ONWI are indicated. A common suggestion for computer code verification and validation is the need for specific and unambiguous documentation of the results of performance assessment activities. A major portion of this report consists of status summaries of 27 model codes indicated as potentially useful by ONWI. The code summaries focus on three main areas: (1) the code's purpose, capabilities, and limitations; (2) status of the elements of documentation and review essential for code verification and validation; and (3) proposed application of the code for performance assessment of salt repository systems. 15 references, 6 figures, 4 tables.

Ditmars, J.D.; Walbridge, E.W.; Rote, D.M.; Harrison, W.; Herzenberg, C.L.

1983-10-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

Technology of high-level nuclear waste disposal. Advances in the science and engineering of the management of high-level nuclear wastes. Volume 1  

SciTech Connect (OSTI)

The papers in this volume cover the following subjects: waste isolation and the natural geohydrologic system; repository perturbations of the natural system; radionuclide migration through the natural system; and repository design technology. Individual papers are abstracted.

Hofmann, P.L.; Breslin, J.J. (eds.)

1981-01-01T23:59:59.000Z

362

Radioactive waste isolation in salt: Peer review of the Office of Nuclear Waste Isolation's draft report on an issues hierarchy and data needs for site characterization  

SciTech Connect (OSTI)

At the request of the Salt Repository Project (SRPO), Argonne National Laboratory conducted an independent peer review of a report by the Battelle Office of Nuclear Waste Isolation entitled ''Salt Repository Project Issues Hierarchy and Data Needs for Site Characterization (Draft).'' This report provided a logical structure for evaluating the outstanding questions (issues) related to selection and licensing of a site as a high-level waste repository. It also provided a first estimate of the information and data necessary to answer or resolve those questions. As such, this report is the first step in developing a strategy for site characterization. Microfiche copies of ''Draft Issues Hierarchy, Resolution Strategy, and Information Needs for Site Characterization and Environmental/Socioeconomic Evaluation - July, 1986'' and ''Issues Hierarchy and Data Needs for Site Characterization - February, 1985'' are included in the back pocket of this report.

Harrison, W.; Fenster, D.F.; Ditmars, J.D.; Paddock, R.A.; Rote, D.M.; Hambley, D.F.; Seitz, M.G.; Hull, A.B.

1986-12-01T23:59:59.000Z

363

Sodium-Bearing Waste Treatment Alternatives Implementation Study  

SciTech Connect (OSTI)

The purpose of this document is to discuss issues related to the implementation of each of the five down-selected INEEL/INTEC radioactive liquid waste (sodium-bearing waste - SBW) treatment alternatives and summarize information in three main areas of concern: process/technical, environmental permitting, and schedule. Major implementation options for each treatment alternative are also identified and briefly discussed. This report may touch upon, but purposely does not address in detail, issues that are programmatic in nature. Examples of these include how the SBW will be classified with respect to the Nuclear Waste Policy Act (NWPA), status of Waste Isolation Pilot Plant (WIPP) permits and waste storage availability, available funding for implementation, stakeholder issues, and State of Idaho Settlement Agreement milestones. It is assumed in this report that the SBW would be classified as a transuranic (TRU) waste suitable for disposal at WIPP, located in New Mexico, after appropriate treatment to meet transportation requirements and waste acceptance criteria (WAC).

Charles M. Barnes; James B. Bosley; Clifford W. Olsen

2004-07-01T23:59:59.000Z

364

Toward Understanding the Effect of Nuclear Waste Glass Composition of Sulfur Solubility  

SciTech Connect (OSTI)

The concentration of sulfur in nuclear waste glass melter feed must be maintained below the point where salt accumulates on the melt surface. The allowable concentrations may range from 0.37 to over 2.05 weight percent (of SO3 on a calcined oxide basis) depending on the composition of the melter feed and processing conditions. If the amount of sulfur exceeds the melt tolerance level, a molten salt will accumulate, which may upset melter operations and potentially shorten the useful life of the melter. At the Hanford site, relatively conservative limits have been placed on sulfur loading in melter feed, which in turn significantly increases the amount of glass that will be produced. Crucible-scale sulfur solubility data and scaled melter sulfur tolerance data have been collected on simulated Hanford waste glasses over the last 15 years. These data were compiled and analyzed. A model was developed to predict the solubility of SO3 in glass based on 252 simulated Hanford low-activity waste (LAW) glass compositions. This model represents the data well, accounting for over 85% of the variation in data, and was well validated. The model was also found to accurately predict the tolerance for sulfur in melter feed for 13 scaled melter tests of simulated LAW glasses. The model can be used to help estimate glass volumes and make informed decisions on process options. The model also gives quantitative estimates of component concentration effects on sulfur solubility. The components that most increase sulfur solubility are Li2O > V2O5> CaO ? P2O5 > Na2O ? B2O3 > K2O. The components that most decrease sulfur solubility are Cl > Cr2O3 > Al2O3 > ZrO2 ? SnO2 > Others ? SiO2. The order of component effects is similar to previous literature data, in most cases.

Vienna, John D.; Kim, Dong-Sang; Muller, I. S.; Kruger, Albert A.; Piepel, Gregory F.

2014-02-13T23:59:59.000Z

365

Modeling Solute Thermokinetics in LiCI-KCI Molten Salt for Nuclear Waste Separation  

SciTech Connect (OSTI)

Recovery of actinides is an integral part of a closed nuclear fuel cycle. Pyrometallurgical nuclear fuel recycling processes have been developed in the past for recovering actinides from spent metallic and nitride fuels. The process is essentially to dissolve the spent fuel in a molten salt and then extract just the actinides for reuse in a reactor. Extraction is typically done through electrorefining, which involves electrochemical reduction of the dissolved actinides and plating onto a cathode. Knowledge of a number of basic thermokinetic properties of salts and salt-fuel mixtures is necessary for optimizing present and developing new approaches for pyrometallurgical waste processing. The properties of salt-fuel mixtures are presently being studied, but there are so many solutes and varying concentrations that direct experimental investigation is prohibitively time consuming and expensive (particularly for radioactive elements like Pu). Therefore, there is a need to reduce the number of required experiments through modeling of salt and salt-fuel mixture properties. This project will develop first-principles-based molecular modeling and simulation approaches to predict fundamental thermokinetic properties of dissolved actinides and fission products in molten salts. The focus of the proposed work is on property changes with higher concentrations (up to 5 mol%) of dissolved fuel components, where there is still very limited experimental data. The properties predicted with the modeling will be density, which is used to assess the amount of dissolved material in the salt; diffusion coefficients, which can control rates of material transport during separation; and solute activity, which determines total solubility and reduction potentials used during electrorefining. The work will focus on La, Sr, and U, which are chosen to include the important distinct categories of lanthanides, alkali earths, and actinides, respectively. Studies will be performed using LiCl-KCl salt at the eutectic composition (58 mol% LiCl, 42 mol% KCl), which is used for treating spent EBR-II fuel. The same process being used for EBRII fuel is currently being studied for widespread international implementation. The methods will focus on first-principles and first- principles derived interatomic potential based simulations, primarily using molecular dynamics. Results will be validated against existing literature and parallel ongoing experimental efforts. The simulation results will be of value for interpreting experimental results, validating analytical models, and for optimizing waste separation by potentially developing new salt configurations and operating conditions.

Morgan, Dane; Eapen, Jacob

2013-10-01T23:59:59.000Z

366

1993 International conference on nuclear waste management and environmental remediation, Prague, Czech Republic, September 5--11, 1993. Combined foreign trip report  

SciTech Connect (OSTI)

The purpose of the trip was to attend the 1993 International Conference on Nuclear Waste Management and Environmental Remediation. The principal objective of this conference was to facilitate a truly international exchange of information on the management of nuclear wastes as well as contaminated facilities and sites emanating from nuclear operations. The conference was sponsored by the American Society of Mechanical Engineers, the Czech and Slovak Mechanical Engineering Societies, and the Czech and Slovak Nuclear Societies in cooperation with the Commission of the European Communities, the International Atomic Energy Agency, and the OECD Nuclear Agency. The conference was cosponsored by the American Nuclear Society, the Atomic Energy Society of Japan, the Canadian Nuclear Society, the (former USSR) Nuclear Society, and the Japan Society of Mechanical Engineers. This was the fourth in a series of biennial conferences, which started in Hong Kong, in 1987. This report summarizes shared aspects of the trip; however, each traveler`s observations and recommendations are reported separately.

Slate, S.C. [comp.; Allen, R.E. [ed.

1993-12-01T23:59:59.000Z

367

Department of Energy plan for recovery and utilization of nuclear byproducts from defense wastes. Volume 1. Executive summary  

SciTech Connect (OSTI)

Nuclear byproducts are a major national resource that has yet to be incorporated into the economy. The current Defense Byproducts Program is designed to match specific military and commercial needs with the availability of valuable products which are currently treated as waste at considerable expense in waste management costs. This program plan focuses on a few specific areas with the greatest potential for near-term development and application. It also recognizes the need for a continuing effort to develop new applications for byproducts and to continue to assess the impacts on waste management. The entire program has been, and will continue to be structured so as to ensure the safety of the public and maintain the purity of the environment. Social and institutional concerns have been recognized and will be handled appropriately. A significant effort will be undertaken to inform the public of the benefits of byproduct use and of the care being taken to ensure safe, efficient operation.

None

1983-08-01T23:59:59.000Z

368

NWTS program criteria for mined geologic disposal of nuclear waste: program objectives, functional requirements, and system performance criteria  

SciTech Connect (OSTI)

At the present time, final repository criteria have not been issued by the responsible agencies. This document describes general objectives, requirements, and criteria that the DOE intends to apply in the interim to the National Waste Terminal Storage (NWTS) Program. These objectives, requirements, and criteria have been developed on the basis of DOE's analysis of what is needed to achieve the National objective of safe waste disposal in an environmentally acceptable and economic manner and are expected to be consistent with anticipated regulatory standards. The qualitative statements in this document address the broad issues of public and occupational health and safety, institutional acceptability, engineering feasibility, and economic considerations. A comprehensive set of criteria, general and project specific, of which these are a part, will constitute a portion of the technical basis for preparation and submittal by the DOE of formal documents to support future license applications for nuclear waste repositories.

None

1981-04-01T23:59:59.000Z

369

Application of evolved gas analysis to cold-cap reactions of melter feeds for nuclear waste vitrification  

SciTech Connect (OSTI)

In the vitrification of nuclear wastes, the melter feed (a mixture of nuclear waste and glass-forming and modifying additives) experiences multiple gas-evolving reactions in an electrical glass-melting furnace. We employed the thermogravimetry-gas chromatography-mass spectrometry (TGA-GC-MS) combination to perform evolved gas analysis (EGA). Apart from identifying the gases evolved, we performed quantitative analysis relating the weighed sum of intensities of individual gases linearly proportional with the differential themogravimetry. The proportionality coefficients were obtained by three methods based on the stoichiometry, least squares, and calibration. The linearity was shown to be a good first-order approximation, in spite of the complicated overlapping reactions.

Kruger, Albert A.; Chun, Jaehun; Hrma, Pavel R.; Rodriguez, Carmen P.; Schweiger, Michael J.

2014-04-30T23:59:59.000Z

370

Application of evolved gas analysis to cold-cap reactions of melter feeds for nuclear waste vitrification  

SciTech Connect (OSTI)

In the vitrification of nuclear wastes, the melter feed (a mixture of nuclear waste and glass forming and modifying additives) experiences multiple gas-evolving reactions in an electrical glass-melting furnace. Foams from the residual gases can significantly alter the melting rate through mass and heat transfers. We employed the thermogravimetry-gas chromatography-mass spectrometry (TGA-GC-MS) combination to perform quantitative evolved gas analysis (EGA) and developed a simple calibration model which correlates the overall mass loss rate with the evolution rates for individual gases. The model parameters are obtained from the least squares analysis, assuming that the gas-evolving reactions are independent. Thus, the EGA adds the chemical identity to the reactions indicated by the phenomenological kinetic model.

Rodriguez, Carmen P.; Chun, Jaehun; Schweiger, Michael J.; Kruger, Albert A.; Hrma, Pavel R.

2014-09-01T23:59:59.000Z

371

Proposed rulemaking on the storage and disposal of nuclear waste. Cross-statement of the United States Department of Energy  

SciTech Connect (OSTI)

The US DOE cross-statement in the matter of proposed rulemaking in the storage and disposal of nuclear wastes is presented. It is concluded from evidence contained in the document that: (1) spent fuel can be disposed of in a manner that is safe and environmentally acceptable; (2) present plans for establishing geological repositories are an effective and reasonable means of disposal; (3) spent nuclear fuel from licensed facilities can be stored in a safe and environmentally acceptable manner on-site or off-site until disposal facilities are ready; (4) sufficient additional storage capacity for spent fuel will be established; and (5) the disposal and interim storage systems for spent nuclear fuel will be integrated into an acceptable operating system. It was recommended that the commission should promulgate a rule providing that the safety and environmental implications of spent nuclear fuel remaining on site after the anticipated expiration of the facility licenses involved need not be considered in individual facility licensing proceedings. A prompt finding of confidence in the nuclear waste disposal and storage area by the commission is also recommeded. (DMC)

None

1980-09-05T23:59:59.000Z

372

ORNL nuclear waste programs annual progress report for period ending September 30, 1982  

SciTech Connect (OSTI)

Research progress is reported in 20 activities under the headings: spent fuels, defense waste management, commercial waste management, remedial action, and conventional reactors. Separate entries were prepared for each activity.

Not Available

1983-05-01T23:59:59.000Z

373

Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act: Volume 2  

SciTech Connect (OSTI)

The Yucca Mountain site in Nevada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in accordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site; to describe the conceptual designs for the repository and the waste package and to present the plans for obtaining the geologic information necessary to demonstrate the suitability of the site for a repository, to design the repository and the waste package, to prepare an environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. Chapter 3 summarizes present knowledge of the regional and site hydrologic systems. The purpose of the information presented is to (1) describe the hydrology based on available literature and preliminary site-exploration activities that have been or are being performed and (2) provide information to be used to develop the hydrologic aspects of the planned site characterization program. Chapter 4 contains geochemical information about the Yucca Mountain site. The chapter references plan for continued collection of geochemical data as a part of the site characterization program. Chapter 4 describes and evaluates data on the existing climate and site meterology, and outlines the suggested procedures to be used in developing and validating methods to predict future climatic variation. 534 refs., 100 figs., 72 tabs.

NONE

1988-01-01T23:59:59.000Z

374

Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act: Volume 1  

SciTech Connect (OSTI)

The Yucca Mountain site in Nevada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in acordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site;to describe the conceptual designs for the repository and the waste package and to present the plans for obtaining the geologic information necessary to demonstrate the suitability of the site for a repository, to design the repository and the waste package, to prepare an environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. This introduction begins with a brief section on the process for siting and eveloping a repository, followed by a discussion of the pertinent legislation and regulations. A description of site characterization is presented next;it describes the facilities to be constructed for the site characterization program and explains the principal activities to be conducted during the program. Finally, the purpose, content, organizing prinicples, and organization of this site characterization plan are outlined, and compliance with applicable regulations is discussed. 880 refs., 130 figs., 25 tabs.

NONE

1988-01-01T23:59:59.000Z

375

Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act: Volume 4  

SciTech Connect (OSTI)

The Yucca Mountain site in Nevada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended and approved by the President for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in accordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site; to describe the conceptual designs for the repository and the waste package; and to present the plans for obtaining the geologic information necessary to demonstate the suitability of the site for a repository, to desin the repository and the waste package, to prepare an environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. This introduction begins with a brief section on the process for siting and developing a repository, followed by a discussion of the pertinent legislation and regulations. A description of site characterization is presented next; it describes the facilities to be constructed for the site characterization program and explains the principal activities to be conducted during the program. Finally, the purpose, content, organizing principles, and organization of this site characterization plan are outlined, and compliance with applicable regulations is discussed.

NONE

1988-01-01T23:59:59.000Z

376

Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act: Volume 7  

SciTech Connect (OSTI)

The Yucca Mountain site in Neavada is one of three candidate sites for the first geologic repository for radioactive waste. On May 28, 1986, it was recommended and approved for detailed study in a program of site characterization. This site characterization plan (SCP) has been prepared in accordance with the requirements of the Nuclear Waste Policy Act to summarize the information collected to date about the geologic conditions at the site;to describe the conceptual designs for the repository and the waste package;and to present the plans for obtaining hte geologic information necessary to demonstrate the suitability of the site for a repository, to design the repository and the waste package, to prepare and environmental impact statement, and to obtain from the US Nuclear Regulatory Commission (NRC) an authorization to construct the repository. This introduction begins with a brief section on the process for siting and developing a repository, followed by a discussion of the pertinent legislation and regulations. A description of site characterization is presented next;it describes the facilities to be constructed for the site characterization program and explains the principal activities to be conducted during the program. Finally, the purpose, content, organizing principles, and organization of this site characterization plan are outlined, and compliance with applicable regulations is discussed.

NONE

1988-01-01T23:59:59.000Z

377

Surfactants containing radioactive run-offs: Ozone treatment, influence on nuclear power plants water waste special treatment  

SciTech Connect (OSTI)

The authors discuss the problems encountered in the efficiency of radioactive waste treatment in nuclear power plants in Kursk. The ozonization of aqueous solutions of surfactants was carried out in the laboratory`s ozonization system. The surfactants which are discharged to the ion exchangers deteriorate resins, clog up the ion exchangers, and decrease filtration velocity. Therefore, this investigation focused on finding a method to increase the efficiency of this treatment process.

Prokudina, S.A.; Grachok, M.A. [Belarussian State Economic Univ., Minsk (Belarus)

1993-12-31T23:59:59.000Z

378

Geothermal reservoir simulation to enhance confidence in predictions for nuclear waste disposal  

SciTech Connect (OSTI)

Numerical simulation of geothermal reservoirs is useful and necessary in understanding and evaluating reservoir structure and behavior, designing field development, and predicting performance. Models vary in complexity depending on processes considered, heterogeneity, data availability, and study objectives. They are evaluated using computer codes written and tested to study single and multiphase flow and transport under nonisothermal conditions. Many flow and heat transfer processes modeled in geothermal reservoirs are expected to occur in anthropogenic thermal (AT) systems created by geologic disposal of heat-generating nuclear waste. We examine and compare geothermal systems and the AT system expected at Yucca Mountain, Nevada, and their modeling. Time frames and spatial scales are similar in both systems, but increased precision is necessary for modeling the AT system, because flow through specific repository locations will affect long-term ability radionuclide retention. Geothermal modeling experience has generated a methodology, used in the AT modeling for Yucca Mountain, yielding good predictive results if sufficient reliable data are available and an experienced modeler is involved. Codes used in geothermal and AT modeling have been tested extensively and successfully on a variety of analytical and laboratory problems.

Kneafsey, Timothy J.; Pruess, Karsten; O'Sullivan, Michael J.; Bodvarsson, Gudmundur S.

2002-06-15T23:59:59.000Z

379

Oxidative Alkaline leaching of Americium from simulated high-level nuclear waste sludges  

SciTech Connect (OSTI)

Oxidative alkaline leaching has been proposed to pre-treat the high-level nuclear waste sludges to remove some of the problematic (e.g., Cr) and/or non-radioactive (e.g., Na, Al) constituents before vitrification. It is critical to understand the behavior of actinides, americium and plutonium in particular, in oxidative alkaline leaching. We have studied the leaching behavior of americium from four different sludge simulants (BiPO{sub 4}, BiPO{sub 4 modified}, Redox, PUREX) using potassium permanganate and potassium persulfate in alkaline solutions. Up to 60% of americium sorbed onto the simulants is leached from the sludges by alkaline persulfate and permanganate. The percentage of americium leached increases with [NaOH] (between 1.0 and 5.0 M). The initial rate of americium leaching by potassium persulfate increases in the order BiPO{sub 4} sludge < Redox sludge < PUREX sludge. The data are most consistent with oxidation of Am{sup 3+} in the sludge to either AmO{sub 2}{sup +} or AmO{sub 2}{sup 2+} in solution. Though neither of these species is expected to exhibit long-term stability in solution, the potential for mobilization of americium from sludge samples would have to be accommodated in the design of any oxidative leaching process for real sludge samples.

Reed, Wendy A.; Garnov, Alexander Yu.; Rao, Linfeng; Nash, Kenneth L.; Bond, Andrew H.

2004-01-23T23:59:59.000Z

380

Aerosol particle transport modeling for preclosure safety studies of nuclear waste repositories  

SciTech Connect (OSTI)

An important concern for preclosure safety analysis of a nuclear waste repository is the potential release to the environment of respirable aerosol particles. Such particles, less than 10 {mu}m in aerodynamic diameter, may have significant adverse health effects if inhaled. To assess the potential health effects of these particles, it is not sufficient to determine the mass fraction of respirable aerosol. The chemical composition of the particles is also of importance since different radionuclides may pose vastly different health hazards. Thus, models are needed to determine under normal and accident conditions the particle size and the chemical composition distributions of aerosol particles as a function of time and of position in the repository. In this work a multicomponent sectional aerosol model is used to determine the aerosol particle size and composition distributions in the repository. A range of aerosol mass releases with varying mean particle sizes and chemical compositions is used to demonstrate the sensitivities and uncertainties of the model. Decontamination factors for some locations in the repository are presented. 8 refs., 1 tab.

Gelbard, F. [Sandia National Labs., Albuquerque, NM (USA)

1989-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "nuclear waste fund" 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

Obsidians and tektites: Natural analogues for water diffusion in nuclear waste glasses  

SciTech Connect (OSTI)

Projected scenarios for the proposed Yucca Mountain repository include significant periods of time when high relative humidity atmospheres will be present, thus the reaction processes of interest will include those known to occur under these conditions. The ideal natural analog for the proposed Yucca Mountain repository would consist of natural borosilicate glasses exposed to expected repository conditions for thousands of years; however, the prospects for identifying such an analog are remote, but an important caveat for using natural analog studies is to relate the reaction processes in the analog to those in the system of interest, rather than a strict comparison of the glass compositions. In lieu of this, identifying natural glasses that have reacted via reaction processes expected in the repository is the most attractive option. The goal of this study is to quantify molecular water diffusion in the natural analogs obsidian and tektites. Results from this study can be used in assessing the importance of factors affecting molecular water diffusion in nuclear waste glasses, relative to other identified reaction processes. In this way, a better understanding of the long-term reaction mechanism can be developed and incorporated into performance assessment models. 17 refs., 4 figs.

Mazer, J.J.; Bates, J.K.; Bradley, C.R. [Argonne National Lab., IL (United States); Stevenson, C.M. [Archaeological Services Consultants, Inc., Columbus, OH (United States)

1991-11-01T23:59:59.000Z

382

Spent Fuel Disposal Trust Fund (Maine)  

Broader source: Energy.gov [DOE]

Any licensee operating a nuclear power plant in this State shall establish a segregated Spent Nuclear Fuel Disposal Trust Fund in accordance with this subchapter for the eventual disposal of spent...

383

Initial performance assessment of the disposal of spent nuclear fuel and high-level waste stored at Idaho National Engineering Laboratory. Volume 1, Methodology and results  

SciTech Connect (OSTI)

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

384

Steelmaker Matches Recovery Act Funds to Save Energy & Reduce...  

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

and installed with DOE Recovery Act Funding. Blast Furnace Gas Recovery Boiler Provides Steam and Power at Steel Mill More Documents & Publications Capturing Waste Gas: Saves...

385

THE POSITIVE IMPACTS OF AMERICAN REINVESTMENT AND RECOVERY ACT (ARRA) FUNDING TO THE WASTE MANAGEMENT PROGRAM ON HANFORD'S PLATEAU REMEDIATION PROJECT  

SciTech Connect (OSTI)

In April 2009, the Department of Energy (DOE) Richland Operations Office (RL) was allocated $1.6 billion (B) in ARRA funding to be applied to cleanup projects at the Hanford Site. DOE-RL selected projects to receive ARRA funding based on 3-criteria: creating/saving jobs, reducing the footprint of the Hanford Site, and reducing life-cycle costs for cleanup. They further selected projects that were currently covered under regulatory documents and existing prime contracts, which allowed work to proceed quickly. CH2M HILL Plateau Remediation Company (CHPRC) is a prime contractor to the DOE focused on the environmental cleanup of the DOE Hanford Site Central Plateau. CHPRC was slated to receive $1.36B in ARRA funding. As of January, 2010, CHPRC has awarded over $200 million (M) in subcontracts (64% to small businesses), created more that 1,100 jobs, and touched more than 2,300 lives - all in support of long-term objectives for remediation of the Central Plateau, on or ahead of schedule. ARRA funding is being used to accelerate and augment cleanup activities already underway under the baseline Plateau Remediation Contract (PRC). This paper details challenges and accomplishments using ARRA funding to meet DOE-RL objectives of creating/saving jobs, expediting cleanup, and reducing lifecycle costs for cleanup during the first months of implementation.

BLACKFORD LT

2010-01-19T23:59:59.000Z

386

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 (OSTI)

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

387

World nuclear outlook 1995  

SciTech Connect (OSTI)

As part of the EIA program to provide energy information, this analysis report presents the current status and projections through 2015 of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the uranium market. Long-term projections of US nuclear capacity, generation, and spent fuel discharges for two different scenarios through 2040 are developed for the Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM). In turn, the OCRWM provides partial funding for preparation of this report. The projections of uranium requirements are provided to the Organization for Economic Cooperation and Development (OECD) for preparation of the Nuclear Energy Agency/OECD report, Summary of Nuclear Power and Fuel Cycle Data in OECD Member Countries.

NONE

1995-09-29T23:59:59.000Z

388

World nuclear outlook 1994  

SciTech Connect (OSTI)

As part of the EIA program to provide energy information, this analysis report presents the current status and projections through 2010 of nuclear capacity, generation, and fuel cycle requirements for all countries in the world using nuclear power to generate electricity for commercial use. It also contains information and forecasts of developments in the uranium market. Long-term projections of US nuclear capacity, generation, and spent fuel discharges for three different scenarios through 2040 are developed for the Department of Energy`s Office of Civilian Radioactive Waste Management (OCRWM). In turn, the OCRWM provides partial funding for preparation of this report. The projections of uranium requirements are provided to the Organization for Economic Cooperation and Development (OECD) for preparation of the Nuclear Energy Agency/OECD report, Summary of Nuclear Power and Fuel Cycle Data in OECD Member Countries.

NONE

1994-12-01T23:59:59.000Z

389

Nuclear Energy Governance and the Politics of Social Justice: Technology, Public Goods, and Redistribution in Russia and France  

E-Print Network [OSTI]

for nuclear waste disposal and decommissioning whilethe cost of decommissioning and nuclear waste disposal on

Grigoriadis, Theocharis N

2009-01-01T23:59:59.000Z

390

Requested Funding Categories  

E-Print Network [OSTI]

Requested Funding Categories: Department's Funding Priority Request (in the event partial funding is granted): Committee recommends the following funding: Wages for adjunct or part- time faculty or admin Factors: Has unit received previous internship grant funding? _______ ifso

Martinez, Tony R.

391

Waste and Recycling  

ScienceCinema (OSTI)

Nuclear engineer Dr. Kathy McCarthy talks about nuclear energy, the challenge of nuclear waste and the research aimed at solutions. For more information about nuclear energy research, visit http://www.facebook.com/idahonationallaboratory.

McCarthy, Kathy

2013-05-28T23:59:59.000Z

392

Slovak Nuclear Regulatory Body Position in the Transport of Radioactive Waste  

SciTech Connect (OSTI)

This paper describes safety requirements for transport of radioactive waste in Slovakia and the role of regulatory body in the transport licensing and assessment processes. Importance of radioactive waste shipments have been increased since 1999 by starting of NPP A-1 decommissioning and operation of near surface disposal facility. Also some information from history of shipment as well as future activities are given. Legal basis for radioactive waste transport is resulting from IAEA recommendations in this area. Different types of transport equipment were approved by regulatory body for both liquid and solid waste and transportation permits were issued to their shipment. Regulatory body attention during evaluation of transport safety is focused mainly on ability of individual packages to withstand different transport conditions and on safety analyses performed for transport equipment for liquid waste with high frequency of shipments. During past three years no event was occurred in connection with radioactive waste transport in Slovakia.

Homola, J.

2003-02-27T23:59:59.000Z

393

Results from NNWSI [Nevada Nuclear Waste Storage Investigations] Series 2 bare fuel dissolution tests  

SciTech Connect (OSTI)

The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Two bare spent fuel specimens plus the empty cladding hulls were tested in NNWSI J-13 well water in unsealed fused silica vessels under ambient hot cell air conditions (25{degree}C) in the currently reported tests. One of the specimens was prepared from a rod irradiated in the H. B. Robinson Unit 2 reactor and the other from a rod irradiated in the Turkey Point Unit 3 reactor. Results indicate that most radionuclides of interest fall into three groups for release modeling. The first group principally includes the actinides (U, Np, Pu, Am, and Cm), all of which reached solubility-limited concentrations that were orders of magnitude below those necessary to meet the NRC 10 CFR 60.113 release limits for any realistic water flux predicted for the Yucca Mountain repository site. The second group is nuclides of soluble elements such as Cs, Tc, and I, for which release rates do not appear to be solubility-limited and may depend on the dissolution rate of fuel. In later test cycles, {sup 137}Cs, {sup 90}Sr, {sup 99}Tc, and {sup 129}I were continuously released at rates between about 5 {times} 10{sup {minus}5} and 1 {times} 10{sup {minus}4} of inventory per year. The third group is radionuclides that may be transported in the vapor phase, of which {sup 14}C is of primary concern. Detailed test results are presented and discussed. 17 refs., 15 figs., 21 tabs.

Wilson, C.N.

1990-09-01T23:59:59.000Z

394

Natural glass analogues to alteration of nuclear waste glass: A review and recommendations for further study  

SciTech Connect (OSTI)

The purpose of this report is to review previous work on the weathering of natural glasses; and to make recommendations for further work with respect to studying the alteration of natural glasses as it relates quantifying rates of dissolution. the first task was greatly simplified by the published papers of Jercinovic and Ewing (1987) and Byers, Jercinovic, and Ewing (1987). The second task is obviously the more difficult of the two and the author makes no claim of completeness in this regard. Glasses weather in the natural environment by reacting with aqueous solutions producing a rind of secondary solid phases. It had been proposed by some workers that the thickness of this rind is a function of the age of the glass and thus could be used to estimate glass dissolution rates. However, Jercinovic and Ewing (1987) point out that in general the rind thickness does not correlate with the age of the glass owing to the differences in time of contact with the solution compared to the actual age of the sample. It should be noted that the rate of glass dissolution is also a function of the composition of both the glass and the solution, and the temperature. Quantification of the effects of these parameters (as well as time of contact with the aqueous phase and flow rates) would thus permit a prediction of the consequences of glass-fluid interactions under varying environmental conditions. Defense high- level nuclear waste (DHLW), consisting primarily of liquid and sludge, will be encapsulated by and dispersed in a borosilicate glass before permanent storage in a HLW repository. This glass containing the DHLW serves to dilute the radionuclides and to retard their dispersion into the environment. 318 refs.

McKenzie, W.F.

1990-01-01T23:59:59.000Z

395

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

SciTech Connect (OSTI)

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

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

1987-12-01T23:59:59.000Z

396

Radioactive Waste Management (Minnesota)  

Broader source: Energy.gov [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...

397

A Characteristics-Based Approach to Radioactive Waste Classification in Advanced Nuclear Fuel Cycles  

E-Print Network [OSTI]

05-?20 Regarding Depleted Uranium. Rulemaking Issue (Department of Energy depleted uranium Environmental waste (MLLW) and depleted uranium (DU) from

Djokic, Denia

2013-01-01T23:59:59.000Z

398

The use of carbonation and fractional evaporative crystallization in the pretreatment of Hanford nuclear wastes.  

E-Print Network [OSTI]

??The purpose of this work was to explore the use of fractional evaporative crystallization as a technology that can be used to separate medium-curie waste (more)

Dumont, George Pierre, Jr.

2007-01-01T23:59:59.000Z

399

Bentonite alteration due to thermal-hydro-chemical processes during the early thermal period in a nuclear waste repository  

SciTech Connect (OSTI)

After closure of an underground nuclear waste repository, the decay of radionuclides will raise temperature in the repository, and the bentonite buffer will resaturate by water inflow from the surrounding host rock. The perturbations from these thermal and hydrological processes are expected to dissipate within hundreds to a few thousand years. Here, we investigate coupled thermal-hydro-chemical processes and their effects on the short-term performance of a potential nuclear waste repository located in a clay formation. Using a simplified geometric configuration and abstracted hydraulic parameters of the clayey formation, we examine geochemical processes, coupled with thermo-hydrologic phenomena, and potential changes in porosity near the waste container during the early thermal period. The developed models were used for evaluating the mineral alterations and potential changes in porosity of the buffer, which can affect the repository performance. The results indicate that mineral alteration and associated changes in porosity induced by early thermal and hydrological processes are relatively small and are expected to not significantly affect flow and transport properties. Chlorite precipitation was obtained in all simulation cases. A maximum of one percent volume fraction of chlorite could be formed, whose process may reduce swelling and sorption capacity of bentonite clay, affecting the performance of the repository. llitisation process was not obtained from the present simulations.

Xu, T.; Senger, R.; Finsterle, S.

2011-02-01T23:59:59.000Z

400

Geological Constraints on High-Level Nuclear Waste Disposal and their Relationship to Possible  

E-Print Network [OSTI]

nuclear energy. The U.S. government has recognized geologic disposal as a solution since the mid-1950s of plants produces about 20% of the United States' total energy consumption [EPA website, Nuclear Energy radioactivity produced in the process of electricity generation by nuclear fission [World Nuclear Association

Polly, David

Note: This page contains sample records for the topic "nuclear waste fund" 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.


401

Evaluation of nuclear facility decommissioning projects. Three Mile Island Unit 2. Radioactive waste and laundry shipments. Volume 9. Summary status report  

SciTech Connect (OSTI)

This document summarizes information concerning radioactive waste and laundry shipments from the Three Mile Island Nuclear Station Unit 2 to radioactive waste disposal sites and to protective clothing decontamination facilities (laundries) since the loss of coolant accident experienced on March 28, 1979. Data were collected from radioactive shipment records, summarized, and placed in a computerized data information retrieval/manipulation system which permits extraction of specific information. This report covers the period of April 9, 1979 to May 5, 1985. Included in this report are: waste disposal site locations, dose rates, curie content, waste description, container type and number, volumes and weights. This information is presented in two major categories: protective clothing (laundry) and radioactive waste. Each of the waste shipment reports is in chronological order.

Doerge, D. H.; Miller, R. L.; Scotti, K. S.

1986-05-01T23:59:59.000Z

402

Analyzing Nuclear Fuel Cycles from Isotopic Ratios of Waste Products Applicable to Measurement by Accelerator Mass Spectrometry  

SciTech Connect (OSTI)

An extensive study was conducted to determine isotopic ratios of nuclides in spent fuel that may be utilized to reveal historical characteristics of a nuclear reactor cycle. This forensic information is important to determine the origin of unknown nuclear waste. The distribution of isotopes in waste products provides information about a nuclear fuel cycle, even when the isotopes of uranium and plutonium are removed through chemical processing. Several different reactor cycles of the PWR, BWR, CANDU, and LMFBR were simulated for this work with the ORIGEN-ARP and ORIGEN 2.2 codes. The spent fuel nuclide concentrations of these reactors were analyzed to find the most informative isotopic ratios indicative of irradiation cycle length and reactor design. Special focus was given to long-lived and stable fission products that would be present many years after their creation. For such nuclides, mass spectrometry analysis methods often have better detection limits than classic gamma-ray spectroscopy. The isotopic ratios {sup 151}Sm/{sup 146}Sm, {sup 149}Sm/{sup 146}Sm, and {sup 244}Cm/{sup 246}Cm were found to be good indicators of fuel cycle length and are well suited for analysis by accelerator mass spectroscopy.

Biegalski, S R; Whitney, S M; Buchholz, B

2005-08-24T23:59:59.000Z

403

Spent nuclear fuel as a waste form for geologic disposal: Assessment and recommendations on data and modeling needs  

SciTech Connect (OSTI)

This study assesses the status of knowledge pertinent to evaluating the behavior of spent nuclear fuel as a waste form in geologic disposal systems and provides background information that can be used by the DOE to address the information needs that pertain to compliance with applicable standards and regulations. To achieve this objective, applicable federal regulations were reviewed, expected disposal environments were described, the status of spent-fuel modeling was summarized, and information regarding the characteristics and behavior of spent fuel was compiled. This compiled information was then evaluated from a performance modeling perspective to identify further information needs. A number of recommendations were made concerning information still needed to enhance understanding of spent-fuel behavior as a waste form in geologic repositories. 335 refs., 22 figs., 44 tabs.

Van Luik, A.E.; Apted, M.J.; Bailey, W.J.; Haberman, J.H.; Shade, J.S.; Guenther, R.E.; Serne, R.J.; Gilbert, E.R.; Peters, R.; Williford, R.E.

1987-09-01T23:59:59.000Z

404

Use of relativistic hadronic mechanics for the exact representation of nuclear magnetic moments and the prediction of new recycling of nuclear waste  

E-Print Network [OSTI]

We present a new realization of relativistic hadronic me- chanics and its underlying iso-Poincar'e symmetry specifically constructed for nuclear physics which: 1) permits the representation of nucleons as ex- tended, nonspherical and deformable charge distributions with alterable mag- netic moments yet conventional angular momentum and spin; 2) results to be a nonunitary ``completion'' of relativistic quantum mechanics much along the EPR argument; yet 3) is axiom-preserving, thus preserves conventional quantum laws and the axioms of the special relativity. We show that the proposed new formalism permits the apparently first exact representation of the total magnetic moments of new-body nuclei under conventional physical laws. We then point out that, if experimentally confirmed the alterability of the intrinsic characteristics of nucleons would imply new forms of recycling nuclear waste by the nuclear power plants in their own site, thus avoiding its transportation and storage in a (yet unidentified) dumping area. A number of possible, additional basic advances are also indicated, such as: new un- derstanding of nuclear forces with nowel nonlinear, nonlocal and nonunitary terms due to mutual penetrations of the hyperdense nucleons; consequential new models of nuclear structures; new magnetic confinement of the controlled fusion taking into account the possible alterability of the intrinsic magnetic moments of nucleons at the initiation of the fusion process; new sources of en- ergy based on subnuclear processes; and other possible advances. The paper ends with the proposal of three experiments, all essential for the continuation of scientific studies and all of basic character, relatively moderate cost and full feasibility in any nuclear physical laboratory.

Ruggero Maria Santilli

1997-04-09T23:59:59.000Z

405

Strategic Nuclear Research Collaboration - FY99 Annual Report  

SciTech Connect (OSTI)

The Idaho National Engineering and Environmental Laboratory (INEEL) has created the Strategic Nuclear Research Collaboration. The SNRC brings together some of America's finest laboratory and university nuclear researchers in a carefully focused research program intended to produce ''breakthrough'' solutions to the difficult issues of nuclear economics, safety, non-proliferation, and nuclear waste. This integrated program aims to address obstacles that stand in the way of nuclear power development in the US These include fuel cycle concerns related to waste and proliferation, the need for more efficient regulatory practices, and the high cost of constructing and operating nuclear power plants. Funded at an FY99 level of $2.58M, the SNRC is focusing the efforts of scientists and engineers from the INEEL and the Massachusetts Institute of Technology to solve complex nuclear energy challenges in a carefully chosen, integrated portfolio of research topics. The result of this collaboration will be research that serves as a catalyst for future direct-funded nuclear research and technology development and which preserves and enhances the INEEL's role as America's leading national laboratory for nuclear power research. In its first year, the SNRC has focused on four research projects each of which address one or more of the four issues facing further nuclear power development (economics, safety, waste disposition and proliferation-resistance). This Annual Report describes technical work and accomplishments during the first year of the SNRC's existence.

T. J. Leahy

1999-07-01T23:59:59.000Z

406

The Application of Performance Assessment to Make Regulatory and Operational Changes in an Operating Nuclear Waste Repository  

SciTech Connect (OSTI)

This paper describes how performance assessment (PA) is used to support changes to the regulatory basis of the Waste Isolation Pilot Plant (WIPP). The WIPP, located near Carlsbad, New Mexico is operated by the U.S. Department of Energy (DOE) as the nation's only deep geologic repository for the disposal of transuranic nuclear waste. In 1998, the Environmental Protection Agency (EPA) certified that the WIPP met the performance requirements of 40 CFR Part 191, Environmental Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes. A PA analysis of long term (10,000 year) repository performance successfully demonstrated that the probability and consequences of potential long-term releases of radionuclides to the accessible environment would be well below the established limits. These results were key in obtaining WIPP's initial certification, allowing the first shipment and disposal of nuclear waste in March of 1999. As disposal operations have taken place over the last eight years, changes have been identified in the regulatory and operational realms of the facility that would enhance waste disposal efficiency. Such changes, however, cannot be made without prior consent of the EPA. Therefore, changes planned by the DOE must be thoroughly described and supported by varying degrees of the same type of analyses that were conducted to demonstrate the WIPP's containment capabilities as presented in the initial compliance application submitted to EPA in 1996. Such analyses are used to identify the impacts or benefits of implementing the planned change. The DOE has successfully used performance assessment analyses for the approval of changes such as: 1) the disposal of super-compacted waste forms, and; 2) the adoption of new parameters and modeling assumptions In some cases the planned changes are simpler in nature than those listed above, and therefore only require targeted or simplified PA analyses to demonstrate the effect on performance. Targeted analyses have been used to successfully gain approval of the following: 1) a reduction in the amount of magnesium oxide (MgO) chemical buffer backfill that must be emplaced in the repository 2) a change in the repository mining/disposal horizon In addition to these approved changes, the DOE has used PA analyses to support the following planned change requests that await EPA's approval: 1) panel closure redesign 2) further reduction in the MgO-to-waste ratio Finally, this paper will discuss some of the changes that the DOE is currently preparing and plans to submit to the EPA for approval in the near future. This paper will describe how a set of analytical tools initially used to open the WIPP continues to have a role in making the repository more efficient and adaptable as variations in waste streams, operational demands, and other dynamic forces change the operating environment over time. (authors)

Patterson, R. [Department of Energy, Carlsbad Field Office, Carlsbad, NM (United States); Kirkes, R. [John Hart and Associates, P.A., Albuquerque, NM (United States)

2008-07-01T23:59:59.000Z

407

Removal of pertechnetate from simulated nuclear waste streams using supported zerovalent iron  

E-Print Network [OSTI]

be to use a technetium ion- exchange or sorbent materialtechnetium problem is to remove pertechnetate from the waste using commercially available polymeric ion exchangetechnetium-containing eluate that would be generated in removing pertechnetate from the commercial ion-exchange

Darab, John

2008-01-01T23:59:59.000Z

408

GEOTECHNICAL ASSESSMENT AND INSTRUMENTATION NEEDS FOR NUCLEAR WASTE ISOLATION IN CRYSTALLINE AND ARGILLACEOUS ROCKS SYMPOSIUM  

E-Print Network [OSTI]

active waste storage glass. AERE-R 8706, May. Hall, A. R. ,from dispersed blocks, AERE-R 8763, June. Holdoway, M. J. (to the FINGAL process. AERE-R 6418, May. Jenkins, I. L. (

Authors, Various

2011-01-01T23:59:59.000Z

409

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

410

Removal of pertechnetate from simulated nuclear waste streams using supported zerovalent iron  

E-Print Network [OSTI]

3. ?? TM/JD 2. Westinghouse Hanford Co. , Report WHC-SD-WM-Department of Energys Hanford Site: (1) the direct removaltypical of those found in Hanford tank waste; and (2) the

Darab, John

2008-01-01T23:59:59.000Z

411

Removal of Radionuclides from Waste Water at Fukushima Daiichi Nuclear Power Plant: Desalination and Adsorption Methods - 13126  

SciTech Connect (OSTI)

Waste water containing high levels of radionuclides due to the Fukushima Daiichi Nuclear Power Plant accident, has been treated by the adsorption removal and reverse-osmosis (RO) desalination to allow water re-use for cooling the reactors. Radionuclides in the waste water are collected in the adsorbent medium and the RO concentrate (RO brine) in the water treatment system currently operated at the Fukushima Daiichi site. In this paper, we have studied the behavior of radionuclides in the presently applied RO desalination system and the removal of radionuclides in possible additional adsorption systems for the Fukushima Daiichi waste water treatment. Regarding the RO desalination system, decontamination factors (DFs) of the elements present in the waste water were obtained by lab-scale testing using an RO unit and simulated waste water with non-radioactive elements. The results of the lab-scale testing using representative elements showed that the DF for each element depended on its hydrated ionic radius: the larger the hydrated ionic radius of the element, the higher its DF is. Thus, the DF of each element in the waste water could be estimated based on its hydrated ionic radius. For the adsorption system to remove radionuclides more effectively, we studied adsorption behavior of typical elements, such as radioactive cesium and strontium, by various kinds of adsorbents using batch and column testing. We used batch testing to measure distribution coefficients (K{sub d}s) for cesium and strontium onto adsorbents under different brine concentrations that simulated waste water conditions at the Fukushima Daiichi site. For cesium adsorbents, K{sub d}s with different dependency on the brine concentration were observed based on the mechanism of cesium adsorption. As for strontium, K{sub d}s decreased as the brine concentration increased for any adsorbents which adsorbed strontium by intercalation and by ion exchange. The adsorbent titanium oxide had higher K{sub d}s and it was used for the column testing to obtain breakthrough curves under various conditions of pH and brine concentration. The breakthrough point had a dependency on pH and the brine concentration. We found that when the pH was higher or the brine concentration was lower, the longer it took to reach the breakthrough point. The inhibition of strontium adsorption by alkali earth metals would be diminished for conditions of higher pH and lower brine concentration. (authors)

Kani, Yuko; Kamosida, Mamoru; Watanabe, Daisuke [Hitachi Research Laboratory, Hitachi, Ltd., 7-2-1 Omika-cho, Hitachi, Ibaraki, 319-1221 (Japan)] [Hitachi Research Laboratory, Hitachi, Ltd., 7-2-1 Omika-cho, Hitachi, Ibaraki, 319-1221 (Japan); Asano, Takashi; Tamata, Shin [Hitachi Works, Hitachi-GE Nuclear Energy, Ltd. (Japan)] [Hitachi Works, Hitachi-GE Nuclear Energy, Ltd. (Japan)

2013-07-01T23:59:59.000Z

412

Development of Vitrification Process and Glass Formulation for Nuclear Waste Conditioning  

SciTech Connect (OSTI)

The vitrification of high-level waste is the internationally recognized standard to minimize the impact to the environment resulting from waste disposal as well as to minimize the volume of conditioned waste to be disposed of. COGEMA has been vitrifying high-level waste industrially for over 20 years and is currently operating three commercial vitrification facilities based on a hot metal crucible technology, with outstanding records of safety, reliability and product quality. To further increase the performance of vitrification facilities, CEA and COGEMA have been developing the cold crucible melter technology since the beginning of the 1980s. This type of melter is characterized by a virtually unlimited equipment service life and a great flexibility in dealing with various types of waste and allowing development of high temperature matrices. In complement of and in parallel with the vitrification process, a glass formulation methodology has been developed by the CEA in order to tailor matrices for the wastes to be conditioned while providing the best adaptation to the processing technology. The development of a glass formulation is a trade-off between material properties and qualities, technical feasibility, and disposal safety criteria. It involves non-radioactive and radioactive laboratories in order to achieve a comprehensive matrix qualification. Several glasses and glass ceramics have thus been studied by the CEA to be compliant with industrial needs and waste characteristics: glasses or other matrices for a large spectrum of fission products, or for high contents of specifics elements such as sodium, phosphate, iron, molybdenum, or actinides. New glasses or glass-ceramics designed to minimize the final wasteform volume for solutions produced during the reprocessing of high burnup fuels or to treat legacy wastes are now under development and take benefit from the latest CEA hot-laboratories and technology development. The paper presents the CEA state-of-the-art in developing matrices or glasses and provides several examples.

Petitjean, V.; Fillet, C.; Boen, R.; Veyer, C.; Flament, T.

2002-02-26T23:59:59.000Z

413

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

SciTech Connect (OSTI)

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

Not Available

1990-03-01T23:59:59.000Z

414

ASME AG-1 REQUIREMENT EXEMPTION JUSTIFICATIONS FOR VENTILATION SYSTEMS AT NUCLEAR WASTE STORAGE TANKS AT THE HANFORD SITE  

SciTech Connect (OSTI)

Washington State Department of Health regulations require compliance with the American Society of Mechanical Engineers (ASME) AG-1, ''Code on Nuclear Air and Gas Treatment,'' for all new radioactive air emission units. As a result, these requirements have been applied to systems that ventilate the radioactive waste storage tanks in the tank farm facilities on the U.S. Department of Energy's Hanford Site. ASME AG-1 is applied as a regulatory constraint to waste tank ventilation systems at the Hanford Site, even though the code was not intended for these systems. An assessment was performed to identify which requirements should be exempted for waste tank ventilation systems. The technical justifications for requirement exemptions were prepared and presented to the regulator. The technical justifications were documented so that select requirement exemptions for specific projects and systems can be sought through the regulator's permitting process. This paper presents the rationale for attempting to receive requirement exemption and presents examples of the technical justifications that form the basis for these exemptions.

GUSTAVSON, R.D.

2004-09-03T23:59:59.000Z

415

Materials Characterization Center workshop on leaching mechanisms of nuclear waste forms, May 19-21, 1982, Gaithersburg, Maryland. Summary report  

SciTech Connect (OSTI)

This is a report of the second workshop on the leaching mechanism of nuclear waste forms, which was held at Geithersburg, Maryland, May 19-21, 1982. The first session of the workshop was devoted to progress reports by participants in the leaching mechanisms program. These progress reports, as prepared by the participants, are given in Section 3.0. The goal of the remainder of the workshop was to exchange information on the development of repository-relevant leach testing techniques, often called interactions testing. To this end, a wide spectrum of investigators, many of whose work is sponsored by DOE's Nuclear Waste Terminal Storage (NWTS) project, made presentations at the workshop. These presentations were a significant and beneficial part of the workshop and are summarized in Sections 4.0, 5.0 and 6.0 according to the workshop agenda topics. In many cases, the presenters provided a written version of their presentation which has been included verbatim; in the other cases, the workshop chairman has supplied a brief synopsis. Twenty-one papers have been abstracted and indexed for inclusion in the data base.

Mendel, J.E. (comp.)

1982-08-01T23:59:59.000Z

416

Status of iron-enriched basalt as a medium for nuclear waste immobilization: a report by an independent peer review panel  

SciTech Connect (OSTI)

The purpose of the Peer Review Panel was to provide an independent review by experts in nuclear waste processing and materials on the adequacy of the existing data base for the iron-enriched basalt waste form developed by EG and G Idaho, and to evaluate the broad range of proposed applications for this waste form. It was not the purpose of this review to specifically rank iron-enriched basalt against other nuclear waste forms. It was the concensus of the Peer Review Panel that the concept, experimental research, and identification of potential applications of the iron-enriched basalt waste form were of high quality. Iron-enriched basalt is a primarily ceramic waste form with a residual glass phase. It has a broad range of composition, permitting the incorporation of a wide variety of nuclear wastes. The product has good mechanical strength and produces very low quantities of respirable particles under impact conditions. Matrix dissolution rates under neutral pH conditions are comparable to or lower than those of borosilicate glass. In the area of waste form characterization, the Panel recommended additional static and dynamic leaching tests as a function of pH and CO/sub 2/ in solution, and in brine solutions of varying composition. The panel also recommended that unprocessed transuranic (TRU) wastes be subjected to leach tests. Large-scale iron-enriched basalt castings in which the grain growth was uncontrolled have been observed to be less durable than controlled-grain-growth laboratory-scale castings. Therefore, the Panel also recommended leaching tests as a function of microstructure to determine ranges of acceptable microstructure. In the area of the IEB production process, the Panel recommended a variety of laboratory-scale and pilot plant-scale research.

Palmour, H. III; Dosch, R.G.; Macedo, P.B.; Machiels, A.J.; Owen, D.E.

1981-09-01T23:59:59.000Z

417

Compression and immersion tests and leaching of radionuclides, stable metals, and chelating agents from cement-solidified decontamination waste collected from nuclear power stations  

SciTech Connect (OSTI)

A study was performed for the Nuclear Regulatory Commission (NRC) to evaluate structural stability and leachability of radionuclides, stable metals, and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from seven commercial boiling water reactors and one pressurized water reactor. The decontamination methods used at the reactors were the Can-Decon, AP/Citrox, Dow NS-1, and LOMI processes. Samples of untreated resin waste and solidified waste forms were subjected to immersion and compressive strength testing. Some waste-form samples were leach-tested using simulated groundwaters and simulated seawater for comparison with the deionized water tests that are normally performed to assess waste-form leachability. This report presents the results of these tests and assesses the effects of the various decontamination methods, waste form formulations, leachant chemical compositions, and pH of the leachant on the structural stability and leachability of the waste forms. Results indicate that releases from intact and degraded waste forms are similar and that the behavior of some radionuclides such as {sup 55}Fe, {sup 60}Co, and {sup 99}Tc were similar. In addition, the leachability indexes are greater than 6.0, which meets the requirement in the NRC`s ``Technical Position on Waste Form,`` Revision 1.

Akers, D.W.; Kraft, N.C.; Mandler, J.W. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1994-06-01T23:59:59.000Z

418

Qualitative and Quantitative Assessment of Nuclear Materials Contained in High-Activity Waste Arising from the Operations at the 'SHELTER' Facility  

SciTech Connect (OSTI)

As a result of the nuclear accident at the Chernobyl NPP in 1986, the explosion dispeesed nuclear materials contained in the nuclear fuel of the reactor core over the destroyed facilities at Unit No. 4 and over the territory immediately adjacent to the destroyed unit. The debris was buried under the Cascade Wall. Nuclear materials at the SHELTER can be characterized as spent nuclear fuel, fresh fuel assemblies (including fuel assemblies with damaged geometry and integrity, and individual fuel elements), core fragments of the Chernobyl NPP Unit No. 4, finely-dispersed fuel (powder/dust), uranium and plutonium compounds in water solutions, and lava-like nuclear fuel-containing masses. The new safe confinement (NSC) is a facility designed to enclose the Chernobyl NPP Unit No. 4 destroyed by the accident. Construction of the NSC involves excavating operations, which are continuously monitored including for the level of radiation. The findings of such monitoring at the SHELTER site will allow us to characterize the recovered radioactive waste. When a process material categorized as high activity waste (HAW) is detected the following HLW management operations should be involved: HLW collection; HLW fragmentation (if appropriate); loading HAW into the primary package KT-0.2; loading the primary package filled with HAW into the transportation cask KTZV-0.2; and storing the cask in temporary storage facilities for high-level solid waste. The CDAS system is a system of 3He tubes for neutron coincidence counting, and is designed to measure the percentage ratio of specific nuclear materials in a 200-liter drum containing nuclear material intermixed with a matrix. The CDAS consists of panels with helium counter tubes and a polyethylene moderator. The panels are configured to allow one to position a waste-containing drum and a drum manipulator. The system operates on the add a source basis using a small Cf-252 source to identify irregularities in the matrix during an assay. The platform with the source is placed under the measurement chamber. The platform with the source material is moved under the measurement chamber. The design allows one to move the platform with the source in and out, thus moving the drum. The CDAS system and radioactive waste containers have been built. For each drum filled with waste two individual measurements (passive/active) will be made. This paper briefly describes the work carried out to assess qualitatively and quantitatively the nuclear materials contained in high-level waste at the SHELTER facility. These efforts substantially increased nuclear safety and security at the facility.

Cherkas, Dmytro

2011-10-01T23:59:59.000Z

419

Conservation Trust Funds  

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than $50,000, such as Suriname's Forest People's Fund, "levels. 94 For example, the Suriname trust fund operates at

Guerin-McManus, Marianne

2001-01-01T23:59:59.000Z

420

Radiation effects in moist-air systems and the influence of radiolytic product formation on nuclear waste glass corrosion  

SciTech Connect (OSTI)

Ionizing radiation may affect the performance of glass in an unsaturated repository site by interacting with air, water vapor, or liquid water to produce a variety of radiolytic products. Tests were conducted to examine the effects of radiolysis under high gas/liquid ratios. Results indicate that nitrate is the predominant radiolytic product produced following both gamma and alpha radiation exposure, with lesser amounts of nitrite and carboxylic acids. The formation of nitrogen acids during exposure to long-lived, alpha-particle-emitting transuranic elements indicates that these acids may play a role in influencing nuclear waste form reactions in a long-term unsaturated disposal scenario. Experiments were also conducted with samples that simulate the composition of Savannah River Plant nuclear waste glasses. Radiolytic product formation in batch tests (340 m{sup {minus}1}, 90 C) resulted in a small increase in the release rates of many glass components, such as alkali and alkaline earth elements, although silicon and uranium release rates were slightly reduced indicating an overall beneficial effect of radiation on waste form stability. The radiolytic acids increased the rate of ion exchange between the glass and the thin film of condensate, resulting in accelerated corrosion rates for the glass. The paragenetic sequence of alteration phases formed on both the irradiated and nonirradiated glass samples reacted in the vapor hydration tests matches closely with those developed during volcanic glass alteration in naturally occurring saline-alkaline lake systems. This correspondence suggests that the high temperatures used in these tests have not changed the underlying glass reaction mechanism relate to that which controls glass reactions under ambient surficial conditions.

Wronkiewicz, D.J.; Bates, J.K.; Buck, E.C.; Hoh, J.C.; Emery, J.W. [Argonne National Lab., IL (United States). Chemical Technology Div.; Wang, L.M. [Univ. of New Mexico, Albuquerque, NM (United States). Dept. of Geology

1997-07-01T23:59:59.000Z

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421

Volume reduction/solidification of liquid radioactive waste using bitumen at Ontario hydro`s Bruce nuclear generating station {open_quotes}A{close_quotes}  

SciTech Connect (OSTI)

Ontario Hydro at the Bruce Nuclear Generating Station {open_quotes}A{close_quotes} has undertaken a program to render the station`s liquid radioactive waste suitable for discharge to Lake Huron by removing sufficient radiological and chemical contaminants from five different plant waste streams. The contaminants will be immobilized and stored at on-site radioactive waste storage facilities and the purified streams will be discharged. The discharge targets established by Ontario Hydro are set well below the limits established by the Ontario Ministry of Environment (MOE) and are based on the Best Available Technology Economically Achievable Approach (B.A.T.E.A.). ADTECHS Corporation has been selected by Ontario Hydro to provide volume reduction/solidification technology for one of the five waste streams. The system will dry and immobilize the contaminants from a liquid waste stream in emulsified asphalt using thin film evaporation technology.

Day, J.E.; Baker, R.L. [ADTECHS Corporation, Herndon, VA (United States)

1994-12-31T23:59:59.000Z

422

Removal of pertechnetate from simulated nuclear waste streams using supported zerovalent iron  

SciTech Connect (OSTI)

The application of nanoparticles of predominantly zerovalent iron (nanoiron), either unsupported or supported, to the separation and reduction of pertechnetate anions (TcO4-) from complex waste mixtures was investigated as an alternative approach to current waste-processing schemes. Although applicable to pertechnetate-containing waste streams in general, the research discussed here was directed at two specific potential applications at the U.S. Department of Energy's Hanford Site: (1) the direct removal of pertechnetate from highly alkaline solutions, typical of those found in Hanford tank waste, and (2) the removal of dilute pertechnetate from near-neutral solutions, typical of the eluate streams from commercial organic ion-exchange resins that may be used to remediate Hanford tank wastes. It was envisioned that both applications would involve the subsequent encapsulation of the loaded sorbent material into a separate waste form. A high surface area (>200 M2/g) base-stable, nanocrystalline zirconia was used as a support for nanoiron for tests with highly alkaline solutions, while a silica gel support was used for tests with near-neutral solutions. It was shown that after 24 h of contact time, the high surface area zirconia supported nanoiron sorbent removed about 50percent (K-d = 370 L/kg) of the pertechnetate from a pH 14 tank waste simulant containing 0.51 mM TCO4- and large concentrations of Na+, OH-, NO3-, and CO32- for a phase ratio of 360 L simulant per kg of sorbent. It was also shown that after 18 h of contact time, the silica-supported nanoiron removed>95percent pertechnetate from a neutral pH eluate simulant containing 0.076 mM TcO4_ for a phase ratio of 290 L/kg. It was determined that in all cases, nanoiron reduced the Tc(VII) to Tc(IV), or possibly to Tc(V), through a redox reaction. Finally, it was demonstrated that a mixture of 20 mass percent of the solid reaction products obtained from contacting zirconia- supported nanoiron with an alkaline waste solution containing Re(VII), a surrogate for Tc(VII), with 80 mass percent alkali borosilicate based frit heat-treated at 700 degrees C for 4 h sintered into an easily handled glass composite waste form.

Darab, John; Amonette, Alexandra; Burke, Deborah; Orr, Robert; Ponder, Sherman; Schrick, Bettina; Mallouk, Thomas; Lukens, Wayne; Caulder, Dana; Shuh, David

2007-07-11T23:59:59.000Z

423

The Decline and Death of Nuclear Power  

E-Print Network [OSTI]

funding, causing nuclear power to simply fall off the energyor ambivalent about nuclear power to firmly against it.

Melville, Jonathan

2013-01-01T23:59:59.000Z

424

E-Print Network 3.0 - affecting nuclear waste Sample Search Results  

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

studies (whose results could affect industry profits)--may create a COI, a risk... ORIGINAL PAPER Climate Change, Nuclear Economics, and Conflicts of Interest...

425

Organic components of nuclear wastes and their potential for altering radionuclide distribution when released to soil  

SciTech Connect (OSTI)

Normal waste processing at the Hanford operations requires the use of many organic materials, chiefly in the form of complexing agents and diluents. These organic materials and their chemical and radiolytic degradation products, have potential for complexing fission products and transuranium elements, both in the waste streams and upon infiltration into soil, perhaps influencing future sorption or migration of the nuclides. Particular complexation characteristics of various nuclides which constitute the major fission products, long-lived isotopes, and the most mobile in radioactive wastes are discussed briefly with regards to their anticipated sorption or mobility in soils. Included in the discussion are Am, Sb, Ce, Cs, Co, Cm, Eu, I, Np, Pm, Pu, Ra, Ru, Sr, Tc, U, and Zr. 107 references.

McFadden, K.M.

1980-08-01T23:59:59.000Z

426

Stakeholder Transportation Scorecard: Reviewing Nevada's Recommendations for Enhancing the Safety and Security of Nuclear Waste Shipments - 13518  

SciTech Connect (OSTI)

As a primary stakeholder in the Yucca Mountain program, the state of Nevada has spent three decades examining and considering national policy regarding spent nuclear fuel and high-level radioactive waste transportation. During this time, Nevada has identified 10 issues it believes are critical to ensuring the safety and security of any spent nuclear fuel transportation program, and achieving public acceptance. These recommendations are: 1) Ship the oldest fuel first; 2) Ship mostly by rail; 3) Use dual-purpose (transportable storage) casks; 4) Use dedicated trains for rail shipments; 5) Implement a full-scale cask testing program; 6) Utilize a National Environmental Policy Act (NEPA) process for the selection of a new rail spur to the proposed repository site; 7) Implement the Western Interstate Energy Board (WIEB) 'straw man' process for route selection; 8) Implement Section 180C assistance to affected States, Tribes and localities through rulemaking; 9) Adopt safety and security regulatory enhancements proposed states; and 10) Address stakeholder concerns about terrorism and sabotage. This paper describes Nevada's proposals in detail and examines their current status. The paper describes the various forums and methods by which Nevada has presented its arguments and sought to influence national policy. As of 2012, most of Nevada's recommendations have been adopted in one form or another, although not yet implemented. If implemented in a future nuclear waste program, the State of Nevada believes these recommendations would form the basis for a successful national transportation plan for shipments to a geologic repository and/or centralized interim storage facility. (authors)

Dilger, Fred C. [Black Mountain Research, Henderson, NV 81012 (United States)] [Black Mountain Research, Henderson, NV 81012 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge, CA 91330 (United States)] [Department of Sociology, California State University, Northridge, CA 91330 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects, Carson City, NV 80906 (United States)] [State of Nevada Agency for Nuclear Projects, Carson City, NV 80906 (United States)

2013-07-01T23:59:59.000Z

427

Modelling and Numerical Simulation of Gas Migration in a Nuclear Waste Repository  

E-Print Network [OSTI]

We present a compositional compressible two-phase, liquid and gas, flow model for numerical simulations of hydrogen migration in deep geological radioactive waste repository. This model includes capillary effects and the gas diffusivity. The choice of the main variables in this model, Total or Dissolved Hydrogen Mass Concentration and Liquid Pressure, leads to a unique and consistent formulation of the gas phase appearance and disappearance. After introducing this model, we show computational evidences of its adequacy to simulate gas phase appearance and disappearance in different situations typical of underground radioactive waste repository.

Bourgeat, Alain; Smai, Farid

2010-01-01T23:59:59.000Z

428

Radioactive and chemotoxic wastes: Only radioactive wastes?  

SciTech Connect (OSTI)

Radioactive waste arising from Italian Nuclear Power Plants and Research Centers, classified as 1st and 2nd Category wastes, are managed only as radioactive wastes following the Technical Guide No. 26 issued by the Italian Regulatory Body: ENEA DISP on 1987. A very important Regulatory Regime revision for Italian Nuclear Activities started at the end of 1991. This paper considers the need to develop a new strategy dedicated to mixed waste in line with current international trends.

Eletti, G.F.; Tocci, M. [ENEA DISP, Rome (Italy)

1993-12-31T23:59:59.000Z

429

Geochemical assessment of nuclear waste isolation. Report of activities during fiscal year 1982  

SciTech Connect (OSTI)

The status of the following investigations is reported: canister/overpack-backfill chemical interactions and mechanisms; backfill and near-field host rock chemical interactions mechanisms; far-field host rock geochemical interactions; verification and improvement of predictive algorithms for radionuclide migration; and geologic systems as analogues for long-term radioactive waste isolation.

Not Available

1983-07-01T23:59:59.000Z

430

The development of a management strategy for interim storage and final disposal of nuclear wastes  

SciTech Connect (OSTI)

The overall waste management strategy for alternative routes from reactor to final disposal, including dry interim storage, is discussed. Within the framework of a preliminary structure plan possible technical solutions must be investigated, and with sufficient relevant information available the future progress of the project, can be addressed on the base of a decision analysis.

Engelmann, H.J.; Popp, F.W. [Deutsche Gesellschaft zum Bau and Betrieb von Endglagern fuer Abfallostofe mbH, Peine (Germany); Arntzen, P.; Botzem, W. [NUKEM GmbH, Alzenau (Germany); Soucek, B. [Czech Power Board, Prague (Czech Republic)

1993-12-31T23:59:59.000Z

431

Preliminary neutronics design of china lead-alloy cooled demonstration reactor (CLEAR-III) for nuclear waste transmutation  

SciTech Connect (OSTI)

China Lead-Alloy cooled Demonstration Reactor (CLEAR-III), which is the concept of lead-bismuth cooled accelerator driven sub-critical reactor for nuclear waste transmutation, was proposed and designed by FDS team in China. In this study, preliminary neutronics design studies have primarily focused on three important performance parameters including Transmutation Support Ratio (TSR), effective multiplication factor and blanket thermal power. The constraint parameters, such as power peaking factor a