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Note: This page contains sample records for the topic "nwpa nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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1

RS-NWPA [National Waste Policy Act] | Department of Energy  

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

Waste Policy Act Basalt Waste & Salt River projects RS-NWPA National Waste Policy Act More Documents & Publications OFFICE OF PROCUREMENT & ASSISTANCE MANAGEMENT...

2

Nuclear Waste Fund fee adequacy: An assessment  

SciTech Connect

The purpose of this report is to present the Department of Energy`s (the Department) analysis of the adequacy of the 1.00 mill per kilowatt-hour (kWh) fee being paid by the utilities generating nuclear power for the permanent disposal of their spent nuclear fuel (SNF). In accordance with the Nuclear Waste Policy Act (NWPA), the SNF would be disposed of in a geologic repository to be developed by the Department. An annual analysis of the fee`s adequacy is required by the NWPA.

NONE

1990-11-01T23:59:59.000Z

3

Nuclear Waste Assessment System for Technical Evaluation (NUWASTE)  

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

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

4

2011 SECRETARIAL DETERMINATION OF THE ADEQUACY OF THE NUCLEAR WASTE FUND  

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

2011 SECRETARIAL DETERMINATION OF THE ADEQUACY OF THE NUCLEAR WASTE 2011 SECRETARIAL DETERMINATION OF THE ADEQUACY OF THE NUCLEAR WASTE FUND FEE 2011 SECRETARIAL DETERMINATION OF THE ADEQUACY OF THE NUCLEAR WASTE FUND FEE As required by the Nuclear Waste Policy Act of 1982 (NWPA), DOE has completed its annual review of the adequacy of the Nuclear Waste Fund fee. Based on that review, the Office of Standard Contract Management has determined that there is no basis to propose an adjustment to the fee to Congress. The Secretary of Energy has adopted and approved this determination. As a result, the fee will remain at the amount specified in the NWPA pending the next annual review. The Secretary's determination is available here. 2011 Secretarial Fee Adequacy Determination.PDF More Documents & Publications Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee

5

2011 SECRETARIAL DETERMINATION OF THE ADEQUACY OF THE NUCLEAR WASTE FUND  

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

2011 SECRETARIAL DETERMINATION OF THE ADEQUACY OF THE NUCLEAR WASTE 2011 SECRETARIAL DETERMINATION OF THE ADEQUACY OF THE NUCLEAR WASTE FUND FEE 2011 SECRETARIAL DETERMINATION OF THE ADEQUACY OF THE NUCLEAR WASTE FUND FEE As required by the Nuclear Waste Policy Act of 1982 (NWPA), DOE has completed its annual review of the adequacy of the Nuclear Waste Fund fee. Based on that review, the Office of Standard Contract Management has determined that there is no basis to propose an adjustment to the fee to Congress. The Secretary of Energy has adopted and approved this determination. As a result, the fee will remain at the amount specified in the NWPA pending the next annual review. The Secretary's determination is available here. 2011 Secretarial Fee Adequacy Determination.PDF More Documents & Publications Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee

6

DOE Completes Annual Determination of the Adequacy of the Nuclear Waste  

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

Completes Annual Determination of the Adequacy of the Nuclear Completes Annual Determination of the Adequacy of the Nuclear Waste Fund Fee DOE Completes Annual Determination of the Adequacy of the Nuclear Waste Fund Fee November 2, 2010 - 7:41pm Addthis As required by the Nuclear Waste Policy Act of 1982 (NWPA), DOE has completed its annual review of the adequacy of the Nuclear Waste Fund fee. Based on that review, the Office of Standard Contract Management has determined that there is no basis to propose an adjustment to the fee to Congress. The Secretary of Energy has adopted and approved this determination. As a result, the fee will remain at the amount specified in the NWPA pending the next annual review. The Secretary's determination is available here. Addthis Related Articles GC Commits to Transparency on Nuclear Waste Fund Fee Adequacy Decisions

7

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

Science Conference Proceedings (OSTI)

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

2010-05-27T23:59:59.000Z

8

RS-NWPA [National Waste Policy Act]  

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

Minning (OCRWM) Minning (OCRWM) I 586-4349 Ed Nugent DOE 586-3288 6. C k K T : F I C A T E O F A G E N C Y R E P R E S E N T A T I V E r - . . REQUEST FOR RECORDS DISPOSITION AUTHORITY (See tnstructions on reverse) GENERAL SERVICES ADMINISTRATION NATIONAL ARCHIVES AND RECORDS SERVICE. WASHINGTON, DC 20408 i F ROha (AECnc? or rstcblishmcnt) a % - e ~ - - - Office of C i v i l f a n R a d i o a c t i v e Naste Mar -gement - 3 M I N O R S U B D l V l S l O N - - 6 N A M E O F P E R S O N W l T P W H O M T O C O N F E R I 5. TELEPHONE E X T . I hereby cert~fy that I am authorized to act for thls agency In ms.ter: perta~ning to the disposal of the aoency's record: that the records proposed for drsposal In this Request of ' 2 ' page(s) are not now needed for the business o f th agency or w ~ i l not be needed after ;he retention per~ods speclf~ed; and that written concurrence from the Gener:

9

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

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

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

10

Nuclear Waste Management  

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

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

11

Immobilization of Nuclear Wastes  

Science Conference Proceedings (OSTI)

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

12

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

National Nuclear Security Administration (NNSA)

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

13

Chapter 19 - Nuclear Waste Fund  

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

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

14

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

15

Pioneering Nuclear Waste Disposal  

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

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

16

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

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

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

17

Development of Cementitious Waste Forms for Nuclear Waste ...  

Science Conference Proceedings (OSTI)

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

18

Swedish nuclear waste efforts  

SciTech Connect

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

Rydberg, J.

1981-09-01T23:59:59.000Z

19

WEB RESOURCE: Nuclear Waste Disposal  

Science Conference Proceedings (OSTI)

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

20

Pioneering Nuclear Waste Disposal  

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

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

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


21

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

Science Conference Proceedings (OSTI)

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

2010-06-29T23:59:59.000Z

22

Pioneering Nuclear Waste Disposal  

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

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

23

Site characterization progress report: Yucca Mountain, Nevada, April 1, 1990--September 30, 1990, Number 3; Nuclear Waste Policy Act (Section 113)  

SciTech Connect

In accordance with the requirements of Section 113(b)(3) of the Nuclear Waste Policy Act of 1982 (NWPA), as amended, the US Department of Energy has prepared this report on the progress of site characterization activities at Yucca Mountain, Nevada, for the period April 1 through September 30, 1990. This report is the third of a series of reports that are issued at intervals of approximately six months during site characterization. The report covers a number of new initiatives to improve the effectiveness of the site characterization program and covers continued efforts related to preparatory activities, study plans, and performance assessment. 85 refs., 2 figs., 3 tabs.

NONE

1991-03-01T23:59:59.000Z

24

NWPA-830G U.S. DEPARTMENT OF ENERGY  

U.S. Energy Information Administration (EIA) Indexed Site

U.S. DEPARTMENT OF ENERGY OMB No. 1901-0260 (Revised 03/12) Germantown, MD 20874-1290 Expires: 3-31-2016 Burden: 5 Hours Appendix G -Standard Remittance Advice for Payment of Fees This information is being collected under mandatory authorities vested in the U.S. Department of Energy under Public Law 97-425. Late filing, failure to file or otherwise comply with instructions provided may result in interest penalties as provided by Article VIII C of the Standard Contract of Spent Nuclear Fuel and/or High Level Radioactive Waste. For information concerning confidentiality of information see Item 6 of the instructions. Public reporting burden for this collection of information is estimated to average 5 hours per response, including

25

Waste canister for storage of nuclear wastes  

DOE Patents (OSTI)

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

Duffy, James B. (Fullerton, CA)

1977-01-01T23:59:59.000Z

26

Nuclear Waste Policy Act | Department of Energy  

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

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

27

Chemotoxicity of nuclear waste repositories  

Science Conference Proceedings (OSTI)

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

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

1992-03-01T23:59:59.000Z

28

Small businesses selected for nuclear waste services  

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

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

29

UNITED STATES DEPARTMENT OF ENERGY OFFICE OF CIVILIAN RADIOACTIVE WASTE  

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

UNITED STATES DEPARTMENT OF ENERGY OFFICE OF CIVILIAN RADIOACTIVE UNITED STATES DEPARTMENT OF ENERGY OFFICE OF CIVILIAN RADIOACTIVE WASTE MANAGEMENT Annual Financial Report Years Ended September 30, 2009 and 2008 UNITED STATES DEPARTMENT OF ENERGY OFFICE OF CIVILIAN RADIOACTIVE WASTE MANAGEMENT Annual Financial Report Years Ended September 30, 2009 and 2008 As required by Section 304(c) of the Nuclear Waste Policy Act (NWPA) of 1982, as amended, Public Law 97-425, the following document is the United States Department of Energy's (DOE) Office of Civilian Radioactive Waste Management's Annual Financial Report for the years ended September 30, 2009 and 2008 as required by Section 302(c)(l) ofNWPA. The information in this report is current only as of September 30, 2009, and does not reflect actions or changes that have occurred since then.

30

Nuclear Waste Management using Electrometallurgical Technology - Nuclear  

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

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

31

Nuclear waste solidification  

DOE Patents (OSTI)

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

Bjorklund, William J. (Richland, WA)

1977-01-01T23:59:59.000Z

32

Nuclear Waste Management Policy in France  

Science Conference Proceedings (OSTI)

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

Jean F. Lefevre

33

Advancements in nuclear waste assay.  

E-Print Network (OSTI)

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

Curtis, Deborah Claire

2008-01-01T23:59:59.000Z

34

Pioneering Nuclear Waste Disposal  

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

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

35

Guidelines for Fabrication, Examination, Testing and Oversight of Spent Nuclear Fuel Dry Storage Systems  

Science Conference Proceedings (OSTI)

The Nuclear Waste Policy Act (NWPA) of 1982 and subsequent amendments require the U. S. Department of Energy (DOE) to receive and be responsible for disposal of spent commercial nuclear power plant fuel from U.S. utilities. However, because of delays in the siting of a permanent federal repository, and with no federal interim storage facilities designated, U.S. utilities have been forced to provide additional spent nuclear fuel (SNF) storage capability to accommodate spent fuel discharge requirements. At...

1999-12-10T23:59:59.000Z

36

Waste Management | Department of Energy  

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

July 27, 2011 July 27, 2011 End of Year 2010 SNF & HLW Inventories Map of the United States of America that shows the location of approximately 64,000 MTHM of Spent Nuclear Fuel (SNF) & 275 High-Level Radioactive Waste (HLW) Canisters. July 27, 2011 FY 2007 Total System Life Cycle Cost, Pub 2008 The Analysis of the Total System Life Cycle Cost (TSLCC) of the Civilian Radioactive Waste Management Program presents the Office of Civilian Radioactive Waste Management's (OCRWM) May 2007 total system cost estimate for the disposal of the Nation's spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The TSLCC analysis provides a basis for assessing the adequacy of the Nuclear Waste Fund (NWF) Fee as required by Section 302 of the Nuclear Waste Policy Act of 1982 (NWPA), as amended.

37

Pioneering Nuclear Waste Disposal  

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

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

38

The Social and Ethical Aspects of Nuclear Waste  

E-Print Network (OSTI)

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

Marshall, Alan

2005-01-01T23:59:59.000Z

39

Materials Science of Nuclear Waste Management I  

Science Conference Proceedings (OSTI)

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

40

Materials and Processes to Immobilize Nuclear Waste  

Science Conference Proceedings (OSTI)

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

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


41

Pioneering Nuclear Waste Disposal  

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

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

42

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

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

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

43

The Social and Ethical Aspects of Nuclear Waste  

E-Print Network (OSTI)

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

Marshall, Alan

2005-01-01T23:59:59.000Z

44

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network (OSTI)

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

45

Environmental Degradation of Nuclear Waste Storage Canister ...  

Science Conference Proceedings (OSTI)

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

46

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

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

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

47

NUCLEAR WASTE CONSULTANTS, INC.  

E-Print Network (OSTI)

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

Technical Review Branch

1988-01-01T23:59:59.000Z

48

The Social and Ethical Aspects of Nuclear Waste  

E-Print Network (OSTI)

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

Marshall, Alan

2005-01-01T23:59:59.000Z

49

Nuclear Utility Mixed Waste Stream Characterization Study  

Science Conference Proceedings (OSTI)

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

1994-12-31T23:59:59.000Z

50

Uranium immobilization and nuclear waste  

SciTech Connect

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

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

1982-02-01T23:59:59.000Z

51

Basic Research for Evaluating Nuclear Waste Form Performance  

Science Conference Proceedings (OSTI)

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

Don J. Bradley

52

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.

53

January 16, 2013 Secretarial Determination of the Adequacy of the Nuclear  

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

January 16, 2013 Secretarial Determination of the Adequacy of the January 16, 2013 Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee January 16, 2013 Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee As required by the Nuclear Waste Policy Act of 1982 (NWPA), DOE has completed its review of the adequacy of the Nuclear Waste Fund fee. Based on that review, the Secretary has determined not to propose an adjustment to the fee at this time. The Secretary's determination is available here. January 16 2013 Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee.pdf More Documents & Publications CIVILIAN RADIOACTIVE WASTE MANAGEMENT 2008 FEE ADEQUACY ASSESSMENT LETTER REPORT FY 2007 Fee Adequacy, Pub 2008 Fiscal Year 2007 Civilian Radioactive Waste Management Fee Adequacy

54

January 16, 2013 Secretarial Determination of the Adequacy of the Nuclear  

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

January 16, 2013 Secretarial Determination of the Adequacy of the January 16, 2013 Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee January 16, 2013 Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee As required by the Nuclear Waste Policy Act of 1982 (NWPA), DOE has completed its review of the adequacy of the Nuclear Waste Fund fee. Based on that review, the Secretary has determined not to propose an adjustment to the fee at this time. The Secretary's determination is available here. January 16 2013 Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee.pdf More Documents & Publications FY 2007 Fee Adequacy, Pub 2008 Fiscal Year 2007 Civilian Radioactive Waste Management Fee Adequacy Assessment Report CIVILIAN RADIOACTIVE WASTE MANAGEMENT 2008 FEE ADEQUACY ASSESSMENT LETTER

55

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

Science Conference Proceedings (OSTI)

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

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

1985-06-01T23:59:59.000Z

56

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

E-Print Network (OSTI)

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

57

Nuclear Waste Policy Act.doc  

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

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

58

Plasma Mass Filters For Nuclear Waste Reprocessing  

SciTech Connect

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

Abraham J. Fetterman and Nathaniel J. Fisch

2011-05-26T23:59:59.000Z

59

Accelerated Chemical Aging of Crystalline Nuclear Waste Forms  

Science Conference Proceedings (OSTI)

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

60

Removing nuclear waste, one shipment at a time  

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

Stories 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,...

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


61

GNEP Element:Minimize Nuclear Waste | Department of Energy  

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

Documents & Publications Global Nuclear Energy Partnership Fact Sheet - Minimize Nuclear Waste The Global Nuclear Energy Partnership: Greater Energy Security in a Cleaner, Safer...

62

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

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

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

63

Nuclear Fuel Cycle and Waste Management Technologies - Nuclear Engineering  

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

Nuclear Fuel Cycle and Nuclear Fuel Cycle and Waste Management Technologies Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Fuel Cycle and Waste Management Technologies Overview Bookmark and Share Much of the NE Division's research is directed toward developing software and performing analyses, system engineering design, and experiments to support the demonstration and optimization of the electrometallurgical

64

Sodium-Bearing Waste Treatment Alternatives Implementation Study  

SciTech Connect

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

65

Natural analogues of nuclear waste glass corrosion.  

SciTech Connect

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

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

1999-01-06T23:59:59.000Z

66

System Engineering Design [Nuclear Waste Management using  

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

System Engineering System Engineering Design Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Waste Management using Electrometallurgical Technology System Engineering Design Bookmark and Share Two major pieces of electrometallurgical process equipment are the Electrorefiner and the Cathode Processor. NE personnel have been involved in the conceptual design, final design, procurement, manufacture,

67

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network (OSTI)

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

68

Nuclear waste package fabricated from concrete  

Science Conference Proceedings (OSTI)

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

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

1987-03-01T23:59:59.000Z

69

Nuclear waste incineration technology status  

Science Conference Proceedings (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

70

Challenges of Nuclear Waste Vitrification  

Science Conference Proceedings (OSTI)

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

71

WIPP - Pioneering Nuclear Waste Disposal  

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

Waste Disposal Cover Page and Table of Contents Closing the Circle The Long Road to WIPP - Part 1 The Long Road to WIPP - Part 2 Looking to the Future Related Reading and The...

72

Ceramics for Nuclear Waste Disposition  

Science Conference Proceedings (OSTI)

Oct 27, 2009 ... Materials Solutions for the Nuclear Renaissance: Ceramics for .... In-situ Decommissioning of Heavy Water Reactor Disassembly Basin: ...

73

Chemical aspects of nuclear waste treatment  

SciTech Connect

The chemical aspects of the treatment of gaseous, liquid, and solid wastes are discussed in overview. The role of chemistry and the chemical reactions in waste treatment are emphasized. Waste treatment methods encompass the chemistry of radioactive elements from every group of the periodic table. In most streams, the radioactive elements are present in relatively low concentrations and are often associated with moderately large amounts of process reagents, or materials. In general, it is desirable that waste treatment methods are based on chemistry that is selective for the concentration of radionuclides and does not require the addition of reagents that contribute significantly to the volume of the treated waste. Solvent extraction, ion exchange, and sorbent chemistry play a major role in waste treatment because of the high selectivity provided for many radionuclides. This paper deals with the chemistry of the onsite treatment methods that is typically used at nuclear installations and is not concerned with the chemistry of the various alternative materials proposed for long-term storage of nuclear wastes. The chemical aspects are discussed from a generic point of view in which the chemistry of important radionuclides is emphasized.

Bond, W.D.

1980-01-01T23:59:59.000Z

74

Corrosion Behavior of Container Alloys in Nuclear Waste Repositories  

Science Conference Proceedings (OSTI)

Symposium, Materials Issues in Nuclear Waste Management in the 21st Century ... Abstract Scope, Commercial nuclear energy has been around for six ...

75

Experience in Using Fills for Spent Nuclear Fuel Waste Packages  

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

Fills for SNF Waste Packages Experience in Using Fills for Spent Nuclear Fuel Waste Packages The use of other fill materials in waste packages has been investigated by several...

76

Recovery of fissile materials from nuclear wastes  

DOE Patents (OSTI)

A process is described 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.

1997-12-01T23:59:59.000Z

77

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

78

Materials for Nuclear Waste Disposal and Environmental Cleanup  

Science Conference Proceedings (OSTI)

Symposium, Materials for Nuclear Waste Disposal and Environmental Cleanup ... Secure and Certify Studies to Work on Production of Spiked Plutonium.

79

Neutron pulse simulation in nuclear waste for waste characterization  

SciTech Connect

The numerical simulations discussed in this paper show how analysis with computer-generated illustrations can be used to explain the concepts and advantages of pulsed neutron systems for tank waste evaluations. Furthermore, the analysis-illustration approach lends itself to parametric studies evaluating design features of hardware before it is fabricated. Nuclear material characteristics of hazardous or toxic simulants can be evaluated before preparing them or finding nontoxic or nonhazardous substitutes that will exhibit similar nuclear properties. Pulsed neutron systems hold significant promise for partial characterization of tank waste. The device could operate in a high background gamma radiation field and provide important information on moisture concentrations, fissionable material contents, and material interfaces quickly and at considerably less cost than obtainable from sample analyses.

Toffer, H.; Watson, W.T.; Roetman, V.E.

1993-12-01T23:59:59.000Z

80

An Overview of Waste Classification for Disposal Summary  

E-Print Network (OSTI)

Radioactive waste is a byproduct of nuclear weapons production, commercial nuclear power generation, and the naval reactor program. Waste byproducts also result from radioisotopes used for scientific, medical, and industrial purposes. The legislative definitions adopted for radioactive wastes, for the most part, refer to the processes that generated the wastes. Thus, waste disposal policies have tended to link the processes to uniquely tailored disposal solutions. Consequently, the origin of the waste, rather than its radiologic characteristics, often determines its fate. Plutonium and enriched uranium-235 were first produced by the Manhattan Project during World War II. These materials were later defined by the Atomic Energy Act of 1954 as special nuclear materials, along with other materials that the former Atomic Energy Commission (AEC) determined were capable of releasing energy through nuclear fission. Reprocessing of irradiated nuclear fuel to extract special nuclear material generated highly radioactive liquid and solid byproducts. The Nuclear Waste Policy Act of 1982 (NWPA) defined irradiated fuel as spent nuclear fuel, and the byproducts as high-level waste. Uranium ore processing

Anthony Andrews

2006-01-01T23:59:59.000Z

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


81

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

SciTech Connect

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

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

1980-04-01T23:59:59.000Z

82

Nuclear Waste Fund Activities Management Team | Department of Energy  

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

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

83

Sulfate Retention in High Level Nuclear Waste Glasses  

Science Conference Proceedings (OSTI)

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

84

The Social and Ethical Aspects of Nuclear Waste  

E-Print Network (OSTI)

level waste and spent nuclear fuel: The continuing societal1999). Transportation of spent nuclear fuel and high-levelfor the disposal of spent nuclear fuel in Finland, 15-16

Marshall, Alan

2005-01-01T23:59:59.000Z

85

YUCCA MOUNTAIN PROJECT 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  

Science Conference Proceedings (OSTI)

For more than half a century, since nuclear science helped us win World War II and ring in the Atomic Age, scientists have known that !he Nation would need a secure, permanent facility in which to dispose of radioactive wastes. Twenty years ago, when Congress adopted the Nuclear Waste Policy Act of 1982 (NWPA or ''the Act''), it recognized the overwhelming consensus in the scientific community that the best option for such a facility would be a deep underground repository. Fifteen years ago, Congress directed the Secretary of Energy to investigate and recommend to the President whether such a repository could be located safely at Yucca Mountain, Nevada. Since then, our country has spent billions of dollars and millions of hours of research endeavoring to answer this question. I have carefully reviewed the product of this study. In my judgment, it constitutes sound science and shows that a safe repository can be sited there. I also believe that compelling national interests counsel in favor of proceeding with this project. Accordingly, consistent with my responsibilities under the NWPA, today I am recommending that Yucca Mountain be developed as the site for an underground repository for spent fuel and other radioactive wastes. The first consideration in my decision was whether the Yucca Mountain site will safeguard the health and safety of the people, in Nevada and across the country, and will be effective in containing at minimum risk the material it is designed to hold. Substantial evidence shows that it will. Yucca Mountain is far and away the most thoroughly researched site of its kind in the world. It is a geologically stable site, in a closed groundwater basin, isolated on thousands of acres of Federal land, and farther from any metropolitan area than the great majority of less secure, temporary nuclear waste storage sites that exist in the country today. This point bears emphasis. We are not confronting a hypothetical problem. We have a staggering amount of radioactive waste in this country--nearly 100,000,000 gallons of high-level nuclear waste and more than 40,000 metric tons of spent nuclear fuel with more created every day. Our choice is not between, on the one hand, a disposal site with costs and risks held to a minimum, and, on the other, a magic disposal system with no costs or risks at all. Instead, the real choice is between a single secure site, deep under the ground at Yucca Mountain, or making do with what we have now or some variant of it--131 aging surface sites, scattered across 39 states. Every one of those sites was built on the assumption that it would be temporary. As time goes by. every one is closer to the limit of its safe life span. And every one is at least a potential security risk--safe for today, but a question mark in decades to come.

NA

2002-03-26T23:59:59.000Z

86

International nuclear waste management fact book  

Science Conference Proceedings (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

87

Seal welded cast iron nuclear waste container  

SciTech Connect

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

88

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

Science Conference Proceedings (OSTI)

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

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

1994-05-01T23:59:59.000Z

89

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

SciTech Connect

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

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

1993-11-01T23:59:59.000Z

90

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

E-Print Network (OSTI)

Anthony   V.   Guide  Nuclear  Reactors.   University   of  of   fuel   for   nuclear   reactors—create   wastes  Level  Waste   nuclear reactors, and subsequent utilization

Djokic, Denia

2013-01-01T23:59:59.000Z

91

Current Status and Potential Impacts Regarding the Proposed Development of a Rail Line to the Yucca Mountain Nuclear Waste Repository  

Science Conference Proceedings (OSTI)

This paper provides a description of the current status regarding the proposed development of a rail line to the Yucca Mountain Nuclear Waste Repository in Nye County, Southern Nevada, which includes potential impacts analyzed during the National Environmental Policy Act (NEPA) process, and the subsequent creation of an Environmental Impact Statement (EIS) for the rail line. Potential impacts are addressed within the context of impacts to natural and human environmental resources found within the geographic area of the proposed federal project. Potential impacts to these resources have been fully analyzed in the Rail Alignment Draft EIS (DEIS). This paper includes a summary of the potential impacts analyzed in the DEIS. Examples of potential impacts include land use conflicts, air quality, water use, and impacts to biological and cultural resources, among others. In conclusion: Based on its obligations under the NWPA and its decision to select the mostly rail scenario for the transportation of spent nuclear fuel and high-level radioactive waste, DOE needs to ship these materials by rail in Nevada to a repository at Yucca Mountain. DOE prepared the Rail Alignment EIS to provide the background, data, information, and analyses to help decision makers and the public understand the potential environmental impacts that could result from constructing and operating a railroad for shipment of spent nuclear fuel, high-level radioactive waste, and other materials from an existing rail line in Nevada to a repository at Yucca Mountain. This railroad would consist of a rail line, railroad operations support facilities, and other related infrastructure. DOE will use the Rail Alignment EIS to decide whether to construct and operate the proposed railroad, and if so, to: - Select a rail alignment (Caliente rail alignment or Mina rail alignment) in which to construct the railroad; - Select the common segments and alternative segments within either a Caliente rail alignment or a Mina rail alignment. The Department would use the selected common segments and alternative segments to identify the public lands to be included in right-of-way applications; - Decide where to construct proposed railroad operations support facilities; - Decide whether to restrict use of the rail line to DOE trains, or whether to allow commercial shippers to operate over the rail line; and - Determine what mitigation measures to implement. (authors)

Lanthrum, G. [U.S. Department of Energy, Office of Civilian Radioactive Waste Management, Washington, DC (United States); Gunnerson, J. [Booz Allen Hamilton, Las Vegas, NV (United States)

2008-07-01T23:59:59.000Z

92

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

93

WIPP Uses Recovery Act Funding to Reduce Nuclear Waste Footprint |  

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

Uses Recovery Act Funding to Reduce Nuclear Waste Footprint Uses Recovery Act Funding to Reduce Nuclear Waste Footprint WIPP Uses Recovery Act Funding to Reduce Nuclear Waste Footprint August 1, 2011 - 12:00pm Addthis Media Contact Deb Gill www.wipp.energy.gov 575-234-7270 CARLSBAD, N.M. - The U.S. Department of Energy's (DOE's) Carlsbad Field Office (CBFO) reduced the nuclear waste footprint by using American Recovery and Reinvestment Act funds to expedite the clean up of five transuranic (TRU) waste storage sites and to make important infrastructure improvements at the Waste Isolation Pilot Plant (WIPP). Expediting TRU waste shipments supports DOE's goal to dispose of 90 percent of legacy TRU waste by 2015, saving taxpayers million of dollars in storage and maintenance costs. Recovery Act funds allowed highly trained teams to safely prepare and load

94

Disposing of nuclear waste in a salt bed  

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

Disposing of nuclear waste in a salt bed Disposing of nuclear waste in a salt bed 1663 Los Alamos science and technology magazine Latest Issue:November 2013 All Issues » submit Disposing of nuclear waste in a salt bed Decades' worth of transuranic waste from Los Alamos is being laid to rest at the Waste Isolation Pilot Plant in southeastern New Mexico March 25, 2013 Disposing of nuclear waste in a salt bed Depending on the impurities embedded within it, the salt from WIPP can be anything from a reddish, relatively opaque rock to a clear crystal like the one shown here. Ordinary salt effectively seals transuranic waste in a long-term repository Transuranic waste, made of items such as lab coats and equipment that have been contaminated by radioactive elements heavier than uranium, is being shipped from the Los Alamos National Laboratory to a long-term storage

95

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

96

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

97

Secretarial Determination of the Adequacy of the Nuclear Waste...  

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

next annual review. Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee More Documents & Publications The Report To The President And The Congress By The...

98

Congressional Preferences and the Advancement of American Nuclear Waste Policy.  

E-Print Network (OSTI)

??The problem of nuclear waste disposal has existed since the time of the Manhattan Project in World War II. Although there exist a number of… (more)

Ternate, Rhoel Gonzales

2013-01-01T23:59:59.000Z

99

Development of Ceramic Waste Forms for an Advanced Nuclear ...  

Science Conference Proceedings (OSTI)

Presentation Title, Development of Ceramic Waste Forms for an Advanced Nuclear Fuel Cycle. Author(s), James C. Marra, Amanda Billings, Kyle Brinkman,  ...

100

Geopolymers in Nuclear Waste Immobilization: Past, Present, and ...  

Science Conference Proceedings (OSTI)

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

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


101

Developments in Nuclear Waste Forms: University/International ...  

Science Conference Proceedings (OSTI)

Symposium, Materials for Nuclear Waste Disposal and Environmental Cleanup ... to proceed albeit with even greater care over security and safety aspects.

102

Project characteristics monitoring report: BWIP (Basalt Waste Isolation Program) repository project  

Science Conference Proceedings (OSTI)

This monitoring report has been prepared to show compliance with provisions of the Nuclear Waste Policy Act of 1982 (NWPA) and to provide local and state government agencies with information concerning the Basalt Waste Isolation Program (BWIP). This report contains data for the time period May 26, 1986 to February 1988. The data include employment figures, salaries, project purchases, taxes and fees paid, worker survey results, and project closedown personal interview summaries. This information has become particularly important since the decision in December 1987 to stop all BWIP activities except those for site reclamation. The Nuclear Waste Policy Amendments Act of 1987 requires nonreclamation work at the Hanford Site to stop as of March 22, 1988. 7 refs., 6 figs., 28 tabs.

Friedli, E.A.; Herborn, D.I.; Taylor, C.D.; Tomlinson, K.M.

1988-03-01T23:59:59.000Z

103

Waste form product characteristics  

SciTech Connect

The Department of Energy has operated nuclear facilities at the Idaho National Engineering Laboratory (INEL) to support national interests for several decades. Since 1953, it has supported the development of technologies for the storage and reprocessing of spent nuclear fuels (SNF) and the resultant wastes. However, the 1992 decision to discontinue reprocessing of SNF has left nearly 768 MT of SNF in storage at the INEL with unspecified plans for future dispositioning. Past reprocessing of these fuels for uranium and other resource recovery has resulted in the production of 3800 M{sup 3} calcine and a total inventory of 7600 M{sup 3} of radioactive liquids (1900 M{sup 3} destined for immediate calcination and the remaining sodium-bearing waste requiring further treatment before calcination). These issues, along with increased environmental compliance within DOE and its contractors, mandate operation of current and future facilities in an environmentally responsible manner. This will require satisfactory resolution of spent fuel and waste disposal issues resulting from the past activities. A national policy which identifies requirements for the disposal of SNF and high level wastes (HLW) has been established by the Nuclear Waste Policy Act (NWPA) Sec.8,(b) para(3)) [1982]. The materials have to be conditioned or treated, then packaged for disposal while meeting US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. The spent fuel and HLW located at the INEL will have to be put into a form and package that meets these regulatory criteria. The emphasis of Idaho Chemical Processing Plant (ICPP) future operations has shifted toward investigating, testing, and selecting technologies to prepare current and future spent fuels and waste for final disposal. This preparation for disposal may include mechanical, physical and/or chemical processes, and may differ for each of the various fuels and wastes.

Taylor, L.L.; Shikashio, R.

1995-01-01T23:59:59.000Z

104

GEOTECHNICAL ASSESSMENT AND INSTRUMENTATION NEEDS FOR NUCLEAR WASTE ISOLATION IN CRYSTALLINE AND ARGILLACEOUS ROCKS SYMPOSIUM  

E-Print Network (OSTI)

Radioactive Wastes from the Nuclear Fuel Cycle. ProceedingsRadioactive Wastes from the Nuclear Fuel Cycle. ProceedingsWastes from the Nuclear Fuel Cycle, Vienna, March 22-26,

Authors, Various

2011-01-01T23:59:59.000Z

105

U.S. Nuclear Waste Technical Review Board: Roles and Priorities...  

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

Nuclear Waste Technical Review Board: Roles and Priorities U.S. Nuclear Waste Technical Review Board: Roles and Priorities Presentation made by Nigel Mote for the NTSF annual...

106

Nuclear waste vitrification efficiency: cold cap reactions  

SciTech Connect

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

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

2012-12-15T23:59:59.000Z

107

A QUARTER CENTURY OF NUCLEAR WASTE MANAGEMENT IN JAPAN  

Science Conference Proceedings (OSTI)

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

Masuda, S.

2002-02-25T23:59:59.000Z

108

Method of preparing nuclear wastes for tansportation and interim storage  

SciTech Connect

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

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

1984-01-01T23:59:59.000Z

109

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

SciTech Connect

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

Jacobson, Victor Levon

2002-08-01T23:59:59.000Z

110

WEB RESOURCE: Nuclear Materials and Nuclear Fuel/Waste  

Science Conference Proceedings (OSTI)

Feb 12, 2007 ... Select, Sandbox, Open Discussion Regarding Materials for Nuclear ... Trends in Nuclear Power, The Nuclear Fuel Cycle, Nuclear Science ...

111

Removing nuclear waste, one shipment at a time  

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

Stories » Stories » 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 to a permanent repository near Carlsbad, NM. June 26, 2012 Governor Martinez applauding the 1014th TRU waste shipment New Mexico Governor Susana Martinez and other dignitaries applaud as the 1,014th shipment of transuranic waste leaves Los Alamos National Laboratory. Contact Environmental Communication & Public Involvement P.O. Box 1663 MS M996 Los Alamos, NM 87545 (505) 667-0216 Email "The milestone we're celebrating is one that has been a long-term environmental commitment." Removing nuclear waste, one shipment at a time Elected officials and other dignitaries recently gathered at Los Alamos

112

Nepheline Crystallization in Nuclear Waste Glasses: Progress ...  

Science Conference Proceedings (OSTI)

Abstract Scope, One significant limitation to waste loading in glass for Hanford defense wastes is the commonly high Al concentrations. The primary concern is ...

113

Nuclear waste treatment program: Annual report for FY 1987  

SciTech Connect

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

114

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

115

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 isotonically-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.

1997-12-01T23:59:59.000Z

116

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

117

Spent Nuclear Fuel Transportation: An Examination of Potential Lessons Learned From Prior Shipping Campaigns  

Science Conference Proceedings (OSTI)

The Nuclear Waste Policy Act of 1982 (NWPA), as amended, assigned the Department of Energy (DOE) responsibility for developing and managing a Federal system for the disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW). The Office of Civilian Radioactive Waste Management (OCRWM) is responsible for accepting, transporting, and disposing of SNF and HLW at the Yucca Mountain repository in a manner that protects public health, safety, and the environment; enhances national and energy security; and merits public confidence. OCRWM faces a near-term challenge—to develop and demonstrate a transportation system that will sustain safe and efficient shipments of SNF and HLW to a repository. To better inform and improve its current planning, OCRWM has extensively reviewed plans and other documents related to past high-visibility shipping campaigns of SNF and other radioactive materials within the United States. This report summarizes the results of this review and, where appropriate, lessons learned.

Marsha Keister; Kathryn McBride

2006-08-01T23:59:59.000Z

118

Systems and Components Development Expertise [Nuclear Waste Management  

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

Systems and Components Systems and Components Development Expertise Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Waste Management using Electrometallurgical Technology Systems and Components Development Expertise Bookmark and Share Electrorefiner The electrorefiner: an apparatus used for electrometallurgical treatment of spent nuclear fuel to facilitate storage and ultimate disposal. Click on

119

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

Science Conference Proceedings (OSTI)

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

Dietmar P. F. Möller

2007-07-01T23:59:59.000Z

120

Fuel Cycle Technologies Near Term Planning for Storage and Transportation of Used Nuclear Fuel  

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

of Section 180(c) of the Nuclear of Section 180(c) of the Nuclear Waste Policy Act, as amended National Transportation Stakeholder's Forum Buffalo, NY May 15, 2013 Section 180(c) Mandate "The Secretary shall provide technical assistance and funds to States for training for public safety officials of appropriate units of local government and Indian tribes through whose jurisdiction the Secretary plans to transport spent nuclear fuel or high-level radioactive waste [to an NWPA-authorized facility]. * The training shall cover procedures for safe routine transportation of these materials and procedures for dealing with emergency response situations. * Covers all modes of transport 2 Section 180(c) - Background  DOE nearly implemented Section 180(c) in the mid-

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


121

The Middle Ground for Nuclear Waste Management: Social and Ethical Aspects of Shallow Storage  

Science Conference Proceedings (OSTI)

The 2001 terrorist attacks in the USA and the 2011 seismic events in Japan have brought into sharp relief the vulnerabilities involved in storing nuclear waste on the land's surface. Nuclear engineers and waste managers are deciding that disposing nuclear ... Keywords: Ethics, Inter-Generational Equity, Nuclear Waste, Shallow Storage, Waste Disposal

Alan Marshall

2011-04-01T23:59:59.000Z

122

GEOTECHNICAL ASSESSMENT AND INSTRUMENTATION NEEDS FOR NUCLEAR WASTE ISOLATION IN CRYSTALLINE AND ARGILLACEOUS ROCKS SYMPOSIUM  

E-Print Network (OSTI)

Summary • • INSTRUMENTATION NEEDS FOR DETERMINING ROCKDepository Design . . • • • . . INSTRUMENTATION NEEDS FORASSESSMENT AND INSTRUMENTATION NEEDS FOR NUCLEAR WASTE

Authors, Various

2011-01-01T23:59:59.000Z

123

Linear and nonlinear transient heat conduction in nuclear waste repositories  

Science Conference Proceedings (OSTI)

Analytical solutions of thermal problems connected with the disposal of nuclear wastes are presented. Linear and nonlinear diffusion problems are analyzed considering time-dependent heat sources. Comparisons between the temperature distributions at a ...

C. A. Estrada-Gasca; M. H. Cobble

1988-10-01T23:59:59.000Z

124

Nuclear waste disposal in New Mexico and Nevada  

Science Conference Proceedings (OSTI)

The author describes the Waste Isolation Pilot Project (WIPP) for safely discarded nuclear wates is discussed. WIPP, now essentially complete, near Carlsbad, NM, and Yucca Mountain, in the early stages of construction northwest of Las Vegas, Nev. The ...

J. Beard

1997-11-01T23:59:59.000Z

125

Nuclear Waste Partnership LLC PO Box 2078 GSA-207  

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

072012 1 of 3 A URS-led partnership with B&W and AREVA Note: Form must be filled out completely and signed prior to submittal to the following address: Nuclear Waste Partnership...

126

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

127

The Properties of Spent Nuclear Fuel under Waste Disposal ...  

Science Conference Proceedings (OSTI)

Symposium, Materials Issues in Nuclear Waste Management in the 21st Century ... UO2 in the form of a ceramic pellet with a density close to theoretical. ... On discharge fro reactor the pellets have undergone a number of physical and ...

128

Method for calcining nuclear waste solutions containing zirconium and halides  

DOE Patents (OSTI)

A reduction in the quantity of gelatinous solids which are formed in aqueous zirconium-fluoride nuclear reprocessing waste solutions by calcium nitrate added to suppress halide volatility during calcination of the solution while further suppressing chloride volatility is achieved by increasing the aluminum to fluoride mole ratio in the waste solution prior to adding the calcium nitrate.

Newby, Billie J. (Idaho Falls, ID)

1979-01-01T23:59:59.000Z

129

Method for forming microspheres for encapsulation of nuclear waste  

SciTech Connect

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

130

Sulfate Volatilization in Simulated Nuclear Waste Glasses  

Science Conference Proceedings (OSTI)

Radioactive Demonstrations of Fluidized Bed Steam Reforming (FBSR) with Hanford Low Activity Wastes · Radionuclide Behavior and Geochemistry in Boom

131

Potential dispositioning flowsheets for ICPP SNF and wastes  

SciTech Connect

The Idaho Chemical Processing Plant (ICPP), located at the Idaho National Laboratory (INEL), has reprocessed irradiated nuclear fuels for the US Department of Energy (DOE) since 1953. This activity resulted mainly in the recovery of uranium and the management of the resulting wastes. The acidic radioactive high-level liquid waste was routinely stored in stainless steel tanks and then calcined to form a dry granular solid. The calcine is stored in stainless steel bins that are housed in underground concrete vaults. In April 1992, the DOE discontinued the practice of reprocessing irradiated nuclear fuels. This decision has left a legacy of 1.8 million gallons of radioactive liquid wastes (1.5 million gallons of radioactive sodium-bearing liquid wastes and 0.3 million gallons of high-level liquid waste), 3800 cubic meters of calcine waste, and 289 metric tons of heavy metal within unprocessed spent nuclear fuel (SNF) left in inventory at the ICPP. The nation`s radioactive waste policy has been established by the Nuclear Waste Policy Act (NWPA), which requires the final disposal of SNF and radioactive waste in accordance with US Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) standards. In accordance with these regulations and other legal agreements between the State of Idaho and the DOE, the DOE must, among other requirements, (1) complete a final Environmental Impact Statement by April 30, 1995, (2) evaluate and test sodium-bearing waste pre-treatment technologies, (3) select the sodium-bearing and calcine waste pre-treatment technology, if necessary, by June 1, 1995, and (4) select a technology for converting calcined waste into an appropriate disposal form by June 1, 1995.

Olson, A.L. [ed.; Anderson, P.A.; Bendixsen, C.L. [and others

1995-11-01T23:59:59.000Z

132

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

133

Microbial Effects on Nuclear Waste Packaging Materials  

DOE Green Energy (OSTI)

Microorganisms may enhance corrosion of components of planned engineered barriers within the proposed nuclear waste repository at Yucca Mountain (YM). Corrosion could occur either directly, through processes collectively known as Microbiologically Influenced Corrosion (MIC), or indirectly, by adversely affecting the composition of water or brines that come into direct contact with engineered barrier surfaces. Microorganisms of potential concern (bacteria, archea, and fungi) include both those indigenous to Yucca Mountain and those that infiltrate during repository construction and after waste emplacement. Specific aims of the experimental program to evaluate the potential of microorganisms to affect damage to engineered barrier materials include the following: Indirect Effects--(1) Determine the limiting factors to microbial growth and activity presently in the YM environment. (2) Assess these limiting factors to aid in determining the conditions and time during repository evolution when MIC might become operant. (3) Evaluate present bacterial densities, the composition of the YM microbial community, and determining bacterial densities if limiting factors are overcome. During a major portion of the regulatory period, environmental conditions that are presently extant become reestablished. Therefore, these studies ascertain whether biomass is sufficient to cause MIC during this period and provide a baseline for determining the types of bacterial activities that may be expected. (4) Assess biogenic environmental effects, including pH, alterations to nitrate concentration in groundwater, the generation of organic acids, and metal dissolution. These factors have been shown to be those most relevant to corrosion of engineered barriers. Direct Effects--(1) Characterize and quantify microbiological effects on candidate containment materials. These studies were carried out in a number of different approaches, using whole YM microbiological communities, a subset of YM bacteria, and select reference organisms. Studies were carried out to determine morphological alterations to materials surfaces and using electrochemical methods to help quantify effects and modes of MIC, and to provide additional alternative means of evaluating MIC effects. They were carried out only under conservative conditions (low temperature, saturated conditions); thus, resulting conclusions may be considered an upper bound of potential biological effects on tested materials.

Horn, J; Martin, S; Carrillo, C; Lian, T

2005-07-22T23:59:59.000Z

134

Safety Analysis, Hazard and Risk Evaluations [Nuclear Waste Management  

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

Safety Analysis, Hazard Safety Analysis, Hazard and Risk Evaluations Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Waste Management using Electrometallurgical Technology Safety Analysis, Hazard and Risk Evaluations Bookmark and Share NE Division personnel had a key role in the creation of the FCF Final Safety Analysis Report (FSAR), FCF Technical Safety Requirements (TSR)

135

Nuclear waste/nuclear power: their futures are linked  

SciTech Connect

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

Skoblar, L.T.

1981-01-01T23:59:59.000Z

136

Expected brine movement at potential nuclear waste repository salt sites  

SciTech Connect

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

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

1987-08-01T23:59:59.000Z

137

Spherical tokamak (ST) transmutation of nuclear wastes  

SciTech Connect

The concept for an ST fusion core that drives a He-cooled, actinide-bearing, molten-salt blanket of moderate power density to generate electricity is examined for the first time. The results show that the fusion core is suited for this purpose and require a level of plasma, power density, engineering, and material performances moderate in comparison with what has been considered desirable for fusion-only power plants. The low aspect ratio of ST introduces a relatively thick, diverted scrape-off layer which leads to reduced heat fluxes at the limiter and divertor tiles. The use of a demountable, water-cooled, single-turn copper center leg for the toroidal field coils enables simplifications of the fusion core configuration and improves overall practicality for future power applications. These result in much reduced size and cost of the fusion core for the transmutation power plant relative to an optimized fusion-only fusion core. Surrounded by a separate tritium-breeding zone, the molten-salt blanket concept is in principle less complex and costly than the thermal breeding blankets for fusion. These combine to effect major reductions in the cost and weight of the power core equipment for the transmutation power plant. The minimum cost of electricity for such a power plant is thus reduced from the best fusion-only counterpart by more than 30%, based on consistent but approximate modeling. The key issues, development steps, and the potential value inherent in the ST fusion core in addressing the world needs for nuclear waste energy production are discussed.

Peng, Y.-K.M.; Galambos, J.D. [Oak Ridge National Lab., TN (United States); Cheng, E.T.; Cerbone, R.J. [TSI Research, Inc., Solana Beach, CA (United States)

1995-12-31T23:59:59.000Z

138

Spherical tokamak (ST) transmutation of nuclear wastes  

SciTech Connect

The concept for an ST fusion core that drives a He-cooled, actinide-bearing, molten-salt blanket of moderate power density to generate electricity is examined for the first time. The results show that the fusion core is suited for this purpose and require a level of plasma, power density, engineering, and material performances moderate in comparison with what has been considered desirable for fusion-only power plants. The low aspect ratio of ST introduces a relatively thick, diverted scrape-off layer which leads to reduced heat fluxes at the limiter and divertor tiles. The use of a demountable, water-cooled, single-turn copper center leg for the toroidal field coils enables simplifications of the fusion core configuration and improves overall practicality for future power applications. These result in much reduced size and cost of the fusion core for the transmutation power plant relative to an optimized fusion-only fusion core. Surrounded by a separate tritium-breeding zone, the molten-salt blanket concept is in principle less complex and costly than the thermal breeding blankets for fusion. These combine to effect major reductions in the cost and weight of the power core equipment for the transmutation power plant. The minimum cost of electricity for such a power plant is thus reduced from the best fusion-only counterpart by more than 30%, based on consistent but approximate modeling. The key issues, development steps, and the potential value inherent in the ST fusion core in addressing the world needs for nuclear waste reduction and energy production are discussed.

Peng, Yueng Kay Martin [ORNL; Cheng, E.T. [TSI Research Inc.; Galambos, John D [ORNL; Cerbone, R. J. [TSI Research Inc.

1995-01-01T23:59:59.000Z

139

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

Science Conference Proceedings (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

140

Energy Department and Catholic University Improve Safety of Nuclear Waste |  

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

Catholic University Improve Safety of Nuclear Catholic University Improve Safety of Nuclear Waste Energy Department and Catholic University Improve Safety of Nuclear Waste January 30, 2013 - 12:51pm Addthis Secretary of Energy Steven Chu participates in a tour of Catholic University's Vitreous State Laboratory. | Photo courtesy of the Office of Environmental Management. Secretary of Energy Steven Chu participates in a tour of Catholic University's Vitreous State Laboratory. | Photo courtesy of the Office of Environmental Management. David Sheeley David Sheeley Editor/Writer What does this project do? Hanford treats and immobilizes significant quantities of legacy nuclear waste left from the manufacture of plutonium during World War II and the Cold War. Secretary Steven Chu recently visited Catholic University's Vitreous

Note: This page contains sample records for the topic "nwpa nuclear waste" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
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141

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

Science Conference Proceedings (OSTI)

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

Not Available

1994-03-01T23:59:59.000Z

142

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

143

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

144

Approach for enhancing nuclear materials tracking and reporting in waste  

SciTech Connect

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

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

2001-06-01T23:59:59.000Z

145

National briefing summaries: Nuclear fuel cycle and waste management  

SciTech Connect

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

146

Chemical digestion of low level nuclear solid waste material  

DOE Patents (OSTI)

A chemical digestion for treatment of low level combustible nuclear solid waste material is provided and comprises reacting the solid waste material with concentrated sulfuric acid at a temperature within the range of 230.degree.-300.degree.C and simultaneously and/or thereafter contacting the reacting mixture with concentrated nitric acid or nitrogen dioxide. In a special embodiment spent ion exchange resins are converted by this chemical digestion to noncombustible gases and a low volume noncombustible residue.

Cooley, Carl R. (Richland, WA); Lerch, Ronald E. (Richland, WA)

1976-01-01T23:59:59.000Z

147

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

SciTech Connect

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

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

1981-06-01T23:59:59.000Z

148

Materials Science of Nuclear Waste Management  

Science Conference Proceedings (OSTI)

The intent is to provide a forum for researchers from national laboratories, universities, and nuclear industry to discuss current understanding of materials ...

149

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

150

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

151

Recovery of transplutonium elements from nuclear reactor waste  

DOE Patents (OSTI)

A method of separating actinide values from nitric acid waste solutions resulting from reprocessing of irradiated nuclear fuels comprises oxalate precipitation of the major portion of actinide and lanthanide values to provide a trivalent fraction suitable for subsequent actinide/lanthanide partition, exchange of actinide and lanthanide values in the supernate onto a suitable cation exchange resin to provide an intermediate-lived raffinate waste stream substantially free of actinides, and elution of the actinide values from the exchange resin. The eluate is then used to dissolve the trivalent oxalate fraction prior to actinide/lanthanide partition or may be combined with the reprocessing waste stream and recycled.

Campbell, David O. (Oak Ridge, TN); Buxton, Samuel R. (Wartburg, TN)

1977-05-24T23:59:59.000Z

152

Working Group Report on - Space Nuclear Power Systems and Nuclear Waste  

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

Working Group Report on - Space Nuclear Power Systems and Nuclear Working Group Report on - Space Nuclear Power Systems and Nuclear Waste Technology R&D Working Group Report on - Space Nuclear Power Systems and Nuclear Waste Technology R&D "Even though one cannot anticipate the answers in basic research, the return on the public's investment can be maximized through long-range planning of the most promising avenues to explore and the resources needed to explore them." (p. v) "Pursuit of this goal entails developing new technologies and advanced facilities, educating young scientists, training a technical workforce, and contributing to the broader science and technology enterprise?." (p. vi) Ref:: "Nuclear Science: A Long Range Plan", DOE/NSF, Feb. 1996. The purpose of this effort is to develop the first iteration of a

153

Working Group Report on - Space Nuclear Power Systems and Nuclear Waste  

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

Working Group Report on - Space Nuclear Power Systems and Nuclear Working Group Report on - Space Nuclear Power Systems and Nuclear Waste Technology R&D Working Group Report on - Space Nuclear Power Systems and Nuclear Waste Technology R&D "Even though one cannot anticipate the answers in basic research, the return on the public's investment can be maximized through long-range planning of the most promising avenues to explore and the resources needed to explore them." (p. v) "Pursuit of this goal entails developing new technologies and advanced facilities, educating young scientists, training a technical workforce, and contributing to the broader science and technology enterprise?." (p. vi) Ref:: "Nuclear Science: A Long Range Plan", DOE/NSF, Feb. 1996. The purpose of this effort is to develop the first iteration of a

154

Mesoscale to plant-scale models of nuclear waste reprocessing.  

Science Conference Proceedings (OSTI)

Imported oil exacerabates our trade deficit and funds anti-American regimes. Nuclear Energy (NE) is a demonstrated technology with high efficiency. NE's two biggest political detriments are possible accidents and nuclear waste disposal. For NE policy, proliferation is the biggest obstacle. Nuclear waste can be reduced through reprocessing, where fuel rods are separated into various streams, some of which can be reused in reactors. Current process developed in the 1950s is dirty and expensive, U/Pu separation is the most critical. Fuel rods are sheared and dissolved in acid to extract fissile material in a centrifugal contactor. Plants have many contacts in series with other separations. We have taken a science and simulation-based approach to develop a modern reprocessing plant. Models of reprocessing plants are needed to support nuclear materials accountancy, nonproliferation, plant design, and plant scale-up.

Noble, David Frederick; O'Hern, Timothy John; Moffat, Harry K.; Nemer, Martin B.; Domino, Stefan Paul; Rao, Rekha Ranjana; Cipiti, Benjamin B.; Brotherton, Christopher M.; Jove-Colon, Carlos F.; Pawlowski, Roger Patrick

2010-09-01T23:59:59.000Z

155

Salt disposal of heat-generating nuclear waste.  

SciTech Connect

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

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

2011-01-01T23:59:59.000Z

156

Risk analysis and solving the nuclear waste siting problem  

SciTech Connect

In spite of millions of dollars and countless human resources being expended on finding nuclear wastes sites, the search has proved extremely difficult for the nuclear industry. This may be due to the approach followed, rather than inadequacies in research or funding. A new approach to the problem, the reverse Dutch auction, is suggested. It retains some of the useful elements of the present system, but it also adds new ones.

Inhaber, H.

1993-12-01T23:59:59.000Z

157

Idaho Site Taps Old World Process to Treat Nuclear Waste | Department...  

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

Idaho Site Taps Old World Process to Treat Nuclear Waste Idaho Site Taps Old World Process to Treat Nuclear Waste September 9, 2013 - 12:00pm Addthis The Idaho site's sodium...

158

UNITED STATES NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network (OSTI)

waste management program in addition to the site-characterization efforts at Yucca Mountain Upwelling at Yucca Mountain The Board completed its review of material (11 re- ports) first submitted that there is evidence of ongoing, intermittent hydrothermal upwelling at Yucca Mountain and that large earth- quake

159

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

160

Multiple-code simulation study of the long-term EDZ evolution of geological nuclear waste repositories  

E-Print Network (OSTI)

SKB TR-06-09. Swedish Nuclear Fuel and Waste Management Co,and tunnel boring. Swedish Nuclear Fuel and Waste Management

Rutqvist, J.

2008-01-01T23:59:59.000Z

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


161

Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Second technical progress report  

Science Conference Proceedings (OSTI)

OAK-B135 Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Second technical progress report

Ulm, Franz-Josef

1999-12-31T23:59:59.000Z

162

Locations of Spent Nuclear Fuel and High-Level Radioactive Waste  

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

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

163

Communication Between U.S. Nuclear Waste Technical Review Board  

E-Print Network (OSTI)

Moving Beyond the Yucca Mountain Viability Assessment U.S. Nuclear Waste Technical Review Board the Yucca Mountain site in Nevada as the sole location to be studied for possi- ble development of the Yucca Mountain site. The U.S. Department of Energy (DOE) recently published Viability As- sessment

164

U.S. NUCLEAR WASTE TECHNICAL REVIEW BOARD  

E-Print Network (OSTI)

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

165

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 must be stored for around 60 years in underwater storage pools before permanent disposal. These underwater storage environments must be carefully monitored and controlled to avoid an environmental

Jeavons, Peter

166

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

167

Mass Tracking System Software [Nuclear Waste Management using  

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

Mass Tracking System Mass Tracking System Software Nuclear Fuel Cycle and Waste Management Technologies Overview Modeling and analysis Unit Process Modeling Mass Tracking System Software Waste Form Performance Modeling Safety Analysis, Hazard and Risk Evaluations Development, Design, Operation Overview Systems and Components Development Expertise System Engineering Design Other Major Programs Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE Division on Flickr Nuclear Waste Management using Electrometallurgical Technology Mass Tracking System Software Bookmark and Share The NE Division has developed a computer-based Mass Tracking (MTG) system, which is used at the Idaho National Laboratory Fuel Conditioning Facility (FCF) to maintain a real-time accounting of the inventory of containers and

168

Waste management plan for Hanford spent nuclear fuel characterization activities  

SciTech Connect

A joint project was initiated between Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratory (PNL) to address critical issues associated with the Spent Nuclear Fuel (SNF) stored at the Hanford Site. Recently, particular attention has been given to remediation of the SNF stored in the K Basins. A waste management plan (WMP) acceptable to both parties is required prior to the movement of selected material to the PNL facilities for examination. N Reactor and Single Pass Reactor (SPR) fuel has been stored for an extended period of time in the N Reactor, PUREX, K-East, and K-West Basins. Characterization plans call for transport of fuel material form the K Basins to the 327 Building Postirradiation Testing Laboratory (PTL) in the 300 Area for examination. However, PNL received a directive stating that no examination work will be started in PNL hot cell laboratories without an approved disposal route for all waste generated related to the activity. Thus, as part of the Characterization Program Management Plan for Hanford Spent Nuclear Fuel, a waste management plan which will ensure that wastes generated as a result of characterization activities conducted at PNL will be accepted by WHC for disposition is required. This document contains the details of the waste handling plan that utilizes, to the greatest extent possible, established waste handling and disposal practices at Hanford between PNL and WHC. Standard practices are sufficient to provides for disposal of most of the waste materials, however, special consideration must be given to the remnants of spent nuclear fuel elements following examination. Fuel element remnants will be repackaged in an acceptable container such as the single element canister and returned to the K Basins for storage.

Chastain, S.A. [Westinghouse Hanford Co., Richland, WA (United States); Spinks, R.L. [Pacific Northwest Lab., Richland, WA (United States)

1994-10-17T23:59:59.000Z

169

Existing nuclear sites can be used for new powerplants and nuclear waste storage  

SciTech Connect

Locating future nuclear powerplants at existing sites offers important advantages which warrant consideration by the Nuclear Regulatory Commission. The number of locations committed to long-term restricted use and periodic surveillance and maintenance could be limited. The burden of long-term care and final disposition of retired nuclear powerplants could be eased. Overall environmental impacts from the construction and operation of the powerplants could be reduced. Time and money in completing licensing proceedings could be saved. GAO also found that low-level wastes can be stored at nuclear powerplant sites, but such storage only postpones the inevitable need for disposal. Finally, permanent waste disposal at powerplant sites should only be permitted when sites conform to the national low-level waste disposal plan being prepared by the Department of Energy.

Staats, E.B.

1980-04-01T23:59:59.000Z

170

Melting of foaming batches: Nuclear waste glass  

Science Conference Proceedings (OSTI)

A simple model is presented for the rate of melting of a batch blanket in an electric glassmelting furnace. The melting process is assumed to be jointly controlled by the heat transfer from the pool of molten glass and the batch-to-glass conversion kinetics. Factors affecting the melting rate in the conversion-controlled regime are discussed. Attention is paid to gas evolution from redox reactions in waste glass batches and component accumulation within the blanket. It is suggested that the high rate of the blanket-free melting in a mechanically agitated furnace is made possible by increasing the rate of melt surface renewal. 27 refs.

Hrma, P.

1990-10-01T23:59:59.000Z

171

Case histories of EA documents for nuclear waste  

SciTech Connect

Nuclear power programs and policies in the United States have been subject to environmental assessment under the National Environmental Policy Act (NEPA) since 1971. NEPA documentation prepared for programmatic policy decision-making within the nuclear fuel cycle and concurrent federal policy are examined as they relate to radioactive waste management in this paper. Key programmatic environmental impact statements that address radioactive waste management include: the Atomic Energy Commission document on management of commercial high-level and transuranium-contaminated radioactive waste, which focussed on development of engineered retrievable surface storage facilities (RSSF); the Nuclear Regulatory Commission (NRC) document on use of recycled plutonium in mixed oxide fuel in light water cooled reactors, which focussed on plutonium recycle and RSSF; the NRC statement on handling of spent light water power reactor fuel, which focussed on spent fuel storage; and the Department of Energy (DOE) statement on management of commercially generated radioactive wastes, which focussed on development of deep geologic repositories. DOE is currently pursuing the deep geologic repository option, with monitored retrievable storage as a secondary option.

Vocke, R.W.

1985-01-01T23:59:59.000Z

172

Graphite matrix materials for nuclear waste isolation  

SciTech Connect

At low temperatures, graphites are chemically inert to all but the strongest oxidizing agents. The raw materials from which artificial graphites are produced are plentiful and inexpensive. Morover, the physical properties of artificial graphites can be varied over a very wide range by the choice of raw materials and manufacturing processes. Manufacturing processes are reviewed herein, with primary emphasis on those processes which might be used to produce a graphite matrix for the waste forms. The approach, recommended herein, involves the low-temperature compaction of a finely ground powder produced from graphitized petroleum coke. The resultant compacts should have fairly good strength, low permeability to both liquids and gases, and anisotropic physical properties. In particular, the anisotropy of the thermal expansion coefficients and the thermal conductivity should be advantageous for this application. With two possible exceptions, the graphite matrix appears to be superior to the metal alloy matrices which have been recommended in prior studies. The two possible exceptions are the requirements on strength and permeability; both requirements will be strongly influenced by the containment design, including the choice of materials and the waste form, of the multibarrier package. Various methods for increasing the strength, and for decreasing the permeability of the matrix, are reviewed and discussed in the sections in Incorporation of Other Materials and Elimination of Porosity. However, it would be premature to recommend a particular process until the overall multi-barrier design is better defined. It is recommended that increased emphasis be placed on further development of the low-temperature compacted graphite matrix concept.

Morgan, W.C.

1981-06-01T23:59:59.000Z

173

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-Martínez, A

2004-01-01T23:59:59.000Z

174

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

175

Control of Nepheline Crystallization in Nuclear Waste Glass  

SciTech Connect

Glass frits with a high B{sub 2}O{sub 3} concentration have been designed which, when combined with high-alumina concentration nuclear waste streams, will form glasses with durabilities that are acceptable for repository disposal and predictable using a free energy of hydration model. Two glasses with nepheline discriminator values closest to 0.62 showed significant differences in normalized boron release between the quenched and heat treated versions of each glass. X-ray diffraction confirmed that nepheline crystallized in the glass with the lowest nepheline discriminator value, and nepheline may also exist in the second glass as small nanocrystals. The high-B{sub 2}O{sub 3} frit was successful in producing simulated waste glasses with no detectable nepheline crystallization at waste loadings of up to 45 wt%. The melt rate of this frit was also considerably better than other frits with increased concentrations of Na{sub 2}O.

Fox, Kevin

2008-07-01T23:59:59.000Z

176

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

SciTech Connect

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

Magoulas, V.

2013-06-03T23:59:59.000Z

177

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

178

Potential applications of nanostructured materials in nuclear waste management.  

Science Conference Proceedings (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

179

Tank waste remediation system nuclear criticality safety program management review  

SciTech Connect

This document provides the results of an internal management review of the Tank Waste Remediation System (TWRS) criticality safety program, performed in advance of the DOE/RL assessment for closure of the TWRS Nuclear Criticality Safety Issue, March 1994. Resolution of the safety issue was identified as Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) Milestone M-40-12, due September 1999.

BRADY RAAP, M.C.

1999-06-24T23:59:59.000Z

180

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

Science Conference Proceedings (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

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


181

Assessment of nuclear safety and nuclear criticality potential in the Defense Waste Processing Facility  

SciTech Connect

A panel of experts in the fields of process engineering, process chemistry, and safety analysis met together on January 26, 1993, and February 19, 1993, to discuss nuclear safety and nuclear criticality potential in the Defense Waste Processing Facility (DWPF) processes. Nuclear safety issues and possibilities of nuclear criticality incidents in the DWPF were examined in depth. The discussion started at the receipt of slurry feeds: The Low Point Pump Pit Precipitate Tank (LPPPPT) and the Low Point Pump Pit Sludge Tank (LPPPST), and went into detail the whole DWPF processes. This report provides discussion of each of the areas and processes of the DWPF in terms of potential nuclear safety issues and nuclear criticality concerns.

Ha, B.C.

1993-05-10T23:59:59.000Z

182

Review of Concrete Biodeterioration in Relation to Buried Nuclear Waste  

Science Conference Proceedings (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

183

HEAT TRANSFER ANALYSIS FOR NUCLEAR WASTE SOLIDIFICATION CONTAINER  

SciTech Connect

The Nuclear Nonproliferation Programs Design Authority is in the design stage of the Waste Solidification Building (WSB) for the treatment and solidification of the radioactive liquid waste streams generated by the Pit Disassembly and Conversion Facility (PDCF) and Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). The waste streams will be mixed with a cementitious dry mix in a 55-gallon waste container. Savannah River National Laboratory (SRNL) has been performing the testing and evaluations to support technical decisions for the WSB. Engineering Modeling & Simulation Group was requested to evaluate the thermal performance of the 55-gallon drum containing hydration heat source associated with the current baseline cement waste form. A transient axi-symmetric heat transfer model for the drum partially filled with waste form cement has been developed and heat transfer calculations performed for the baseline design configurations. For this case, 65 percent of the drum volume was assumed to be filled with the waste form, which has transient hydration heat source, as one of the baseline conditions. A series of modeling calculations has been performed using a computational heat transfer approach. The baseline modeling results show that the time to reach the maximum temperature of the 65 percent filled drum is about 32 hours when a 43 C initial cement temperature is assumed to be cooled by natural convection with 27 C external air. In addition, the results computed by the present model were compared with analytical solutions. The modeling results will be benchmarked against the prototypic test results. The verified model will be used for the evaluation of the thermal performance for the WSB drum. Detailed results and the cases considered in the calculations will be discussed here.

Lee, S.

2009-06-01T23:59:59.000Z

184

Nuclear waste repository transparency technology test bed demonstrations at WIPP  

SciTech Connect

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

185

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

Science Conference Proceedings (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

186

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

187

Idaho Site Taps Old World Process to Treat Nuclear Waste | Department of  

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

Idaho Site Taps Old World Process to Treat Nuclear Waste Idaho Site Taps Old World Process to Treat Nuclear Waste Idaho Site Taps Old World Process to Treat Nuclear Waste September 9, 2013 - 12:00pm Addthis The Idaho site's sodium distillation system. The Idaho site's sodium distillation system. The top of a sodium distillation vessel, where waste enters the system. The top of a sodium distillation vessel, where waste enters the system. The Idaho site's sodium distillation system. The top of a sodium distillation vessel, where waste enters the system. IDAHO FALLS, Idaho - The EM program at the Idaho site is using an age-old process to treat transuranic (TRU) waste left over from nuclear reactor experiments. Developed in the first century and perfected by moonshiners in the 19th century, distillation will be used at the Idaho Nuclear Technology and

188

WASTE TREATMENT AND DISPOSAL PROBLEMS OF THE FUTURE NUCLEAR POWER INDUSTRY  

SciTech Connect

The elements of waste treatment and disposal are assessed which are expected to become important in the development of the nuclear power industry of the future. Growth of the nuclear power economy is considered along with composition and quantities of anticipated waste. In addition, the economic implications of waste disposal are considered. It is concluded that research should be concentrated on decontaminating off-gases and on conversion of wastes to a more suitable form than liquid for storage. (J.R.D.)

Bruce, F.R.

1959-01-28T23:59:59.000Z

189

National briefing summaries: Nuclear fuel cycle and waste management  

SciTech Connect

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

190

Flammability Control In A Nuclear Waste Vitrification System  

SciTech Connect

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

191

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

Science Conference Proceedings (OSTI)

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

NONE

1991-12-31T23:59:59.000Z

192

Hydrogen speciation in hydrated layers on nuclear waste glass  

DOE Green Energy (OSTI)

The hydration of an outer layer on nuclear waste glasses is known to occur during leaching, but the actual speciation of hydrogen (as water or hydroxyl groups) in these layers has not been determined. As part of the Nevada Nuclear Waste Storage Investigations Project, we have used infrared spectroscopy to determine hydrogen speciations in three nuclear waste glass compositions (SRL-131 & 165, and PNL 76-68), which were leached at 90{sup 0}C (all glasses) or hydrated in a vapor-saturated atmosphere at 202{sup 0}C (SRL-131 only). Hydroxyl groups were found in the surface layers of all the glasses. Molecular water was found in the surface of SRL-131 and PNL 76-68 glasses that had been leached for several months in deionized water, and in the vapor-hydrated sample. The water/hydroxyl ratio increases with increasing reaction time; molecular water makes up most of the hydrogen in the thick reaction layers on vapor-phase hydrated glass while only hydroxyl occurs in the least reacted samples. Using the known molar absorptivities of water and hydroxyl in silica-rich glass the vapor-phase layer contained 4.8 moles/liter of molecular water, and 0.6 moles water in the form hydroxyl. A 15 {mu}m layer on SRL-131 glass formed by leaching at 90{sup 0}C contained a total of 4.9 moles/liter of water, 2/3 of which was as hydroxyl. The unreacted bulk glass contains about 0.018 moles/liter water, all as hydroxyl. The amount of hydrogen added to the SRL-131 glass was about 70% of the original Na + Li content, not the 300% that would result from alkali=hydronium ion interdiffusion. If all the hydrogen is then assumed to be added as the result of alkali-H{sup +} interdiffusion, the molecular water observed may have formed from condensation of the original hydroxyl groups.

Aines, R.D.; Weed, H.C.; Bates, J.K.

1987-01-15T23:59:59.000Z

193

Thermodynamic data management system for nuclear waste disposal performance assessment  

Science Conference Proceedings (OSTI)

Thermodynamic property values for use in assessing the performance of a nuclear waste repository are described. More emphasis is on a computerized data base management system which facilitates use of the thermodynamic data in sensitivity analysis and other studies which critically assess the performance of disposal sites. Examples are given of critical evaluation procedures; comparison of apparent equilibrium constants calculated from the data base, with other work; and of correlations useful in estimating missing values of both free energy and enthalpy of formation for aqueous species. 49 refs., 11 figs., 6 tabs.

Phillips, S.L.; Hale, F.V.; Siegel, M.D.

1988-04-01T23:59:59.000Z

194

Ultrafiltration treatment for liquid laundry wastes from nuclear power stations  

SciTech Connect

The authors conduct a comprehensive analysis of the waste constituents--radioactive and organic--of the laundry water resulting from the on-site laundering and decontamination of clothing worn in nuclear power plants. The primary isotope contaminants consist of niobium and zirconium 95, manganese 54, cobalt 60, iron 59, and cesium 134 and 137. A variety of filter and adsorbent materials used in an ultrafiltration process are comparatively tested for their effectiveness in removing not only these isotopes but also the organic contaminants in the process of recycling the water. Those materials consist of copper hexacyanoferrate, polyacrylophosphonic acid, and several metal-polymer complexes.

Kichik, V.A.; Maslova, M.N.; Svittsov, A.A.; Kuleshov, N.F.

1988-03-01T23:59:59.000Z

195

Double Diffusive Natural Convection in a Nuclear Waste Repository  

Science Conference Proceedings (OSTI)

In this study, we conduct a two-dimensional numerical analysis of double diffusive natural convection in an emplacement drift for a nuclear waste repository. In-drift heat and moisture transport is driven by combined thermal- and compositional-induced buoyancy forces. Numerical results demonstrate buoyancy-driven convective flow patterns and configurations during both repository heat-up and cool-down phases. It is also shown that boundary conditions, particularly on the drip-shield surface, have strong impacts on the in-drift convective flow and transport.

Y. Hao; J. Nitao; T.A. Buscheck; Y. Sun

2006-03-28T23:59:59.000Z

196

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

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

GC-52 provides legal advice to DOE regarding the long-term management of high-level radioactive waste (HLW) and spent nuclear fuel (SNF). SNF is nuclear fuel that has been used as fuel in a reactor...

197

Liquidus Temperature Studies for High Level Nuclear Waste Glasses  

Science Conference Proceedings (OSTI)

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

198

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

199

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

200

Disposition of nuclear waste using subcritical accelerator-driven systems  

Science Conference Proceedings (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

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


201

Separation of technetium from nuclear waste stream simulants. Final report  

Science Conference Proceedings (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

202

Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project  

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

Oversight Assessment of Oversight Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project May 2011 January 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Independent Oversight Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project

203

Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project  

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

Oversight Assessment of Oversight Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project May 2011 January 2013 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Independent Oversight Assessment of Nuclear Safety Culture at the Salt Waste Processing Facility Project

204

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center  

Science Conference Proceedings (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 M. C. Swenson

2007-06-01T23:59:59.000Z

205

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

206

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

207

Application of rock melting to construction of storage holes for nuclear waste  

Science Conference Proceedings (OSTI)

Rock melting technology can provide in-situ glass liners in nuclear waste package emplacement holes to reduce permeability and increase borehole stability. Reduction of permeability would reduce the time and probability of groundwater contacting the waste packages. Increasing the stability of the storage boreholes would enhance the retrievability of the nuclear waste packages. The rock melting hole forming technology has already been tested in volcanic tuff similar to the geology at the proposed nuclear waste repository at Yucca Mountain, Nevada. 6 refs., 5 figs., 2 tabs.

Neudecker, J.W. Jr.

1988-12-31T23:59:59.000Z

208

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

209

GC Commits to Transparency on Nuclear Waste Fund Fee Adequacy Decisions |  

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

GC Commits to Transparency on Nuclear Waste Fund Fee Adequacy GC Commits to Transparency on Nuclear Waste Fund Fee Adequacy Decisions GC Commits to Transparency on Nuclear Waste Fund Fee Adequacy Decisions February 26, 2010 - 3:17pm Addthis 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 will be made available to the public on DOE's website shortly after DOE makes a determination. The report relied upon in determining fee adequacy for 2008, the most recent year for which DOE has made a determination, is available here: (2008 Fee Adequacy Letter Report). Addthis Related Articles DOE Completes Annual Determination of the Adequacy of the Nuclear Waste Fund Fee Department of Energy Files Motion to Withdraw Yucca Mountain License

210

EXPERIMENTAL METHODS TO ESTIMATE ACCUMULATED SOLIDS IN NUCLEAR WASTE TANKS  

SciTech Connect

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

211

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

SciTech Connect

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

Bratzel, D.R.

1997-01-07T23:59:59.000Z

212

Assessment of nuclear safety and nuclear criticality potential in the Defense Waste Processing Facility. Revision 1  

SciTech Connect

The S-Area Defense Waste Processing Facility (DWPF) will initially process Batch 1 sludge in the sludge-only processing mode, with simulated non-radioactive Precipitate Hydrolysis, Aqueous (PHA) product, without the risk of nuclear criticality. The dilute concentration of fissile material in the sludge combined with excess of neutron absorbers during normal operations make criticality throughout the whole process incredible. Subsequent batches of the DWPF involving radioactive precipitate slurry and PHA will require additional analysis. Any abnormal or upset process operations, which are not considered in this report and could potentially separate fissile material, must be individually evaluated. Scheduled maintenance operation procedures are not considered to be abnormal.

Ha, B.C.

1993-07-20T23:59:59.000Z

213

Nuclear energy and radioactive waste disposal in the age of recycling  

Science Conference Proceedings (OSTI)

The magnitude of humanity's energy needs requires that we embrace a multitude of various energy sources and applications. For a variety of reasons, nuclear energy must be a major portion of the distribution, at least one third. The often-cited strategic hurdle to this approach is nuclear waste disposal. Present strategies concerning disposal of nuclear waste need to be changed if the world is to achieve both a sustainable energy distribution by 2040 and solve the largest environmental issue of the 21. century - global warming. It is hoped that ambitious proposals to replace fossil fuel power generation by alternatives will drop the percentage of fossil fuel use substantially, but the absolute amount of fossil fuel produced electricity must be kept at or below its present 10 trillion kW-hrs/year. Unfortunately, the rapid growth in consumption to over 30 trillion kW-hrs/year by 2040, means that 20 trillion kW-hrs/yr of non-fossil fuel generated power has to come from other sources. If half of that comes from alternative non-nuclear, non-hydroelectric sources (an increase of 3000%), then nuclear still needs to increase by a factor of four worldwide to compensate. Many of the reasons nuclear energy did not expand after 1970 in North America (proliferation, capital costs, operational risks, waste disposal, and public fear) are no longer a problem. The WIPP site in New Mexico, an example of a solution to the nuclear waste disposal issue, and also to public fear, is an operating deep geologic nuclear waste repository in the massive bedded salt of the Salado Formation. WIPP has been operating for eight years, and as of this writing, has disposed of over 50,000 m{sup 3} of transuranic waste (>100 nCi/g but <23 Curie/liter) including high activity waste. The Salado Formation is an ideal host for any type of nuclear waste, especially waste from recycled spent fuel. (authors)

Conca, James L. [New Mexico State University, CEMRC IEE, 1400 University Drive, Carlsbad New Mexico 88220 (United States); Apted, Michael [Monitor Scientific, 3900 S. Wadsworth, Denver, CO 80235 (United States)

2007-07-01T23:59:59.000Z

214

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

SciTech Connect

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

215

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

SciTech Connect

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

216

Idaho Chemical Processing Plant Spent Fuel and Waste Management Technology Development Program Plan  

SciTech Connect

The Department of Energy (DOE) has received spent nuclear fuel (SNF) at the Idaho Chemical Processing Plant (ICPP) for interim storage and reprocessing since 1953. Reprocessing of SNF has resulted in an existing inventory of 1.5 million gallons of radioactive sodium-bearing liquid waste and 3800 cubic meters (m{sup 3}) of calcine, in addition to the 768 metric tons (MT) of SNF and various other fuel materials in inventory. To date, the major activity of the ICPP has been the reprocessing of SNF to recover fissile uranium; however, recent changes in world events have diminished the demand to recover and recycle this material. As a result, DOE has discontinued reprocessing SNF for uranium recovery, making the need to properly manage and dispose of these and future materials a high priority. In accordance with the Nuclear Waste Policy Act (NWPA) of 1982, as amended, disposal of SNF and high-level waste (HLW) is planned for a geological repository. Preparation of SNF, HLW, and other radioactive wastes for disposal may include mechanical, physical, and/or chemical processes. This plan outlines the program strategy of the ICPP Spent Fuel and Waste Management Technology Development Program (SF&WMTDP) to develop and demonstrate the technology required to ensure that SNF and radioactive waste will properly stored and prepared for final disposal. Program elements in support of acceptable interim storage and waste minimization include: developing and implementing improved radioactive waste treatment technologies; identifying and implementing enhanced decontamination and decommissioning techniques; developing radioactive scrap metal (RSM) recycle capabilities; and developing and implementing improved technologies for the interim storage of SNF.

1993-09-01T23:59:59.000Z

217

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

218

Depleted Uraniuim Dioxide as a Spent-Nuclear-Fuel-Waste Package...  

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

15 DEPLETED URANIUM DIOXIDE AS A SPENT-NUCLEAR-FUEL WASTE-PACKAGE PARTICULATE FILL: FILL BEHAVIOR Charles W. Forsberg Oak Ridge National Laboratory * P.O. Box 2008 Oak Ridge,...

219

Nuclear Safety R&D in the Waste Processing Technology Development & Deployment Program  

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

R&D in the Waste Processing R&D in the Waste Processing Technology Development & Deployment Program Presentation to the DOE High Level Waste Corporate Board July 29, 2009 Al Baione Office of Waste Processing DOE-EM Office of Engineering & Technology 2 Outline Nuclear Safety Research & Development Overview Summary of EM- NSR&D Presentations from February 2009 Evaluating Performance of Nuclear Grade HEPA Filters under Fire/Smoke Challenge Conditions Structural Integrity Initiative for HLW Tanks Pipeline Plugging and Prevention Advanced Mixing Models Basic Science Opportunities in HLW Storage and Processing Safety Cementitious Barriers Partnership 3 Nuclear Safety Research & Development Overview DNFSB 2004-1 identified need for renewed DOE attention to nuclear safety R&D

220

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

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


221

Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 3  

SciTech Connect

OAK-B135 Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 3(NOTE: Part II A item 1 indicates ''PAPER'', but a report is attached electronically)

Ulm, Franz-Josef

2000-03-31T23:59:59.000Z

222

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

223

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

224

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

225

Method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions  

DOE Patents (OSTI)

A method for recovering palladium and technetium values from nuclear fuel reprocessing waste solutions containing these and other values by contacting the waste solution with an extractant of tricaprylmethylammonium nitrate in an inert hydrocarbon diluent which extracts the palladium and technetium values from the waste solution. The palladium and technetium values are recovered from the extractant and from any other coextracted values with a strong nitric acid strip solution.

Horwitz, E. Philip (Elmhurst, IL); Delphin, Walter H. (Woodridge, IL)

1979-07-24T23:59:59.000Z

226

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

SciTech Connect

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

NONE

1995-02-01T23:59:59.000Z

227

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

Science Conference Proceedings (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

228

Disposition of nuclear waste using subcritical accelerator-driven systems  

Science Conference Proceedings (OSTI)

Studies have shown that the repository long-term radiological risk is from the long-lived transuranics and the fission products Tc-99 and I-129, thermal loading concerns arise mainly form the short-lived fission products Sr-90 and Cs-137. In relation to the disposition of nuclear waste, ATW is expected to accomplish the following: (1) destroy over 99.9% of the actinides; (2) destroy over 99.9% of the Tc and I; (3) separate Sr and Cs (short half-life isotopes); (4) separate uranium; (5) produce electricity. In the ATW concept, spent fuel would be shipped to a ATW site where the plutonium, other transuranics and selected long-lived fission products would be destroyed by fission or transmutation in their only pass through the facility. This approach contrasts with the present-day reprocessing practices in Europe and Japan, during which high purity plutonium is produced and used in the fabrication of fresh mixed-oxide fuel (MOX) that is shipped off-site for use in light water reactors.

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

1998-12-01T23:59:59.000Z

229

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

SciTech Connect

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, J.H. Jr.; Clark, D.E.

1981-04-01T23:59:59.000Z

230

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

231

Revised rail-stop exposure model for incident-free transport of nuclear waste  

SciTech Connect

This report documents a model for estimating railstop doses that occur during incident-free transport of nuclear waste by rail. The model, which has been incorporated into the RADTRAN III risk assessment code, can be applied to general freight and dedicated train shipments of waste.

Ostmeyer, R.M.

1986-02-01T23:59:59.000Z

232

Modeling of Sulfate Double-salts in Nuclear Wastes  

Science Conference Proceedings (OSTI)

Due to limited tank space at Hanford and Savannah River, the liquid nuclear wastes or supernatants have been concentrated in evaporators to remove excess water prior to the hot solutions being transferred to underground storage tanks. As the waste solutions cooled, the salts in the waste exceeded the associated solubility limits and precipitated in the form of saltcakes. The initial step in the remediation of these saltcakes is a rehydration process called saltcake dissolution. At Hanford, dissolution experiments have been conducted on small saltcake samples from five tanks. Modeling of these experimental results, using the Environmental Simulation Program (ESP), are being performed at the Diagnostic Instrumentation and Analysis Laboratory (DIAL) at Mississippi State University. The River Protection Project (RPP) at Hanford will use these experimental and theoretical results to determine the amount of water that will be needed for its dissolution and retrieval operations. A comprehensive effort by the RPP and the Tank Focus Area continues to validate and improve the ESP and its databases for this application. The initial effort focused on the sodium, fluoride, and phosphate system due to its role in the formation of pipeline plugs. In FY 1999, an evaluation of the ESP predictions for sodium fluoride, trisodium phosphate dodecahydrate, and natrophosphate clearly indicated that improvements to the Public database of the ESP were needed. One of the improvements identified was double salts. The inability of any equilibrium thermodynamic model to properly account for double salts in the system can result in errors in the predicted solid-liquid equilibria (SLE) of species in the system. The ESP code is evaluated by comparison with experimental data where possible. However, data does not cover the range of component concentrations and temperatures found in many tank wastes. Therefore, comparison of ESP with another code is desirable, and may illuminate problems with both. For this purpose, the SOLGASMIX code was used in conjunction with a small private database developed at ORNL. This code calculates thermodynamic equilibria through minimization of Gibbs Energy, and utilizes the Pitzer model for activity coefficients. The sodium nitrate-sulfate double salt and the sodium fluoride-sulfate double salt were selected for the FY 2000 validation study of ESP. Even though ESP does not include the sulfate-nitrate double salt, this study found that this omission does not appear to be a major consequence. In this case, the solubility predictions with and without the sulfate-nitrate double salt are comparable. In contrast, even though the sulfate-fluoride double salt is included within the ESP databank, comparison to previous experimental results indicates that ESP underestimates solubility. Thus, the prediction for the sulfate-fluoride system needs to be improved. A main consequence of the inability to accurately predict the SLE of double salts is its impact on the predicted ionic strength of the solution. The ionic strength has been observed to be an important factor in the formation of pipeline plugs. To improve the ESP prediction, solubility tests on the sulfate-fluoride system are underway at DIAL, and these experimental results will be incorporated into the Public database by OLI System, Inc. Preliminary ESP simulations also indicated difficulties with the SLE prediction for anhydrous sodium sulfate. The Public database for the ESP does not include fundamental parameters for this solid in mixed solutions below 32.4 C. The limitation, in the range of anhydrous sodium sulfate, leads to convergence problems in ESP and to inaccurate predictions of solubility near the invariant point when sodium sulfate decahydrate and other salts, such as sodium nitrate, were present. These difficulties were partially corrected through the use of an additional database. In conclusion, these results indicate the need for experimental data at temperatures above 25 C and in solutions containing both nitrate and hydroxide. Furthermore, the validation and do

Toghiani, B.

2000-10-30T23:59:59.000Z

233

Nuclear Solid Waste Processing Design at the Idaho Spent Fuels Facility  

Science Conference Proceedings (OSTI)

A spent nuclear fuels (SNF) repackaging and storage facility was designed for the Idaho National Engineering and Environmental Laboratory (INEEL), with nuclear solid waste processing capability. Nuclear solid waste included contaminated or potentially contaminated spent fuel containers, associated hardware, machinery parts, light bulbs, tools, PPE, rags, swabs, tarps, weld rod, and HEPA filters. Design of the nuclear solid waste processing facilities included consideration of contractual, regulatory, ALARA (as low as reasonably achievable) exposure, economic, logistical, and space availability requirements. The design also included non-attended transfer methods between the fuel packaging area (FPA) (hot cell) and the waste processing area. A monitoring system was designed for use within the FPA of the facility, to pre-screen the most potentially contaminated fuel canister waste materials, according to contact- or non-contact-handled capability. Fuel canister waste materials which are not able to be contact-handled after attempted decontamination will be processed remotely and packaged within the FPA. Noncontact- handled materials processing includes size-reduction, as required to fit into INEEL permitted containers which will provide sufficient additional shielding to allow contact handling within the waste areas of the facility. The current design, which satisfied all of the requirements, employs mostly simple equipment and requires minimal use of customized components. The waste processing operation also minimizes operator exposure and operator attendance for equipment maintenance. Recently, discussions with the INEEL indicate that large canister waste materials can possibly be shipped to the burial facility without size-reduction. New waste containers would have to be designed to meet the drop tests required for transportation packages. The SNF waste processing facilities could then be highly simplified, resulting in capital equipment cost savings, operational time savings, and significantly improved ALARA exposure.

Dippre, M. A.

2003-02-25T23:59:59.000Z

234

Calcine Waste Storage at the Idaho Nuclear Technology and Engineering Center  

SciTech Connect

This report documents an inventory of calcined waste produced at the Idaho Nuclear Technology and Engineering Center during the period from December 1963 to May 2000. The report was prepared based on calciner runs, operation of the calcined solids storage facilities, and miscellaneous operational information that establishes the range of chemical compositions of calcined waste stored at Idaho Nuclear Technology and Engineering Center. The report will be used to support obtaining permits for the calcined solids storage facilities, possible treatment of the calcined waste at the Idaho National Engineering and Environmental Laboratory, and to ship the waste to an off-site facility including a geologic repository. The information in this report was compiled from calciner operating data, waste solution analyses and volumes calcined, calciner operating schedules, calcine temperature monitoring records, and facility design of the calcined solids storage facilities. A compact disk copy of this report is provided to facilitate future data manipulations and analysis.

Staiger, Merle Daniel; M. C. Swenson

2005-01-01T23:59:59.000Z

235

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

SciTech Connect

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

Ha, B.C.

1994-10-21T23:59:59.000Z

236

An Evaluation of Alternative Classification Methods for Routine Low Level Waste from the Nuclear Power Industry  

Science Conference Proceedings (OSTI)

This report investigates the feasibility of classifying all routine nuclear power plant low level waste, including Class B and Class C waste, as Class A low level waste within the framework of NRC regulatory requirements. A change in classification could expand disposal venues and reduce the uncertainty of future disposal. The report shows that all of the waste, when managed as a composite stream, will meet the requirements for Class A disposal without leaving a portion of the stream orphaned to on-site ...

2007-11-19T23:59:59.000Z

237

Coating crystalline nuclear waste forms to improve inertness  

Science Conference Proceedings (OSTI)

Crystalline waste forms of high simulated waste loading were successfully coated with layers of pyrolytic carbon and silicon carbide. Sol-gel technology was used to produce microspheres that contained simulated waste. A separate process for cesium immobilization was developed, which loads 5 wt % Cs onto zeolite particles for subsequent coating. The chemical vapor deposition process was developed for depositing thin layers of carbon and silicon carbide onto particles in a fluidized-bed coater. Pyrolytic carbon-coated particles were extremely inert in numerous leach tests. Aqueous leach test results of coated waste forms were below detection limits of such sensitive analytical techniques as atomic absorption and inductively coupled plasma atomic emission.

Stinton, D.P.; Angelini, P.; Caputo, A.J.; Lackey, W.J.

1981-01-01T23:59:59.000Z

238

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

SciTech Connect

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

239

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

SciTech Connect

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

240

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

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

The Department of Energy's Nuclear The Department of Energy's Nuclear Waste Fund's Fiscal Year 2011 Financial Statements OAS-FS-12-03 November 2011 Department of Energy Washington, DC 20585 November 21, 2011 MEMORANDUM FOR THE DIRECTOR, OFFICE OF STANDARD CONTRACT MANAGEMENT, OFFICE OF GENERAL COUNSEL FROM: Rickey R. Hass Deputy Inspector General for Audits and Inspections Office of Inspector General SUBJECT: INFORMATION: Report on "The Department of Energy's Nuclear Waste Fund's Fiscal Year 2011 Financial Statements" The attached report presents the results of the independent certified public accountants' audit of the Department of Energy's Nuclear Waste Fund's (Fund) Fiscal Year 2011 balance sheet and the related statements of net cost, changes in net position, and budgetary resources.

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


241

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

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

Nuclear Nuclear Waste Fund's Fiscal Year 2012 Financial Statements OAS-FS-13-05 November 2012 U.S. Department of Energy Office of Inspector General Office of Audits & Inspections Department of Energy Washington, DC 20585 November 28, 2012 MEMORANDUM FOR THE DIRECTOR, OFFICE OF STANDARD CONTRACT MANAGEMENT, OFFICE OF GENERAL COUNSEL FROM: Rickey R. Hass Deputy Inspector General for Audits and Inspections Office of Inspector General SUBJECT: INFORMATION: Audit Report on "Department of Energy's Nuclear Waste Fund's Fiscal Year 2012 Financial Statement Audit" The attached report presents the results of the independent certified public accountants' audit of the Department of Energy's Nuclear Waste Fund's (Fund) Fiscal Year 2012 balance sheet and the

242

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

SciTech Connect

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

243

Secondary Waste Considerations for Vitrification of Sodium-Bearing Waste at the Idaho Nuclear Technology and Engineering Center FY-2001 Status Report  

SciTech Connect

The Idaho Nuclear Technology and Engineering Center (INTEC) is considering vitrification to process liquid sodium-bearing waste. Preliminary studies were completed to evaluate the potential secondary wastes from the melter off-gas clean up systems. Projected secondary wastes comprise acidic and caustic scrubber solutions, HEPA filters, activated carbon, and ion exchange media. Possible treatment methods, waste forms, and disposal sites are evaluated from radiological and mercury contamination estimates.

Herbst, A.K.; Kirkham, R.J.; Losinski, S.J.

2002-09-26T23:59:59.000Z

244

Secondary Waste Considerations for Vitrification of Sodium-Bearing Waste at the Idaho Nuclear Techology and Engineering Center FY-2001 Status Report  

SciTech Connect

The Idaho Nuclear Technology and Engineering Center (INTEC) is considering vitrification to process liquid sodium-bearing waste. Preliminary studies were completed to evaluate the potential secondary wastes from the melter off-gas clean up systems. Projected secondary wastes comprise acidic and caustic scrubber solutions, HEPA filters, activated carbon, and ion exchange media. Possible treatment methods, waste forms, and disposal sites are evaluated from radiological and mercury contamination estimates.

Herbst, Alan Keith; Kirkham, Robert John; Losinski, Sylvester John

2001-09-01T23:59:59.000Z

245

Solid radioactive waste management facility design for managing CANDU{sup R} 600 MW nuclear generating station re-tube/refurbishment Waste Streams  

Science Conference Proceedings (OSTI)

The main design features of the re-tube canisters, waste handling equipment and waste containers designed by Atomic Energy of Canada Limited (AECL{sup R}) and implemented in support of the re-tube/refurbishment activities for Candu 600 MW nuclear generating stations are described in this paper. The re-tube/refurbishment waste characterization and the waste management principles, which form the basis of the design activities, are also briefly outlined. (authors)

Pontikakis, N.; Hopkins, J.; Scott, D.; Bajaj, V.; Nosella, L. [AECL, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

2007-07-01T23:59:59.000Z

246

Intermediate long-lived nuclear waste management: an ...  

management: an integrated ... Nuclear Energy Division / Department of Physico-Chemistry. ... CIMETAL Project Air carbonation phenomenology Transport ...

247

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

SciTech Connect

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

248

Method for utilizing decay heat from radioactive nuclear wastes  

DOE Patents (OSTI)

Management of radioactive heat-producing waste material while safely utilizing the heat thereof is accomplished by encapsulating the wastes after a cooling period, transporting the capsules to a facility including a plurality of vertically disposed storage tubes, lowering the capsules as they arrive at the facility into the storage tubes, cooling the storage tubes by circulating a gas thereover, employing the so heated gas to obtain an economically beneficial result, and continually adding waste capsules to the facility as they arrive thereat over a substantial period of time.

Busey, H.M.

1974-10-14T23:59:59.000Z

249

Nuclear data for nuclear transmutation  

Science Conference Proceedings (OSTI)

Current status on nuclear data for the study of nuclear transmutation of radioactive wastes is reviewed

Hideo Harada

2009-01-01T23:59:59.000Z

250

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

Science Conference Proceedings (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

251

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

Science Conference Proceedings (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

252

Thermal impact of waste emplacement and surface cooling associated with geologic disposal of nuclear waste  

Science Conference Proceedings (OSTI)

The thermal effects associated with the emplacement of aged radioactive wastes in a geologic repository were studied, with emphasis on the following subjects: the waste characteristics, repository structure, and rock properties controlling the thermally induced effects; the current knowledge of the thermal, thermomechanical, and thermohydrologic impacts, determined mainly on the basis of previous studies that assume 10-year-old wastes; the thermal criteria used to determine the repository waste loading densities; and the technical advantages and disadvantages of surface cooling of the wastes prior to disposal as a means of mitigating the thermal impacts. The waste loading densities determined by repository designs for 10-year-old wastes are extended to older wastes using the near-field thermomechanical criteria based on room stability considerations. Also discussed are the effects of long surface cooling periods determined on the basis of far-field thermomechanical and thermohydrologic considerations. The extension of the surface cooling period from 10 years to longer periods can lower the near-field thermal impact but have only modest long-term effects for spent fuel. More significant long-term effects can be achieved by surface cooling of reprocessed high-level waste.

Wang, J.S.Y.; Mangold, D.C.; Spencer, R.K.; Tsang, C.F.

1982-08-01T23:59:59.000Z

253

Disposition of Nuclear Waste Using Subcritical Accelerator-Driven Systems: Technology Choices and Implementation Scenarios  

SciTech Connect

Los Alamos National Laboratory has led the development of accelerator-driven transmutation of waste (ATW) to provide an alternative technological solution to the disposition of nuclear waste. While ATW will not eliminate the need for a high-level waste repository, it offers a new technology option for altering the nature of nuclear waste and enhancing the capability of a repository. The basic concept of ATW focuses on reducing the time horizon for the radiological risk from hundreds of thousands of years to a few hundred years and on reducing the thermal loading. As such, ATW will greatly reduce the amount of transuranic elements that will be disposed of in a high-level waste repository. The goal of the ATW nuclear subsystem is to produce three orders of magnitude reduction in the long-term radiotoxicity of the waste sent to a repository, including losses through processing. If the goal is met, the radiotoxicity of ATW-treated waste after 300 yr would be less than that of untreated waste after 100 000 yr.These objectives can be achieved through the use of high neutron fluxes produced in accelerator-driven subcritical systems. While critical fission reactors can produce high neutron fluxes to destroy actinides and select fission products, the effectiveness of the destruction is limited by the criticality requirement. Furthermore, a substantial amount of excess reactivity would have to be supplied initially and compensated for by control poisons. To overcome these intrinsic limitations, we searched for solutions in subcritical systems freed from the criticality requirement by taking advantage of the recent breakthroughs in accelerator technology and the release of liquid lead/bismuth nuclear coolant technology from Russia. The effort led to the selection of an accelerator-driven subcritical system that results in the destruction of the actinides and fission products of concern as well as permitting easy operational control through the external control of the neutron source.

Venneri, Francesco; Williamson, Mark A.; Li Ning; Houts, Michael G.; Morley, Richard A.; Beller, Denis E.; Sailor, William; Lawrence, George [Los Alamos National Laboratory (United States)

2000-10-15T23:59:59.000Z

254

Method for forming microspheres for encapsulation of nuclear waste. [Patent application  

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. Fuel particles were also produced using this method.

Angelini, P.; Caputo, A.J.; Hutchens, R.E.; Lackey, W.J.; Stinton, D.P.

1982-01-29T23:59:59.000Z

255

Review and Demonstration of Korea Hydro & Nuclear Power (KHNP) Vitrification Technology for Low Level Waste Treatment  

Science Conference Proceedings (OSTI)

Vitrification is the process of stabilizing nuclides in a glass matrix in order to enhance disposal options. A mature technology, vitrification has been applied to high level radioactive waste (HLW) for more than 40 years. As disposal costs and public concern for the environment increase, vitrification is considered to be a promising technology for low level waste (LLW) stabilization. This report covers the characteristics of LLW generated from nuclear power plants, current melter technologies ...

2013-08-14T23:59:59.000Z

256

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

SciTech Connect

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

257

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

SciTech Connect

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

258

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

SciTech Connect

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

Dees, L.A.

1994-08-15T23:59:59.000Z

259

Assessment of Nuclear Safety Culture at the Salt Waste Processing...  

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

BARS Behavioral Anchored Rating Scales DNFSB Defense Nuclear Facilities Safety Board DOE U.S. Department of Energy DPO Differing Professional Opinion ECP Employee Concern...

260

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

DOE Green Energy (OSTI)

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

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

1992-01-01T23:59:59.000Z

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


261

A STUDY OF CORROSION AND STRESS CORROSION CRACKING OF CARBON STEEL NUCLEAR WASTE STORAGE TANKS  

SciTech Connect

The Hanford reservation Tank Farms in Washington State has 177 underground storage tanks that contain approximately 50 million gallons of liquid legacy radioactive waste from cold war plutonium production. These tanks will continue to store waste until it is treated and disposed. These nuclear wastes were converted to highly alkaline pH wastes to protect the carbon steel storage tanks from corrosion. However, the carbon steel is still susceptible to localized corrosion and stress corrosion cracking. The waste chemistry varies from tank to tank, and contains various combinations of hydroxide, nitrate, nitrite, chloride, carbonate, aluminate and other species. The effect of each of these species and any synergistic effects on localized corrosion and stress corrosion cracking of carbon steel have been investigated with electrochemical polarization, slow strain rate, and crack growth rate testing. The effect of solution chemistry, pH, temperature and applied potential are all considered and their role in the corrosion behavior will be discussed.

BOOMER, K.D.

2007-08-21T23:59:59.000Z

262

Management of radioactive waste gases from the nuclear fuel cycle. Volume I. Comparison of alternatives  

SciTech Connect

Alternatives were compared for collection and fixation of radioactive waste gases released during normal operation of the nuclear fuel cycle, and for transportation and storage/disposal of the resulting waste forms. The study used a numerical rating scheme to evaluate and compare the alternatives for krypton-85, iodine-129, and carbon-14; whereas a subjective evaluation, based on published reports and engineering judgement, was made for transportation and storage/disposal options. Based on these evaluations, certain alternatives are recommended for an integrated scheme for waste management of each of the subject waste gases. Phase II of this project, which is concerned with the development of performance criteria for the waste forms associated with the subject gases, will be completed by the end of 1980. This work will be documented as Volume II of this report.

Evans, A.G.; Prout, W.E.; Buckner, J.T.; Buckner, M.R.

1980-12-01T23:59:59.000Z

263

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

264

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

SciTech Connect

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

265

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

266

Supplemnental Volume - Independent Oversight Assessment of the Nuclear Safety Culture and Management of Nuclear Safety Concerns at the Hanford Site Waste Treatment and Immobilization Plant, January 2012  

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

Volume Volume Independent Oversight Assessment of Nuclear Safety Culture and Management of Nuclear Safety Concerns at the Hanford Site Waste Treatment and Immobilization Plant January 2012 Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Office of Health, Safety and Security HSS i Independent Oversight Assessment of Safety Culture and Management of Nuclear Safety Concerns at the Hanford Site Waste Treatment and Immobilization Plant Supplemental Volume Table of Contents Foreword ...................................................................................................................................................... iii Acronyms ...................................................................................................................................................... v

267

U.S. Nuclear Waste Technical Review Board Correspondence with  

E-Print Network (OSTI)

Symposium on Zirconium in the Nuclear Industry: Fourteenth International Symposium on 13­17 June 2004 deformation, thus also avoiding friction effects. In this study, the steel punch had a diameter of 25.4 mm Zircaloy-4," 11th International Symposium on Zr in the Nuclear Industry, ASTM STP 1295, Garmisch

268

NEAMS Nuclear Waste Management IPSC : evaluation and selection of tools for the quality environment.  

Science Conference Proceedings (OSTI)

The objective of the U.S. Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation Nuclear Waste Management Integrated Performance and Safety Codes (NEAMS Nuclear Waste Management 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. These M&S capabilities are to be managed, verified, and validated within the NEAMS Nuclear Waste Management IPSC quality environment. M&S capabilities and the supporting analysis workflow and simulation data management tools will be distributed to end-users from this same quality environment. The same analysis workflow and simulation data management tools that are to be distributed to end-users will be used for verification and validation (V&V) activities within the quality environment. This strategic decision reduces the number of tools to be supported, and increases the quality of tools distributed to end users due to rigorous use by V&V activities. This report documents an evaluation of the needs, options, and tools selected for the NEAMS Nuclear Waste Management IPSC quality environment. The objective of the U.S. Department of Energy (DOE) Office of Nuclear Energy Advanced Modeling and Simulation Nuclear Waste Management Integrated Performance and Safety Codes (NEAMS Nuclear Waste Management IPSC) program element is to provide an integrated suite of computational modeling and simulation (M&S) capabilities to assess quantitatively the long-term performance of waste forms in the engineered and geologic environments of a radioactive-waste storage facility or disposal repository. This objective will be fulfilled by acquiring and developing M&S capabilities, and establishing a defensible level of confidence in these M&S capabilities. The foundation for assessing the level of confidence is based upon the rigor and results from verification, validation, and uncertainty quantification (V&V and UQ) activities. M&S capabilities are to be managed, verified, and validated within the NEAMS Nuclear Waste Management IPSC quality environment. M&S capabilities and the supporting analysis workflow and simulation data management tools will be distributed to end-users from this same quality environment. The same analysis workflow and simulation data management tools that are to be distributed to end-users will be used for verification and validation (V&V) activities within the quality environment. This strategic decision reduces the number of tools to be supported, and increases the quality of tools distributed to end users due to rigorous use by V&V activities. NEAMS Nuclear Waste Management IPSC V&V and UQ practices and evidence management goals are documented in the V&V Plan. This V&V plan includes a description of the quality environment into which M&S capabilities are imported and V&V and UQ activities are managed. The first phase of implementing the V&V plan is to deploy an initial quality environment through the acquisition and integration of a set of software tools. An evaluation of the needs, options, and tools selected for the quality environment is given in this report.

Bouchard, Julie F.; Stubblefield, William Anthony; Vigil, Dena M.; Edwards, Harold Carter (Org. 1444 : Multiphysics Simulation Technology)

2011-05-01T23:59:59.000Z

269

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

E-Print Network (OSTI)

for each of T M W three nuclear fuel cycles Decay heat powerfor d i f f e r e n t nuclear fuel cycles for a PWR. Decayd i f f e r e n t nuclear fuel cycles for a BWR. Relative

Wang, J.S.Y.

2010-01-01T23:59:59.000Z

270

Present and future roles of solvent extraction in treatment of nuclear wastes  

SciTech Connect

Solvent extraction has played a major role in development of the nuclear industry and has recovered much of the uranium from raw materials and essentially all of the plutonium and uranium from spent fuels. These operations produced a wide variety of radioactive wastes as well as the uranium and plutonium products. Solvent extraction worked well in the earlier nuclear facilities and should play a significant role in future cleanup operations.

Watson, J.S.

1995-12-31T23:59:59.000Z

271

Depleted uranium oxides as spent-nuclear-fuel waste-package invert and backfill materials  

SciTech Connect

A new technology has been proposed in which depleted uranium, in the form of oxides or silicates, is placed around the outside of the spent nuclear fuel waste packages in the geological repository. This concept may (1) reduce the potential for repository nuclear criticality events and (2) reduce long-term release of radionuclides from the repository. As a new concept, there are significant uncertainties.

Forsberg, C.W.; Haire, M.J.

1997-07-07T23:59:59.000Z

272

Nuclear waste transportation: Case studies of identifying stakeholder risk information needs  

E-Print Network (OSTI)

The U.S. Department of Energy (DOE) is responsible for the cleanup of our nation’s nuclear legacy, involving complex decisions about how and where to dispose of nuclear waste and how to transport it to its ultimate disposal site. It is widely recognized that a broad range of stakeholders and tribes should be involved in this kind of decision. All too frequently, however, stakeholders and tribes are only invited to participate by commenting on processes and activities that are near completion; they are not included in the problem formulation stages. Moreover, it is often assumed that high levels of complexity and uncertainty prevent meaningful participation by these groups. Considering the types of information that stakeholders and tribes need to be able to participate in the full life cycle of decision making is critical for improving participation and transparency of decision making. Toward this objective, the Consortium for Risk Evaluation with Stakeholder Participation (CRESP) participated in three public processes relating to nuclear waste transportation and disposal in 1997–1998. First, CRESP organized focus groups to identify concerns about nuclear waste transportation. Second, CRESP conducted exit surveys at regional public workshops held by DOE to get input from stakeholders on intersite waste transfer issues. Third, CRESP developed visual tools to synthesize technical information and allow stakeholders and tribes with varying levels of knowledge about nuclear waste to participate in meaningful discussion. In this article we share the results of the CRESP findings, discuss common themes arising from these interactions, and comment on special considerations needed to facilitate stakeholder and tribal participation in similar decision-making processes. Key words: environmental information, hazardous waste, risk communication, risk perception, stakeholders. Environ Health Perspect

Christina H. Drew; Deirdre A. Grace; Susan M. Silbernagel; Erin S. Hemmings; Alan Smith; William C. Griffith; Timothy K. Takaro; Elaine M. Faustman

2003-01-01T23:59:59.000Z

273

Plan for waste package environment for NNWSI [Nevada Nuclear Waste Storage Investigations  

SciTech Connect

The purpose and objective of the Waste Package Environment task is to establish and characterize the environmental processes affecting the near-field repository host rock after waste package emplacement. These processes, which reflect the perturbation induces in the environment by engineering effects and by the waste package decay heat and radiation, will influence chemical, mineralogical and hydrological features of the environment. The thermal and radiation output of the waste packages will change with time, resulting in an environment in which the chemical, mineralogical and physical attributes may also change through time. To assure that waste package design considerations reflect the characteristics of this evolving environment, it is necessary to determine the range of conditions that may develop in the pre- and post-emplacement waste package environment. To assure that the emplacement configurations do not compromise the lifetime of the repository or the waste packages, the design of the emplacement configuration must also consider the environmental features. Recognition of these requirements resulted in the development of the issue an information needs. 20 refs.

Glassley, W.E.

1988-02-01T23:59:59.000Z

274

Communication Between U.S. Nuclear Waste Technical Review Board  

E-Print Network (OSTI)

Lessons Learned from Yucca Mountain and Other Programs June 2011 A Report to Congress and the Secretary's efforts to develop a deep geologic repository for high-activity waste1 at Yucca Mountain in Nevada Commission (NRC) to construct a repository at Yucca Mountain. An important part of the Board's mission

275

Management of Salt Waste from Electrochemical Processing of Used Nuclear Fuel  

SciTech Connect

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

276

The Use of Transportable Processing Systems for the Treatment of Radioactive Nuclear Wastes  

Science Conference Proceedings (OSTI)

EnergySolutions has developed two major types of radioactive processing plants based on its experience in the USA and UK, and its exclusive North American access to the intellectual property and know-how developed over 50 years at the Sellafield nuclear site in the UK. Passive Secure Cells are a type of hot cell used in place of the Canyons typically used in US-designed radioactive facilities. They are used in permanent, large scale plants suitable for long term processing of large amounts of radioactive material. The more recently developed Transportable Processing Systems, which are the subject of this paper, are used for nuclear waste processing and clean-up when processing is expected to be complete within shorter timescales and when it is advantageous to be able to move the processing equipment amongst a series of geographically spread-out waste treatment sites. Such transportable systems avoid the construction of a monolithic waste processing plant which itself would require extensive decommissioning and clean-up when its mission is complete. This paper describes a range of transportable radioactive waste processing equipment that EnergySolutions and its partners have developed including: the portable MOSS drum-based waste grouting system, the skid mounted MILWPP large container waste grouting system, the IPAN skid-mounted waste fissile content non-destructive assay system, the Wiped Film Evaporator low liquid hold-up transportable evaporator system, the CCPU transportable solvent extraction cesium separation system, and the SEP mobile shielded cells for emptying radioactive debris from water-filled silos. Maximum use is made of proven, robust, and compact processing equipment such as centrifugal contactors, remote sampling systems, and cement grout feed and metering devices. Flexible, elastomer-based Hose-in-Hose assemblies and container-based transportable pump booster stations are used in conjunction with these transportable waste processing units for transferring radioactive waste from its source to the processing equipment. (authors)

Phillips, Ch.; Houghton, D.; Crawford, G. [EnergySolutions LLC., 2345 Stevens Drive, Richland, WA (United States)

2008-07-01T23:59:59.000Z

277

Related Resources - Nuclear Data Program, Nuclear Engineering...  

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

278

Publications: Other Resources - Nuclear Data Program - Nuclear...  

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

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

279

Publications 2005 - Nuclear Data Program - Nuclear Engineering...  

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

280

Publications 2003 - Nuclear Data Program - Nuclear Engineering...  

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

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

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


281

Contacts - Nuclear Data Program, Nuclear Engineering Division...  

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

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

282

Publications 2001 - Nuclear Data Program - Nuclear Engineering...  

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

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

283

Publications 2004 - Nuclear Data Program - Nuclear Engineering...  

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

284

Publications 2009 - Nuclear Data Program - Nuclear Engineering...  

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

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

285

Nuclear Criticality Safety: Current Activities - Nuclear Engineering...  

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

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

286

Nuclear Criticality Safety - Nuclear Engineering Division (Argonne...  

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

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

287

Nuclear Systems Analysis - Nuclear Engineering Division (Argonne...  

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

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

288

Publications 2011 - Nuclear Data Program - Nuclear Engineering...  

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

289

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

E-Print Network (OSTI)

l e of both uranium and plutonium. The overall objective i sthe remaining uranium or plutonium, or both, can be recycledSF) contains both uranium and plutonium as waste components.

Wang, J.S.Y.

2010-01-01T23:59:59.000Z

290

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

SciTech Connect

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

291

Technical support for the Ukrainian State Committee for Nuclear Radiation Safety on specific waste issues  

Science Conference Proceedings (OSTI)

The government of Ukraine, a now-independent former member of the Soviet Union, has asked the United States to assist its State Committee for Nuclear and Radiation Safety (SCNRS) in improving its regulatory control in technical fields for which it has responsibility. The US Nuclear Regulatory Commission (NRC) is providing this assistance in several areas, including management of radioactive waste and spent fuel. Radioactive wastes resulting from nuclear power plant operation, maintenance, and decommissioning must be stored and ultimately disposed of appropriately. In addition, radioactive residue from radioisotopes used in various industrial and medical applications must be managed. The objective of this program is to provide the Ukrainian SCNRS with the information it needs to establish regulatory control over uranium mining and milling activities in the Zheltye Vody (Yellow Waters) area and radioactive waste disposal in the Pripyat (Chernobyl) area among others. The author of this report, head of the Environmental Technology Section, Health Sciences Research Division of Oak Ridge National Laboratory, accompanied NRC staff to Ukraine to meet with SCNRS staff and visit sites in question. The report highlights problems at the sites visited and recommends license conditions that SCNRS can require to enhance safety of handling mining and milling wastes. The author`s responsibility was specifically for the visit to Zheltye Vody and the mining and milling waste sites associated with that facility. An itinerary for the Zheltye Vody portion of the trip is included as Appendix A.

Little, C.A.

1995-07-01T23:59:59.000Z

292

Preliminary characterization of risks in the nuclear waste management system based on information in the literature  

Science Conference Proceedings (OSTI)

This document presents preliminary information on the radiological and nonradiological risks in the nuclear waste management system. The objective of the study was to (1) review the literature containing information on risks in the nuclear waste management system and (2) use this information to develop preliminary estimates of the potential magnitude of these risks. Information was collected on a broad range of risk categories to assist the US Department of Energy (DOE) in communicating information about the risks in the waste management systems. The study examined all of the portions of the nuclear waste management system currently expected to be developed by the DOE. The scope of this document includes the potential repository, the integral MRS facility, and the transportation system that supports the potential repository and the MRS facility. Relevant literature was reviewed for several potential repository sites and geologic media. A wide range of ``risk categories`` are addressed in this report: (1) public and occupational risks from accidents that could release radiological materials, (2) public and occupational radiation exposure resulting from routine operations, (3) public and occupational risks from accidents involving hazards other than radioactive materials, and (4) public and occupational risks from exposure to nonradioactive hazardous materials during routine operations. The report is intended to provide a broad spectrum of risk-related information about the waste management system. This information is intended to be helpful for planning future studies.

Daling, P.M.; Rhoads, R.E.; Van Luick, A.E.; Fecht, B.A.; Nilson, S.A.; Sevigny, N.L. [Pacific Northwest Lab., Richland, WA (United States); Armstrong, G.R. [Westinghouse Hanford Co., Richland, WA (United States); Hill, D.H.; Rowe, M.; Stern, E. [Brookhaven National Lab., Upton, NY (United States)

1992-01-01T23:59:59.000Z

293

Recovery of cesium and palladium from nuclear reactor fuel processing waste  

DOE Patents (OSTI)

A method of recovering cesium and palladium values from nuclear reactor fission product waste solution involves contacting the solution with a source of chloride ions and oxidizing palladium ions present in the solution to precipitate cesium and palladium as Cs.sub.2 PdCl.sub.6.

Campbell, David O. (Oak Ridge, TN)

1976-01-01T23:59:59.000Z

294

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

SciTech Connect

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

295

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

SciTech Connect

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

296

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

SciTech Connect

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

297

Management of radioactive waste from nuclear power plants  

SciTech Connect

Even thought risk assessment is an essential consideration in all projects involving radioactive or hazardous waste, its public role is often unclear, and it is not fully utilized in the decision-making process for public acceptance of such facilities. Risk assessment should be an integral part of such projects and should play an important role from beginning to end, i.e., from planning stages to the closing of a disposal facility. A conceptual model that incorporates all potential pathways of exposure and is based on site-specific conditions is key to a successful risk assessment. A baseline comparison with existing standards determines, along with other factors, whether the disposal site is safe. Risk assessment also plays a role in setting priorities between sites during cleanup actions and in setting cleanup standards for certain contaminants at a site. The applicable technologies and waste disposal designs can be screened through risk assessment.

Not Available

1993-09-01T23:59:59.000Z

298

Nuclear Waste Assessment System for Technical Evaluation (NUWASTE)  

E-Print Network (OSTI)

of depleted uranium within the country. The 15% that emerges as enriched uranium is converted into ceramic (total) 15.51­ 40.75 Depleted uranium reconversion 2.10­6.24 Packaging depleted uranium 0.25­3.11 Sequestration of depleted uranium 0.12­0.35 Sequestration of enrichment waste 0.16­0.44 Sequestration

299

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

SciTech Connect

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

300

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

SciTech Connect

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.

1983-08-01T23:59:59.000Z

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


301

Calcine Waste Storage at the Idaho Nuclear Technology and Engineering Center  

Science Conference Proceedings (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

302

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

Science Conference Proceedings (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

303

Appendix A U.S. Nuclear Waste Technical Review  

E-Print Network (OSTI)

engineering­transportation by MIT in 1980. From 1976 to 1980, he worked as a project manager and research of Westinghouse Hanford Company. Dr. Arnold was elected to the National Academy of Engineering in 1974 and a Ph.D. in nuclear engineering in 1984. He then worked for Lawrence Livermore National Laboratory

304

Other U.S. Nuclear Waste Technical Review Board Correspondence  

E-Print Network (OSTI)

and scientific validity of ac- tivities undertaken by the Secretary of Energy to characterize Yucca Mountain) and spent nuclear fuel (SNF). The U.S. Department of Energy (DOE) began studying Yucca Mountain features of Yucca Moun- tain led to its selection as a potential repository site: · Yucca Mountain now

305

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

306

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

307

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

SciTech Connect

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

308

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

E-Print Network (OSTI)

emplacement in t u f f s . Sandia National Laboratories,e a r waste in s h a l e . Sandia N a t i o n a l L a b o rof a WIPP mine d r i f t . Sandia N a t i o n a l L a b o r

Wang, J.S.Y.

2010-01-01T23:59:59.000Z

309

Disposition of Nuclear Waste Using Subcritical Accelerator-Driven Systems  

SciTech Connect

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

310

Long-term comparison of dissolution behavior between fully radioactive and simulated nuclear waste glasses  

SciTech Connect

The behavior of radioactive sludge-based and simulated nuclear waste glasses has been compared by long-term testing of radioactive and simulated compositions of Savannah River Laboratory 165, 131, and 200 glasses. Static tests at glass surface area-to-solution volume (SA/V) ratios of 340 and 2000 m[sup [minus]1] up to 720 days show little difference in reactivity between radioactive and simulated waste glasses. The same leach trends are observed for both glass types. The differences in reactivity at an SA/V of 2000 m[sup [minus]1] or below are not large enough to alter the order of glass durability for the different compositions nor to change the controlling glass dissolution processes. The small differences in reactivity between fully radioactive and simulated glasses can reasonably be explained if the controlling reaction process and leachate pH values are accounted for. However, at an SA/V of 20,000 m[sup [minus]1], the simulated nuclear waste glass, 200S, leaches faster than the corresponding radioactive glass by a factor of 40 within 1 yr. The accelerated reaction with the simulated glass 200S is associated with the formation of crystalline phases such as clinoptilolite (or K-feldspar), and a pH excursion. The radiation field generated by the fully radioactive glass reduces the solution pH, which, in turn, may retard the onset of the increased reaction rate. This result suggests that the fully radioactive nuclear waste glass 200R may be substantially more durable than the simulated 200S glass if the lower pH in the 200R leachate can be sustained. Meaningful comparison tests between radioactive and simulated nuclear waste glasses should include long-term and high SA/V tests.

Feng, Xiangdong; Bates, J.K.; Buck, E.C.; Bradley, C.R.; Gong, Meiling (Argonne National Lab., Argonne, IL (United States))

1993-11-01T23:59:59.000Z

311

Environmental Statements, Availability, Etc., Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs  

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

8679 8679 Thursday June 1, 1995 Part III Department of Energy Environmental Statements, Availability, Etc.; Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs: Notice 28680 Federal Register / Vol. 60, No. 105 / Thursday, June 1, 1995 / Notices DEPARTMENT OF ENERGY Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs AGENCY: Department of Energy. ACTION: Record of decision. SUMMARY: The Department of Energy has issued a Record of Decision on Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs. The Record of Decision includes a Department-wide decision to

312

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

Science Conference Proceedings (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

313

Truck and rail charges for shipping spent fuel and nuclear waste  

SciTech Connect

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

314

Heat pipe effects in nuclear waste isolation: a review  

SciTech Connect

The existence of fractures favors heat pipe development in a geologic repository as does a partially saturated medium. A number of geologic media are being considered as potential repository sites. Tuff is partially saturated and fractured, basalt and granite are saturated and fractured, salt is unfractured and saturated. Thus the most likely conditions for heat pipe formation occur in tuff while the least likely occur in salt. The relative permeability and capillary pressure dependences on saturation are of critical importance for predicting thermohydraulic behavior around a repository. Mineral redistribution in heat pipe systems near high-level waste packages emplaced in partially saturated formations may significantly affect fluid flow and heat transfer processes, and the chemical environment of the packages. We believe that a combined laboratory, field, and theoretical effort will be needed to identify the relevant physical and chemical processes, and the specific parameters applicable to a particular site. 25 refs., 1 fig.

Doughty, C.; Pruess, K.

1985-12-01T23:59:59.000Z

315

Application of exemption principles to low-level waste disposal and recycle of wastes from nuclear facilities  

Science Conference Proceedings (OSTI)

The International Atomic Energy Agency (IAEA) and other international groups are considering exempting from regulatory control certain radiation sources and practices, initially under the general heading of de minimis. A significant fraction of the wastes from industry, research, medicine, and the nuclear fuel cycle are contaminated to such low levels that the associated risks to health are trivial. IAEA work has been conducted by Advisory Groups to establish principles for exemption, and to apply the principles to various areas of waste management. In the second area, the main objectives have been to illustrate a methodology for developing practical radiological criteria through the application of the IAEA preliminary exemption principles, to establish generic criteria, and to determine the practicability of the preliminary exemption principles. The method used relies on a modeling assessment of the potential radiation exposure pathways and scenarios for individuals and population groups following the unrestricted release of materials. This paper describes the IAEA's assessment methodology and presents the generic results expressed in terms of the limiting activity concentration in municipal waste and in low-activity materials for recycle and reuse. 2 refs., 2 tabs.

Kennedy, W.E. Jr.; Hemming, C.R.; O'Donnell, F.R.; Linsley, G.S.

1988-04-01T23:59:59.000Z

316

Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center  

Science Conference Proceedings (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

317

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

Science Conference Proceedings (OSTI)

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

DOE

2002-12-01T23:59:59.000Z

318

Materials Characterization Center. Second workshop on irradiation effects in nuclear waste forms. Summary report  

SciTech Connect

The purpose of this second workshop on irradiations effects was to continue the discussions initiated at the first workshop and to obtain guidance for the Materials Characterization Center in developing test methods. The following major conclusions were reached: Ion or neutron irradiations are not substitutes for the actinide-doping technique, as described by the MCC-6 Method for Preparation and Characterization of Actinide-Doped Waste Forms, in the final evaluation of any waste form with respect to the radiation effects from actinide decay. Ion or neutron irradiations may be useful for screening tests or more fundamental studies. The use of these simulation techniques as screening tests for actinide decay requires that a correlation between ion or neutron irradiations and actinide decay be established. Such a correlation has not yet been established and experimental programs in this area are highly recommended. There is a need for more fundamental studies on dose-rate effects, temperature dependence, and the nature and importance of alpha-particle effects relative to the recoil nucleus in actinide decay. There are insufficient data presently available to evaluate the potential for damage from ionizing radiation in nuclear waste forms. No additional test methods were recommended for using ion or neutron irradiations to simulate actinide decay or for testing ionization damage in nuclear waste forms. It was recognized that additional test methods may be required and developed as more data become available. An American Society for Testing and Materials (ASTM) Task Group on the Simulation of Radiation Effects in Nuclear Waste Forms (E 10.08.03) was organized to act as a continuing vehicle for discussions and development of procedures, particularly with regard to ion irradiations.

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

1982-01-01T23:59:59.000Z

319

Using Single-Camera 3-D Imaging to Guide Material Handling Robots in a Nuclear Waste Package Closure System  

SciTech Connect

Nuclear reactors for generating energy and conducting research have been in operation for more than 50 years, and spent nuclear fuel and associated high-level waste have accumulated in temporary storage. Preparing this spent fuel and nuclear waste for safe and permanent storage in a geological repository involves developing a robotic packaging system—a system that can accommodate waste packages of various sizes and high levels of nuclear radiation. During repository operation, commercial and government-owned spent nuclear fuel and high-level waste will be loaded into casks and shipped to the repository, where these materials will be transferred from the casks into a waste package, sealed, and placed into an underground facility. The waste packages range from 12 to 20 feet in height and four and a half to seven feet in diameter. Closure operations include sealing the waste package and all its associated functions, such as welding lids onto the container, filling the inner container with an inert gas, performing nondestructive examinations on welds, and conducting stress mitigation. The Idaho National Laboratory is designing and constructing a prototype Waste Package Closure System (WPCS). Control of the automated material handling is an important part of the overall design. Waste package lids, welding equipment, and other tools must be moved in and around the closure cell during the closure process. These objects are typically moved from tool racks to a specific position on the waste package to perform a specific function. Periodically, these objects are moved from a tool rack or the waste package to the adjacent glovebox for repair or maintenance. Locating and attaching to these objects with the remote handling system, a gantry robot, in a loosely fixtured environment is necessary for the operation of the closure cell. Reliably directing the remote handling system to pick and place the closure cell equipment within the cell is the major challenge.

Rodney M. Shurtliff

2005-09-01T23:59:59.000Z

320

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

Science Conference Proceedings (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

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


321

Quantitative assessment of in situ microbial communities affecting nuclear waste disposal  

Science Conference Proceedings (OSTI)

Microbes in the environments surrounding nuclear waste depositories pose several questions regarding the protection of the surrounding communities. microbes can facilitate microbially influenced corrosion (MIC), mobilize and facilitate the transport of nuclides as well as produce gaseous emissions which can compromise containment. We have developed an analysis of the extant microbiota that is independent of quantitative recovery and subsequent growth, based on signature biomarkers analysis (SBA).

White, D.C. [Tennessee Univ., Knoxville, TN (United States)]|[Oak Ridge National Lab., TN (United States)

1996-05-01T23:59:59.000Z

322

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 and characterization of the problems and summary of results  

E-Print Network (OSTI)

071, Power Reactor and Nuclear Fuel Development Corporation,safety assessment of a nuclear fuel waste repository in aPower Reactor and Nuclear Fuel Development Corporation.

2004-01-01T23:59:59.000Z

323

Nuclear heat-load limits for above-grade storage of solid transuranium wastes  

SciTech Connect

Nuclear safety and heat load limits were established for above-grade storage of transuranium (TRU) wastes. Nuclear safety limits were obtained from a study by J.L. Forstner and are summarized. Heat load limits are based on temperature calculations for TRU waste drums stored in concrete containers (hats), and results are summarized. Waste already in storage is within these limits. The limiting factors for individual drum heat load limits were (1) avoidance of temperatures in excess of 190/sup 0/F (decomposition temperature of anion resin) when anion resin is present in a concrete hat, and (2) avoidance of temperatures in excess of 450/sup 0/F (ignition temperature of paper) at any point inside a waste drum. The limiting factor for concrete had heat load limits was avoidance of temperatures in excess of 265/sup 0/F (melt point of high density polyethylene) at the drum liners. A temperature profile for drums and hats filled to recommended limits is shown. Equations and assumptions used were conservative.

Clontz, B.G.

1978-06-01T23:59:59.000Z

324

Geochemical modeling of the nuclear-waste repository system. A status report  

Science Conference Proceedings (OSTI)

The primary objective of the geochemical modeling task is to develop an understanding of the waste-repository geochemical system and provide a valuable tool for estimating future states of that system. There currently exists a variety of computer codes which can be used in geochemical modeling studies. Some available codes contain the framework for simulating a natural chemical system and estimating, within limits, the response of that system to environmental changes. By data-base enhancement and code development, this modeling technique can be even more usefully applied to a nuclear-waste repository. In particular, thermodynamic data on elements not presently in the data base but identified as being of particular hazard in the waste-repository system, need to be incorporated into the code to estimate the near-field as well as the far-field reactions during a hypothetical breach. A reaction-path-simulation code, which estimates the products of specific rock/water reactions, has been tested using basalt and ground water. Results show that the mass-transfer capabilities of the code will be useful in chemical-evolution studies and scenario analyses. The purpose of this report is to explain the status of geochemical modeling as it currently applies to the chemical system of a hypothetical nuclear-waste repository in basalt and to present the plan proposed for further developmet and application.

Deutsch, W.J.

1980-12-01T23:59:59.000Z

325

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

SciTech Connect

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

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

1992-07-01T23:59:59.000Z

326

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

SciTech Connect

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

327

Addendum to the Calcined Waste Storage at the Idaho Nuclear Technology Center  

Science Conference Proceedings (OSTI)

This report is an addendum to the report Calcined Waste Storage at the Idaho Nuclear Technology and Engineering Center, INEEL/EXT-98-00455 Rev. 1, June 2003. The original report provided a summary description of the Calcined Solids Storage Facilities (CSSFs). It also contained dozens of pages of detailed data tables documenting the volume and composition (chemical content and radionuclide activity) of the calcine stored in the CSSFs and the liquid waste from which the calcine was derived. This addendum report compiles the calcine composition data from the original report. It presents the compiled data in a graphical format with units (weight percent, curies per cubic meter, and nanocuries per gram) that are commonly used in regulatory and waste acceptance criteria documents. The compiled data are easier to use and understand when comparing the composition of the calcine with potential regulatory or waste acceptance criteria. This addendum report also provides detailed explanations for the large variability in the calcine composition among the CSSFs. The calcine composition varies as a result of reprocessing different types of fuel that had different cladding materials. Different chemicals were used to dissolve the various types of fuel, extract the uranium, and calcine the resulting waste. This resulted in calcine with variable compositions. This addendum report also identifies a few trace chemicals and radionuclides for which the accuracy of the amounts estimated to be in the calcine could be improved by making adjustments to the assumptions and methods used in making the estimates.

M. D. Staiger; Michael Swenson; T. R. Thomas

2004-05-01T23:59:59.000Z

328

DEVELOPMENT OF CERAMIC WASTE FORMS FOR AN ADVANCED NUCLEAR FUEL CYCLE  

SciTech Connect

A series of ceramic waste forms were developed and characterized for the immobilization of a Cesium/Lanthanide (CS/LN) waste stream anticipated to result from nuclear fuel reprocessing. Simple raw materials, including Al{sub 2}O{sub 3} and TiO{sub 2} were combined with simulated waste components to produce multiphase ceramics containing hollandite-type phases, perovskites (particularly BaTiO{sub 3}), pyrochlores and other minor metal titanate phases. Three fabrication methodologies were used, including melting and crystallizing, pressing and sintering, and Spark Plasma Sintering (SPS), with the intent of studying phase evolution under various sintering conditions. X-Ray Diffraction (XRD) and Scanning Electron Microscopy coupled with Energy Dispersive Spectroscopy (SEM/EDS) results showed that the partitioning of the waste elements in the sintered materials was very similar, despite varying stoichiometry of the phases formed. Identification of excess Al{sub 2}O{sub 3} via XRD and SEM/EDS in the first series of compositions led to a Phase II study, with significantly reduced Al{sub 2}O{sub 3} concentrations and increased waste loadings. The Phase II compositions generally contained a reduced amount of unreacted Al{sub 2}O{sub 3} as identified by XRD. Chemical composition measurements showed no significant issues with meeting the target compositions. However, volatilization of Cs and Mo was identified, particularly during melting, since sintering of the pressed pellets and SPS were performed at lower temperatures. Partitioning of some of the waste components was difficult to determine via XRD. SEM/EDS mapping showed that those elements, which were generally present in small concentrations, were well distributed throughout the waste forms.

Marra, J.; Billings, A.; Brinkman, K.; Fox, K.

2010-11-30T23:59:59.000Z

329

Electrical and electronic subsystems of a nuclear waste tank annulus inspection system  

SciTech Connect

The nuclear waste tank annulus inspection system is designed specifically for use at the Nuclear Regulatory Commission's Nuclear Fuel Services Facility at West Valley, New York. This system sends a television and photographic camera into the space between the walls of a double-shell nuclear waste tank to obtain images of the inner and outer walls at precisely known locations. The system is capable of inspecting a wall section 14 ft wide by 27 ft high. Due to the high temperature and radiation of the annulus environment, the operating life for the inspection device is uncertain, but is expected to be at least 100 h, with 1000 R/h at 82/sup 0/C. The film camera is shielded with 1/2 in. of lead to minimize radiation fogging of the film during a 25-min picture taking excursion. The operation of the inspection system is semiautomated with remote manual prepositioning of the camera, followed by a computer controlled wall scan. This apparatus is currently set up to take an array of contiguous pictures, but is adaptable to other modes of operation.

Evenson, R.J.

1981-06-01T23:59:59.000Z

330

MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY  

SciTech Connect

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 it’s 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

331

Report of the committee to review the use of J-13 well water in Nevada Nuclear Waste Storage Investigations  

SciTech Connect

The Waste Management Project Office of the Department of Energy conducted a special audit of the activities of the Nevada Nuclear Waste Storage Investigation Project at Livermore. It was noted that there never has been a comprehensive, well-documented examination of the basis for the use of J-13 water in the nuclear waste storage investigations. In each of the sections of This Report, an issue relating to the use of J-13 water has been addressed. 58 refs., 19 figs., 8 tabs.

Harrar, J.E.; Carley, J.F.; Isherwood, W.F.; Raber, E.

1990-01-01T23:59:59.000Z

332

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

E-Print Network (OSTI)

Mountain in Nevada -- adjacent to the nuclear weapons test site -- was subsequently chosen by CongressA 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

Deutch, John

333

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

SciTech Connect

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

334

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

SciTech Connect

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

335

Understanding Cement Waste Forms  

Science Conference Proceedings (OSTI)

Oct 29, 2009 ... Ongoing nuclear operations, decontamination and decommissioning, salt waste disposal, and closure of liquid waste tanks result in ...

336

Nuclear-waste isolation in the unsaturated zone of arid regions  

Science Conference Proceedings (OSTI)

The vadose zone in arid regions is considered as a possible environment for geologic isolation of nuclear waste. There are several topographic and lithologic combinations in the vadose zone of arid regions that may lend themselves to waste isolation considerations. In some cases, topographic highs such as mesas and interbasin ranges - comprised of several rock types, may contain essentially dry or partially saturated conditions favorable for isolation. The adjacent basins, especially in the far western and southwestern US, may have no surface or subsurface hydrologic connections with systems ultimately leading to the ocean. Some rock types may have the favorable characteristics of very low permeability and contain appropriate minerals for the strong chemical retardation of radionuclides. Environments exhibiting these hydrologic and geochemical attributes are the areas underlain by tuffaceous rocks, relatively common in the Basin and Range geomorphic province. Adjacent valley areas, where tuffaceous debris makes up a significant component of valley fill alluvium, may also contain thick zones of unsaturated material, and as such also lend themselves to strong consideration as respository environments. This paper summarizes the aspects of nuclear waste isolation in unsaturated regimes in alluvial-filled valleys and tuffaceous rocks of the Basin and Range province.

Wollenberg, H.A.; Wang, J.S.Y.; Korbin, G.

1982-05-01T23:59:59.000Z

337

Economic impacts of the total nuclear waste management program envisioned for the United States  

SciTech Connect

This paper presents information on the costs of nuclear waste management and on the impacts of those costs on the price of power and on the capital and labor markets. It is assumed that the LWR would be the sole commercial reactor used through the year 2000. Two fuel cycle options are considered: the throwaway mode (spent fuel is waste), and the full recycle for comparison. Total costs are calculated for all facilities needed to store, package, and reposit all the spent fuel through the lifetime of 380 GW capacity installed by 2000 and operating for 30 y. The economic impact is: the price of power produced by the reactors would be increased by 1.4%; the capital for nuclear plants would apply to waste management; the average annual labor effort needed over the next 50 to 75 years is 3000 to 5000 man years; and the unit cost of spent fuel disposal is $129/kg ($119/kg for full recycle). 7 tables. (DLC)

Busch, L.; Zielen, A.J.; Parry, S.J.S.

1978-01-01T23:59:59.000Z

338

Impact of nuclear fuel cycle centers on shipping special nuclear materials and wastes  

SciTech Connect

The impact of integrated nuclear fuel cycle facilities on the transportation sector appears from this admittedly rather narrow study to be of only marginal significance. However, there are other factors which must be taken into account such as nuclear safeguards, economics, and radiological, ecological, institutional, and sociological impacts. Unless more clear-cut advantages can be shown by on-going studies for some of these other considerations, the regimentation and control of industry that would result from the imposition of the integrated fuel cycle facility concept probably could not be justified. (auth)

Blomeke, J.O.

1975-01-01T23:59:59.000Z

339

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

Science Conference Proceedings (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

340

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

SciTech Connect

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

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


341

Integrated Waste Treatment Facility Fact Sheet | Department of...  

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

Waste Management Tank Waste and Waste Processing Integrated Waste Treatment Facility Fact Sheet Integrated Waste Treatment Facility Fact Sheet Waste Management Nuclear...

342

NDAA Section 3116 Waste Determinations with Related Disposal...  

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

NDAA Section 3116 Waste Determinations with Related Disposal Performance Assessments Waste Management Nuclear Materials & Waste Tank Waste and Waste Processing Waste...

343

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

SciTech Connect

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

344

International Cooperation on Safety of Nuclear Plants - Nuclear...  

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

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

345

Current R&D Activities in Nuclear Criticality Safety - Nuclear...  

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

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

346

NUCLEAR DATA AND MEASUREMENTS REPORTS 161-180 - Nuclear Data...  

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

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

347

Analysis Tools for Nuclear Criticality Safety - Nuclear Engineering...  

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

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

348

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

SciTech Connect

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

349

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  

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

350

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

E-Print Network (OSTI)

Framework   for   Nuclear   Fuel   Cycle   Concepts,”  Of   Used   Nuclear   Fuel”,   Nuclear  Engineering  and  Radiotoxicity  of  Spent  Nuclear   Fuel,”   Integrated  

Djokic, Denia

2013-01-01T23:59:59.000Z

351

Global Nuclear Energy Partnership Fact Sheet - Minimize Nuclear...  

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

Global Nuclear Energy Partnership Fact Sheet - Minimize Nuclear Waste Global Nuclear Energy Partnership Fact Sheet - Minimize Nuclear Waste GNEP will increase the efficiency in the...

352

Organization - Nuclear Engineering Division (Argonne)  

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

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

353

Achievements: Nuclear Engineering Division (Argonne)  

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

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

354

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

Science Conference Proceedings (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

355

Nevada Nuclear Waste Storage Investigations: exploratory shaft. Phase I. Conceptual design report  

SciTech Connect

It is proposed that an Exploratory Shaft (ES) be constructed in Yucca Mountain on or near the southwest portion of the Nevada Test Site (NTS) as part of the Nevada Nuclear Waste Storage Investigations. This document describes a conceptual design for an ES and a cost estimate based on a set of construction assumptions. Included in this document are appendixes consisting of supporting studies done at NTS by Fenix and Scisson, Inc. and Holmes and Narver, Inc. These appendixes constitute a history of the development of the design and are included as part of the record.

Nelson, D.C.; Merson, T.J.; McGuire, P.L.; Sibbitt, W.L.

1982-06-01T23:59:59.000Z

356

COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLY WASTE VOLUME AND WEIGHT ESTIMATION FOR 63,000 MTU  

Science Conference Proceedings (OSTI)

The purpose of this calculation is to create a high-level estimation of the weights and volume of the commercial spent nuclear fuel (SNF) assemblies, at the time of repository receipt, that will comprise 63,000 metric tons of uranium (MTU) waste. The results of this calculation are for informational purposes only and are not intended to be used as input to design documents. This calculation was prepared in accordance with procedure AP-3.12Q REV 00 ICN 0, Calculations.

B.A. Colton

1999-07-22T23:59:59.000Z

357

Nuclear Power and the Environment  

Reports and Publications (EIA)

This Nuclear Issue Paper discusses Nuclear Plant Wastes, Interactions of Fossil Fuel and Nuclear Power Waste Decisions, and the Environmental Position of Nuclear Power.

2013-05-30T23:59:59.000Z

358

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

SciTech Connect

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

359

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

SciTech Connect

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

360

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

E-Print Network (OSTI)

water  reactor   mixed  low-­?level  waste   viii   MOX  includes   mixed   low-­?level   waste   (MLLW)   and  These  include  mixed  low-­?level  wastes  (MLLW),  for  

Djokic, Denia

2013-01-01T23:59:59.000Z

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


361

Assessment of Nuclear Safety Culture at the Idaho Cleanup Project Sodium Bearing Waste Treatment Project, November 2012  

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

Idaho Cleanup Project Idaho Cleanup Project Sodium Bearing Waste Treatment Project May 2011 November 2012 Office of Safety and Emergency Management Evaluations Office of Enforcement and Oversight Office of Health, Safety and Security U.S. Department of Energy Independent Oversight Assessment of Nuclear Safety Culture at the Idaho Cleanup Project Sodium Bearing Waste Treatment Project Table of Contents 1.0 Introduction........................................................................................................................................... 1 2.0 Scope and Methodology ....................................................................................................................... 2

362

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

SciTech Connect

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

363

Double liquid membrane system for the removal of actinides and lanthanides from acidic nuclear wastes  

SciTech Connect

Supported liquid membranes (SLM), consisting of an organic solution of n-octyl-(phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) and tributyl-phosphate (TBP) in decalin are able to perform selective separation and concentration of actinide and lanthanide ions from aqueous nitrate feed solutions and synthetic nuclear wastes. In the membrane process a possible strip solution is a mixture of formic acid and hydroxylammonium formate (HAF). The effectiveness of this strip solution is reduced and eventually nullified by the simultaneous transfer through the SLM of nitric acid which accumulates in the strip solution. A possible way to overcome this drawback is to make use of a second SLM consisting of a primary amine which is able to extract only HNO/sub 3/ from the strip solution. In this work the results obtained by experimentally studying the membrane system: synthetic nuclear waste/CMPO-TBP membrane/HCOOH-HAF strip solution/primary amine membrane/NaOH solution, are reported. They show that the use of a second liquid membrane is effective in controlling the HNO/sub 3/ concentration in the strip solution, thus allowing the actinide and lanthanide ions removal from the feed solution to proceed to completion. 15 refs., 10 figs., 1 tab.

Chiarizia, R.; Danesi, P.R.

1985-01-01T23:59:59.000Z

364

Blending mining and nuclear industries at the Waste Isolation Pilot Plant  

SciTech Connect

At the Waste Isolation Pilot Plant (WIPP) traditional procedures for underground mining activities have been significantly altered in order to assure underground safety and project adherence to numerous regulatory requirements. Innovative techniques have been developed for WIPP underground procedures, mining equipment, and operating environments. The mining emphasis at WIPP is upon the quality of the excavation, not (as in conventional mines) on the production of ore. The WIPP is a United States Department of Energy (DOE) project that is located 30 miles southeast of Carlsbad, New Mexico, where the nation's first underground engineered nuclear repository is being constructed. The WIPP site was selected because of its location amidst a 607 meter thick salt bed, which provides a remarkably stable rock formation for the permanent storage of nuclear waste. The underground facility is located 655 meters below the earth's surface, in the Salado formation, which comprises two-hundred million year old halites with minor amounts of clay and anhydrites. When completed, the WIPP underground facility will consist of two components: approximately 81 square kilometers of experimental areas, and approximately 405 square kilometers of repository. 3 figs.

Walls, J.R.

1990-01-01T23:59:59.000Z

365

Polyphase ceramic for consolidating nuclear waste compositions with high Zr-Cd-Na content  

Science Conference Proceedings (OSTI)

The development of dense polyphase tailored ceramic forms for the immobilization of high-level nuclear wastes has been extended to an Idaho Chemical Processing Plant Fluorinel composition. The ceramic was designed to maximize waste loading and subsequent waste volume reduction without sacrificing chemical durability in aqueous environments. The ceramic, fabricated by hot isostatic pressing, consists of four main crystalline phases, calcium fluoride, zirconia, an apatite-structured solid-solution phase, and sphene. The form also contains a designed borosilicate glass phase, a Ni-Cd alloy, and a minor amount of crystalline zircon. The crystalline apatite solid-solution phase is the major host for incorporating the actinide simulants U, Ce, and Y, while the glass phase contains Cs and Sr. The calcium fluoride and sphene phases provide microstructural isolation of the radionuclide-containing phases. Since the glass and crystalline components of the ceramic are not phase compatible at all temperatures, the exact phase content is determined by the tailoring additives, consolidation temperature, and oxidation state control during processing.

Harker, A.B.; Flintoff, J.F. (Rockwell International Corp., Thousand Oaks, CA (USA). Science Center)

1990-07-01T23:59:59.000Z

366

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

Science Conference Proceedings (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

367

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

Science Conference Proceedings (OSTI)

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

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

1983-11-01T23:59:59.000Z

368

Recycle of Zirconium from Used Nuclear Fuel Cladding: A Major Element of Waste Reduction  

SciTech Connect

Feasibility tests were initiated to determine if the zirconium in commercial used nuclear fuel (UNF) cladding can be recovered in sufficient purity to permit re-use, and if the recovery process can be operated economically. Initial tests are being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Early results indicate that quantitative recovery can be accomplished and product contamination with alloy constituents can be controlled sufficiently to meet purification requirements. Future tests with actual radioactive UNF cladding are planned. The objective of current research is to determine the feasibility of recovery and recycle of zirconium from used fuel cladding wastes. Zircaloy cladding, which contains 98+% of hafnium-free zirconium, is the second largest mass, on average {approx}25 wt %, of the components in used U.S. light-water-reactor fuel assemblies. Therefore, recovery and recycle of the zirconium would enable a large reduction in geologic waste disposal for advanced fuel cycles. Current practice is to compact or grout the cladding waste and store it for subsequent disposal in a geologic repository. This paper describes results of initial tests being performed with unirradiated, non-radioactive samples of various types of Zircaloy materials that are used in UNF cladding to develop the recovery process and determine the degree of purification that can be obtained. Future tests with actual radioactive UNF cladding are planned.

Collins, Emory D [ORNL; DelCul, Guillermo D [ORNL; Terekhov, Dmitri [ORNL; Emmanuel, N. V. [Chemical Vapor Metal Refining, Inc.

2011-01-01T23:59:59.000Z

369

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

SciTech Connect

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

Arthur, E.D.

1993-10-01T23:59:59.000Z

370

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

DOE Green Energy (OSTI)

Corrosion of steel canisters, stored in a repository for spent fuel and high-level nuclear wastes, leads to the generation and accumulation of hydrogen gas in the backfilled emplacement tunnels, which may significantly affect long-term repository safety. Previous studies used H{sub 2} generation rates based on the volume of the waste or canister material and the stoichiometry of the corrosion reaction. However, iron corrosion and H{sub 2} generation rates vary with time, depending on factors such as amount of iron, water availability, water contact area, and aqueous and solid chemistry. To account for these factors and feedback mechanisms, we developed a chemistry model related to iron corrosion, coupled with two-phase (liquid and gas) flow phenomena that are driven by gas-pressure buildup associated with H{sub 2} generation and water consumption. Results indicate that by dynamically calculating H{sub 2} generation rates based on a simple model of corrosion chemistry, and by coupling this corrosion reaction with two-phase flow processes, the degree and extent of gas pressure buildup could be much smaller compared to a model that neglects the coupling between flow and reactive transport mechanisms. By considering the feedback of corrosion chemistry, the gas pressure increases initially at the canister, but later decreases and eventually returns to a stabilized pressure that is slightly higher than the background pressure. The current study focuses on corrosion under anaerobic conditions for which the coupled hydrogeochemical model was used to examine the role of selected physical parameters on the H{sub 2} gas generation and corresponding pressure buildup in a nuclear waste repository. The developed model can be applied to evaluate the effect of water and mineral chemistry of the buffer and host rock on the corrosion reaction for future site-specific studies.

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

2008-10-15T23:59:59.000Z

371

Predicted temperature/time histories resulting from the burial of nuclear waste canisters in bedded salt  

SciTech Connect

This report provides computed thermal mappings for bedded salt surrounding canisters containing nuclear waste. This information can be used to study the possible migration of fluids within bedded salt under the influence of thermal gradients created by the heat-generating nuclear waste. The results presented were obtained from CINDA thermal models. Three different drift/canister configurations were modeled. The thermal conductivity of the salt was assumed to be temperature dependent while both the density and specific heat were assumed to be constant. Thermal power densities of 30, 75, and 150 kW/acre were examined with canister powers of 0.581 kW (51.6 canisters/acre), 3.5 kW (21.4 canisters/acre), and 3.5 kW (42.9 canisters/acre) at emplacement, respectively. These three cases resulted in maximum salt temperatures of 55/sup 0/C, 117/sup 0/C, and 176/sup 0/C, respectively; and maximum thermal gradients of -15/sup 0/C/m, -63/sup 0/C/m, and -101/sup 0/C/m, respectively. Computer-generated plots of temperature versus distance in horizontal planes at the top, midpoint, and bottom of the canister were made for several times after emplacement. Logarithmic or linear equations (whichever provided the better fit) were used to describe these curves. Derivatives of temperature with respect to distance were then taken and results of the form x(dT/dx) and dT/dx for the logarithmic and linear equations, respectively, were plotted against time. For the two cases where the waste thermal outputs decayed exponentially, it was found that x(dT/dx) and dT/dx were linear functions of time over a large period of years.

George, O.L. Jr.

1980-07-01T23:59:59.000Z

372

Radioactive waste shipments to Hanford retrievable storage from Westinghouse Advanced Reactors and Nuclear Fuels Divisions, Cheswick, Pennsylvania  

Science Conference Proceedings (OSTI)

During the next two decades the transuranic (TRU) waste now stored in the burial trenches and storage facilities at the Hanford Sits in southeastern Washington State is to be retrieved, processed at the Waste Receiving and Processing Facility, and shipped to the Waste Isolation Pilot Plant (WIPP), near Carlsbad, New Mexico for final disposal. Approximately 5.7 percent of the TRU waste to be retrieved for shipment to WIPP was generated by the decontamination and decommissioning (D&D) of the Westinghouse Advanced Reactors Division (WARD) and the Westinghouse Nuclear Fuels Division (WNFD) in Cheswick, Pennsylvania and shipped to the Hanford Sits for storage. This report characterizes these radioactive solid wastes using process knowledge, existing records, and oral history interviews.

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

1994-04-01T23:59:59.000Z

373

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

Science Conference Proceedings (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

374

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

SciTech Connect

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

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

1992-05-01T23:59:59.000Z

375

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

E-Print Network (OSTI)

DOE  FCRD   Office  of  Nuclear  Energy,  Department  of  for   Sustainable   Nuclear   Energy   (PATEROS),   Annex  Training   on   Nuclear   Energy   /   Management  of  

Djokic, Denia

2013-01-01T23:59:59.000Z

376

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

E-Print Network (OSTI)

a   Geologic   Repository”,   Nuclear  Technology,   154,  in  decommissioned  U.S.  nuclear   facilities,  German  Framework   for   Nuclear   Fuel   Cycle   Concepts,”  

Djokic, Denia

2013-01-01T23:59:59.000Z

377

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

Science Conference Proceedings (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

378

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

SciTech Connect

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

Rechard, R.P. [ed.

1993-12-01T23:59:59.000Z

379

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

SciTech Connect

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

380

A compilation of reports of the Advisory Committee on Nuclear Waste, July 1993--June 1995. Volume 5  

Science Conference Proceedings (OSTI)

This compilation contains 13 reports issued by the Advisory Committee on Nuclear Waste (ACNW) during the sixth and seventh years of its operation. The reports, submitted to the chairman and commissioners of US NRC, are the recommendations and comments of ACNW during the period July 1, 1993--June 30, 1995.

NONE

1995-08-01T23:59:59.000Z

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


381

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

SciTech Connect

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

382

Accelerator technology for the Los Alamos ATW (accelerator transmutation of nuclear waste) system  

SciTech Connect

The Los Alamos concept for accelerator transmutation of nuclear waste (ATW) employs a high-power proton linear accelerator to generate intense fluxes of thermal neutrons (>10{sup 16} n/cm{sup 2}-s) through spallation on a lead-bismuth target. The nominal beam energy for an ATW accelerator is 1.6 GeV, with average current requirements ranging from 250 mA to 30 mA, depending on application specifics. A recent study of accelerator production of tritium (APT) led to the development of a detailed point design for a 1.6 GeV, 250 mA cw proton linac. The accelerator design was reviewed by the Energy Research Advisory Board (ERAB) and found to be technically sound. The Panel concluded that linac of this power level could now be implemented within the existing technology base, given an adequate component development program and an integrated engineering demonstration of the front end.

Lawrence, G.P.

1991-01-01T23:59:59.000Z

383

Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 4  

Science Conference Proceedings (OSTI)

OAK-B135 Monitoring the Durability Performance of Concrete in Nuclear Waste Containment. Technical Progress Report No. 4. The analysis of the effect of cracks on the acceleration of the calcium leaching process of cement-based materials has been pursued. During the last period (Technical Progress Report No 3), we have introduced a modeling accounting for the high diffusivity of fractures in comparison with the weak solid material diffusivity. It has been shown through dimensional and asymptotic analysis that small fractures do not significantly accelerate the material aging process. This important result for the overall structural aging kinetics of containment structure has been developed in a paper submitted to the international journal ''Transport in Porous Media''.

Ulm, Franz-Josef

2000-06-30T23:59:59.000Z

384

Seismic design and analysis considerations for high level nuclear waste repositories  

SciTech Connect

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

Hossain, Q.A.

1993-09-30T23:59:59.000Z

385

Results from Nevada Nuclear Waste Storage Investigations (NNWSI) Series 3 spent fuel dissolution tests  

SciTech Connect

The dissolution and radionuclide release behavior of spent fuel in groundwater is being studied by the Yucca Mountain Project (YMP), formerly the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. Specimens prepared from pressurized water reactor fuel rod segments were tested in sealed stainless steel vessels in Nevada Test Site J-13 well water at 85{degree}C and 25{degree}C. The test matrix included three specimens of bare-fuel particles plus cladding hulls, two fuel rod segments with artificially defected cladding and water-tight end fittings, and an undefected fuel rod section with watertight end fittings. Periodic solution samples were taken during test cycles with the sample volumes replenished with fresh J-13 water. Test cycles were periodically terminated and the specimens restarted in fresh J-13 water. The specimens were run for three cycles for a total test duration of 15 months. 22 refs., 32 figs., 26 tabs.

Wilson, C.N.

1990-06-01T23:59:59.000Z

386

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

SciTech Connect

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

387

Characteristics of colloids generated during the corrosion of nuclear waste glasses in groundwater  

SciTech Connect

Aqueous colloidal suspensions were generated by reacting nuclear waste glasses with groundwater at 90{degrees}C at different ratios of the glass surface area to solution volume (S/V). The colloids have been characterized in terms of size, charge, identity, and stability with respect to salt concentration, pH, and time, by examination using dynamic light scattering, electrophoretic mobility, and transmission electron microscopy. The colloids are predominately produced by precipitation from solution, possibly with contribution from reacted layers that have spallated from the glass. These colloids are silicon-rich minerals. The colloidal suspensions agglomerate when the salinity of the solutions increase. The following implications for modeling the colloidal transport of contaminants have been derived from this study: (1) The sources of the colloids are not only solubility-limited real colloids and the pseudo colloids formed by adsorption of radionuclides onto a groundwater colloid, but also from the spalled surface layers of reacted waste glasses. (2) In a repository, the local environment is likely to be glass-reaction dominated and the salt concentration is likely to be high, leading to rapid colloid agglomeration and settling; thus, colloid transport may be insignificant. (3) If large volumes of groundwater contact the glass reaction site, the precipitated colloids may become resuspended, and colloid transport may become important. (4) Under most conditions, the colloids are negatively charged and will deposit readily on positively charged surfaces. Negatively charged surfaces will, in general, facilitate colloid stability and transport.

Feng, X.; Buck, E.C.; Mertz, C.; Bates, J.K.; Cunnane, J.C.; Chaiko, D.

1993-10-01T23:59:59.000Z

388

Modeling Hydrogen-Induced Cracking of Titanium Alloys in Nuclear Waste Repository Environments  

DOE Green Energy (OSTI)

This paper reviews the current understanding of hydrogen-induced cracking (HIC) of Ti Grade 7 and other relevant titanium alloys within the context of the current waste package design for the repository environmental conditions anticipated within the Yucca Mountain repository. The review concentrates on corrosion processes possible in the aqueous environments expected within this site. A brief background discussion of the relevant properties of titanium alloys, the hydrogen absorption process, and the properties of passive film on titanium alloys is presented as the basis for the subsequent discussion of model developments. The key corrosion processes that could occur are addressed individually. Subsequently, the expected corrosion performance of these alloys under the specific environmental conditions anticipated at Yucca Mountain is considered. It can be concluded that, based on the conservative modeling approaches adopted, hydrogen-induced cracking of titanium alloys will not occur under nuclear waste repository conditions since there will not be sufficient hydrogen in the alloy after 10,000 years of emplacement.

F. Hua; K. Mon; P. Pasupathi; G. Gordon

2004-09-08T23:59:59.000Z

389

Method for the recovery of actinide elements from nuclear reactor waste  

DOE Green Energy (OSTI)

A process for partitioning and recovering actinide values from acidic waste solutions resulting from reprocessing of irradiated nuclear fuels by adding hydroxylammonium nitrate and hydrazine to the waste solution to adjust the valence of the neptunium and plutonium values in the solution to the +4 oxidation state, thus forming a feed solution and contacting the feed solution with an extractant of dihexoxyethyl phosphoric acid in an organic diluent whereby the actinide values, most of the rare earth values and some fission product values are taken up by the extractant. Separation is achieved by contacting the loaded extractant with two aqueous strip solutions, a nitric acid solution to selectively strip the americium, curium and rare earth values and an oxalate solution of tetramethylammonium hydrogen oxalate and oxalic acid or trimethylammonium hydrogen oxalate to selectively strip the neptunium, plutonium and fission product values. Uranium values remain in the extractant and may be recovered with a phosphoric acid strip. The neptunium and plutonium values are recovered from the oxalate by adding sufficient nitric acid to destroy the complexing ability of the oxalate, forming a second feed, and contacting the second feed with a second extractant of tricaprylmethylammonium nitrate in an inert diluent whereby the neptunium and plutonium values are selectively extracted. The values are recovered from the extractant with formic acid.

Horwitz, E. Philip (Elmhurst, IL); Delphin, Walter H. (Woodridge, IL); Mason, George W. (Clarendon Hills, IL)

1979-01-01T23:59:59.000Z

390

Geotechnical support and topical studies for nuclear waste geologic repositories: Annual report, fiscal year 1987  

Science Conference Proceedings (OSTI)

This multidisciplinary project was initiated in fiscal year 1986. It comprises 11 reports in two major interrelated tasks: The technical assistance part of the project includes reviewing the progress of the major projects in the DOE Office of Civilian Radioactive waste Management (OCRWM) Program and advising the Engineering and Geotechnology Division on significant technical issues facing each project; analyzing geotechnical data, reports, tests, surveys and plans for the different projects; reviewing and commenting on major technical reports and other program documents such as Site Characterization Plans (SCP) and Study Plans; and providing scientific and technical input at technical meetings. The topical studies activity comprises studies on scientific and technical ions and issues of significance to in-situ testing, test analysis methods, and site characterization of nuclear waste geologic repositories. The subjects of study were selected based on discussions with DOE staff. One minor topic is a preliminary consideration and planning exercise for postclosure monitoring studies. The major task, with subtasks involving various geoscience disciplines, is a study of the mechanical, hydraulic, geophysical and geochemical properties of fractures in geologic rock masses.

Not Available

1988-01-01T23:59:59.000Z

391

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

Science Conference Proceedings (OSTI)

The safe disposal of radioactive waste requires that the waste be isolated from the environment until radioactive decay has reduced its toxicity to innocuous levels for plants, animals, and humans. All of the countries currently studying the options for disposing of high-level nuclear waste (HLW) have selected deep geologic formations to be the primary barrier for accomplishing this isolation. In U.S.A., the Nuclear Waste Policy Act of 1982 (as amended in 1987) designated Yucca Mountain in Nevada as the potential site to be characterized for high-level nuclear waste (HLW) disposal. Long-term containment of waste and subsequent slow release of radionuclides into the geosphere will rely on a system of natural and engineered barriers including a robust waste containment design. The waste package design consists of a highly corrosion resistant Ni-based Alloy 22 cylindrical barrier surrounding a Type 316 stainless steel inner structural vessel. The waste package is covered by a mailbox-shaped drip shield composed primarily of Ti Grade 7 with Ti Grade 24 structural support members. The U.S. Yucca Mountain Project has been studying and modeling the degradation issues of the relevant materials for some 20 years. This paper reviews the state-of-the-art understanding of the degradation processes based on the past 20 years studies on Yucca Mountain Project (YMP) materials degradation issues with focus on interaction between the in-drift environmental conditions and long-term materials degradation of waste packages and drip shields within the repository system during the 10,000 years regulatory period. This paper provides an overview of the current understanding of the likely degradation behavior of the waste package and drip shield in the repository after the permanent closure of the facility. The degradation scenario discussed in this paper include aging and phase instability, dry oxidation, general and localized corrosion, stress corrosion cracking and hydrogen induced cracking of Alloy 22 and titanium alloys. The effects of microbial activity and radiation on degradation of Alloy 22 and titanium alloys are also discussed. Further, for titanium alloys, the effects of fluorides, bromides, calcium ions, and galvanic coupling to less noble metals are further considered. It is concluded that, as far as materials degradation is concerned, the materials and design adopted in the U.S. Yucca Mountain Project will provide sufficient safety margins within the 10,000-years regulatory period.

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

2005-09-19T23:59:59.000Z

392

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

SciTech Connect

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

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

1984-04-01T23:59:59.000Z

393

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

SciTech Connect

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

394

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

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

395

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

SciTech Connect

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

396

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

E-Print Network (OSTI)

Konrad  Mine,  TRU   waste  in  WIPP.   Low  Long-­?Term  short  lived  waste   ix   WAC   WIPP   Xe   YMR   yr   Zr  Isolation  Pilot  Plant  (WIPP),  the  former  Yucca  

Djokic, Denia

2013-01-01T23:59:59.000Z

397

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

E-Print Network (OSTI)

remote-­?handled  TRU   waste,  based  on  the  radiation  dose  on  the  surface  of  the  waste  package,  in  order  to  inform   packaging  and  handling  

Djokic, Denia

2013-01-01T23:59:59.000Z

398

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

SciTech Connect

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

399

Major Programs - Nuclear Engineering Division (Argonne)  

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

Assistance Program International Nuclear Safety Materials Disposition Decontamination & Decommissioning Nuclear Criticality Safety Nuclear Data Program Nuclear Waste Form...

400

Executive Bios: Christopher Grandy - Nuclear Engineering Division...  

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

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

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


401

Nuclear Engineering Division of Argonne National Laboratory:...  

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

402

Fuel Cycle Technologies Program - Nuclear Engineering Division...  

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

403

International Safety Projects - Nuclear Engineering Division...  

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

404

The Dawn of the Nuclear Age  

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

405

Facility Safety Assessment - Nuclear Engineering Division (Argonne...  

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

406

Computer Facilities - Nuclear Engineering Division (Argonne)  

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

407

Advanced Computation & Visualization - Nuclear Engineering Division...  

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

408

Steam Generator Tube Integrity Facilities - Nuclear Engineering...  

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

409

Safety - Vulnerability Assessment Team - Nuclear Engineering...  

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

410

Nuclear waste form risk assessment for US defense waste at Savannah River Plant. Annual report fiscal year 1980  

SciTech Connect

Waste form dissolution studies and preliminary performance analyses were carried out to contribute a part of the data needed for the selection of a waste form for the disposal of Savannah River Plant defense waste in a deep geologic repository. The first portion of this work provides descriptions of the chemical interactions between the waste form and the geologic environment. We reviewed critically the dissolution/leaching data for borosilicate glass and SYNROC. Both chemical kinetic and thermodynamic models were developed to describe the dissolution process of these candidate waste forms so as to establish a fundamental basis for interpretation of experimental data and to provide directions for future experiments. The complementary second portion of this work is an assessment of the impacts of alternate waste forms upon the consequences of disposal in various proposed geological media. Employing systems analysis methodology, we began to evaluate the performance of a generic waste form for the case of a high risk scenario for a bedded salt repository. Results of sensitivity analysis, uncertainty analyses, and sensitivity to uncertainty analysis are presented.

Cheung, H.; Jackson, D.D.; Revelli, M.A.

1981-07-01T23:59:59.000Z

411

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

SciTech Connect

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.

2013-01-01T23:59:59.000Z

412

The Secretary  

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

Secretary Secretary of Energy Washington, D.C. 20585 Secretarial Determination of the Adequacy of the Nuclear Waste Fund Fee The Nuclear Waste Policy Act (NWPA) establishes a Nuclear Waste Fund to be used to pay for the disposition of commercial spent nuclear fuel and high-level radioactive waste. Section 302(a)(2) of the NWPA establishes a fee of 1 mill (1/10-cent) per kilowatt-hour of electricity generated and sold that must be paid by nuclear utilities and deposited in the Fund. The NWPA also requires the Secretary to review the adequacy of this fee annually and, upon a determination that either insufficient or excess funds are being collected, to propose an adjustment to the fee to ensure that the full costs of the Federal Government's disposal program will be fully recovered from generators and owners of high-level radioactive waste or

413

Capabilities - Nuclear Engineering Division (Argonne)  

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

Nuclear Waste Form and Repository Performance Modeling Nuclear Systems Technologies Nuclear Criticality Safety Research Reactor Analysis System Process Monitoring,...

414

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

SciTech Connect

This assessment studied the performance of high-level radioactive waste and spent nuclear fuel in a hypothetical repository in unsaturated tuff. The results of this 10-month study are intended to help guide the Office of Environment Management of the US Department of Energy (DOE) on how to prepare its wastes for eventual permanent disposal. The waste forms comprised spent fuel and high-level waste currently stored at the Idaho National Engineering Laboratory (INEL) and the Hanford reservations. About 700 metric tons heavy metal (MTHM) of the waste under study is stored at INEL, including graphite spent nuclear fuel, highly enriched uranium spent fuel, low enriched uranium spent fuel, and calcined high-level waste. About 2100 MTHM of weapons production fuel, currently stored on the Hanford reservation, was also included. The behavior of the waste was analyzed by waste form and also as a group of waste forms in the hypothetical tuff repository. When the waste forms were studied together, the repository was assumed also to contain about 9200 MTHM high-level waste in borosilicate glass from three DOE sites. The addition of the borosilicate glass, which has already been proposed as a final waste form, brought the total to about 12,000 MTHM.

Rechard, R.P. [ed.

1995-03-01T23:59:59.000Z

415

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

SciTech Connect

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

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

1980-08-01T23:59:59.000Z

416

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

SciTech Connect

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

417

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

SciTech Connect

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

418

NUCLEAR DATA AND MEASUREMENTS REPORTS 61-80 - Nuclear Data Program...  

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

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

419

NUCLEAR DATA AND MEASUREMENTS REPORTS 81-100 - Nuclear Data Program...  

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

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

420

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

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

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


421

Financing Strategies For A Nuclear Fuel Cycle Facility  

SciTech Connect

To help meet the nation’s energy needs, recycling of partially used nuclear fuel is required to close the nuclear fuel cycle, but implementing this step will require considerable investment. This report evaluates financing scenarios for integrating recycling facilities into the nuclear fuel cycle. A range of options from fully government owned to fully private owned were evaluated using DPL (Decision Programming Language 6.0), which can systematically optimize outcomes based on user-defined criteria (e.g., lowest lifecycle cost, lowest unit cost). This evaluation concludes that the lowest unit costs and lifetime costs are found for a fully government-owned financing strategy, due to government forgiveness of debt as sunk costs. However, this does not mean that the facilities should necessarily be constructed and operated by the government. The costs for hybrid combinations of public and private (commercial) financed options can compete under some circumstances with the costs of the government option. This analysis shows that commercial operations have potential to be economical, but there is presently no incentive for private industry involvement. The Nuclear Waste Policy Act (NWPA) currently establishes government ownership of partially used commercial nuclear fuel. In addition, the recently announced Global Nuclear Energy Partnership (GNEP) suggests fuels from several countries will be recycled in the United States as part of an international governmental agreement; this also assumes government ownership. Overwhelmingly, uncertainty in annual facility capacity led to the greatest variations in unit costs necessary for recovery of operating and capital expenditures; the ability to determine annual capacity will be a driving factor in setting unit costs. For private ventures, the costs of capital, especially equity interest rates, dominate the balance sheet; and the annual operating costs, forgiveness of debt, and overnight costs dominate the costs computed for the government case. The uncertainty in operations, leading to lower than optimal processing rates (or annual plant throughput), is the most detrimental issue to achieving low unit costs. Conversely, lowering debt interest rates and the required return on investments can reduce costs for private industry.

David Shropshire; Sharon Chandler

2006-07-01T23:59:59.000Z

422

Report to the American Physical Society by the study group on nuclear fuel cycles and waste management  

SciTech Connect

Utilization of nuclear fuels and management of nuclear wastes have become major topics of public discussion. Under the auspices of the American Physical Society this study was undertaken as an independent evaluation of technical issues in the use of fissionable materials in nuclear fuel cycles, together with their principal economic, environmental, health and safety implications. Reprocessing and recycling in light water reactors were examined, along with technical measures proposed as possible safeguards; advanced reactor fuel cycles were also studied for their resource and safeguards implications. Much of the work of the group centered on the principal alternatives for disposal of radioactive wastes and control of effluents. The group examined the research and development programs sponsored by government agencies along with associated relationships among agencies and between government and private industry. Available information was also considered on nuclear fuel resources, and on important economic and environmental aspects of the various fuel cycles in order to strive for a balanced comparative study. The report presents many conclusions on various aspects of the nuclear fuel cycles and also provides recommendations concerning present utilization and future improvement of fuel cycle technology.

APS Study Group Participants; Hebel, L.C. Chairman; Christensen, E.L.; Donath, F.A.; Falconer, W.E.; Lidofsky, L.J.; Moniz, E.J.; Moss, T.H.; Pigford, R.L.; Pigford, T.H.; Rochlin, G.I.; Silsbee, R.H.; Wrenn, M.E.

1978-01-01T23:59:59.000Z

423

Public acceptability of the use of gamma rays from spent nuclear fuel as a hazardous waste treatment process  

Science Conference Proceedings (OSTI)

Three methods were used to estimate public reaction to the use of gamma irradiation of hazardous wastes as a hazardous waste treatment process. The gamma source of interest is spent nuclear fuel. The first method is Benefit-Risk Decision Making, where the benefits of the proposed technology are compared to its risks. The second analysis compares the proposed technology to the other, currently used nuclear technologies and estimates public reaction based on that comparison. The third analysis is called Analysis of Public Consent, and is based on the professional methods of the Institute for Participatory Management and Planning. The conclusion of all three methods is that the proposed technology should not result in negative public reaction sufficient to prevent implementation.

Mincher, B.J.; Wells, R.P.; Reilly, H.J.

1992-01-01T23:59:59.000Z

424

Geology and hydrogeology of the proposed nuclear waste repository at Yucca Mountain, Nevada and the surrounding area  

SciTech Connect

In late 1987 Congress issued an amendment to the Nuclear Waste Policy Act of 1982 which directed the characterization of Yucca Mountain, Nevada as the only remaining potential site for the Nation`s first underground high-level radioactive waste repository. The evaluation of a potential underground repository is guided and regulated by policy established by the Department of Energy (DOE), Nuclear Regulatory Commission (NRC), Environmental Protection Agency (EPA), Department of Transportation (DOT), and the US Congress. The Yucca Mountain Project is the responsibility of the DOE. The purpose of this field trip is to introduce the present state of geologic and hydrologic knowledge concerning this site. This report describes the field trip. 108 refs., 6 figs., 1 tab.

Mattson, S.R.; Broxton, D.E.; Crowe, B.M.; Buono, A.; Orkild, P.P.

1989-07-01T23:59:59.000Z

425

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

SciTech Connect

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

426

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

Science Conference Proceedings (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

427

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

Science Conference Proceedings (OSTI)

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

2001-12-13T23:59:59.000Z

428

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

Science Conference Proceedings (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

429

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

Science Conference Proceedings (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

430

Site characterization plan: Yucca Mountain site, Nevada research and development area, Nevada: Consultation draft, Nuclear Waste Policy Act: Volume 6  

Science Conference Proceedings (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 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 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

431

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

Science Conference Proceedings (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

432

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

SciTech Connect

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

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

1986-03-01T23:59:59.000Z

433

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

SciTech Connect

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

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

1980-05-01T23:59:59.000Z

434

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

SciTech Connect

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

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

1980-01-01T23:59:59.000Z

435

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

DOE Green Energy (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

436

ON THE ANODIC POLARIZATION BEHAVIOR OF CARBON STEEL IN HANFORD NUCLEAR WASTES  

Science Conference Proceedings (OSTI)

The effect of the important chemical constituents in the Hanford nuclear waste simulant on the anodic behavior of carbon steel was studied. Specifically, the effect of pH, nitrite concentration, nitrite/nitrate concentration ratios, total organic carbon and the chloride concentration on the open circuit potential, pitting potential and repassivation potential was evaluated. It was found that pH adjusting, although capable of returning the tank chemistry back to specification, did not significantly reduce the corrosivity of the stimulant compared to the present condition. Nitrite was found to be a potent inhibitor for carbon steel. A critical concentration of approximately 1.2M appeared to be beneficial to increase the difference of repassivation potential and open circuit potential considerably and thus prevent pitting corrosion from occurring. No further benefit was gained when increasing nitrite concentration to a higher level. The organic compounds were found to be weak inhibitors in the absence of nitrite and the change of chloride from 0.05M to 0.2M did not alter the anodic behavior dramatically.

BOOMER, K.D.

2007-01-31T23:59:59.000Z

437

Ceramic Coatings for Corrosion Resistant Nuclear Waste Container Evaluated in Simulated Ground Water at 90?C  

SciTech Connect

Ceramic materials have been considered as corrosion resistant coatings for nuclear waste containers. Their suitability can be derived from the fully oxidized state for selected metal oxides. Several types of ceramic coatings applied to plain carbon steel substrates by thermal spray techniques have been exposed to 90 C simulated ground water for nearly 6 years. In some cases no apparent macroscopic damage such as coating spallation was observed in coatings. Thermal spray processes examined in this work included plasma spray, High Velocity Oxy Fuel (HVOF), and Detonation Gun. Some thermal spray coatings have demonstrated superior corrosion protection for the plain carbon steel substrate. In particular the HVOF and Detonation Gun thermal spray processes produced coatings with low connected porosity, which limited the growth rate of corrosion products. It was also demonstrated that these coatings resisted spallation of the coating even when an intentional flaw (which allowed for corrosion of the carbon steel substrate underneath the ceramic coating) was placed in the coating. A model for prediction of the corrosion protection provided by ceramic coatings is presented. The model includes the effect of the morphology and amount of the porosity within the thermal spray coating and provides a prediction of the exposure time needed to produce a crack in the ceramic coating.

Haslam, J J; Farmer, J C

2004-03-31T23:59:59.000Z

438

Test Operation of Oxygen-Enriched Incinerator for Wastes From Nuclear Fuel Fabrication Facility  

SciTech Connect

The oxygen-enriched combustion concept, which can minimize off-gas production, has been applied to the incineration of combustible uranium-containing wastes from a nuclear fuel fabrication facility. A simulation for oxygen combustion shows the off-gas production can be reduced by a factor of 6.7 theoretically, compared with conventional air combustion. The laboratory-scale oxygen enriched incineration (OEI) process with a thermal capacity of 350 MJ/h is composed of an oxygen feeding and control system, a combustion chamber, a quencher, a ceramic filter, an induced draft fan, a condenser, a stack, an off-gas recycle path, and a measurement and control system. Test burning with cleaning paper and office paper in this OEI process shows that the thermal capacity is about 320 MJ/h, 90 % of design value and the off-gas reduces by a factor of 3.5, compared with air combustion. The CO concentration for oxygen combustion is lower than that of air combustion, while the O2 concentration in off-gas is kept above 25 vol % for a simple incineration process without any grate. The NOx concentration in an off-gas stream does not reduce significantly due to air incoming by leakage, and the volume and weight reduction factors are not changed significantly, which suggests a need for an improvement in sealing.

Kim, J.-G.; Yang, H.cC.; Park, G.-I.; Kim, I.-T.; Kim, J.-K.

2002-02-26T23:59:59.000Z

439

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

SciTech Connect

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.

1983-05-01T23:59:59.000Z

440

The TRansUranium EXtraction (TRUEX) process: A vital tool for disposal of US defense nuclear waste  

SciTech Connect

The TRUEX (TRansUranium EXtraction) process is a generic actinide extraction/recovery process for the removal of all actinides from acidic nitrate and chloride nuclear waste solutions. Because of its high efficiency and flexibility and its compatibility with existing process facilities, TRUEX has now become a vital tool for the disposal of certain US defense nuclear waste. The development of TRUEX is closely coupled to the development of bifunctional extractants belonging to the carbamoylphosphoryl class and CMPO in particular. A brief review of the development of CMPO and its relationship to other bifunctional and monofunctional extractants is presented. The effect of TBP on CMPO, the selectivity of CMPO for actinides extracted from acidic nitrate media, the influence of diluents on CM