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1

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix D, Part B: Naval spent nuclear fuel management  

SciTech Connect (OSTI)

This volume contains the following attachments: transportation of Naval spent nuclear fuel; description of Naval spent nuclear receipt and handling at the Expended Core Facility at the Idaho National Engineering Laboratory; comparison of storage in new water pools versus dry container storage; description of storage of Naval spent nuclear fuel at servicing locations; description of receipt, handling, and examination of Naval spent nuclear fuel at alternate DOE facilities; analysis of normal operations and accident conditions; and comparison of the Naval spent nuclear fuel storage environmental assessment and this environmental impact statement.

Not Available

1994-06-01T23:59:59.000Z

2

EIS-0251: Department of the Navy Final Environmental Impact Statement for a Container System for the Management of Naval Spent Nuclear Fuel (November 1996)  

Broader source: Energy.gov [DOE]

This Final Environmental Impact Statementaddresses six general alternative systems for the loading, storage, transport, and possible disposal of naval spent nuclear fuel following examination.

3

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs, Draft Environmental Impact Statement. Volume 1, Appendix D: Part A, Naval Spent Nuclear Fuel Management  

SciTech Connect (OSTI)

Volume 1 to the Department of Energy`s Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Management Programs Environmental Impact Statement evaluates a range of alternatives for managing naval spent nuclear fuel expected to be removed from US Navy nuclear-powered vessels and prototype reactors through the year 2035. The Environmental Impact Statement (EIS) considers a range of alternatives for examining and storing naval spent nuclear fuel, including alternatives that terminate examination and involve storage close to the refueling or defueling site. The EIS covers the potential environmental impacts of each alternative, as well as cost impacts and impacts to the Naval Nuclear Propulsion Program mission. This Appendix covers aspects of the alternatives that involve managing naval spent nuclear fuel at four naval shipyards and the Naval Nuclear Propulsion Program Kesselring Site in West Milton, New York. This Appendix also covers the impacts of alternatives that involve examining naval spent nuclear fuel at the Expended Core Facility in Idaho and the potential impacts of constructing and operating an inspection facility at any of the Department of Energy (DOE) facilities considered in the EIS. This Appendix also considers the impacts of the alternative involving limited spent nuclear fuel examinations at Puget Sound Naval Shipyard. This Appendix does not address the impacts associated with storing naval spent nuclear fuel after it has been inspected and transferred to DOE facilities. These impacts are addressed in separate appendices for each DOE site.

Not Available

1994-06-01T23:59:59.000Z

4

What is spent nuclear fuel?  

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

What is Spent Nuclear Fuel? Spent nuclear fuel (SNF) is irradiated fuel or targets containing uranium, plutonium, or thorium that is permanently withdrawn from a nuclear reactor or...

5

Nuclear Spent Fuel Program Drivers  

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

was created to plan and coordinate the management of Department of Energy-owned spent nuclear fuel. It was established as a result of a 1992 decision to stop spent nuclear fuel...

6

National Spent Nuclear Fuel Program  

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

need to safely and efficiently manage all DOE-owned spent nuclear fuel and high level waste and prepare it for disposal. The National Spent Nuclear Fuel Program is addressing...

7

Spent Nuclear Fuel Fact Sheets  

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

management needs. By coordinating common needs for research, technology development, and testing programs, the National Spent Nuclear Fuel Program is achieving cost efficiencies...

8

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Volume 1, Appendix B: Idaho National Engineering Laboratory Spent Nuclear Fuel Management Program  

SciTech Connect (OSTI)

The US Department of Energy (DOE) has prepared this report to assist its management in making two decisions. The first decision, which is programmatic, is to determine the management program for DOE spent nuclear fuel. The second decision is on the future direction of environmental restoration, waste management, and spent nuclear fuel management activities at the Idaho National Engineering Laboratory. Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent nuclear fuel management on the quality of the human and natural environment for planning years 1995 through 2035. DOE has derived the information and analysis results in Volume 1 from several site-specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes environmental impacts for various environmental restoration, waste management, and spent nuclear fuel management alternatives for planning years 1995 through 2005. This Appendix B to Volume 1 considers the impacts on the INEL environment of the implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the management of DOE spent nuclear fuel including naval-type fuel.

Not Available

1994-06-01T23:59:59.000Z

9

Spent nuclear fuel sampling strategy  

SciTech Connect (OSTI)

This report proposes a strategy for sampling the spent nuclear fuel (SNF) stored in the 105-K Basins (105-K East and 105-K West). This strategy will support decisions concerning the path forward SNF disposition efforts in the following areas: (1) SNF isolation activities such as repackaging/overpacking to a newly constructed staging facility; (2) conditioning processes for fuel stabilization; and (3) interim storage options. This strategy was developed without following the Data Quality Objective (DQO) methodology. It is, however, intended to augment the SNF project DQOS. The SNF sampling is derived by evaluating the current storage condition of the SNF and the factors that effected SNF corrosion/degradation.

Bergmann, D.W.

1995-02-08T23:59:59.000Z

10

Spent nuclear fuel reprocessing modeling  

SciTech Connect (OSTI)

The long-term wide development of nuclear power requires new approaches towards the realization of nuclear fuel cycle, namely, closed nuclear fuel cycle (CNFC) with respect to fission materials. Plant nuclear fuel cycle (PNFC), which is in fact the reprocessing of spent nuclear fuel unloaded from the reactor and the production of new nuclear fuel (NF) at the same place together with reactor plant, can be one variant of CNFC. Developing and projecting of PNFC is a complicated high-technology innovative process that requires modern information support. One of the components of this information support is developed by the authors. This component is the programme conducting calculations for various variants of process flow sheets for reprocessing SNF and production of NF. Central in this programme is the blocks library, where the blocks contain mathematical description of separate processes and operations. The calculating programme itself has such a structure that one can configure the complex of blocks and correlations between blocks, appropriate for any given flow sheet. For the ready sequence of operations balance calculations are made of all flows, i.e. expenses, element and substance makeup, heat emission and radiation rate are determined. The programme is open and the block library can be updated. This means that more complicated and detailed models of technological processes will be added to the library basing on the results of testing processes using real equipment, in test operating mode. The development of the model for the realization of technical-economic analysis of various variants of technologic PNFC schemes and the organization of 'operator's advisor' is expected. (authors)

Tretyakova, S.; Shmidt, O.; Podymova, T.; Shadrin, A.; Tkachenko, V. [Bochvar Institute, 5 Rogova str., Moscow 123098 (Russian Federation); Makeyeva, I.; Tkachenko, V.; Verbitskaya, O.; Schultz, O.; Peshkichev, I. [Russian Federal Nuclear Center - VNIITF E.I. Zababakhin, p.o.box 245, Snezhinsk, 456770 (Russian Federation)

2013-07-01T23:59:59.000Z

11

Spent Nuclear Fuel (SNF) Project Execution Plan  

SciTech Connect (OSTI)

The Spent Nuclear Fuel (SNF) Project supports the Hanford Site Mission to cleanup the Site by providing safe, economic, environmentally sound management of Site spent nuclear fuel in a manner that reduces hazards by staging it to interim onsite storage and deactivates the 100 K Area facilities.

LEROY, P.G.

2000-11-03T23:59:59.000Z

12

Importance of Nuclear DataImportance of Nuclear Data to the Naval Nuclear Propulsion Program  

E-Print Network [OSTI]

Importance of Nuclear DataImportance of Nuclear Data to the Naval Nuclear Propulsion Program Don Cores · Project Prometheus · Some Very Recent Criticality Analyses #12;Use of Early RPI Measurements · Criticality Analyses of Under-moderated Systemsy y y · Most Reactive Condition ­ Highest Water Density

Danon, Yaron

13

Spent Nuclear Fuel Alternative Technology Decision Analysis  

SciTech Connect (OSTI)

The Westinghouse Savannah River Company (WSRC) made a FY98 commitment to the Department of Energy (DOE) to recommend a technology for the disposal of aluminum-based spent nuclear fuel (SNF) at the Savannah River Site (SRS). The two technologies being considered, direct co-disposal and melt and dilute, had been previously selected from a group of eleven potential SNF management technologies by the Research Reactor Spent Nuclear Fuel Task Team chartered by the DOE''s Office of Spent Fuel Management. To meet this commitment, WSRC organized the SNF Alternative Technology Program to further develop the direct co-disposal and melt and dilute technologies and ultimately provide a WSRC recommendation to DOE on a preferred SNF alternative management technology.

Shedrow, C.B.

1999-11-29T23:59:59.000Z

14

Spent Nuclear Fuel (SNF) Project Product Specification  

SciTech Connect (OSTI)

This document establishes the limits and controls for the significant parameters that could potentially affect the safety and/or quality of the Spent Nuclear Fuel (SNF) packaged for processing, transport, and storage. The product specifications in this document cover the SNF packaged in Multi-Canister Overpacks to be transported throughout the SNF Project.

PAJUNEN, A.L.

2000-01-20T23:59:59.000Z

15

Surrogate Spent Nuclear Fuel Vibration Integrity Investigation  

SciTech Connect (OSTI)

Transportation packages for spent nuclear fuel (SNF) must meet safety requirements under normal and accident conditions as specified by federal regulations. During transportation, SNF experiences unique conditions and challenges to cladding integrity due to the vibrational and impact loading encountered during road or rail shipment. ORNL has been developing testing capabilities that can be used to improve our understanding of the impacts of vibration loading on SNF integrity, especially for high burn-up SNF in normal transportation operation conditions. This information can be used to meet nuclear industry and U.S. Nuclear Regulatory Commission needs in the area of safety of SNF storage and transportation operations.

Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Bevard, Bruce Balkcom [ORNL; Howard, Rob L [ORNL

2014-01-01T23:59:59.000Z

16

Spent nuclear fuel project - criteria document spent nuclear fuel final safety analysis report  

SciTech Connect (OSTI)

The criteria document provides the criteria and planning guidance for developing the Spent Nuclear Fuel (SNF) Final Safety Analysis Report (FSAR). This FSAR will support the US Department of Energy, Richland Operations Office decision to authorize the procurement, installation, installation acceptance testing, startup, and operation of the SNF Project facilities (K Basins, Cold Vacuum Drying Facility, and Canister Storage Building).

MORGAN, R.G.

1999-02-23T23:59:59.000Z

17

Transportation capabilities study of DOE-owned spent nuclear fuel  

SciTech Connect (OSTI)

This study evaluates current capabilities for transporting spent nuclear fuel owned by the US Department of Energy. Currently licensed irradiated fuel shipping packages that have the potential for shipping the spent nuclear fuel are identified and then matched against the various spent nuclear fuel types. Also included are the results of a limited investigation into other certified packages and new packages currently under development. This study is intended to support top-level planning for the disposition of the Department of Energy`s spent nuclear fuel inventory.

Clark, G.L.; Johnson, R.A.; Smith, R.W. [Packaging Technology, Inc., Tacoma, WA (United States); Abbott, D.G.; Tyacke, M.J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)

1994-10-01T23:59:59.000Z

18

Spent Nuclear Fuel project integrated safety management plan  

SciTech Connect (OSTI)

This document is being revised in its entirety and the document title is being revised to ``Spent Nuclear Fuel Project Integrated Safety Management Plan.

Daschke, K.D.

1996-09-17T23:59:59.000Z

19

Pyroprocess for processing spent nuclear fuel  

DOE Patents [OSTI]

This is a pyroprocess for processing spent nuclear fuel. The spent nuclear fuel is chopped into pieces and placed in a basket which is lowered in to a liquid salt solution. The salt is rich in ZrF.sub.4 and containing alkali or alkaline earth fluorides, and in particular, the salt chosen was LiF-50 mol % ZrF.sub.4 with a eutectic melting point of 500.degree. C. Prior to lowering the basket, the salt is heated to a temperature of between 550.degree. C. and 700.degree. C. in order to obtain a molten solution. After dissolution the oxides of U, Th, rare earth and other like oxides, the salt bath solution is subject to hydro-fluorination to remove the oxygen and then to a fluorination step to remove U as gaseous UF.sub.6. In addition, after dissolution, the basket contains PuO.sub.2 and undissolved parts of the fuel rods, and the basket and its contents are processed to remove the Pu.

Miller, William E. (Naperville, IL); Tomczuk, Zygmunt (Lockport, IL)

2002-01-01T23:59:59.000Z

20

Naval Nuclear Propulsion Plants | National Nuclear Security Administration  

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

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

Note: This page contains sample records for the topic "naval spent nuclear" 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

Naval Nuclear Propulsion Plants | National Nuclear Security Administration  

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

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

22

Low Energy Nuclear Reaction Research at the Naval Research Laboratory D.A. Kidwell1  

E-Print Network [OSTI]

Low Energy Nuclear Reaction Research at the Naval Research Laboratory D.A. Kidwell1 , D have explored the field of Low Energy Nuclear Reactions (LENR) for about eight years focusing energy (as heat) is produced ­ neither nuclear products nor transmutations have been firmly established

Noble, James S.

23

Risk and Responsibility Sharing in Nuclear Spent Fuel Management  

E-Print Network [OSTI]

With the Nuclear Waste Policy Act of 1982, the responsibility of American utilities in the long-term management of spent nuclear fuel was limited to the payment of a fee. This narrow involvement did not result in faster ...

De Roo, Guillaume

24

Characterization plan for Hanford spent nuclear fuel  

SciTech Connect (OSTI)

Reprocessing of spent nuclear fuel (SNF) at the Hanford Site Plutonium-Uranium Extraction Plant (PUREX) was terminated in 1972. Since that time a significant quantity of N Reactor and Single-Pass Reactor SNF has been stored in the 100 Area K-East (KE) and K-West (KW) reactor basins. Approximately 80% of all US Department of Energy (DOE)-owned SNF resides at Hanford, the largest portion of which is in the water-filled KE and KW reactor basins. The basins were not designed for long-term storage of the SNF and it has become a priority to move the SNF to a more suitable location. As part of the project plan, SNF inventories will be chemically and physically characterized to provide information that will be used to resolve safety and technical issues for development of an environmentally benign and efficient extended interim storage and final disposition strategy for this defense production-reactor SNF.

Abrefah, J.; Thornton, T.A.; Thomas, L.E.; Berting, F.M.; Marschman, S.C.

1994-12-01T23:59:59.000Z

25

The Impacts of Dry-Storage Canister Designs on Spent Nuclear...  

Office of Environmental Management (EM)

Canister Designs on Spent Nuclear Fuel Handling, Storage, Transportation, and Disposal in the U.S. The Impacts of Dry-Storage Canister Designs on Spent Nuclear Fuel...

26

Calculation of the Naval Long and Short Waste Package Three-Dimensional Thermal Interface Temperatures  

SciTech Connect (OSTI)

The purpose of this calculation is to evaluate the thermal performance of the Naval Long and Naval Short spent nuclear fuel (SNF) waste packages (WP) in the repository emplacement drift. The scope of this calculation is limited to the determination of the temperature profiles upon the surfaces of the Naval Long and Short SNF waste package for up to 10,000 years of emplacement. The temperatures on the top of the outside surface of the naval canister are the thermal interfaces for the Naval Nuclear Propulsion Program (NNPP). The results of this calculation are intended to support Licensing Application design activities.

H. Marr

2006-10-25T23:59:59.000Z

27

Inventory of LWR spent nuclear fuel in the 324 Building  

SciTech Connect (OSTI)

This document contains the results of calculations to estimate the decay heat, neutron source term, photon source term, and radioactive inventory of light-water-reactor spent nuclear fuel in the 324 Building at Pacific Northwest National Laboratory.

Jenquin, U.P.

1996-09-01T23:59:59.000Z

28

Spent nuclear fuel discharges from U.S. reactors 1994  

SciTech Connect (OSTI)

Spent Nuclear Fuel Discharges from US Reactors 1994 provides current statistical data on fuel assemblies irradiated at commercial nuclear reactors operating in the US. This year`s report provides data on the current inventories and storage capacities at these reactors. Detailed statistics on the data are presented in four chapters that highlight 1994 spent fuel discharges, storage capacities and inventories, canister and nonfuel component data, and assembly characteristics. Five appendices, a glossary, and bibliography are also included. 10 figs., 34 tabs.

NONE

1996-02-01T23:59:59.000Z

29

What to Expect When Readying to Move Spent Nuclear Fuel from...  

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

What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power Plants What to Expect When Readying to Move Spent Nuclear Fuel from Commercial Nuclear Power...

30

Spent Nuclear Fuel (SNF) Project Product Specification  

SciTech Connect (OSTI)

The process for removal of Spent Nuclear Fuel (SNF) from the K Basins has been divided into major sub-systems. The Fuel Retrieval System (FRS) removes fuel from the existing storage canisters, cleans it, and places it into baskets. The multi-canister overpack (MCO) loading system places the baskets into an MCO that has been pre-loaded in a cask. The cask, containing a loaded MCO, is then transferred to the Cold Vacuum Drying (CVD) Facility. After drying at the CVD Facility, the cask, and MCO, are transferred to the Canister Storage Building (CSB), where the MCO is removed from the cask, staged, inspected, sealed (by welding), and stored until a suitable permanent disposal option is implemented. The purpose of this document is to specify the process related characteristics of an MCO at the interface between major process systems. The characteristics are derived from the primary technical documents that form the basis for safety analysis and design calculations. This document translates the calculation assumptions into implementation requirements and describes the method of verifying that the requirement is achieved. These requirements are used to define validation test requirements and describe requirements that influence multiple sub-project safety analysis reports. This product specification establishes limits and controls for each significant process parameter at interfaces between major sub-systems that potentially affect the overall safety and/or quality of the SNF packaged for processing, transport, and interim dry storage. The product specifications in this document cover the SNF packaged in MCOs to be transported throughout the SNF Project. The description of the product specifications are organized in the document as follows: Section 2.0--Summary listing of product specifications at each major sub-system interface. Section 3.0--Summary description providing guidance as to how specifications are complied with by equipment design or processing within a major sub-system. Section 4.0--Specific technical basis description for each product specification. The scope of this product specification does not include data collection requirements to support accountability or environmental compliance activities.

PAJUNEN, A.L.

2000-12-07T23:59:59.000Z

31

Spent nuclear fuel discharges from US reactors 1993  

SciTech Connect (OSTI)

The Energy Information Administration (EIA) of the U.S. Department of Energy (DOE) administers the Nuclear Fuel Data Survey, Form RW-859. This form is used to collect data on fuel assemblies irradiated at commercial nuclear reactors operating in the United States, and the current inventories and storage capacities of those reactors. These data are important to the design and operation of the equipment and facilities that DOE will use for the future acceptance, transportation, and disposal of spent fuels. The data collected and presented identifies trends in burnup, enrichment, and spent nuclear fuel discharged form commercial light-water reactor as of December 31, 1993. The document covers not only spent nuclear fuel discharges; but also site capacities and inventories; canisters and nonfuel components; and assembly type characteristics.

Not Available

1995-02-01T23:59:59.000Z

32

DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE  

SciTech Connect (OSTI)

The purpose of this analysis is to demonstrate the design of the naval spent nuclear fuel (SNF) waste package (WP) using the Waste Package Department's (WPD) design methodologies and processes described in the ''Waste Package Design Methodology Report'' (CRWMS M&O [Civilian Radioactive Waste Management System Management and Operating Contractor] 2000b). The calculations that support the design of the naval SNF WP will be discussed; however, only a sub-set of such analyses will be presented and shall be limited to those identified in the ''Waste Package Design Sensitivity Report'' (CRWMS M&O 2000c). The objective of this analysis is to describe the naval SNF WP design method and to show that the design of the naval SNF WP complies with the ''Naval Spent Nuclear Fuel Disposal Container System Description Document'' (CRWMS M&O 1999a) and Interface Control Document (ICD) criteria for Site Recommendation. Additional criteria for the design of the naval SNF WP have been outlined in Section 6.2 of the ''Waste Package Design Sensitivity Report'' (CRWMS M&O 2000c). The scope of this analysis is restricted to the design of the naval long WP containing one naval long SNF canister. This WP is representative of the WPs that will contain both naval short SNF and naval long SNF canisters. The following items are included in the scope of this analysis: (1) Providing a general description of the applicable design criteria; (2) Describing the design methodology to be used; (3) Presenting the design of the naval SNF waste package; and (4) Showing compliance with all applicable design criteria. The intended use of this analysis is to support Site Recommendation reports and assist in the development of WPD drawings. Activities described in this analysis were conducted in accordance with the technical product development plan (TPDP) ''Design Analysis for the Naval SNF Waste Package (CRWMS M&O 2000a).

T.L. Mitchell

2000-05-31T23:59:59.000Z

33

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

34

Application of Neutron-Absorbing Structural-Amorphous metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Controls  

E-Print Network [OSTI]

241 Pu, etc. ). To prevent nuclear criticality in spent fuelto enhance criticality safety for spent nuclear fuel inSpent Nuclear Fuel (SNF) Container to Enhance Criticality

2006-01-01T23:59:59.000Z

35

Air Shipment of Spent Nuclear Fuel from Romania to Russia  

SciTech Connect (OSTI)

Romania successfully completed the worlds first air shipment of spent nuclear fuel transported in Type B(U) casks under existing international laws and without shipment license special exceptions when the last Romanian highly enriched uranium (HEU) spent nuclear fuel was transported to the Russian Federation in June 2009. This air shipment required the design, fabrication, and licensing of special 20 foot freight containers and cask tiedown supports to transport the eighteen TUK 19 shipping casks on a Russian commercial cargo aircraft. The new equipment was certified for transport by road, rail, water, and air to provide multi modal transport capabilities for shipping research reactor spent fuel. The equipment design, safety analyses, and fabrication were performed in the Russian Federation and transport licenses were issued by both the Russian and Romanian regulatory authorities. The spent fuel was transported by truck from the VVR S research reactor to the Bucharest airport, flown by commercial cargo aircraft to the airport at Yekaterinburg, Russia, and then transported by truck to the final destination in a secure nuclear facility at Chelyabinsk, Russia. This shipment of 23.7 kg of HEU was coordinated by the Russian Research Reactor Fuel Return Program (RRRFR), as part of the U.S. Department of Energy Global Threat Reduction Initiative (GTRI), in close cooperation with the Rosatom State Atomic Energy Corporation and the International Atomic Energy Agency, and was managed in Romania by the National Commission for Nuclear Activities Control (CNCAN). This paper describes the planning, shipment preparations, equipment design, and license approvals that resulted in the safe and secure air shipment of this spent nuclear fuel.

Igor Bolshinsky; Ken Allen; Lucian Biro; Alexander Buchelnikov

2010-10-01T23:59:59.000Z

36

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network [OSTI]

Security of the National Nuclear Security Administration, USof Energys National Nuclear Security Administration (NNSA)

Quiter, Brian

2012-01-01T23:59:59.000Z

37

Assessment of Nuclear Resonance Fluorescence for Spent Nuclear Fuel Assay  

E-Print Network [OSTI]

of the Institute of Nuclear Material Management, Tucson, AZ,Assay, Institute of Nuclear Materials Management 51st Annual

Quiter, Brian

2012-01-01T23:59:59.000Z

38

Molten tin reprocessing of spent nuclear fuel elements  

DOE Patents [OSTI]

A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support the liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Heckman, Richard A. (Castro Valley, CA)

1983-01-01T23:59:59.000Z

39

Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements  

SciTech Connect (OSTI)

In 1998, a major change in the technical strategy for managing Multi Canister Overpacks (MCO) while stored within the Canister Storage Building (CSB) occurred. The technical strategy is documented in Baseline Change Request (BCR) No. SNF-98-006, Simplified SNF Project Baseline (MCO Sealing) (FDH 1998). This BCR deleted the hot conditioning process initially adopted for the Spent Nuclear Fuel Project (SNF Project) as documented in WHC-SD-SNF-SP-005, Integrated Process Strategy for K Basins Spent Nuclear Fuel (WHC 199.5). In summary, MCOs containing Spent Nuclear Fuel (SNF) from K Basins would be placed in interim storage following processing through the Cold Vacuum Drying (CVD) facility. With this change, the needs for the Hot Conditioning System (HCS) and inerting/pressure retaining capabilities of the CSB storage tubes and the MCO Handling Machine (MHM) were eliminated. Mechanical seals will be used on the MCOs prior to transport to the CSB. Covers will be welded on the MCOs for the final seal at the CSB. Approval of BCR No. SNF-98-006, imposed the need to review and update the CSB functions and requirements baseline documented herein including changing the document title to ''Spent Nuclear Fuel Project Canister Storage Building Functions and Requirements.'' This revision aligns the functions and requirements baseline with the CSB Simplified SNF Project Baseline (MCO Sealing). This document represents the Canister Storage Building (CSB) Subproject technical baseline. It establishes the functions and requirements baseline for the implementation of the CSB Subproject. The document is organized in eight sections. Sections 1.0 Introduction and 2.0 Overview provide brief introductions to the document and the CSB Subproject. Sections 3.0 Functions, 4.0 Requirements, 5.0 Architecture, and 6.0 Interfaces provide the data described by their titles. Section 7.0 Glossary lists the acronyms and defines the terms used in this document. Section 8.0 References lists the references used for this document.

KLEM, M.J.

2000-10-18T23:59:59.000Z

40

Standard guide for drying behavior of spent nuclear fuel  

E-Print Network [OSTI]

1.1 This guide is organized to discuss the three major components of significance in the drying behavior of spent nuclear fuel: evaluating the need for drying, drying spent nuclear fuel, and confirmation of adequate dryness. 1.1.1 The guide addresses drying methods and their limitations in drying spent nuclear fuels that have been in storage at water pools. The guide discusses sources and forms of water that remain in SNF, its container, or both, after the drying process and discusses the importance and potential effects they may have on fuel integrity, and container materials. The effects of residual water are discussed mechanistically as a function of the container thermal and radiological environment to provide guidance on situations that may require extraordinary drying methods, specialized handling, or other treatments. 1.1.2 The basic issue in drying is to determine how dry the SNF must be in order to prevent issues with fuel retrievability, container pressurization, or container corrosion. Adequate d...

American Society for Testing and Materials. Philadelphia

2008-01-01T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" 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

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

Broader source: Energy.gov [DOE]

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

42

An experiment to simulate the heat transfer properties of a dry, horizontal spent nuclear fuel assembly  

E-Print Network [OSTI]

Nuclear power reactors generate highly radioactive spent fuel assemblies. Initially, the spent fuel assemblies are stored for a period of several years in an on-site storage facility to allow the radioactivity levels of ...

Lovett, Phyllis Maria

1991-01-01T23:59:59.000Z

43

Gamma Ray Mirrors for Direct Measurement of Spent Nuclear Fuel  

SciTech Connect (OSTI)

Direct measurement of the amount of Pu and U in spent nuclear fuel represents a challenge for the safeguards community. Ideally, the characteristic gamma-ray emission lines from different isotopes provide an observable suitable for this task. However, these lines are generally lost in the fierce flux of radiation emitted by the fuel. The rates are so high that detector dead times limit measurements to only very small solid angles of the fuel. Only through the use of carefully designed view ports and long dwell times are such measurements possible. Recent advances in multilayer grazing-incidence gamma-ray optics provide one possible means of overcoming this difficulty. With a proper optical and coating design, such optics can serve as a notch filter, passing only narrow regions of the overall spectrum to a fully shielded detector that does not view the spent fuel directly. We report on the design of a mirror system and a number of experimental measurements.

Pivovaroff, Dr. Michael J. [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL); Ziock, Klaus-Peter [ORNL] [ORNL; Harrison, Mark J [ORNL] [ORNL; Soufli, Regina [Lawrence Livermore National Laboratory (LLNL)] [Lawrence Livermore National Laboratory (LLNL)

2014-01-01T23:59:59.000Z

44

Technical strategy for the management of INEEL spent nuclear fuel  

SciTech Connect (OSTI)

This report presents evaluations, findings, and recommendations of the Idaho National Engineering and Environmental Laboratory (INEEL) Spent Nuclear Fuel Task Team. The technical strategy developed by the Task Team includes stabilization, near term storage, packaging, transport, and ultimate disposal. Key issues identified and discussed include waste characterization, criticality, packaging, waste form performance, and special fuels. Current plans focus on onsite needs, and include three central elements: (1) resolution of near-term vulnerabilities, (2) consolidation of storage locations, and (3) achieving dry storage in transportable packages. In addition to the Task Team report, appendices contain information on the INEEL spent fuel inventory; regulatory decisions and agreements; and analyses of criticality, packaging, storage, transportation, and system performance of a geological repository. 16 refs., 6 figs., 4 tabs.

NONE

1997-03-01T23:59:59.000Z

45

Corrosion of Spent Nuclear Fuel: The Long-Term Assessment  

SciTech Connect (OSTI)

Spent nuclear fuel, essentially U{sub 2}, accounts for over 95% of the total radioactivity of all of the radioactive wastes in the United States that require disposal, disposition or remediation. The UO{sub 2} in SNF is not stable under oxiding conditions and may also be altered under reducing conditions. The alteration of SNF results in the formation of new uranium phases that can cause the release or retardation of actinide and fission product radionuclides. Over the long term, and depending on the extent to which the secondary uranium phases incorporate fission products and actinides, these alteration phases become the near-field source term.

Rodney C. Ewing

2004-10-07T23:59:59.000Z

46

Closure mechanism and method for spent nuclear fuel canisters  

DOE Patents [OSTI]

A canister is provided for storing, transporting, and/or disposing of spent nuclear fuel. The canister includes a canister shell, a top shield plug disposed within the canister, and a leak-tight closure arrangement. The closure arrangement includes a shear ring which forms a containment boundary of the canister, and which is welded to the canister shell and top shield plug. An outer seal plate, forming an outer seal, is disposed above the shear ring and is welded to the shield plug and the canister.

Doman, Marvin J. (Monroeville, PA)

2004-11-23T23:59:59.000Z

47

Instrumented, Shielded Test Canister System for Evaluation of Spent Nuclear Fuel in Dry Storage  

SciTech Connect (OSTI)

This document describes the development of an instrumented, shielded test canister system to store and monitor aluminum-based spent nuclear duel under dry storage conditions.

Sindelar, R.L.

1999-10-21T23:59:59.000Z

48

Report to Congress on Plan for Interim Storage of Spent Nuclear...  

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

Report to Congress on Plan for Interim Storage of Spent Nuclear Fuel from Decommissioned Reactors More Documents & Publications Information Request, "THE REPORT TO THE PRESIDENT...

49

Delayed Gamma-Ray Spectroscopy for Spent Nuclear Fuel Assay  

SciTech Connect (OSTI)

High-energy, beta-delayed gamma-ray spectroscopy is investigated as a non-destructive assay technique for the determination of plutonium mass in spent nuclear fuel. This approach exploits the unique isotope-specific signatures contained in the delayed gamma-ray emission spectra detected following active interrogation with an external neutron source. A high fidelity modeling approach is described that couples radiation transport, analytical decay/depletion, and a newly developed gamma-ray emission source reconstruction code. Initially simulated and analyzed was a one-pass delayed gamma-ray assay that focused on the long-lived signatures. Also presented are the results of an independent study that investigated pulsed mode measurements, to capture the more isotope-specific, short-lived signatures. Initial modeling results outlined in this paper suggest that delayed gamma-ray assay of spent nuclear fuel assemblies can be accomplished with a neutron generator of sufficient strength and currently available gamma-ray detectors.

Campbell, Luke W.; Hunt, Alan W.; Ludewigt, Bernhard A.; Mozin, Vladimir V.

2012-04-01T23:59:59.000Z

50

Spent nuclear fuel recycling with plasma reduction and etching  

DOE Patents [OSTI]

A method of extracting uranium from spent nuclear fuel (SNF) particles is disclosed. Spent nuclear fuel (SNF) (containing oxides of uranium, oxides of fission products (FP) and oxides of transuranic (TRU) elements (including plutonium)) are subjected to a hydrogen plasma and a fluorine plasma. The hydrogen plasma reduces the uranium and plutonium oxides from their oxide state. The fluorine plasma etches the SNF metals to form UF6 and PuF4. During subjection of the SNF particles to the fluorine plasma, the temperature is maintained in the range of 1200-2000 deg K to: a) allow any PuF6 (gas) that is formed to decompose back to PuF4 (solid), and b) to maintain stability of the UF6. Uranium (in the form of gaseous UF6) is easily extracted and separated from the plutonium (in the form of solid PuF4). The use of plasmas instead of high temperature reactors or flames mitigates the high temperature corrosive atmosphere and the production of PuF6 (as a final product). Use of plasmas provide faster reaction rates, greater control over the individual electron and ion temperatures, and allow the use of CF4 or NF3 as the fluorine sources instead of F2 or HF.

Kim, Yong Ho

2012-06-05T23:59:59.000Z

51

Safe Advantage on Dry Interim Spent Nuclear Fuel Storage  

SciTech Connect (OSTI)

This paper aims to present the advantages of dry cask storage in comparison with the wet storage (cooling water pools) for SNF. When the nuclear fuel is removed from the core reactor, it is moved to a storage unit and it wait for a final destination. Generally, the spent nuclear fuel (SNF) remains inside water pools within the reactors facility for the radioactive activity decay. After some period of time in pools, SNF can be sent to a definitive deposition in a geological repository and handled as radioactive waste or to reprocessing facilities, or still, wait for a future solution. Meanwhile, SNF remains stored for a period of time in dry or wet facilities, depending on the method adopted by the nuclear power plant or other plans of the country. Interim storage, up to 20 years ago, was exclusively wet and if the nuclear facility had to be decommissioned another storage solution had to be found. At the present time, after a preliminary cooling of the SNF elements inside the water pool, the elements can be stored in dry facilities. This kind of storage does not need complex radiation monitoring and it is safer then wet one. Casks, either concrete or metallic, are safer, especially on occurrence of earthquakes, like that occurred at Kashiwazaki-Kariwa nuclear power plant, in Japan on July 16, 2007. (authors)

Romanato, L.S. [Centro Tecnologico da Marinha em S.Paulo, Brazilian Navy Technological Center, Sao Paulo (Brazil)

2008-07-01T23:59:59.000Z

52

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect (OSTI)

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

A. Alsaed

2005-07-28T23:59:59.000Z

53

Commercial Spent Nuclear Fuel Waste Package Misload Analysis  

SciTech Connect (OSTI)

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

J.K. Knudson

2003-10-02T23:59:59.000Z

54

Comparative analysis of LWR and FBR spent fuels for nuclear forensics evaluation  

SciTech Connect (OSTI)

Some interesting issues are attributed to nuclide compositions of spent fuels from thermal reactors as well as fast reactors such as a potential to reuse as recycled fuel, and a possible capability to be manage as a fuel for destructive devices. In addition, analysis on nuclear forensics which is related to spent fuel compositions becomes one of the interesting topics to evaluate the origin and the composition of spent fuels from the spent fuel foot-prints. Spent fuel compositions of different fuel types give some typical spent fuel foot prints and can be estimated the origin of source of those spent fuel compositions. Some technics or methods have been developing based on some science and technological capability including experimental and modeling or theoretical aspects of analyses. Some foot-print of nuclear forensics will identify the typical information of spent fuel compositions such as enrichment information, burnup or irradiation time, reactor types as well as the cooling time which is related to the age of spent fuels. This paper intends to evaluate the typical spent fuel compositions of light water (LWR) and fast breeder reactors (FBR) from the view point of some foot prints of nuclear forensics. An established depletion code of ORIGEN is adopted to analyze LWR spent fuel (SF) for several burnup constants and decay times. For analyzing some spent fuel compositions of FBR, some coupling codes such as SLAROM code, JOINT and CITATION codes including JFS-3-J-3.2R as nuclear data library have been adopted. Enriched U-235 fuel composition of oxide type is used for fresh fuel of LWR and a mixed oxide fuel (MOX) for FBR fresh fuel. Those MOX fuels of FBR come from the spent fuels of LWR. Some typical spent fuels from both LWR and FBR will be compared to distinguish some typical foot-prints of SF based on nuclear forensic analysis.

Permana, Sidik; Suzuki, Mitsutoshi; Su'ud, Zaki [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 Nuclear Physics and Bio (Indonesia); Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA), 2-4 Shirane, Shirakata, Tokai Mura, Naka-gun, Ibaraki 319-1195 (Japan); Nuclear Physics and Bio Physics Research Group, Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia)

2012-06-06T23:59:59.000Z

55

Natural convection heat transfer within horizontal spent nuclear fuel assemblies  

SciTech Connect (OSTI)

Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array.

Canaan, R.E.

1995-12-01T23:59:59.000Z

56

Savannah River Site, Spent Nuclear Fuel Management, Draft Environmental Impact Statement  

SciTech Connect (OSTI)

The proposed DOE action described in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets assigned to the Savannah River Site (SRS), including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel (20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some programmatic material stored at SRS for repackaging and dry storage pending shipment offsite).

N /A

1998-12-24T23:59:59.000Z

57

Managing Spent Nuclear Fuel at the Idaho National Laboratory  

SciTech Connect (OSTI)

The Idaho National Laboratory (INL) has a large inventory of diverse types of spent nuclear fuel (SNF). This legacy derives from the history of the INL as the National Reactor Testing Station, and from its mission to recover HEU from SNF and to test and examine SNF after irradiation. The INL also has a large diversity of SNF storage facilities, some 50 years old. SNF at INL has many formsfrom intact assemblies down to metallurgical mounts, and some fuel has been wet stored for over 40 years. SNF is stored bare or in metal cans under water, or dry in vaults, caissons or casks. Inspection shows varying corrosion and degradation of the SNF and its storage cans. SNF has been stored in 10 different facilities: 5 pools, one cask storage pad, one vault, two generations of caisson facilities, and one modular Independent Spent Fuel Storage Installation (ISFSI). The pools range in age from 40 years old to the most modern in the US Department of Energy (DOE) complex. The near-term objective is to move SNF from older pools to interim dry storage, allowing shutdown and decommissioning of the older facilities. This move involves drying methods that are dependent on fuel type. The long-term objective is to have INL SNF in safe dry storage and ready to be shipped to the National Repository. The unique features of the INL SNF requires special treatments and packaging to meet the proposed repository acceptance criteria and SNF will be repackaged in standardized canisters for shipment and disposal in the National Repository. Disposal will use the standardized canisters that can be co-disposed with High Level Waste glass logs to limit the total fissile material in a repository waste package. The DOE standardized canister also simplifies the repository handling of the multitude of DOE SNF sizes and shapes.

Thomas Hill; Denzel L. Fillmore

2005-10-01T23:59:59.000Z

58

Direct Investigations of the Immobilization of Radionuclides in the Alteration Products of Spent Nuclear Fuel  

SciTech Connect (OSTI)

Safe disposal of the nation's nuclear waste in a geological repository involves unique scientific and engineering challenges owing to the very long-lived radioactivity of the waste. The repository must retain a variety of radionuclides that have vastly different chemical characters for several thousand years. Most of the radioactivity that will be housed in the proposed repository at Yucca Mountain will be associated with spent nuclear fuel, much of which is derived from commercial reactors. DOE is custodian of approximately 8000 tons of spent nuclear fuel that is also intended for eventual disposal in a geological repository. Unlike the spent fuel from commercial reactors, the DOE fuel is diverse in composition with more than 250 varieties. Safe disposal of spent fuel requires a detailed knowledge of its long-term behavior under repository conditions, as well as the fate of radionuclides released from the spent fuel as waste containers are breached.

Peter C. Burns; Robert J. Finch; David J. Wronkiewicz

2004-12-27T23:59:59.000Z

59

Plutonium Discharge Rates and Spent Nuclear Fuel Inventory Estimates for Nuclear Reactors Worldwide  

SciTech Connect (OSTI)

This report presents a preliminary survey and analysis of the five primary types of commercial nuclear power reactors currently in use around the world. Plutonium mass discharge rates from the reactors spent fuel at reload are estimated based on a simple methodology that is able to use limited reactor burnup and operational characteristics collected from a variety of public domain sources. Selected commercial reactor operating and nuclear core characteristics are also given for each reactor type. In addition to the worldwide commercial reactors survey, a materials test reactor survey was conducted to identify reactors of this type with a significant core power rating. Over 100 material or research reactors with a core power rating >1 MW fall into this category. Fuel characteristics and spent fuel inventories for these material test reactors are also provided herein.

Brian K. Castle; Shauna A. Hoiland; Richard A. Rankin; James W. Sterbentz

2012-09-01T23:59:59.000Z

60

EIS-0306: Treatment and Management of Sodium-Bonded Spent Nuclear Fuel  

Broader source: Energy.gov [DOE]

This EIS evaluates thepotential environmental impacts of the proposed electrometallurgical treatment of DOE-owned sodium bonded spent nuclear fuel in the Fuel Conditioning Facility at Argonne...

Note: This page contains sample records for the topic "naval spent nuclear" 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

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask  

SciTech Connect (OSTI)

The VSC-17 Spent Nuclear Fuel Storage Cask was surveyed for degradation of the concrete shield by radiation measurement, temperature measurement, and ultrasonic testing. No general loss of shielding function was identified.

Koji Shirai

2006-04-01T23:59:59.000Z

62

EA-1117: Management of Spent Nuclear Fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee  

Broader source: Energy.gov [DOE]

This EA evaluates the environmental impacts of the proposal for the management of spent nuclear fuel on the U.S. Department of Energy's Oak Ridge Reservation to implement the preferred alternative...

63

Advanced dry head-end reprocessing of light water reactor spent nuclear fuel  

SciTech Connect (OSTI)

A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Collins, Emory D.; Delcul, Guillermo D.; Hunt, Rodney D.; Johnson, Jared A.; Spencer, Barry B.

2014-06-10T23:59:59.000Z

64

Advanced dry head-end reprocessing of light water reactor spent nuclear fuel  

DOE Patents [OSTI]

A method for reprocessing spent nuclear fuel from a light water reactor includes the step of reacting spent nuclear fuel in a voloxidation vessel with an oxidizing gas having nitrogen dioxide and oxygen for a period sufficient to generate a solid oxidation product of the spent nuclear fuel. The reacting step includes the step of reacting, in a first zone of the voloxidation vessel, spent nuclear fuel with the oxidizing gas at a temperature ranging from 200-450.degree. C. to form an oxidized reaction product, and regenerating nitrogen dioxide, in a second zone of the voloxidation vessel, by reacting oxidizing gas comprising nitrogen monoxide and oxygen at a temperature ranging from 0-80.degree. C. The first zone and the second zone can be separate. A voloxidation system is also disclosed.

Collins, Emory D; Delcul, Guillermo D; Hunt, Rodney D; Johnson, Jared A; Spencer, Barry B

2013-11-05T23:59:59.000Z

65

Development of Gd-Enriched Alloys for Spent Nuclear Fuel Applications--Part 1: Preliminary Characterization  

E-Print Network [OSTI]

of Gd-containing alloys for storage, transport, and disposal of spent nuclear fuel. However, unlike, the basket materials must be corrosion resistant under the projected stor- age conditions. Recent research

DuPont, John N.

66

SOURCE OF BURNUP VALUES FOR COMMERCIAL SPENT NUCLEAR FUEL ASSEMBLIES  

SciTech Connect (OSTI)

Waste packages are loaded with commercial spent nuclear fuel (SNF) that satisfies the minimum burnup requirements of a criticality loading curve. The burnup value assigned by the originating nuclear utility to each SNF assembly (assigned burnup) is used to load waste packages in compliance with a criticality loading curve. The burnup provided by a nuclear utility has uncertainties, so conservative calculation methods are used to characterize those uncertainties for incorporation into the criticality loading curves. Procedural safety controls ensure that the correct assembly is loaded into each waste package to prevent a misload that could create a condition affecting the safety margins. Probabilistic analyses show that procedural safety controls can minimize the chance of a misload but can not completely eliminate the possibility. Physical measurements of burnup with instrumentation in the surface facility are not necessary due to the conservative calculation methods used to produce the criticality loading curves. The reactor records assigned burnup of a commercial SNF assembly contains about two percent uncertainty, which is increased to five-percent to ensure conservatism. This five-percent uncertainty is accommodated by adjusting the criticality loading curve. Also, the record keeping methods of nuclear utilities are not uniform and the level of detail required by the NRC has varied over the last several decades. Thus, some SNF assemblies may have assigned burnups that are averages for a batch of assemblies with similar characteristics. Utilities typically have access to more detailed core-follow records that allow the batch average burnup to be changed to an assembly specific burnup. Alternatively, an additional safety margin is incorporated into the criticality loading curve to accommodate SNF assemblies with batch average burnups or greater uncertainties due to the methodology used by the nuclear utility. The utility records provide the assembly identifier, initial {sup 235}U enrichment, and time of discharge from the reactor as well as the assigned burnup, but the distribution. of burnup axially along the assembly length is not provided. The axial burnup profile is maintained within acceptable bounds by the operating conditions of the nuclear reactor and is calculated during preparations to reload a reactor, but the actual burnup profile is not measured. The axial burnup profile is important to the determination of the reactivity of a waste package, so a conservative evaluation of the calculated axial profiles for a large database of SNF has been performed. The product of the axial profile evaluation is a profile that is conservative. Thus, there is no need for physical measurement of the axial profile. The assembly identifier is legible on each SNF assembly and the utility records provide the associated characteristics of the assembly. The conservative methodologies used to determine the criticality loading curve for a waste package provide sufficient margin so that criticality safety is assured for preclosure operations even in the event of a misload. Consideration of misload effects for postclosure time periods is provided by the criticality Features, Events, and Processes (FEPs) analysis. The conservative approaches used to develop and apply the criticality loading curve are thus sufficiently robust that the utility assigned burnup is an adequate source of burnup values, and additional means of verification of assigned burnup through physical measurements are not needed.

BSC

2004-12-01T23:59:59.000Z

67

MANAGING SPENT NUCLEAR FUEL WASTES AT THE IDAHO NATIONAL LABORATORY  

SciTech Connect (OSTI)

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

Hill, Thomas J

2005-09-01T23:59:59.000Z

68

Deployment evaluation methodology for the electrometallurgical treatment of DOE-EM spent nuclear fuel  

SciTech Connect (OSTI)

Part of the Department of Energy (DOE) spent nuclear fuel (SNF) inventory may require some type of treatment to meet acceptance criteria at various disposition sites. The current focus for much of this spent nuclear fuel is the electrometallurgical treatment process under development at Argonne National Laboratory. Potential flowsheets for this treatment process are presented. Deployment of the process for the treatment of the spent nuclear fuel requires evaluation to determine the spent nuclear fuel program need for treatment and compatibility of the spent nuclear fuel with the process. The evaluation of need includes considerations of cost, technical feasibility, process material disposition, and schedule to treat a proposed fuel. A siting evaluation methodology has been developed to account for these variables. A work breakdown structure is proposed to gather life-cycle cost information to allow evaluation of alternative siting strategies on a similar basis. The evaluation methodology, while created specifically for the electrometallurgical evaluation, has been written such that it could be applied to any potential treatment process that is a disposition option for spent nuclear fuel. Future work to complete the evaluation of the process for electrometallurgical treatment is discussed.

Dahl, C.A.; Adams, J.P.; Ramer, R.J.

1998-07-01T23:59:59.000Z

69

Spent Nuclear Fuel (SNF) Project Design Basis Capacity Study  

SciTech Connect (OSTI)

This study of the design basis capacity of process systems was prepared by Fluor Federal Services for the Spent Nuclear Fuel Project. The evaluation uses a summary level model of major process sub-systems to determine the impact of sub-system interactions on the overall time to complete fuel removal operations. The process system model configuration and time cycle estimates developed in the original version of this report have been updated as operating scenario assumptions evolve. The initial document released in Fiscal Year (FY) 1996 varied the number of parallel systems and transport systems over a wide range, estimating a conservative design basis for completing fuel processing in a two year time period. Configurations modeling planned operations were updated in FY 1998 and FY 1999. The FY 1998 Base Case continued to indicate that fuel removal activities at the basins could be completed in slightly over 2 years. Evaluations completed in FY 1999 were based on schedule modifications that delayed the start of KE Basin fuel removal, with respect to the start of KW Basin fuel removal activities, by 12 months. This delay resulted in extending the time to complete all fuel removal activities by 12 months. However, the results indicated that the number of Cold Vacuum Drying (CVD) stations could be reduced from four to three without impacting the projected time to complete fuel removal activities. This update of the design basis capacity evaluation, performed for FY 2000, evaluates a fuel removal scenario that delays the start of KE Basin activities such that staffing peaks are minimized. The number of CVD stations included in all cases for the FY 2000 evaluation is reduced from three to two, since the scenario schedule results in minimal time periods of simultaneous fuel removal from both basins. The FY 2000 evaluation also considers removal of Shippingport fuel from T Plant storage and transfer to the Canister Storage Building for storage.

CLEVELAND, K.J.

2000-08-17T23:59:59.000Z

70

Application of ALARA principles to shipment of spent nuclear fuel  

SciTech Connect (OSTI)

The public exposure from spent fuel shipment is very low. In view of this low exposure and the perfect safety record for spent fuel shipment, existing systems can be considered satisfactory. On the other hand, occupational exposure reduction merits consideration and technology improvement to decrease dose should concentrate on this exposure. Practices that affect the age of spent fuel in shipment and the number of times the fuel must be shipped prior to disposal have the largest impact. A policy to encourage a 5-year spent fuel cooling period prior to shipment coupled with appropriate cask redesign to accommodate larger loads would be consistent with ALARA and economic principles. And finally, bypassing high population density areas will not in general reduce shipment dose.

Greenborg, J.; Brackenbush, L.W.; Murphy, D.W. Burnett, R.A.; Lewis, J.R.

1980-05-01T23:59:59.000Z

71

Preparation for the Recovery of Spent Nuclear Fuel (SNF) at Andreeva Bay, North West Russia - 13309  

SciTech Connect (OSTI)

Andreeva Bay is located near Murmansk in the Russian Federation close to the Norwegian border. The ex-naval site was used to de-fuel nuclear-powered submarines and icebreakers during the Cold War. Approximately 22,000 fuel assemblies remain in three Dry Storage Units (DSUs) which means that Andreeva Bay has one of the largest stockpiles of highly enriched spent nuclear fuel (SNF) in the world. The high contamination and deteriorating condition of the SNF canisters has made improvements to the management of the SNF a high priority for the international community for safety, security and environmental reasons. International Donors have, since 2002, provided support to projects at Andreeva concerned with improving the management of the SNF. This long-term programme of work has been coordinated between the International Donors and responsible bodies within the Russian Federation. Options for the safe and secure management of SNF at Andreeva Bay were considered in 2004 and developed by a number of Russian Institutes with international participation. This consisted of site investigations, surveys and studies to understand the technical challenges. A principal agreement was reached that the SNF would be removed from the site altogether and transported to Russia's reprocessing facility at Mayak in the Urals. The analytical studies provided the information necessary to develop the construction plan for the site. Following design and regulatory processes, stakeholders endorsed the technical solution in April 2007. This detailed the processes, facilities and equipment required to safely remove the SNF and identified other site services and support facilities required on the site. Implementation of this strategy is now well underway with the facilities in various states of construction. Physical works have been performed to address the most urgent tasks including weather protection over one of the DSUs, installation of shielding over the cells, provision of radiation protection infrastructure and general preparation of the site for construction of the facilities for the removal of the SNF. This paper describes the development and implementation of the strategy and work to improve the safe and secure management of SNF, preparing it for retrieval and removal from Andreeva Bay. (authors)

Field, D.; McAtamney, N. [Nuvia Limited (United Kingdom)] [Nuvia Limited (United Kingdom)

2013-07-01T23:59:59.000Z

72

Uncanistered Spent Nuclear fuel Disposal Container System Description Document  

SciTech Connect (OSTI)

The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in the emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Multiple boiling water reactor (BWR) and pressurized water reactor (PWR) disposal container designs are needed to accommodate the expected range of spent fuel assemblies and provide long-term confinement of the commercial SNF. The disposal container will include outer and inner cylinder walls, outer cylinder lids (two on the top, one on the bottom), inner cylinder lids (one on the top, one on the bottom), and an internal metallic basket structure. Exterior labels will provide a means by which to identify the disposal container and its contents. The two metal cylinders, in combination with the cladding, Emplacement Drift System, drip shield, and natural barrier, will support the design philosophy of defense-in-depth. The use of materials with different properties prevents a single mode failure from breaching the waste package. The inner cylinder and inner cylinder lids will be constructed of stainless steel and the outer cylinder and outer cylinder lid will be made of high-nickel alloy. The basket will assist criticality control, provide structural support, and improve heat transfer. The Uncanistered SNF Disposal Container System interfaces with the emplacement drift environment and internal waste by transferring heat from the SNF to the external environment and by protecting the SFN assemblies and their contents from damage/degradation by the external environment. The system also interfaces with the SFN by limiting access of moderator and oxidizing agents of the SFN. The waste package interfaces with the Emplacement Drift System's emplacement drift pallets upon which the wasted packages are placed. The disposal container interfaces with the Assembly Transfer System, Waste Emplacement/Retrieval System, Disposal Container Handling System, and Waste Package Remediation System during loading, handling, transfer, emplacement and retrieval of the disposal container/waste package.

NONE

2000-10-12T23:59:59.000Z

73

Naval Waste Package Design Report  

SciTech Connect (OSTI)

A design methodology for the waste packages and ancillary components, viz., the emplacement pallets and drip shields, has been developed to provide designs that satisfy the safety and operational requirements of the Yucca Mountain Project. This methodology is described in the ''Waste Package Design Methodology Report'' Mecham 2004 [DIRS 166168]. To demonstrate the practicability of this design methodology, four waste package design configurations have been selected to illustrate the application of the methodology. These four design configurations are the 21-pressurized water reactor (PWR) Absorber Plate waste package, the 44-boiling water reactor (BWR) waste package, the 5-defense high-level waste (DHLW)/United States (U.S.) Department of Energy (DOE) spent nuclear fuel (SNF) Co-disposal Short waste package, and the Naval Canistered SNF Long waste package. Also included in this demonstration is the emplacement pallet and continuous drip shield. The purpose of this report is to document how that design methodology has been applied to the waste package design configurations intended to accommodate naval canistered SNF. This demonstrates that the design methodology can be applied successfully to this waste package design configuration and support the License Application for construction of the repository.

M.M. Lewis

2004-03-15T23:59:59.000Z

74

RUSSIAN-ORIGIN HIGHLY ENRICHED URANIUM SPENT NUCLEAR FUEL SHIPMENT FROM BULGARIA  

SciTech Connect (OSTI)

In July 2008, the Global Threat Reduction Initiative and the IRT 2000 research reactor in Sofia, Bulgaria, operated by the Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped 6.4 kilograms of Russian origin highly enriched uranium (HEU) spent nuclear fuel (SNF) to the Russian Federation. The shipment, which resulted in the removal of all HEU from Bulgaria, was conducted by truck, barge, and rail modes of transport across two transit countries before reaching the final destination at the Production Association Mayak facility in Chelyabinsk, Russia. This paper describes the work, equipment, organizations, and approvals that were required to complete the spent fuel shipment and provides lessons learned that might assist other research reactor operators with their own spent nuclear fuel shipments.

Kelly Cummins; Igor Bolshinsky; Ken Allen; Tihomir Apostolov; Ivaylo Dimitrov

2009-07-01T23:59:59.000Z

75

A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel  

E-Print Network [OSTI]

A Tomographic Method for Verification of the Integrity of Spent Nuclear Fuel S. Jacobsson.g. in medicine. This thesis describes a tomographic method developed for measurements on nuclear fuel assemblies of the integrity of the assemblies, i.e. for controlling that all fuel rods are present. The application has been

Haviland, David

76

Evaluation of Radiation Impacts of Spent Nuclear Fuel Storage (SNFS-2) of Chernobyl NPP - 13495  

SciTech Connect (OSTI)

Radiation effects are estimated for the operation of a new dry storage facility for spent nuclear fuel (SNFS-2) of Chernobyl NPP RBMK reactors. It is shown that radiation exposure during normal operation, design and beyond design basis accidents are minor and meet the criteria for safe use of radiation and nuclear facilities in Ukraine. (authors)

Paskevych, Sergiy; Batiy, Valiriy; Sizov, Andriy [Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, 36 a Kirova str. Chornobyl, Kiev region, 07200 (Ukraine)] [Institute for Safety Problems of Nuclear Power Plants, National Academy of Sciences of Ukraine, 36 a Kirova str. Chornobyl, Kiev region, 07200 (Ukraine); Schmieman, Eric [Battelle Memorial Institute, PO Box 999 MSIN K6-90, Richland, WA 99352 (United States)] [Battelle Memorial Institute, PO Box 999 MSIN K6-90, Richland, WA 99352 (United States)

2013-07-01T23:59:59.000Z

77

Conceptual design report for the ICPP spent nuclear fuel dry storage project  

SciTech Connect (OSTI)

The conceptual design is presented for a facility to transfer spent nuclear fuel from shipping casks to dry storage containers, and to safely store those containers at ICPP at INEL. The spent fuels to be handled at the new facility are identified and overall design and operating criteria established. Physical configuration of the facility and the systems used to handle the SNF are described. Detailed cost estimate for design and construction of the facility is presented.

NONE

1996-07-01T23:59:59.000Z

78

Savannah River Site Spent Nuclear Fuel Management Final Environmental Impact Statement  

SciTech Connect (OSTI)

The proposed DOE action considered in this environmental impact statement (EIS) is to implement appropriate processes for the safe and efficient management of spent nuclear fuel and targets at the Savannah River Site (SRS) in Aiken County, South Carolina, including placing these materials in forms suitable for ultimate disposition. Options to treat, package, and store this material are discussed. The material included in this EIS consists of approximately 68 metric tons heavy metal (MTHM) of spent nuclear fuel 20 MTHM of aluminum-based spent nuclear fuel at SRS, as much as 28 MTHM of aluminum-clad spent nuclear fuel from foreign and domestic research reactors to be shipped to SRS through 2035, and 20 MTHM of stainless-steel or zirconium-clad spent nuclear fuel and some Americium/Curium Targets stored at SRS. Alternatives considered in this EIS encompass a range of new packaging, new processing, and conventional processing technologies, as well as the No Action Alternative. A preferred alternative is identified in which DOE would prepare about 97% by volume (about 60% by mass) of the aluminum-based fuel for disposition using a melt and dilute treatment process. The remaining 3% by volume (about 40% by mass) would be managed using chemical separation. Impacts are assessed primarily in the areas of water resources, air resources, public and worker health, waste management, socioeconomic, and cumulative impacts.

N /A

2000-04-14T23:59:59.000Z

79

MANAGEMENT OF RESEARCH AND TEST REACTOR ALUMINUM SPENT NUCLEAR FUEL - A TECHNOLOGY ASSESSMENT  

SciTech Connect (OSTI)

The Department of Energy's Environmental Management (DOE-EM) Program is responsible for the receipt and storage of aluminum research reactor spent nuclear fuel or used fuel until ultimate disposition. Aluminum research reactor used fuel is currently being stored or is anticipated to be returned to the U.S. and stored at DOE-EM storage facilities at the Savannah River Site and the Idaho Nuclear Technology and Engineering Center. This paper assesses the technologies and the options for safe transportation/receipt and interim storage of aluminum research reactor spent fuel and reviews the comprehensive strategy for its management. The U.S. Department of Energy uses the Appendix A, Spent Nuclear Fuel Acceptance Criteria, to identify the physical, chemical, and isotopic characteristics of spent nuclear fuel to be returned to the United States under the Foreign Research Reactor Spent Nuclear Fuel Acceptance Program. The fuel is further evaluated for acceptance through assessments of the fuel at the foreign sites that include corrosion damage and handleability. Transport involves use of commercial shipping casks with defined leakage rates that can provide containment of the fuel, some of which are breached. Options for safe storage include wet storage and dry storage. Both options must fully address potential degradation of the aluminum during the storage period. This paper focuses on the various options for safe transport and storage with respect to technology maturity and application.

Vinson, D.

2010-07-11T23:59:59.000Z

80

Fresh and Spent Nuclear Fuel Repatriation from the IRT-2000 Research Reactor Facility, Sofia, Bulgaria  

SciTech Connect (OSTI)

The IRT 2000 research reactor, operated by the Bulgarian Institute for Nuclear Research and Nuclear Energy (INRNE), safely shipped all of their Russian-origin nuclear fuel from the Republic of Bulgaria to the Russian Federation beginning in 2003 and completing in 2008. These fresh and spent fuel shipments removed all highly enriched uranium (HEU) from Bulgaria. The fresh fuel was shipped by air in December 2003 using trucks and a commercial cargo aircraft. One combined spent fuel shipment of HEU and low enriched uranium (LEU) was completed in July 2008 using high capacity VPVR/M casks transported by truck, barge, and rail. The HEU shipments were assisted by the Russian Research Reactor Fuel Return Program (RRRFR) and the LEU spent fuel shipment was funded by Bulgaria. This report describes the work, approvals, organizations, equipment, and agreements required to complete these shipments and concludes with several major lessons learned.

K. J. Allen; T. G. Apostolov; I. S. Dimitrov

2009-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" from the National Library of EnergyBeta (NLEBeta).
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81

Final Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel  

SciTech Connect (OSTI)

DOE is responsible for the safe and efficient management of its sodium-bonded spent nuclear fuel. This fuel contains metallic sodium, a highly reactive material; metallic uranium, which is also reactive; and in some cases, highly enriched uranium. The presence of reactive materials could complicate the process of qualifying and licensing DOE's sodium-bonded spent nuclear fuel inventory for disposal in a geologic repository. Currently, more than 98 percent of this inventory is located at the Idaho National Engineering and Environmental Laboratory (INEEL), near Idaho Falls, Idaho. In addition, in a 1995 agreement with the State of Idaho, DOE committed to remove all spent nuclear fuel from Idaho by 2035. This EIS evaluates the potential environmental impacts associated with the treatment and management of sodium-bonded spent nuclear fuel in one or more facilities located at Argonne National Laboratory-West (ANL-W) at INEEL and either the F-Canyon or Building 105-L at the Savannah River Site (SRS) near Aiken, South Carolina. DOE has identified and assessed six proposed action alternatives in this EIS. These are: (1) electrometallurgical treatment of all fuel at ANL-W, (2) direct disposal of blanket fuel in high-integrity cans with the sodium removed at ANL-W, (3) plutonium-uranium extraction (PUREX) processing of blanket fuel at SRS, (4) melt and dilute processing of blanket fuel at ANL-W, (5) melt and dilute processing of blanket fuel at SRS, and (6) melt and dilute processing of all fuel at ANL-W. In addition, Alternatives 2 through 5 include the electrometallurgical treatment of driver fuel at ANL-W. Under the No Action Alternative, the EIS evaluates both the continued storage of sodium-bonded spent nuclear fuel until the development of a new treatment technology or direct disposal without treatment. Under all of the alternatives, the affected environment is primarily within 80 kilometers (50 miles) of spent nuclear fuel treatment facilities. Analyses indicate little difference in the environmental impacts among alternatives. DOE has identified electrometallurgical treatment as its Preferred Alternative for the treatment and management of all sodium-bonded spent nuclear fuel, except for the Fermi-1 blanket fuel. The No Action Alternative is preferred for the Fermi-1 blanket spent nuclear fuel.

N /A

2000-08-04T23:59:59.000Z

82

Draft Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel  

SciTech Connect (OSTI)

This document summarizes the U.S. Department of Energy's ''Draft Environmental Impact Statement for the Treatment and Management of Sodium-Bonded Spent Nuclear Fuel''. In addition to information concerning the background, purpose and need for the proposed action, and the National Environmental Policy Act process, this summary describes the characteristics of sodium-bonded spent nuclear fuel, the proposed treatment methods, the proposed facilities, the alternatives considered, and the environmental consequences of these alternatives. A glossary is included at the end to assist the reader with some of the technical terms used in this document.

N /A

1999-07-30T23:59:59.000Z

83

Nuclear nonproliferation: Concerns with US delays in accepting foregin research reactors` spent fuel  

SciTech Connect (OSTI)

One key US nonproliferation goal is to discourage use of highly enriched uranium fuel (HEU), which can be used to make nuclear bombs, in civilian nuclear programs worldwide. DOE`s Off-Site Fuels Policy for taking back spent HEU from foreign research reactors was allowed to expire due to environmental reasons. This report provides information on the effects of delays in renewing the Off-Site Fuels Policy on US nonproliferation goals and programs (specifically the reduced enrichment program), DOE`s efforts to renew the fuels policy, and the price to be charged to the operators of foreign reactors for DOE`s activities in taking back spent fuel.

NONE

1994-03-25T23:59:59.000Z

84

Standard guide for characterization of spent nuclear fuel in support of geologic repository disposal  

E-Print Network [OSTI]

1.1 This guide provides guidance for the types and extent of testing that would be involved in characterizing the physical and chemical nature of spent nuclear fuel (SNF) in support of its interim storage, transport, and disposal in a geologic repository. This guide applies primarily to commercial light water reactor (LWR) spent fuel and spent fuel from weapons production, although the individual tests/analyses may be used as applicable to other spent fuels such as those from research and test reactors. The testing is designed to provide information that supports the design, safety analysis, and performance assessment of a geologic repository for the ultimate disposal of the SNF. 1.2 The testing described includes characterization of such physical attributes as physical appearance, weight, density, shape/geometry, degree, and type of SNF cladding damage. The testing described also includes the measurement/examination of such chemical attributes as radionuclide content, microstructure, and corrosion product c...

American Society for Testing and Materials. Philadelphia

2009-01-01T23:59:59.000Z

85

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1, Appendix C, Savannah River Site Spent Nuclear Fuel Mangement Program  

SciTech Connect (OSTI)

The US Department of Energy (DOE) is engaged in two related decision making processes concerning: (1) the transportation, receipt, processing, and storage of spent nuclear fuel (SNF) at the DOE Idaho National Engineering Laboratory (INEL) which will focus on the next 10 years; and (2) programmatic decisions on future spent nuclear fuel management which will emphasize the next 40 years. DOE is analyzing the environmental consequences of these spent nuclear fuel management actions in this two-volume Environmental Impact Statement (EIS). Volume 1 supports broad programmatic decisions that will have applicability across the DOE complex and describes in detail the purpose and need for this DOE action. Volume 2 is specific to actions at the INEL. This document, which limits its discussion to the Savannah River Site (SRS) spent nuclear fuel management program, supports Volume 1 of the EIS. Following the introduction, Chapter 2 contains background information related to the SRS and the framework of environmental regulations pertinent to spent nuclear fuel management. Chapter 3 identifies spent nuclear fuel management alternatives that DOE could implement at the SRS, and summarizes their potential environmental consequences. Chapter 4 describes the existing environmental resources of the SRS that spent nuclear fuel activities could affect. Chapter 5 analyzes in detail the environmental consequences of each spent nuclear fuel management alternative and describes cumulative impacts. The chapter also contains information on unavoidable adverse impacts, commitment of resources, short-term use of the environment and mitigation measures.

Not Available

1994-06-01T23:59:59.000Z

86

Spent nuclear fuel storage -- Performance tests and demonstrations  

SciTech Connect (OSTI)

This report summarizes the results of heat transfer and shielding performance tests and demonstrations conducted from 1983 through 1992 by or in cooperation with the US Department of Energy (DOE), Office of Commercial Radioactive Waste Management (OCRWM). The performance tests consisted of 6 to 14 runs involving one or two loadings, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. A description of the test plan, spent fuel load patterns, results from temperature and dose rate measurements, and fuel integrity evaluations are contained within the report.

McKinnon, M.A.; DeLoach, V.A.

1993-04-01T23:59:59.000Z

87

Spent Fuel Analyses for Nuclear Safeguards | ornl.gov  

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

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

88

Environmental Assessment of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel  

SciTech Connect (OSTI)

The Department of Energy has completed the Environmental Assessment (EA) of Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel and issued a Finding of No Significant Impact (FONSI) for the proposed action. The EA and FONSI are enclosed for your information. The Department has decided to accept a limited number of spent nuclear fuel elements (409 elements) containing uranium that was enriched in the United States from eight research reactors in Austria, Denmark, Germany, Greece, the Netherlands, Sweden, and Switzerland. This action is necessary to maintain the viability of a major US nuclear weapons nonproliferation program to limit or eliminate the use of highly enriched uranium in civil programs. The purpose of the EA is to maintain the cooperation of the foreign research reactor operators with the nonproliferation program while a more extensive Environmental Impact Statement (EIS) is prepared on a proposed broader policy involving the acceptance of up to 15,000 foreign research reactor spent fuel elements over a 10 to 15 year period. Based on an evaluation of transport by commercial container liner or chartered vessel, five eastern seaboard ports, and truck and train modes of transporting the spent fuel overland to the Savannah River Sits, the Department has concluded that no significant impact would result from any combination of port and made of transport. In addition, no significant impacts were found from interim storage of spent fuel at the Savannah River Site.

Not Available

1994-04-01T23:59:59.000Z

89

Characterization program management plan for Hanford K basin spent nuclear fuel  

SciTech Connect (OSTI)

The program management plan for characterization of the K Basin spent nuclear fuel was revised to incorporate corrective actions in response to SNF Project QA surveillance 1K-FY-99-060. This revision of the SNF Characterization PMP replaces Duke Eng.

TRIMBLE, D.J.

1999-07-19T23:59:59.000Z

90

Spent nuclear fuel discharges from US reactors 1992  

SciTech Connect (OSTI)

This report provides current statistical data on every fuel assembly irradiated in commercial nuclear reactors operating in the United States. It also provides data on the current inventories and storage capacities of those reactors to a wide audience, including Congress, Federal and State agencies, the nuclear and electric industries and the general public. It uses data from the mandatory, ``Nuclear Fuel Data`` survey, Form RW-859 for 1992 and historical data collected by the Energy Information Administration (EIA) on previous Form RW-859 surveys. The report was prepared by the EIA under a Memorandum of Understanding with the Office of Civilian Radioactive Waste Management.

Not Available

1994-05-05T23:59:59.000Z

91

Dry Storage of Research Reactor Spent Nuclear Fuel - 13321  

SciTech Connect (OSTI)

Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)

Adams, T.M.; Dunsmuir, M.D.; Leduc, D.R.; Severynse, T.F.; Sindelar, R.L. [Savannah River National Laboratory (United States)] [Savannah River National Laboratory (United States); Moore, E.N. [Moore Nuclear Energy, LLC (United States)] [Moore Nuclear Energy, LLC (United States)

2013-07-01T23:59:59.000Z

92

Spent nuclear fuel integrity during dry storage - performance tests and demonstrations  

SciTech Connect (OSTI)

This report summarizes the results of fuel integrity surveillance determined from gas sampling during and after performance tests and demonstrations conducted from 1983 through 1996 by or in cooperation with the US DOE Office of Commercial Radioactive Waste Management (OCRWM). The cask performance tests were conducted at Idaho National Engineering Laboratory (INEL) between 1984 and 1991 and included visual observation and ultrasonic examination of the condition of the cladding, fuel rods, and fuel assembly hardware before dry storage and consolidation of fuel, and a qualitative determination of the effects of dry storage and fuel consolidation on fission gas release from the spent fuel rods. The performance tests consisted of 6 to 14 runs involving one or two loading, usually three backfill environments (helium, nitrogen, and vacuum backfills), and one or two storage system orientations. The nitrogen and helium backfills were sampled and analyzed to detect leaking spent fuel rods. At the end of each performance test, periodic gas sampling was conducted on each cask. A spent fuel behavior project (i.e., enhanced surveillance, monitoring, and gas sampling activities) was initiated by DOE in 1994 for intact fuel in a CASTOR V/21 cask and for consolidated fuel in a VSC-17 cask. The results of the gas sampling activities are included in this report. Information on spent fuel integrity is of interest in evaluating the impact of long-term dry storage on the behavior of spent fuel rods. Spent fuel used during cask performance tests at INEL offers significant opportunities for confirmation of the benign nature of long-term dry storage. Supporting cask demonstration included licensing and operation of an independent spent fuel storage installation (ISFSI) at the Virginia Power (VP) Surry reactor site. A CASTOR V/21, an MC-10, and a Nuclear Assurance NAC-I28 have been loaded and placed at the VP ISFSI as part of the demonstration program. 13 refs., 14 figs., 9 tabs.

McKinnon, M.A.; Doherty, A.L.

1997-06-01T23:59:59.000Z

93

Spent Fuel Background Report Volume I  

SciTech Connect (OSTI)

This report is an overview of current spent nuclear fuel management in the DOE complex. Sources of information include published literature, internal DOE documents, interviews with site personnel, and information provided by individual sites. Much of the specific information on facilities and fuels was provided by the DOE sites in response to the questionnaire for data for spent fuels and facilities data bases. This information is as accurate as is currently available, but is subject to revision pending results of further data calls. Spent fuel is broadly classified into three categories: (a) production fuels, (b) special fuels, and (c) naval fuels. Production fuels, comprising about 80% of the total inventory, are those used at Hanford and Savannah River to produce nuclear materials for defense. Special fuels are those used in a wide variety of research, development, and testing activities. Special fuels include fuel from DOE and commercial reactors used in research activities at DOE sites. Naval fuels are those developed and used for nuclear-powered naval vessels and for related research and development. Given the recent DOE decision to curtail reprocessing, the topic of main concern in the management of spent fuel is its storage. Of the DOE sites that have spent nuclear fuel, the vast majority is located at three sites-Hanford, INEL, and Savannah River. Other sites with spent fuel include Oak Ridge, West Valley, Brookhaven, Argonne, Los Alamos, and Sandia. B&W NESI Lynchburg Technology Center and General Atomics are commercial facilities with DOE fuel. DOE may also receive fuel from foreign research reactors, university reactors, and other commercial and government research reactors. Most DOE spent fuel is stored in water-filled pools at the reactor facilities. Currently an engineering study is being performed to determine the feasibility of using dry storage for DOE-owned spent fuel currently stored at various facilities. Delays in opening the deep geologic repository and the decision to phase out reprocessing of production fuels are extending the need for interim storage. The report describes the basic storage conditions and the general SNF inventory at individual DOE facilities.

Abbott, D.

1994-03-01T23:59:59.000Z

94

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

SciTech Connect (OSTI)

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

NONE

1995-09-01T23:59:59.000Z

95

Full-Scale Cask Testing and Public Acceptance of Spent Nuclear Fuel Shipments - 12254  

SciTech Connect (OSTI)

Full-scale physical testing of spent fuel shipping casks has been proposed by the National Academy of Sciences (NAS) 2006 report on spent nuclear fuel transportation, and by the Presidential Blue Ribbon Commission (BRC) on America's Nuclear Future 2011 draft report. The U.S. Nuclear Regulatory Commission (NRC) in 2005 proposed full-scale testing of a rail cask, and considered 'regulatory limits' testing of both rail and truck casks (SRM SECY-05-0051). The recent U.S. Department of Energy (DOE) cancellation of the Yucca Mountain project, NRC evaluation of extended spent fuel storage (possibly beyond 60-120 years) before transportation, nuclear industry adoption of very large dual-purpose canisters for spent fuel storage and transport, and the deliberations of the BRC, will fundamentally change assumptions about the future spent fuel transportation system, and reopen the debate over shipping cask performance in severe accidents and acts of sabotage. This paper examines possible approaches to full-scale testing for enhancing public confidence in risk analyses, perception of risk, and acceptance of spent fuel shipments. The paper reviews the literature on public perception of spent nuclear fuel and nuclear waste transportation risks. We review and summarize opinion surveys sponsored by the State of Nevada over the past two decades, which show consistent patterns of concern among Nevada residents about health and safety impacts, and socioeconomic impacts such as reduced property values along likely transportation routes. We also review and summarize the large body of public opinion survey research on transportation concerns at regional and national levels. The paper reviews three past cask testing programs, the way in which these cask testing program results were portrayed in films and videos, and examines public and official responses to these three programs: the 1970's impact and fire testing of spent fuel truck casks at Sandia National Laboratories, the 1980's regulatory and demonstration testing of MAGNOX fuel flasks in the United Kingdom (the CEGB 'Operation Smash Hit' tests), and the 1980's regulatory drop and fire tests conducted on the TRUPACT II containers used for transuranic waste shipments to the Waste Isolation Pilot Plant in New Mexico. The primary focus of the paper is a detailed evaluation of the cask testing programs proposed by the NRC in its decision implementing staff recommendations based on the Package Performance Study, and by the State of Nevada recommendations based on previous work by Audin, Resnikoff, Dilger, Halstead, and Greiner. The NRC approach is based on demonstration impact testing (locomotive strike) of a large rail cask, either the TAD cask proposed by DOE for spent fuel shipments to Yucca Mountain, or a similar currently licensed dual-purpose cask. The NRC program might also be expanded to include fire testing of a legal-weight truck cask. The Nevada approach calls for a minimum of two tests: regulatory testing (impact, fire, puncture, immersion) of a rail cask, and extra-regulatory fire testing of a legal-weight truck cask, based on the cask performance modeling work by Greiner. The paper concludes with a discussion of key procedural elements - test costs and funding sources, development of testing protocols, selection of testing facilities, and test peer review - and various methods of communicating the test results to a broad range of stakeholder audiences. (authors)

Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge Northridge, CA 91330 (United States)

2012-07-01T23:59:59.000Z

96

Molten tin reprocessing of spent nuclear fuel elements. [Patent application; continuous process  

DOE Patents [OSTI]

A method and apparatus for reprocessing spent nuclear fuel is described. Within a containment vessel, a solid plug of tin and nitride precipitates supports a circulating bath of liquid tin therein. Spent nuclear fuel is immersed in the liquid tin under an atmosphere of nitrogen, resulting in the formation of nitride precipitates. The layer of liquid tin and nitride precipitates which interfaces the plug is solidified and integrated with the plug. Part of the plug is melted, removing nitride precipitates from the containment vessel, while a portion of the plug remains solidified to support te liquid tin and nitride precipitates remaining in the containment vessel. The process is practiced numerous times until substantially all of the precipitated nitrides are removed from the containment vessel.

Heckman, R.A.

1980-12-19T23:59:59.000Z

97

Intact and Degraded Component Criticality Calculations of N Reactors Spent Nuclear Fuel  

SciTech Connect (OSTI)

The objective of this calculation is to perform intact and degraded mode criticality evaluations of the Department of Energy's (DOE) N Reactor Spent Nuclear Fuel codisposed in a 2-Defense High-Level Waste (2-DHLW)/2-Multi-Canister Overpack (MCO) Waste Package (WP) and emplaced in a monitored geologic repository (MGR) (see Attachment I). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for both intact and degraded mode internal configurations of the codisposal waste package. This calculation will support the analysis that will be performed to demonstrate the technical viability for disposing of U-metal (N Reactor) spent nuclear fuel in the potential MGR.

L. Angers

2001-01-31T23:59:59.000Z

98

Investigation of Electrochemical Recovery of Zirconium from Spent Nuclear Fuels  

SciTech Connect (OSTI)

This project uses both modeling and experimental studies to design optimal electrochemical technology methods for recovery of zirconium from used nuclear fuel rods for more effective waste management. The objectives are to provide a means of efficiently separating zirconium into metallic high-level waste forms and to support development of a process for decontamination of zircaloy hulls to enable their disposal as low- and intermediate-level waste. Modeling work includes extension of a 3D model previously developed by Seoul National University for uranium electrorefining by adding the ability to predict zirconium behavior. Experimental validation activities include tests for recovery of zirconium from molten salt solutions and aqueous tests using surrogate materials. *This is a summary of the FY 2013 progress for I-NERI project # 2010-001-K provided to the I-NERI office.

Michael Simpson; II-Soon Hwang

2014-06-01T23:59:59.000Z

99

K Basin spent nuclear fuel hot conditioning system functions {ampersand} requirements  

SciTech Connect (OSTI)

The purpose of this F{ampersand}R document is to establish the functional requirements baseline for the Spent Nuclear Fuel Hot Conditioning System (HCS) subproject. This F{ampersand}R documents the: -mission of the HCS, -evolution of the technical baseline leading to the HCS, -functions that must be performed to accomplish the HCS mission, -requirements basis allocated to the HCS mission and functions, -identification and definition of interfaces between the HCS and other SNF subprojects.

Miska, C.R., Westinghouse Hanford

1996-07-08T23:59:59.000Z

100

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

SciTech Connect (OSTI)

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

Forsberg, C.W.

1997-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" 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

Separation of uranium from technetium in recovery of spent nuclear fuel  

DOE Patents [OSTI]

Uranium and technetium in the product stream of the Purex process for recovery of uranium in spent nuclear fuel are separated by (1) contacting the aqueous Purex product stream with hydrazine to reduce Tc/sup +7/ therein to a reduced species, and (2) contacting said aqueous stream with an organic phase containing tributyl phosphate and an organic diluent to extract uranium from said aqueous stream into said organic phase.

Pruett, D.J.; McTaggart, D.R.

1983-08-31T23:59:59.000Z

102

LCEs for Naval Reactor Benchmark Calculations  

SciTech Connect (OSTI)

The purpose of this engineering calculation is to document the MCNP4B2LV evaluations of Laboratory Critical Experiments (LCEs) performed as part of the Disposal Criticality Analysis Methodology program. LCE evaluations documented in this report were performed for 22 different cases with varied design parameters. Some of these LCEs (10) are documented in existing references (Ref. 7.1 and 7.2), but were re-run for this calculation file using more neutron histories. The objective of this analysis is to quantify the MCNP4B2LV code system's ability to accurately calculate the effective neutron multiplication factor (k{sub eff}) for various critical configurations. These LCE evaluations support the development and validation of the neutronics methodology used for criticality analyses involving Naval reactor spent nuclear fuel in a geologic repository.

W.J. Anderson

1999-07-19T23:59:59.000Z

103

Development of a techno-economic model to optimization DOE spent nuclear fuel disposition  

SciTech Connect (OSTI)

The purpose of the National Spent Nuclear Fuel (NSNF) Program conducted by Lockheed Martin Idaho Technology Co. (LMITCO) at the Idaho National Engineering and Environmental Laboratory (INEEL) is to evaluate what to do with the spent nuclear fuel (SNF) in the Department of Energy (DOE) complex. Final disposition of the SNF may require that the fuel be treated to minimize material concerns. The treatments may range from electrometallurgical treatment and chemical dissolution to engineering controls. Treatment options and treatment locations will depend on the fuel type and the current locations of the fuel. One of the first steps associated with selecting one or more sites for treating the SNF in the DOE complex is to determine the cost of each option. An economic analysis will assist in determining which fuel treatment alternative attains the optimum disposition of SNF at the lowest possible cost to the government and the public. For this study, a set of questions was developed for the electrometallurgical treatment process for fuels at several locations. The set of questions addresses all issues associated with the design, construction, and operation of a production facility. A matrix table was developed to determine questions applicable to various fuel treatment options. A work breakdown structure (WBS) was developed to identify a treatment process and costs from initial design to shipment of treatment products to final disposition. Costs will be applied to determine the life-cycle cost of each option. This technique can also be applied to other treatment techniques for treating spent nuclear fuel.

Ramer, R.J.; Plum, M.M.; Adams, J.P.; Dahl, C.A.

1997-11-01T23:59:59.000Z

104

The Impact of Microbially Influenced Corrosion on Spent Nuclear Fuel and Storage Life  

SciTech Connect (OSTI)

A study was performed to evaluate if microbial activity could be considered a threat to spent nuclear fuel integrity. The existing data regarding the impact of microbial influenced corrosion (MIC) on spent nuclear fuel storage does not allow a clear assessment to be made. In order to identify what further data are needed, a literature survey on MIC was accomplished with emphasis on materials used in nuclear fuel fabrication, e.g., A1, 304 SS, and zirconium. In addition, a survey was done at Savannah River, Oak Ridge, Hanford, and the INEL on the condition of their wet storage facilities. The topics discussed were the SNF path forward, the types of fuel, ramifications of damaged fuel, involvement of microbial processes, dry storage scenarios, ability to identify microbial activity, definitions of water quality, and the use of biocides. Information was also obtained at international meetings in the area of biological mediated problems in spent fuel and high level wastes. Topics dis cussed included receiving foreign reactor research fuels into existing pools, synergism between different microbes and other forms of corrosion, and cross contamination.

J. H. Wolfram; R. E. Mizia; R. Jex; L. Nelson; K. M. Garcia

1996-10-01T23:59:59.000Z

105

Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel  

SciTech Connect (OSTI)

This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy`s (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States [CIS]). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

Schneider, K.J.; Mitchell, S.J.

1992-04-01T23:59:59.000Z

106

Foreign experience on effects of extended dry storage on the integrity of spent nuclear fuel  

SciTech Connect (OSTI)

This report summarizes the results of a survey of foreign experience in dry storage of spent fuel from nuclear power reactors that was carried out for the US Department of Energy's (DOE) Office of Civilian Radioactive Waste Management (OCRWM). The report reviews the mechanisms for degradation of spent fuel cladding and fuel materials in dry storage, identifies the status and plans of world-wide experience and applications, and documents the available information on the expected long-term integrity of the dry-stored spent fuel from actual foreign experience. Countries covered in this survey are: Argentina, Canada, Federal Republic of Germany (before reunification with the former East Germany), former German Democratic Republic (former East Germany), France, India, Italy, Japan, South Korea, Spain, Switzerland, United Kingdom, and the former USSR (most of these former Republics are now in the Commonwealth of Independent States (CIS)). Industrial dry storage of Magnox fuels started in 1972 in the United Kingdom; Canada began industrial dry storage of CANDU fuels in 1980. The technology for safe storage is generally considered to be developed for time periods of 30 to 100 years for LWR fuel in inert gas and for some fuels in oxidizing gases at low temperatures. Because it will probably be decades before countries will have a repository for spent fuels and high-level wastes, the plans for expanded use of dry storage have increased significantly in recent years and are expected to continue to increase in the near future.

Schneider, K.J.; Mitchell, S.J.

1992-04-01T23:59:59.000Z

107

Truck and rail charges for shipping spent fuel and nuclear waste  

SciTech Connect (OSTI)

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

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

1986-06-01T23:59:59.000Z

108

Dry halide method for separating the components of spent nuclear fuels  

DOE Patents [OSTI]

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200 C to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400 C; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164 to 2 C; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic. 3 figs.

Christian, J.D.; Thomas, T.R.; Kessinger, G.F.

1998-06-30T23:59:59.000Z

109

Dry halide method for separating the components of spent nuclear fuels  

DOE Patents [OSTI]

The invention is a nonaqueous, single method for processing multiple spent nuclear fuel types by separating the fission- and transuranic products from the nonradioactive and fissile uranium product. The invention has four major operations: exposing the spent fuels to chlorine gas at temperatures preferably greater than 1200.degree. C. to form volatile metal chlorides; removal of the fission product chlorides, transuranic product chlorides, and any nickel chloride and chromium chloride in a molten salt scrubber at approximately 400.degree. C.; fractional condensation of the remaining volatile chlorides at temperatures ranging from 164.degree. C. to 2.degree. C.; and regeneration and recovery of the transferred spent molten salt by vacuum distillation. The residual fission products, transuranic products, and nickel- and chromium chlorides are converted to fluorides or oxides for vitrification. The method offers the significant advantages of a single, compact process that is applicable to most of the diverse nuclear fuels, minimizes secondary wastes, segregates fissile uranium from the high level wastes to resolve potential criticality concerns, segregates nonradioactive wastes from the high level wastes for volume reduction, and produces a common waste form glass or glass-ceramic.

Christian, Jerry Dale (Idaho Falls, ID); Thomas, Thomas Russell (Rigby, ID); Kessinger, Glen F. (Idaho Falls, ID)

1998-01-01T23:59:59.000Z

110

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

E-Print Network [OSTI]

and Nuclear Recoil . . . . . . . . . . . . . . . . . . . . .2 Quantitative Measurements using NRF 2.1 Nuclear ResonanceFuture Work A Transmission Nuclear Resonance Fluorescence

Ludewigt, Bernhard A

2011-01-01T23:59:59.000Z

111

THE ECONOMICS OF REPROCESSING vs DIRECT DISPOSAL OF SPENT NUCLEAR FUEL  

SciTech Connect (OSTI)

This report assesses the economics of reprocessing versus direct disposal of spent nuclear fuel. The breakeven uranium price at which reprocessing spent nuclear fuel from existing light-water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is assessed, using central estimates of the costs of different elements of the nuclear fuel cycle (and other fuel cycle input parameters), for a wide range of range of potential reprocessing prices. Sensitivity analysis is performed, showing that the conclusions reached are robust across a wide range of input parameters. The contribution of direct disposal or reprocessing and recycling to electricity cost is also assessed. The choice of particular central estimates and ranges for the input parameters of the fuel cycle model is justified through a review of the relevant literature. The impact of different fuel cycle approaches on the volume needed for geologic repositories is briefly discussed, as are the issues surrounding the possibility of performing separations and transmutation on spent nuclear fuel to reduce the need for additional repositories. A similar analysis is then performed of the breakeven uranium price at which deploying fast neutron breeder reactors would become competitive compared with a once-through fuel cycle in LWRs, for a range of possible differences in capital cost between LWRs and fast neutron reactors. Sensitivity analysis is again provided, as are an analysis of the contribution to electricity cost, and a justification of the choices of central estimates and ranges for the input parameters. The equations used in the economic model are derived and explained in an appendix. Another appendix assesses the quantities of uranium likely to be recoverable worldwide in the future at a range of different possible future prices.

Matthew Bunn; Steve Fetter; John P. Holdren; Bob van der Zwaan

2003-07-01T23:59:59.000Z

112

Nonproliferation impacts assessment for the management of the Savannah River Site aluminum-based spent nuclear fuel  

SciTech Connect (OSTI)

On May 13, 1996, the US established a new, 10-year policy to accept and manage foreign research reactor spent nuclear fuel containing uranium enriched in the US. The goal of this policy is to reduce civilian commerce in weapons-usable highly enriched uranium (HEU), thereby reducing the risk of nuclear weapons proliferation. Two key disposition options under consideration for managing this fuel include conventional reprocessing and new treatment and packaging technologies. The Record of Decision specified that, while evaluating the reprocessing option, ``DOE will commission or conduct an independent study of the nonproliferation and other (e.g., cost and timing) implications of chemical separation of spent nuclear fuel from foreign research reactors.`` DOE`s Office of Arms Control and Nonproliferation conducted this study consistent with the aforementioned Record of Decision. This report addresses the nonproliferation implications of the technologies under consideration for managing aluminum-based spent nuclear fuel at the Savannah River Site. Because the same technology options are being considered for the foreign research reactor and the other aluminum-based spent nuclear fuels discussed in Section ES.1, this report addresses the nonproliferation implications of managing all the Savannah River Site aluminum-based spent nuclear fuel, not just the foreign research reactor spent nuclear fuel. The combination of the environmental impact information contained in the draft EIS, public comment in response to the draft EIS, and the nonproliferation information contained in this report will enable the Department to make a sound decision regarding how to manage all aluminum-based spent nuclear fuel at the Savannah River Site.

NONE

1998-12-01T23:59:59.000Z

113

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

SciTech Connect (OSTI)

The Next Generation Safeguard Initiative (NGSI) of the U.S Department of Energy is supporting a multi-lab/university collaboration to quantify the plutonium (Pu) mass in spent nuclear fuel (SNF) assemblies and to detect the diversion of pins with non-destructive assay (NDA) methods. The following 14 NDA techniques are being studied: Delayed Neutrons, Differential Die-Away, Differential Die-Away Self-Interrogation, Lead Slowing Down Spectrometer, Neutron Multiplicity, Passive Neutron Albedo Reactivity, Total Neutron (Gross Neutron), X-Ray Fluorescence, {sup 252}Cf Interrogation with Prompt Neutron Detection, Delayed Gamma, Nuclear Resonance Fluorescence, Passive Prompt Gamma, Self-integration Neutron Resonance Densitometry, and Neutron Resonance Transmission Analysis. Understanding and maturity of the techniques vary greatly, ranging from decades old, well-understood methods to new approaches. Nuclear Resonance Fluorescence (NRF) is a technique that had not previously been studied for SNF assay or similar applications. Since NRF generates isotope-specific signals, the promise and appeal of the technique lies in its potential to directly measure the amount of a specific isotope in an SNF assay target. The objectives of this study were to design and model suitable NRF measurement methods, to quantify capabilities and corresponding instrumentation requirements, and to evaluate prospects and the potential of NRF for SNF assay. The main challenge of the technique is to achieve the sensitivity and precision, i.e., to accumulate sufficient counting statistics, required for quantifying the mass of Pu isotopes in SNF assemblies. Systematic errors, considered a lesser problem for a direct measurement and only briefly discussed in this report, need to be evaluated for specific instrument designs in the future. Also, since the technical capability of using NRF to measure Pu in SNF has not been established, this report does not directly address issues such as cost, size, development time, nor concerns related to the use of Pu in measurement systems. This report discusses basic NRF measurement concepts, i.e., backscatter and transmission methods, and photon source and {gamma}-ray detector options in Section 2. An analytical model for calculating NRF signal strengths is presented in Section 3 together with enhancements to the MCNPX code and descriptions of modeling techniques that were drawn upon in the following sections. Making extensive use of the model and MCNPX simulations, the capabilities of the backscatter and transmission methods based on bremsstrahlung or quasi-monoenergetic photon sources were analyzed as described in Sections 4 and 5. A recent transmission experiment is reported on in Appendix A. While this experiment was not directly part of this project, its results provide an important reference point for our analytical estimates and MCNPX simulations. Used fuel radioactivity calculations, the enhancements to the MCNPX code, and details of the MCNPX simulations are documented in the other appendices.

Ludewigt, Bernhard A; Quiter, Brian J.; Ambers, Scott D.

2011-01-14T23:59:59.000Z

114

US Spent (Used) Fuel Status, Management and Likely Directions- 12522  

SciTech Connect (OSTI)

As of 2010, the US has accumulated 65,200 MTU (42,300 MTU of PWR's; 23,000 MTU of BWR's) of spent (irradiated or used) fuel from 104 operating commercial nuclear power plants situated at 65 sites in 31 States and from previously shutdown commercial nuclear power plants. Further, the Department of Energy (DOE) has responsibility for an additional 2458 MTU of DOE-owned defense and non defense spent fuel from naval nuclear power reactors, various non-commercial test reactors and reactor demonstrations. The US has no centralized large spent fuel storage facility for either commercial spent fuel or DOE-owned spent fuel. The 65,200 MTU of US spent fuel is being safely stored by US utilities at numerous reactor sites in (wet) pools or (dry) metal or concrete casks. As of November 2010, the US had 63 'independent spent fuel storage installations' (or ISFSI's) licensed by the US Nuclear Regulatory Commission located at 57 sites in 33 states. Over 1400 casks loaded with spent fuel for dry storage are at these licensed ISFSI's; 47 sites are located at commercial reactor sites and 10 are located 'away' from a reactor (AFR's) site. DOE's small fraction of a 2458 MTU spent fuel inventory, which is not commercial spent fuel, is with the exception of 2 MTU, being stored at 4 sites in 4 States. The decades old US policy of a 'once through' fuel cycle with no recycle of spent fuel was set into a state of 'mass confusion or disruption' when the new US President Obama's administration started in early 2010 stopping the only US geologic disposal repository at the Yucca Mountain site in the State of Nevada from being developed and licensed. The practical result is that US nuclear power plant operators will have to continue to be responsible for managing and storing their own spent fuel for an indefinite period of time at many different sites in order to continue to generate electricity because there is no current US government plan, schedule or policy for taking possession of accumulated spent fuel from the utilities. There are technical solutions for continuing the safe storage of spent fuel for 100 years or more and these solutions will be implemented by the US utilities that need to keep their nuclear power plants operating while the unknown political events are played out to establish future US policy decisions that can remain in place long enough regarding accumulated spent fuel inventories to implement any new US spent fuel centralized storage or disposition policy by the US government. (author)

Jardine, Leslie J. [L. J. Jardine Services, Consultant, Dublin CA, 94568 (United States)

2012-07-01T23:59:59.000Z

115

Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository - Volume 3: Appendices  

SciTech Connect (OSTI)

The United States Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).

Taylor, L.L.; Wilson, J.R. (INEEL); Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K. (SNL); Rath, J.S. (New Mexico Engineering Research Institute)

1998-10-01T23:59:59.000Z

116

Technical data summary supporting the spent nuclear fuel environment impact statement, March 1994  

SciTech Connect (OSTI)

This report has been compiled by the WSRC Nuclear Materials Processing Division`s Planning Section at the request of the Office of Spent Fuel Management and Special Projects (EM-37) to support issuance of the Spent Nuclear Fuel Environmental Impact Statement. Savannah River Site input data evaluates five programmatic options (including {open_quotes}No Action{close_quotes}) ranging up to transfer of all DOE responsibility spent fuel to the SRS. For each option, a range of management/disposition scenarios has been examined. Each case summary provides information relative to the technical proposal, technical issues, environmental impacts, and projected costs for a forty year period (FY-35) when it is assumed that the material will be dispositioned from the SRS. The original issue of the report which was prepared under severe time constraints contained many simplifications and assumptions. Although the revisions have corrected some of the shortcomings of the original report, it is still highly recommended that significant additional study be performed before basing key decisions upon the data contained in this report. The data represents the best effort by a significant group of technical personnel familiar with nuclear materials processing, handling, and storage; but it is likely that careful scrutiny will reveal numerous discrepancies, inconsistencies and omissions. Nor does this report attempt to analyze every potential disposal pathway, but probably establishes the bounds for the most of the viable pathways. The bulk of the effort went into defining the engineering approaches necessary to execute the various mission scenarios which were changed since the last revision. The decision to limit reprocessing to only SRS aluminum clad required a major alteration of the TDS. Collection and/or calculation of much of the various waste, emission, and utility consumption data, so important to an EIS, has been updated since the last revision, but not thoroughly completed.

Geddes, R.L.; Claxton, R.E.; Lengel, J.D. [and others

1994-03-01T23:59:59.000Z

117

Performance Assessment Analyses Unique to Department of Energy Spent Nuclear Fuel  

SciTech Connect (OSTI)

This paper describes the iterative process of grouping and performance assessment that has led to the current grouping of the U.S. Department of Energy (DOE) spent nuclear fuel (SNF). The unique sensitivity analyses that form the basis for incorporating DOE fuel into the total system performance assessment (TSPA) base case model are described. In addition, the chemistry that results from dissolution of DOE fuel and high level waste (HLW) glass in a failed co-disposal package, and the effects of disposal of selected DOE SNF in high integrity cans are presented.

Loo, Henry Hung Yiu; Duguid, J. O.

2000-06-01T23:59:59.000Z

118

Dry air oxidation kinetics of K-Basin spent nuclear fuel  

SciTech Connect (OSTI)

The safety and process analyses of the proposed Integrated Process Strategy (IPS) to move the N-Reactor spent nuclear fuel (SNF) stored at K-Basin to an interim storage facility require information about the oxidation behavior of the metallic uranium. Limited experiments have been performed on the oxidation reaction of SNF samples taken from an N-Reactor outer fuel element in various atmospheres. This report discusses studies on the oxidation behavior of SNF using two independent experimental systems: (1) a tube furnace with a flowing gas mixture of 2% oxygen/98% argon; and (2) a thermogravimetric system for dry air oxidation.

Abrefah, J.; Buchanan, H.C.; Gerry, W.M.; Gray, W.J.; Marschman, S.C.

1998-06-01T23:59:59.000Z

119

Radiation Doses to the Public From the Transport of Spent Nuclear Fuel  

SciTech Connect (OSTI)

This paper reviews issues that have been raised concerning radiological risks and safety of the public exposed to shipments of spent nuclear fuel and high-level radioactive waste to a Yucca Mountain repository. It presents and analyzes the contrasting viewpoints of opponents and proponents, presents facts about radiological exposures and risks, and provides perspective from which to observe the degree of risk that would devolve from the shipments. The paper concludes that the risks to the public's health and safety from being exposed to radiation from the shipments will not be discernable.

Best, R. E.; Maheras, S. J.; Ross, S. S.; Weiner, R.

2003-02-25T23:59:59.000Z

120

Experience With Damaged Spent Nuclear Fuel at U.S. DOE Facilities  

SciTech Connect (OSTI)

This report summarizes some of the challenges encountered and solutions implemented to ensure safe storage and handling of damaged spent nuclear fuels (SNF). It includes a brief summary of some SNF storage environments and resulting SNF degradation, experience with handling and repackaging significantly degraded SNFs, and the associated lessons learned. This work provides useful insight and resolutions to many engineering challenges facing SNF handling and storage facilities. The context of this report is taken from a report produced at Idaho National Laboratory and further detailed information, such as equipment design and usage, can be found in the appendices to that report. (authors)

Carlsen, Brett; Fillmore, Denzel; Woolstenhulme, Eric [Idaho National Laboratory, P.O. Box 1625 Idaho Falls, ID 83415 (United States); McCormack, Roger L. [Fluor Hanford Site, Richland, Wash. (United States); Sindelar, Robert; Spieker, Timothy [Savannah River National Laboratory, Savannah River Site Aiken, SC 29808 (United States)

2006-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" 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

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

SciTech Connect (OSTI)

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

NONE

1996-12-01T23:59:59.000Z

122

Spent Nuclear Fuel Project technical baseline document. Fiscal year 1995: Volume 1, Baseline description  

SciTech Connect (OSTI)

This document is a revision to WHC-SD-SNF-SD-002, and is issued to support the individual projects that make up the Spent Nuclear Fuel Project in the lower-tier functions, requirements, interfaces, and technical baseline items. It presents results of engineering analyses since Sept. 1994. The mission of the SNFP on the Hanford site is to provide safety, economic, environmentally sound management of Hanford SNF in a manner that stages it to final disposition. This particularly involves K Basin fuel, although other SNF is involved also.

Womack, J.C. [Westinghouse Hanford Co., Richland, WA (United States); Cramond, R. [TRW (United States); Paedon, R.J. [SAIC (United States)] [and others

1995-03-13T23:59:59.000Z

123

Realization of the German Concept for Interim Storage of Spent Nuclear Fuel - Current Situation and Prospects  

SciTech Connect (OSTI)

The German government has determined a phase out of nuclear power. With respect to the management of spent fuel it was decided to terminate transports to reprocessing plants by 2005 and to set up interim storage facilities on power plant sites. This paper gives an overview of the German concept for spent fuel management focused on the new on-site interim storage concept and the applied interim storage facilities. Since the end of the year 1998, the utilities have applied for permission of on-site interim storage in 13 storage facilities and 5 storage areas; one application for the interim storage facility Stade was withdrawn due to the planned final shut down of Stade nuclear power plant in autumn 2003. In 2001 and 2002, 3 on-site storage areas and 2 on-site storage facilities for spent fuel were licensed by the Federal Office for Radiation Protection (BfS). A main task in 2002 and 2003 has been the examination of the safety and security of the planned interim storage facilities and the verification of the licensing prerequisites. In the aftermath of September 11, 2001, BfS has also examined the attack with a big passenger airplane. Up to now, these aircraft crash analyses have been performed for three on-site interim storage facilities; the fundamental results will be presented. It is the objective of BfS to conclude the licensing procedures for the applied on-site interim storage facilities in 2003. With an assumed construction period for the storage buildings of about two years, the on-site interim storage facilities could then be available in the year 2005.

Thomauske, B. R.

2003-02-25T23:59:59.000Z

124

Electrochemical separation of aluminum from uranium for research reactor spent nuclear fuel applications.  

SciTech Connect (OSTI)

Researchers at Argonne National Laboratory (ANL) are developing an electrorefining process to treat aluminum-based spent nuclear fuel by electrochemically separating aluminum from uranium. The aluminum electrorefiner is modeled after the high-throughput electrorefiner developed at ANL. Aluminum is electrorefined, using a fluoride salt electrolyte, in a potential range of -0.1 V to -0.2 V, while uranium is electrorefined in a potential range of -0.3 V to -0.4 V; therefore, aluminum can be selectively separated electrochemically from uranium. A series of laboratory-scale experiments was performed to demonstrate the aluminum electrorefining concept. These experiments involved selecting an electrolyte (determining a suitable fluoride salt composition); selecting a crucible material for the electrochemical cell; optimizing the operating conditions; determining the effect of adding alkaline and rare earth elements to the electrolyte; and demonstrating the electrochemical separation of aluminum from uranium, using a U-Al-Si alloy as a simulant for aluminum-based spent nuclear fuel. Results of the laboratory-scale experiments indicate that aluminum can be selectively electrotransported from the anode to the cathode, while uranium remains in the anode basket.

Slater, S. A.; Willit, J. L.; Gay, E. C.; Chemical Engineering

1999-01-01T23:59:59.000Z

125

Incorporation of radionuclides in the alteration phases of spent nuclear fuel.  

SciTech Connect (OSTI)

Alteration may be expected for spent nuclear fuel exposed to groundwater under oxidizing conditions such as that which exist at the proposed nuclear waste repository at Yucca Mountain, Nevada. The actinide elements released during the corrosion of spent fuel may be incorporated into the structures of secondary U{sup 6+} phases. The incorporation of transuranics into the crystal structures of the alteration products may significantly decrease their mobility. A series of precipitation tests were conducted at 90 C to determine the potential incorporation of Ce{sup 4+} and Nd{sup 3+} (surrogates for Pu{sup 4+} and Am{sup 3+}, respectively) into uranyl phase. Dehydrated schoepite (UO{sub 3}{center_dot}0.8-1.0HP{sub 2}O) was produced by hydrolysis of a uranium oxyacetate solution containing either cerium or neodymium. ICP-MS analysis of the leachant, leachate, and solid phase reaction products which were dissolved in a HNO{sub 3} solution indicates that 26 ppm of Ce was incorporated into dehydrated schoepite. ICP-MS results from the Nd-doped tests indicate significant neodymium incorporation as well, however, the heterogeneous distribution of Nd in the solid phase noted during the AEM/EELS examination implies that neodymium may not incorporate into the structure of dehydrated schoepite.

Buck, E. C.; Kim, C.-W.; Wronkiewicz, D. J.

1999-08-25T23:59:59.000Z

126

The corrosion of aluminum-clad spent nuclear fuel in wet basin storage  

SciTech Connect (OSTI)

Large quantities of Defense related spent nuclear fuels are being stored in water basins around the United States. Under the non-proliferation policy, there has been no processing since the late 1980`s and these fuels are caught in the pipeline awaiting stabilization or other disposition. At the Savannah River Site, over 200 metric tons of aluminum clad fuel are being stored in four water filled basins. Some of this fuel has experienced visible pitting corrosion. An intensive effort is underway at SRS to understand the corrosion problems and to improve the basin storage conditions for extended storage requirements. Significant improvements have been accomplished during 1993-1996. This paper presents a discussion of the fundamentals of aluminum alloy corrosion as it pertains to the wet storage of spent nuclear fuel. It examines the effects of variables on corrosion in the storage environment and presents the results of corrosion surveillance testing activities at SRS, as well as discussions of fuel storage basins at other production sites of the Department of Energy.

Howell, J.P.; Burke, S.D.

1996-02-20T23:59:59.000Z

127

Concrete Shield Performance of the VSC-17 Spent Nuclear Fuel Cask  

SciTech Connect (OSTI)

In 2003, representatives from the Central Research Institute of Electric Power Industry (CRIEPI) requested development of a project with the objective of determining the performance of a concrete spent nuclear fuel storage cask. Radiation and environmental effects may cause chemical alteration of the concrete that could result in excessive cracking, spalling, and loss of compressive strength. The Idaho National Laboratory (INL) project team and CRIEPI representatives identified the Ventilated Storage Cask (VSC-17) spent nuclear fuel storage cask as a candidate to study cask performance, because it had been used to store fuel as part of a dry cask storage demonstration project for more than 15 years. The project involved investigating the properties of the concrete shield. INL performed a survey of the cask in the summers of 2003 and 2004. Preliminary cask evaluations performed in 2003 indicated that the cask has no visual degradation. However, a 4-5 mrem/hr step-change in the radiation levels about halfway up the cask and a localized hot spot beneath an upper air vent indicate that there may be variability in the density of the concrete or localized cracking. In 2005, INL and CRIEPI scientists performed additional surveys on the VSC-17 cask. This document summarizes the methods used on the VSC-17 to evaluate the cask for compressive strength, concrete cracking, concrete thickness, and temperature distribution.

Sheryl L. Morton; Philip L. Winston; Toshiari Saegusa; Koji Shirai; Akihiro Sasahara; Takatoshi Hattori

2006-04-01T23:59:59.000Z

128

Development of a Techno-Economic Model to Optimize DOE Spent Nuclear Fuel Disposition  

SciTech Connect (OSTI)

The National Spent Nuclear Fuel (NSNF) Program is evaluating final disposition of spent nuclear fuel (SNE) in the Department of Energy (DOE) complex. Final disposition of SNF may require that the fuel be treated to minimize material concerns. The treatments may range from electrometallurgical treatment (EMT) and chemical dissolution to engineering controls. Treatment options and treatment locations will depend on fuel type and location of the fuel. One of the first steps associated with selecting one or more sites for treating SNF in the DOE complex is to determine the cost of each option. An economic analysis will assist in determining which fuel treatment alternative attains the optimum disposition of SNF at the lowest possible cost to the government and the public. For this study, a set of questions was developed for the EMT process for fuels at several locations. The set of questions addresses all issues associated with design, construction, and operation of a production facility. A matrix table was developed to determine questions applicable to various fuel treatment options. A work breakdown structure (WBS) was developed to identify a treatment process and costs from initial design to shipment of treatment products to final disposition. Costs can be applied to determine the life cycle cost of each option. This technique can also be applied to other treatment techniques for treating SNF.

Ramer, R. J.; Plum, M. M.; Adams, J. P.; Dahl, C. A.

1998-02-01T23:59:59.000Z

129

Analysis of dose consequences arising from the release of spent nuclear fuel from dry storage casks.  

SciTech Connect (OSTI)

The resulting dose consequences from releases of spent nuclear fuel (SNF) residing in a dry storage casks are examined parametrically. The dose consequences are characterized by developing dose versus distance curves using simplified bounding assumptions. The dispersion calculations are performed using the MELCOR Accident Consequence Code System (MACCS2) code. Constant weather and generic system parameters were chosen to ensure that the results in this report are comparable with each other and to determine the relative impact on dose of each variable. Actual analyses of site releases would need to accommodate local weather and geographic data. These calculations assume a range of fuel burnups, release fractions (RFs), three exposure scenarios (2 hrs and evacuate, 2 hrs and shelter, and 24 hrs exposure), two meteorological conditions (D-4 and F-2), and three release heights (ground level - 1 meter (m), 10 m, and 100 m). This information was developed to support a policy paper being developed by U.S. Nuclear Regulatory Commission (NRC) staff on an independent spent fuel storage installation (ISFSI) and monitored retrievable storage installation (MRS) security rulemaking.

Durbin, Samuel G.; Morrow, Charles W.

2013-01-01T23:59:59.000Z

130

Final Report - Spent Nuclear Fuel Retrieval System Manipulator System Cold Validation Testing  

SciTech Connect (OSTI)

Manipulator system cold validation testing (CVT) was performed in support of the Fuel Retrieval System (FRS) Sub-Project, a subtask of the Spent Nuclear Fuel Project at the Hanford Site in Richland, Washington. The FRS will be used to retrieve and repackage K-Basin Spent Nuclear Fuel (SNF) currently stored in old K-Plant storage basins. The FRS is required to retrieve full fuel canisters from the basin; clean the fuel elements inside the canister to remove excessive uranium corrosion products (or sludge); remove the contents from the canisters; and sort the resulting debris, scrap, and fuel for repackaging. The fuel elements and scrap will be collected in fuel storage and scrap baskets in preparation for loading into a multi canister overpack (MCO), while the debris is loaded into a debris bin and disposed of as solid waste. The FRS is composed of three major subsystems. The Manipulator Subsystem provides remote handling of fuel, scrap, and debris; the In-Pool Equipment subsystem performs cleaning of fuel and provides a work surface for handling materials; and the Remote Viewing Subsystem provides for remote viewing of the work area by operators. There are two complete and identical FRS systems, one to be installed in the K-West basin and one to be installed in the K-East basin. Another partial system will be installed in a cold test facility to provide for operator training.

D.R. Jackson; G.R. Kiebel

1999-08-24T23:59:59.000Z

131

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

SciTech Connect (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 challengeto 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

132

Experimental Investigation of Burnup Credit for Safe Transport, Storage, and Disposal of Spent Nuclear Fuel  

SciTech Connect (OSTI)

This report describes criticality benchmark experiments containing rhodium that were conducted as part of a Department of Energy Nuclear Energy Research Initiative project. Rhodium is an important fission product absorber. A capability to perform critical experiments with low-enriched uranium fuel was established as part of the project. Ten critical experiments, some containing rhodium and others without, were conducted. The experiments were performed in such a way that the effects of the rhodium could be accurately isolated. The use of the experimental results to test neutronics codes is demonstrated by example for two Monte Carlo codes. These comparisons indicate that the codes predict the behavior of the rhodium in the critical systems within the experimental uncertainties. The results from this project, coupled with the results of follow-on experiments that investigate other fission products, can be used to quantify and reduce the conservatism of spent nuclear fuel safety analyses while still providing the necessary level of safety.

Harms, Gary A.; Helmick, Paul H.; Ford, John T.; Walker, Sharon A.; Berry, Donald T.; Pickard, Paul S.

2004-04-01T23:59:59.000Z

133

Enrico Fermi Fast Reactor Spent Nuclear Fuel Criticality Calculations: Degraded Mode  

SciTech Connect (OSTI)

The objective of this calculation is to characterize the nuclear criticality safety concerns associated with the codisposal of the Department of Energy's (DOE) Enrico Fermi (EF) Spent Nuclear Fuel (SNF) in a 5-Defense High-Level Waste (5-DHLW) Waste Package (WP) and placed in a Monitored Geologic Repository (MGR). The scope of this calculation is limited to the determination of the effective neutron multiplication factor (k{sub eff}) for the degraded mode internal configurations of the codisposal WP. The results of this calculation and those of Ref. 8 will be used to evaluate criticality issues and support the analysis that will be performed to demonstrate the viability of the codisposal concept for the Monitored Geologic Repository.

D.R. Moscalu; L. Angers; J. Monroe-Rammsey; H.R. Radulesca

2000-07-21T23:59:59.000Z

134

Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel  

SciTech Connect (OSTI)

The purpose of this calculation report, Range of Applicability and Bias Determination for Postclosure Criticality of Commercial Spent Nuclear Fuel, is to validate the computational method used to perform postclosure criticality calculations. The validation process applies the criticality analysis methodology approach documented in Section 3.5 of the Disposal Criticality Analysis Methodology Topical Report. The application systems for this validation consist of waste packages containing transport, aging, and disposal canisters (TAD) loaded with commercial spent nuclear fuel (CSNF) of varying assembly types, initial enrichments, and burnup values that are expected from the waste stream and of varying degree of internal component degradation that may occur over the 10,000-year regulatory time period. The criticality computational tool being evaluated is the general-purpose Monte Carlo N-Particle (MCNP) transport code. The nuclear cross-section data distributed with MCNP 5.1.40 and used to model the various physical processes are based primarily on the Evaluated Nuclear Data File/B Version VI (ENDF/B-VI) library. Criticality calculation bias and bias uncertainty and lower bound tolerance limit (LBTL) functions for CSNF waste packages are determined based on the guidance in ANSI/ANS 8.1-1998 (Ref. 4) and ANSI/ANS 8.17-2004 (Ref. 5), as described in Section 3.5.3 of Ref. 1. The development of this report is consistent with Test Plan for: Range of Applicability and Bias Determination for Postclosure Criticality. This calculation report has been developed in support of licensing activities for the proposed repository at Yucca Mountain, Nevada, and the results of the calculation may be used in the criticality evaluation for CSNF waste packages based on a conceptual TAD canister.

Radulescu, Georgeta [ORNL; Mueller, Don [ORNL; Goluoglu, Sedat [ORNL; Hollenbach, Daniel F [ORNL; Fox, Patricia B [ORNL

2007-10-01T23:59:59.000Z

135

1996 environmental monitoring report for the Naval Reactors Facility  

SciTech Connect (OSTI)

The results of the radiological and nonradiological environmental monitoring programs for 1996 at the Naval Reactors Facility (NRF) are presented in this report. The NRF is located on the Idaho National Engineering and Environmental Laboratory and contains three naval reactor prototypes and the Expended Core Facility, which examines developmental nuclear fuel material samples, spent naval fuel, and irradiated reactor plant components/materials. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE).

NONE

1996-12-31T23:59:59.000Z

136

DEVELOPMENT OF METHODOLOGY AND FIELD DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE  

SciTech Connect (OSTI)

This project developed methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of the fuel storage medium and determine the oxide thickness on the spent fuel basin materials. The overall objective of this project was to determine the amount of time fuel has spent in a storage basin to determine if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations. This project developed and validated forensic tools that can be used to predict the age and condition of spent nuclear fuels stored in liquid basins based on key physical, chemical and microbiological basin characteristics. Key parameters were identified based on a literature review, the parameters were used to design test cells for corrosion analyses, tools were purchased to analyze the key parameters, and these were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The key parameters identified in the literature review included chloride concentration, conductivity, and total organic carbon level. Focus was also placed on aluminum based cladding because of their application to weapons production. The literature review was helpful in identifying important parameters, but relationships between these parameters and corrosion rates were not available. Bench scale test systems were designed, operated, harvested, and analyzed to determine corrosion relationships between water parameters and water conditions, chemistry and microbiological conditions. The data from the bench scale system indicated that corrosion rates were dependent on total organic carbon levels and chloride concentrations. The highest corrosion rates were observed in test cells amended with sediment, a large microbial inoculum and an organic carbon source. A complete characterization test kit was field tested to characterize the SRS L-Area spent fuel basin. The sampling kit consisted of a TOC analyzer, a YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization was done over a two day period in June 2011, and confirmed that the L Area basin is a well operated facility with low corrosion potential.

Berry, T.; Milliken, C.; Martinez-Rodriguez, M.; Hathcock, D.; Heitkamp, M.

2012-06-04T23:59:59.000Z

137

Naval Civil Engineering Laboratory  

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

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138

Development of Enriched Borated Aluminum Alloy for Basket Material of Cask for Spent Nuclear Fuel  

SciTech Connect (OSTI)

Concrete cask system is focused as the candidate one for spent fuel dry storage facilities from economic potential in Japan. Concrete cask consists of a concrete storage cask and a steel canister. A canister containing nuclear spent fuel is shipped by a transportation cask from a nuclear power plant to an interim storage facility. The canister is transferred from the transportation cask to a storage cask by a transfer cask in the storage facility. IHI has developed a concrete cask horizontal transfer system. This transfer system indicates that a canister is transferred to a storage cask horizontally. This transfer system has a merit against canister drop accident in transfer operation, i.e. spent fuel assemblies can be kept safe during the transfer operation. There are guide rails inside of the concrete cask, and the canister is installed into the storage cask with sliding on the rails. To develop the horizontal transfer system, IHI carried out a heat load test and numerical analyses by CFD. Heat load experiment was carried out by using a full-scale prototype canister, storage cask and transfer vessel. The decay heat was simulated by an electric heater installed in the canister. Assuming high burn-up spent fuel storage, heat generation was set between 20.0 kW and 25.0 kW. This experiment was focused on the concrete temperature distribution. We confirmed that the maximum concrete temperature in transfer operation period was lower than 40 deg. C (Heat generation 22.5 kW). Moreover we confirmed the maximum concrete temperature passed 24 hours with horizontal orientation was below 90 deg. C (Heat generation 22.5 kW). We analyzed the thermal performance of the concrete cask with horizontal transfer condition and normal storage condition. Thermal analyses for horizontal transfer operation were carried out based on the experimental conditions. The tendency of the analytical results was in good agreement with experimental results. The purpose of vertical thermal analysis was to estimate the concrete temperature increase in the case a canister contacts with guide rails in normal storage. It has a possibility that a canister contacts with guide rails during storage period after concrete cask is upended from transfer operation. The temperature increase due to this contact was calculated 5 deg. C at small local area. This result implies that the affect of the contact is very small. This paper addresses that the storage cask concrete is kept its integrity in transfer operation period and normal storage period. (authors)

Mikio Sakai; Tadatsugu Sakaya; Hiroaki Fujiwara; Akira Sakai [Ishikawajima-Harima Heavy Industries Company Ltd., 1 Shin-Nakaharacho, Isogoku, Yokohama 235-8501 (Japan)

2002-07-01T23:59:59.000Z

139

Standardized DOE Spent Nuclear Fuel Canister and Transportation System for Shipping to the National Repository  

SciTech Connect (OSTI)

The U.S.Department of Energys (DOE) National Spent Nuclear Fuel Program (NSNFP), located at the Idaho National Engineering and Environmental Laboratory (INEEL), has been chartered with the responsibility for developing spent nuclear fuel (SNF) standardized canisters and a transportation cask system for shipping DOE SNF to the national repository. The mandate for this development is outlined in the Memorandum of Agreement for Acceptance of Department of Energy Spent Nuclear Fuel and High-Level Radioactive Waste that states, EM shall design and fabricate DOE SNF canisters for shipment to RW. (1) It also states, EM shall be responsible for the design, NRC certification, and fabrication of the transportation cask system for DOE SNF canisters or bare DOE SNF in accordance with 10 CFR Part 71. (2) In fulfillment of these requirements, the NSNFP has developed four SNF standardized canister configurations and has conceptually designed a versatile transportation cask system for shipping the canisters to the national repository.1 The standardized canister sizes were derived from the national repository waste package design for co-disposal of SNF with high-level waste (HLW). One SNF canister can be placed in the center of the waste package or one can be placed in one of five radial positions, replacing a HLW canister. The internal cavity of the transportation cask was derived using the same logic, matching the size of the internal cavity of the waste package. The size of the internal cavity for the transportation cask allows the shipment of multiple canister configurations with the application of a removable basket design. The standardized canisters have been designed to be loaded with DOE SNF, placed into interim storage, shipped to the national repository, and placed in a waste package without having to be reopened. Significant testing has been completed that clearly demonstrates that the standardized canisters can safely achieve their intended design goals. The transportation cask system will include all of the standard design features, with the addition of dual containment for the shipment of failed fuel. The transportation cask system will also meet the rigorous licensing requirements of the Nuclear Regulatory Commission (NRC) to ensure that the design and the methods of fabrication employed will result in a shipping cask that will safely contain the radioactive materials under all credible accident scenarios. The standardization of the SNF canisters and the versatile design of the transportation cask system will eliminate a proliferation of designs and simplify the operations at the user sites and the national repository.

Pincock, David Lynn; Morton, Dana Keith; Lengyel, Arpad Leslie

2001-02-01T23:59:59.000Z

140

Nuclear Resonance Fluorescence to Measure Plutonium Mass in Spent Nuclear Fuel  

E-Print Network [OSTI]

Level Scheme Library Depleted Uranium Evaluated Nuclear Datafuel and the 238 U in depleted uranium (DU) was used as a

Ludewigt, Bernhard A

2011-01-01T23:59:59.000Z

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


141

Spent nuclear fuel as a waste form for geologic disposal: Assessment and recommendations on data and modeling needs  

SciTech Connect (OSTI)

This study assesses the status of knowledge pertinent to evaluating the behavior of spent nuclear fuel as a waste form in geologic disposal systems and provides background information that can be used by the DOE to address the information needs that pertain to compliance with applicable standards and regulations. To achieve this objective, applicable federal regulations were reviewed, expected disposal environments were described, the status of spent-fuel modeling was summarized, and information regarding the characteristics and behavior of spent fuel was compiled. This compiled information was then evaluated from a performance modeling perspective to identify further information needs. A number of recommendations were made concerning information still needed to enhance understanding of spent-fuel behavior as a waste form in geologic repositories. 335 refs., 22 figs., 44 tabs.

Van Luik, A.E.; Apted, M.J.; Bailey, W.J.; Haberman, J.H.; Shade, J.S.; Guenther, R.E.; Serne, R.J.; Gilbert, E.R.; Peters, R.; Williford, R.E.

1987-09-01T23:59:59.000Z

142

Report on the Savannah River Site aluminum-based spent nuclear fuel alternatives cost study  

SciTech Connect (OSTI)

Initial estimates of costs for the interim management and disposal of aluminum-based spent nuclear fuel (SNF) were developed during preparation of the Environmental Impact Statement (EIS) on the Nuclear Weapons Nonproliferation Policy Concerning Foreign Research Reactor Spent Nuclear Fuel. The Task Team evaluated multiple alternatives, assessing programmatic, technical, and schedule risks, and generated life-cycle cost projections for each alternative. The eight technology alternatives evaluated were: direct co-disposal; melt and dilute; reprocessing; press and dilute; glass material oxidation dissolution system (GMODS); electrometallurgical treatment; dissolve and vitrify; and plasma arc. In followup to the Business Plan that was developed to look at SNF dry storage, WSRC prepared an addendum to the cost study. This addendum estimated the costs for the modification and use of an existing (105L) reactor facility versus a greenfield approach for new facilities (for the Direct Co-Disposal and Melt and Dilute alternatives). WSRC assessed the impacts of a delay in reprocessing due to the potential reservation of H-Canyon for other missions (i.e., down blending HEU for commercial use or the conversion of plutonium to either MOX fuel or an immobilized repository disposal form). This report presents the relevant results from these WSRC cost studies, consistent with the most recent project policy, technology implementation, canyon utilization, and inventory assumptions. As this is a summary report, detailed information on the technical alternatives or the cost assumptions raised in each of the above-mentioned cost studies is not provided. A comparison table that briefly describes the bases used for the WSRC analyses is included as Appendix A.

NONE

1998-12-01T23:59:59.000Z

143

Fusion option to dispose of spent nuclear fuel and transuranic elements  

SciTech Connect (OSTI)

The fusion option is examined to solve the disposition problems of the spent nuclear fuel and the transuranic elements. The analysis of this report shows that the top rated solution, the elimination of the transuranic elements and the long-lived fission products, can be achieved in a fusion reactor. A 167 MW of fusion power from a D-T plasma for sixty years with an availability factor of 0.75 can transmute all the transuranic elements and the long-lived fission products of the 70,000 tons of the US inventory of spent nuclear fuel generated up to the year 2015. The operating time can be reduced to thirty years with use of 334 MW of fusion power, a system study is needed to define the optimum time. In addition, the fusion solution eliminates the need for a geological repository site, which is a major advantage. Meanwhile, such utilization of the fusion power will provide an excellent opportunity to develop fusion energy for the future. Fusion blankets with a liquid carrier for the transuranic elements can achieve a transmutation rate for the transuranic elements up to 80 kg/MW.y of fusion power with k{sub eff} of 0.98. In addition, the liquid blankets have several advantages relative to the other blanket options. The energy from this transmutation is utilized to produce revenue for the system. Molten salt (Flibe) and lithium-lead eutectic are identified as the most promising liquids for this application, both materials are under development for future fusion blanket concepts. The Flibe molten salt with transuranic elements was developed and used successfully as nuclear fuel for the molten salt breeder reactor in the 1960's.

Gohar, Y.

2000-02-10T23:59:59.000Z

144

Evaluation of fission product worth margins in PWR spent nuclear fuel burnup credit calculations.  

SciTech Connect (OSTI)

Current criticality safety calculations for the transportation of irradiated LWR fuel make the very conservative assumption that the fuel is fresh. This results in a very substantial overprediction of the actual k{sub eff} of the transportation casks; in certain cases, this decreases the amount of spent fuel which can be loaded in a cask, and increases the cost of transporting the spent fuel to the repository. Accounting for the change of reactivity due to fuel depletion is usually referred to as ''burnup credit.'' The US DOE is currently funding a program aimed at establishing an actinide only burnup credit methodology (in this case, the calculated reactivity takes into account the buildup or depletion of a limited number of actinides). This work is undergoing NRC review. While this methodology is being validated on a significant experimental basis, it implicitly relies on additional margins: in particular, the absorption of neutrons by certain actinides and by all fission products is not taken into account. This provides an important additional margin and helps guarantee that the methodology is conservative provided these neglected absorption are known with reasonable accuracy. This report establishes the accuracy of fission product absorption rate calculations: (1) the analysis of European fission product worth experiments demonstrates that fission product cross-sections available in the US provide very good predictions of fission product worth; (2) this is confirmed by a direct comparison of European and US cross section evaluations; (3) accuracy of Spent Nuclear Fuel (SNF) fission product content predictions is established in a recent ORNL report where several SNF isotopic assays are analyzed; and (4) these data are then combined to establish in a conservative manner the fraction of the predicted total fission product absorption which can be guaranteed based on available experimental data.

Blomquist, R.N.; Finck, P.J.; Jammes, C.; Stenberg, C.G.

1999-02-17T23:59:59.000Z

145

Fate of Noble Metals during the Pyroprocessing of Spent Nuclear Fuel  

SciTech Connect (OSTI)

During the pyroprocessing of spent nuclear fuel by electrochemical techniques, fission products are separated as the fuel is oxidized at the anode and refined uranium is deposited at the cathode. Those fission products that are oxidized into the molten salt electrolyte are considered active metals while those that do not react are considered noble metals. The primary noble metals encountered during pyroprocessing are molybdenum, zirconium, ruthenium, rhodium, palladium, and technetium. Pyroprocessing of spent fuel to date has involved two distinctly different electrorefiner designs, in particular the anode to cathode configuration. For one electrorefiner, the anode and cathode collector are horizontally displaced such that uranium is transported across the electrolyte medium. As expected, the noble metal removal from the uranium during refining is very high, typically in excess of 99%. For the other electrorefiner, the anode and cathode collector are vertically collocated to maximize uranium throughput. This arrangement results in significantly less noble metals removal from the uranium during refining, typically no better than 20%. In addition to electrorefiner design, operating parameters can also influence the retention of noble metals, albeit at the cost of uranium recovery. Experiments performed to date have shown that as much as 100% of the noble metals can be retained by the cladding hulls while affecting the uranium recovery by only 6%. However, it is likely that commercial pyroprocessing of spent fuel will require the uranium recovery to be much closer to 100%. The above mentioned design and operational issues will likely be driven by the effects of noble metal contamination on fuel fabrication and performance. These effects will be presented in terms of thermal properties (expansion, conductivity, and fusion) and radioactivity considerations. Ultimately, the incorporation of minor amounts of noble metals from pyroprocessing into fast reactor metallic fuel will be shown to be of no consequence to reactor performance.

B.R. Westphal; D. Vaden; S.X. Li; G.L. Fredrickson; R.D. Mariani

2009-09-01T23:59:59.000Z

146

Measurement of Atmospheric Sea Salt Concentration in the Dry Storage Facility of the Spent Nuclear Fuel  

SciTech Connect (OSTI)

Spent nuclear fuel coming from a Japanese nuclear power plant is stored in the interim storage facility before reprocessing. There are two types of the storage methods which are wet and dry type. In Japan, it is anticipated that the dry storage facility will increase compared with the wet type facility. The dry interim storage facility using the metal cask has been operated in Japan. In another dry storage technology, there is a concrete overpack. Especially in USA, a lot of concrete overpacks are used for the dry interim storage. In Japan, for the concrete cask, the codes of the Japan Society of Mechanical Engineers and the governmental technical guidelines are prepared for the realization of the interim storage as well as the code for the metal cask. But the interim storage using the concrete overpack has not been in progress because the evaluation on the stress corrosion cracking (SCC) of the canister is not sufficient. Japanese interim storage facilities would be constructed near the seashore. The metal casks and concrete overpacks are stored in the storage building in Japan. On the other hand, in USA they are stored outside. It is necessary to remove the decay heat of the spent nuclear fuel in the cask from the storage building. Generally, the heat is removed by natural cooling in the dry storage facility. Air including the sea salt particles goes into the dry storage facility. Concerning the concrete overpack, air goes into the cask body and cools the canister. Air goes along the canister surface and is in contact with the surface directly. In this case, the sea salt in the air attaches to the surface and then there is the concern about the occurrence of the SCC. For the concrete overpack, the canister including the spent fuel is sealed by the welding. The loss of sealability caused by the SCC has to be avoided. To evaluate the SCC for the canister, it is necessary to make clear the amount of the sea salt particles coming into the storage building and the concentration on the canister. In present, the evaluation on that point is not sufficient. In this study, the concentration of the sea salt particles in the air and on the surface of the storage facility are measured inside and outside of the building. For the measurement, two sites of the dry storage facility using the metal cask are chosen. This data is applicable for the evaluation on the SCC of the canister to realize the interim storage using the concrete overpack. (authors)

Masumi Wataru; Hisashi Kato; Satoshi Kudo; Naoko Oshima; Koji Wada [Central Research Institute of Electric Power Industry - CRIEPI (Japan); Hirofumi Narutaki [Electric Power Engineering Systems Co. Ltd. (Japan)

2006-07-01T23:59:59.000Z

147

Radiation Exposures Associated with Shipments of Foreign Research Reactor Spent Nuclear Fuel  

SciTech Connect (OSTI)

Experience has shown that the analyses of marine transport of spent fuel in the Environmental Impact Statement (EIS) were conservative. It is anticipated that for most shipments. The external dose rate for the loaded transportation cask will be more in line with recent shipments. At the radiation levels associated with these shipments, we would not expect any personnel to exceed radiation exposure limits for the public. Package dose rates usually well below the regulatory limits and personnel work practices following ALARA principles are keeping human exposures to minimal levels. However, the potential for Mure shipments with external dose rates closer to the exclusive-use regulatory limit suggests that DOE should continue to provide a means to assure that individual crew members do not receive doses in excess of the public dose limits. As a minimum, the program will monitor cask dose rates and continue to implement administrative procedures that will maintain records of the dose rates associated with each shipment, the vessel used, and the crew list for the vessel. DOE will continue to include a clause in the contract for shipment of the foreign research reactor spent nuclear fuel requiring that the Mitigation Action Plan be followed.

MASSEY,CHARLES D.; MESSICK,C.E.; MUSTIN,T.

1999-11-01T23:59:59.000Z

148

Spent Fuel Test-Climax: An evaluation of the technical feasibility of geologic storage of spent nuclear fuel in granite: Final report  

SciTech Connect (OSTI)

In the Climax stock granite on the Nevada Test Site, eleven canisters of spent nuclear reactor fuel were emplaced, and six electrical simulators were energized. When test data indicated that the test objectives were met during the 3-year storage phase, the spent-fuel canisters were retrieved and the thermal sources were de-energized. The project demonstrated the feasibility of packaging, transporting, storing, and retrieving highly radioactive fuel assemblies in a safe and reliable manner. In addition to emplacement and retrieval operations, three exchanges of spent-fuel assemblies between the SFT-C and a surface storage facility, conducted during the storage phase, furthered this demonstration. The test led to development of a technical measurements program. To meet these objectives, nearly 1000 instruments and a computer-based data acquisition system were deployed. Geotechnical, seismological, and test status data were recorded on a continuing basis for the three-year storage phase and six-month monitored cool-down of the test. This report summarizes the engineering and scientific endeavors which led to successful design and execution of the test. The design, fabrication, and construction of all facilities and handling systems are discussed, in the context of test objectives and a safety assessment. The discussion progresses from site characterization and experiment design through data acquisition and analysis of test data in the context of design calculations. 117 refs., 52 figs., 81 tabs.

Patrick, W.C. (comp.)

1986-03-30T23:59:59.000Z

149

Calculation Method for the Projection of Future Spent Nuclear Fuel Discharges  

SciTech Connect (OSTI)

This report describes the calculation method developed for the projection of future utility spent nuclear fuel (SNF) discharges in regard to their timing, quantity, burnup, and initial enrichment. This projection method complements the utility-supplied RW-859 data on historic discharges and short-term projections of SNF discharges by providing long-term projections that complete the total life cycle of discharges for each of the current U.S. nuclear power reactors. The method was initially developed in mid-1999 to update the SNF discharge projection associated with the 1995 RW-859 utility survey (CRWMS M&O 1996). and was further developed as described in Rev. 00 of this report (CRWMS M&O 2001a). Primary input to the projection of SNF discharges is the utility projection of the next five discharges from each nuclear unit, which is provided via the revised final version of the Energy Information Administration (EIA) 1998 RW-859 utility survey (EIA 2000a). The projection calculation method is implemented via a set of Excel 97 spreadsheets. These calculations provide the interface between receipt of the utility five-discharge projections that are provided in the RW-859 survey, and the delivery of projected life-cycle SNF discharge quantities and characteristics in the format requisite for performing logistics analysis to support design of the Civilian Radioactive Waste Management System (CRWMS). Calculation method improvements described in this report include the addition of a reactor-specific maximum enrichment-based discharge burnup limit. This limit is the consequence of the enrichment limit, currently 5 percent. which is imposed as a Nuclear Regulatory Commission (NRC) license condition on nuclear fuel fabrication plants. In addition, the calculation method now includes the capability for projecting future nuclear plant power upratings, consistent with many such recent plant uprates and the prospect of additional future uprates. Finally. this report summarizes the results of the 2002 Reference SNF Discharge Projection.

B. McLeod

2002-02-28T23:59:59.000Z

150

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel  

SciTech Connect (OSTI)

This report presents the technical details of RISIUND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, semiinteractive program that can be run on an IBM or equivalent personal computer. The program language is FORTRAN-77. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incidentfree models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionudide inventory and dose conversion factors.

Yuan, Y.C. [Square Y, Orchard Park, NY (United States); Chen, S.Y.; LePoire, D.J. [Argonne National Lab., IL (United States). Environmental Assessment and Information Sciences Div.; Rothman, R. [USDOE Idaho Field Office, Idaho Falls, ID (United States)

1993-02-01T23:59:59.000Z

151

AIR SHIPMENT OF SPENT NUCLEAR FUEL FROM THE BUDAPEST RESEARCH REACTOR  

SciTech Connect (OSTI)

The shipment of spent nuclear fuel is usually done by a combination of rail, road or sea, as the high activity of the SNF needs heavy shielding. Air shipment has advantages, e.g. it is much faster than any other shipment and therefore minimizes the transit time as well as attention of the public. Up to now only very few and very special SNF shipments were done by air, as the available container (TUK6) had a very limited capacity. Recently Sosny developed a Type C overpack, the TUK-145/C, compliant with IAEA Standard TS-R-1 for the VPVR/M type Skoda container. The TUK-145/C was first used in Vietnam in July 2013 for a single cask. In October and November 2013 a total of six casks were successfully shipped from Hungary in three air shipments using the TUK-145/C. The present paper describes the details of these shipments and formulates the lessons learned.

Dewes, J.

2014-02-24T23:59:59.000Z

152

Use of Integrated Decay Heat Limits to Facilitate Spent Nuclear Fuel Loading to Yucca Mountain  

SciTech Connect (OSTI)

As an alternative to the use of the linear loading or areal power density (APD) concept, using integrated decay heat limits based on the use of mountain-scale heat transfer analysis is considered to represent the thermal impact from the deposited spent nuclear fuel (SNF) to the Yucca Mountain repository. Two different integrated decay heat limits were derived to represent both the short-term (up to 50 years from the time of repository closure) and the long-term decay heat effect (up to 1500 years from the time of repository closure). The derived limits were found to appropriately represent the drift wall temperature limit (200 deg. C) and the midway between adjacent drifts temperature limit (96 deg. C) as long as used fuel is uniformly loaded into the mountain. These limits can be a useful practical guide to facilitate the loading of used fuel into Yucca Mountain. (authors)

Li, Jun; Yim, Man-Sung; McNelis, David [Department of Nuclear Engineering, North Carolina State University (United States); Piet, Steven [Idaho National Laboratory (United States)

2007-07-01T23:59:59.000Z

153

Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual  

SciTech Connect (OSTI)

This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-SD-SNF-SAR-002, Safety Analysis Report for the Cold Vacuum Drying Facility, Phase 2, Supporting Installation of the Processing Systems (Garvin 1998) and, the HNF-SD-SNF-DRD-002, 1997, Cold Vacuum Drying Facility Design Requirements, Rev. 3a. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence, and has been developed for the spent nuclear fuel project (SNFP) Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

IRWIN, J.J.

2000-02-03T23:59:59.000Z

154

Spent Nuclear Fuel (SNF) Cold Vacuum Drying (CVD) Facility Operations Manual  

SciTech Connect (OSTI)

This document provides the Operations Manual for the Cold Vacuum Drying Facility (CVDF). The Manual was developed in conjunction with HNF-553, Spent Nuclear Fuel Project Final Safety Analysis Report Annex B--Cold Vacuum Drying Facility. The HNF-SD-SNF-DRD-002, 1999, Cold Vacuum Drying Facility Design Requirements, Rev. 4, and the CVDF Final Design Report. The Operations Manual contains general descriptions of all the process, safety and facility systems in the CVDF, a general CVD operations sequence and references to the CVDF System Design Descriptions (SDDs). This manual has been developed for the SNFP Operations Organization and shall be updated, expanded, and revised in accordance with future design, construction and startup phases of the CVDF until the CVDF final ORR is approved.

IRWIN, J.J.

1999-07-02T23:59:59.000Z

155

Radiation Imaging of Dry-Storage Casks for Spent Nuclear Fuel  

SciTech Connect (OSTI)

The authors report the results of a measurement campaign conducted on six dry-storage, spent-nuclear-fuel storage casks at the Idaho National Laboratory. a gamma-ray imager, a thermal-neutron imager and a Ge-spectrometer were used to collect data. The campaign was conducted to examine the feasibility of using cask radiation signatures as unique identifiers for individual casks as part of a safeguards regime. The results clearly show different morphologies for the various cask types although the signatures are deemed insufficient to uniquely identify individual casks of the same type. Based on results with the Ge-spectrometer and differences between thermal neutron images and neutron-dose meters, this result is attributed to the limitations of the extant imagers used, rather than of the basic concept.

Ziock, K; Caffrey, G; Lebrun, A; Forman, L; Vanier, P; Wharton, J

2005-11-08T23:59:59.000Z

156

Management Of Hanford KW Basin Knockout Pot Sludge As Spent Nuclear Fuel  

SciTech Connect (OSTI)

CH2M HILL Plateau Remediation Company (CHPRC) and AREVA Federal Services, LLC (AFS) have been working collaboratively to develop and deploy technologies to remove, transport, and interim store remote-handled sludge from the 10S-K West Reactor Fuel Storage Basin on the U.S. Department of Energy (DOE) Hanford Site near Richland, WA, USA. Two disposal paths exist for the different types of sludge found in the K West (KW) Basin. One path is to be managed as Spent Nuclear Fuel (SNF) with eventual disposal at an SNF at a yet to be licensed repository. The second path will be disposed as remote-handled transuranic (RH-TRU) waste at the Waste Isolation Pilot Plant (WIPP) in Carlsbad, NM. This paper describes the systems developed and executed by the Knockout Pot (KOP) Disposition Subproject for processing and interim storage of the sludge managed as SNF, (i.e., KOP material).

Raymond, R. E. [CH2M HIll Plateau Remediation Company, Richland, WA (United States); Evans, K. M. [AREVA, Avignon (France)

2012-10-22T23:59:59.000Z

157

Management of spent nuclear fuel on the Oak Ridge Reservation, Oak Ridge, Tennessee: Environmental assessment  

SciTech Connect (OSTI)

On June 1, 1995, DOE issued a Record of Decision [60 Federal Register 28680] for the Department-wide management of spent nuclear fuel (SNF); regionalized storage of SNF by fuel type was selected as the preferred alternative. The proposed action evaluated in this environmental assessment is the management of SNF on the Oak Ridge Reservation (ORR) to implement this preferred alternative of regional storage. SNF would be retrieved from storage, transferred to a hot cell if segregation by fuel type and/or repackaging is required, loaded into casks, and shipped to off-site storage. The proposed action would also include construction and operation of a dry cask SNF storage facility on ORR, in case of inadequate SNF storage. Action is needed to enable DOE to continue operation of the High Flux Isotope Reactor, which generates SNF. This report addresses environmental impacts.

NONE

1996-02-01T23:59:59.000Z

158

Container for reprocessing and permanent storage of spent nuclear fuel assemblies  

DOE Patents [OSTI]

A single canister process container is described for reprocessing and permanent storage of spent nuclear fuel assemblies comprising zirconium-based cladding and fuel, which process container comprises a collapsible container, having side walls that are made of a high temperature alloy and an array of collapsible support means wherein the container is capable of withstanding temperature necessary to oxidize the zirconium-based cladding and having sufficient ductility to maintain integrity when collapsed under pressure. The support means is also capable of maintaining its integrity at a temperature necessary to oxidize the zirconium-based cladding. The process container also has means to introduce and remove fluids to and from the container. 10 figs.

Forsberg, C.W.

1992-03-24T23:59:59.000Z

159

Overview of Requirements for Using Overweight Vehicles to Ship Spent Nuclear Fuel  

SciTech Connect (OSTI)

The U.S. Department of Energy's (DOE's) Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada, considered a range of options for transportation. In evaluating the impacts of the mostly-legal weight truck scenario, DOE assumed that some shipments would use overweight trucks. The use of overweight trucks is also considered in the Draft Supplemental Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada, issued for public comment in Fall 2007. With the exception of permit requirements and operating restrictions, the vehicles for overweight shipments would be similar to legal-weight truck shipments but might weigh as much as 52,200 kilograms (115,000 pounds). The use of overweight trucks was determined to be acceptable for the Office of Civilian Radioactive Waste Management (OCRWM) Program because the payload is not divisible and the packaging alone may make shipments overweight. Overweight truck shipments are common, and states routinely issue overweight permits, some for vehicles with a gross vehicle weight up to 58,500 kilograms (129,000 pounds). This paper will present an overview of state overweight truck permitting policies and national and regional approaches to promote safety and uniformity. In conclusion: Overweight truck shipments are made routinely by carriers throughout the country. State permits are obtained by the carriers or by companies that provide permitting services to the carriers. While varying state permit restrictions may add complexity to OCRWM's planning activities, the well-established experience of commercial carriers and efforts to bring uniformity to the permitting process should allow the overweight shipment of SNF to be a viable option. (authors)

Thrower, A.W. [U.S. Department of Energy, Office of Civilian Radioactive Waste Management, Washington, DC (United States); Offner, J. [Booz Allen Hamilton, Washington, DC (United States); Bolton, P. [Booz Allen Hamilton, Santa Fe, NM (United States)

2008-07-01T23:59:59.000Z

160

Spent Nuclear Fuel Dry Transfer System Cold Demonstration Project Final Report  

SciTech Connect (OSTI)

The spent nuclear fuel dry transfer system (DTS) provides an interface between large and small casks and between storage-only and transportation casks. It permits decommissioning of reactor pools after shutdown and allows the use of large storage-only casks for temporary onsite storage of spent nuclear fuel irrespective of reactor or fuel handling limitations at a reactor site. A cold demonstration of the DTS prototype was initiated in August 1996 at the Idaho National Engineering and Environmental Laboratory (INEEL). The major components demonstrated included the fuel assembly handling subsystem, the shield plug/lid handling subsystem, the cask interface subsystem, the demonstration control subsystem, a support frame, and a closed circuit television and lighting system. The demonstration included a complete series of DTS operations from source cask receipt and opening through fuel transfer and closure of the receiving cask. The demonstration included both normal operations and recovery from off-normal events. It was designed to challenge the system to determine whether there were any activities that could be made to jeopardize the activities of another function or its safety. All known interlocks were challenged. The equipment ran smoothly and functioned as designed. A few "bugs" were corrected. Prior to completion of the demonstration testing, a number of DTS prototype systems were modified to apply lessons learned to date. Additional testing was performed to validate the modifications. In general, all the equipment worked exceptionally well. The demonstration also helped confirm cost estimates that had been made at several points in the development of the system.

Christensen, Max R; McKinnon, M. A.

1999-12-01T23:59:59.000Z

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161

DOE Spent Nuclear Fuel Information In Support of TSPA-VA  

SciTech Connect (OSTI)

RW has started the viability assessment (VA) effort to determine the feasibility of Yucca Mountain as the first geologic repository for spent nuclear fuel (SNF) and high-level waste. One component of the viability assessment will be a total system performance assessment (TSPA), based on the design concept and the scientific data and analysis available, describing the repository's probable behavior relative to the overall system performance standards. Thus, all the data collected from the Exploratory Studies Facility to-date have been incorporated into the latest TSPA model. In addition, the Repository Integration Program, an integrated probabilistic simulator, used in the TSPA has also been updated by Golder Associates Incorporated at December 1997. To ensure that the Department of Energy-owned (DOE-owned) SNF continues to be acceptable for disposal in the repository, it will be included in the TSPA-VA evaluation. A number of parameters are needed in the TSPA-VA models to predict the performance of the DOE-owned SNF materials placed into the potential repository. This report documents all of the basis and/or derivation for each of these parameters. A number of properties were not readily available at the time the TSPA-VA data was requested. Thus, expert judgement and opinion was utilized to determine a best property value. The performance of the DOE-owned SNF will be published as part of the TSPA-VA report. Each DOE site will be collecting better data as the DOE SNF program moves closer to repository license application. As required by the RW-0333P, the National Spent Nuclear Fuel Program will be assisting each site in qualifying the information used to support the performance assessment evaluations.

A. Brewer; D. Cresap; D. Fillmore; H. Loo; M. Ebner; R. McCormack

1998-09-01T23:59:59.000Z

162

Evaluation of Nondestructive Assay/Nondestructive Examination Capabilities for Department of Energy Spent Nuclear Fuel  

SciTech Connect (OSTI)

This report summarizes an evaluation of the potential use of nondestructive assay (NDA) and nondestructive examination (NDE) technologies on DOE spent nuclear fuel (SNF). It presents the NDA/NDE information necessary for the National Spent Nuclear Fuel Program (NSNFP) and the SNF storage sites to use when defining that role, if any, of NDA/NDE in characterization and certification processes. Note that the potential role for NDA/NDE includes confirmatory testing on a sampling basis and is not restricted to use as a primary, item-specific, data collection method. The evaluation does not attempt to serve as a basis for selecting systems for development or deployment. Information was collected on 27 systems being developed at eight DOE locations. The systems considered are developed to some degree, but are not ready for deployment on the full range of DOE SNF and still require additional development. The system development may only involve demonstrating performance on additional SNF, packaging the system for deployment, and developing calibration standards, or it may be as extensive as performing additional basic research. Development time is considered to range from one to four years. We conclude that NDA/NDE systems are capable of playing a key role in the characterization and certification of DOE SNF, either as the primary data source or as a confirmatory test. NDA/NDE systems will be able to measure seven of the nine key SNF properties and to derive data for the two key properties not measured directly. The anticipated performance goals of these key properties are considered achievable except for enrichment measurements on fuels near 20% enrichment. NDA/NDE systems can likely be developed to measure the standard canisters now being considered for co-disposal of DOE SNF. This ability would allow the preparation of DOE SNF for storage now and the characterization and certification to be finalize later.

Luptak, A.J.; Bulmahn, K.D.

1998-09-01T23:59:59.000Z

163

CHARACTERISTICS OF NEXT-GENERATION SPENT NUCLEAR FUEL (SNF) TRANSPORT AND STORAGE CASKS  

SciTech Connect (OSTI)

The design of spent nuclear fuel (SNF) casks used in the present SNF disposition systems has evolved from early concepts about the nuclear fuel cycle. The reality today is much different from that envisioned by early nuclear scientists. Most SNF is placed in pool storage, awaiting reprocessing (as in Russia) or disposal at a geologic SNF repository (as in the United States). Very little transport of SNF occurs. This paper examines the requirements for SNF casks from today's perspective and attempts to answer this question: What type of SNF cask would be produced if we were to start over and design SNF casks based on today's requirements? The characteristics for a next-generation SNF cask system are examined and are found to be essentially the same in Russia and the United States. It appears that the new depleted uranium dioxide (DUO2)-steel cermet material will enable these requirements to be met. Depleted uranium (DU) is uranium in which a portion of the 235U isotope has been removed during a uranium enrichment process. The DUO2-steel cermet material is described. The United States and Russia are cooperating toward the development of a next-generation, dual-purpose, storage and transport SNF system.

Haire, M.J.; Forsberg, C.W.; Matveev, V.Z.; Shapovalov, V.I.

2004-10-03T23:59:59.000Z

164

COMPLETION OF THE FIRST INTEGRATED SPENT NUCLEAR FUEL TRANSSHIPMENT/INTERIM STORAGE FACILITY IN NW RUSSIA  

SciTech Connect (OSTI)

Northwest and Far East Russia contain large quantities of unsecured spent nuclear fuel (SNF) from decommissioned submarines that potentially threaten the fragile environments of the surrounding Arctic and North Pacific regions. The majority of the SNF from the Russian Navy, including that from decommissioned nuclear submarines, is currently stored in on-shore and floating storage facilities. Some of the SNF is damaged and stored in an unstable condition. Existing Russian transport infrastructure and reprocessing facilities cannot meet the requirements for moving and reprocessing this amount of fuel. Additional interim storage capacity is required. Most of the existing storage facilities being used in Northwest Russia do not meet health and safety, and physical security requirements. The United States and Norway are currently providing assistance to the Russian Federation (RF) in developing systems for managing these wastes. If these wastes are not properly managed, they could release significant concentrations of radioactivity to these sensitive environments and could become serious global environmental and physical security issues. There are currently three closely-linked trilateral cooperative projects: development of a prototype dual-purpose transport and storage cask for SNF, a cask transshipment interim storage facility, and a fuel drying and cask de-watering system. The prototype cask has been fabricated, successfully tested, and certified. Serial production is now underway in Russia. In addition, the U.S. and Russia are working together to improve the management strategy for nuclear submarine reactor compartments after SNF removal.

Dyer, R.S.; Barnes, E.; Snipes, R.L.; Hoeibraaten, S.; Gran, H.C.; Foshaug, E.; Godunov, V.

2003-02-27T23:59:59.000Z

165

EA-0912: Urgent-Relief Acceptance of Foreign Research Reactor Spent Nuclear Fuel  

Broader source: Energy.gov [DOE]

This EA evaluates the environmental impacts of a proposal to accept 409 spent fuel elements from eight foreign research reactors in seven European countries. The spent fuel would be shipped across...

166

HOW MANY DID YOU SAY? HISTORICAL AND PROJECTED SPENT NUCLEAR FUEL SHIPMENTS IN THE UNITED STATES, 1964 - 2048  

SciTech Connect (OSTI)

No comprehensive, up-to-date, official database exists for spent nuclear fuel shipments in the United States. The authors review the available data sources, and conclude that the absence of such a database can only be rectified by a major research effort, similar to that carried out by Oak Ridge National Laboratory (ORNL) in the early 1990s. Based on a variety of published references, and unpublished data from the U.S. Nuclear Regulatory Commission (NRC), the authors estimate cumulative U.S. shipments of commercial spent fuel for the period 1964-2001. The cumulative estimates include quantity shipped, number of cask-shipments, and shipment-miles, by truck and by rail. The authors review previous estimates of future spent fuel shipments, including contractor reports prepared for the U.S. Department of Energy (DOE), NRC, and the State of Nevada. The DOE Final Environmental Impact Statement (FEIS) for Yucca Mountain includes projections of spent nuclear fuel and high-level radioactive was te shipments for two inventory disposal scenarios (24 years and 38 years) and two national transportation modal scenarios (''mostly legal-weight truck'' and ''mostly rail''). Commercial spent fuel would compromise about 90 percent of the wastes shipped to the repository. The authors estimate potential shipments to Yucca Mountain over 38 years (2010-2048) for the DOE ''mostly legal-weight truck'' and ''mostly rail'' scenarios, and for an alternative modal mix scenario based on current shipping capabilities of the 72 commercial reactor sites. The cumulative estimates of future spent fuel shipments include quantity shipped, number of cask-shipments, and shipment-miles, by legal-weight truck, heavy-haul truck, rail and barge.

Halstead, Robert J.; Dilger, Fred

2003-02-27T23:59:59.000Z

167

COGEMA operating experience in the transportation of spent fuel, nuclear materials and radioactive waste  

SciTech Connect (OSTI)

Were a spent fuel transportation accident to occur, no matter how insignificant, the public outcry could jeopardize both reprocessing operations and power plant operations for utilities that have elected to reprocess their spent fuel. Aware of this possibility, COGEMA has become deeply involved in spent fuel transportation to ensure that it is performed according to the highest standards of transportation safety. Spent fuel transportation is a vital link between the reactor site and the reprocessing plant. This paper gives an overview of COGEMA`s experience in the transportation of spent fuel.

Bernard, H. [COGEMA, Velizy-Villacoublay (France)

1993-12-31T23:59:59.000Z

168

Naval Civil Engineering Laboratory  

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

Naval Civil Engineering Laboratory Personnel from the Power Systems Department have participated in numerous distribution equipment research, development, demonstration, testing,...

169

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank  

DOE Patents [OSTI]

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps.

Corletti, Michael M. (New Kensington, PA); Lau, Louis K. (Monroeville, PA); Schulz, Terry L. (Murrysville Boro, PA)

1993-01-01T23:59:59.000Z

170

Combined cooling and purification system for nuclear reactor spent fuel pit, refueling cavity, and refueling water storage tank  

DOE Patents [OSTI]

The spent fuel pit of a pressured water reactor (PWR) nuclear power plant has sufficient coolant capacity that a safety rated cooling system is not required. A non-safety rated combined cooling and purification system with redundant branches selectively provides simultaneously cooling and purification for the spent fuel pit, the refueling cavity, and the refueling water storage tank, and transfers coolant from the refueling water storage tank to the refueling cavity without it passing through the reactor core. Skimmers on the suction piping of the combined cooling and purification system eliminate the need for separate skimmer circuits with dedicated pumps. 1 figures.

Corletti, M.M.; Lau, L.K.; Schulz, T.L.

1993-12-14T23:59:59.000Z

171

TOPICAL REPORT ON ACTINIDE-ONLY BURNUP CREDIT FOR PWR SPENT NUCLEAR FUEL PACKAGES  

SciTech Connect (OSTI)

A methodology for performing and applying nuclear criticality safety calculations, for PWR spent nuclear fuel (SNF) packages with actinide-only burnup credit, is described. The changes in the U-234, U-235, U-236, U-238, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, and Am-241 concentration with burnup are used in burnup credit criticality analyses. No credit for fission product neutron absorbers is taken. The methodology consists of five major steps. (1) Validate a computer code system to calculate isotopic concentrations of SNF created during burnup in the reactor core and subsequent decay. A set of chemical assay benchmarks is presented for this purpose as well as a method for assessing the calculational bias and uncertainty, and conservative correction factors for each isotope. (2) Validate a computer code system to predict the subcritical multiplication factor, k{sub eff}, of a spent nuclear fuel package. Fifty-seven UO{sub 2}, UO{sub 2}/Gd{sub 2}O{sub 3}, and UO{sub 2}/PuO{sub 2} critical experiments have been selected to cover anticipated conditions of SNF. The method uses an upper safety limit on k{sub eff} (which can be a function of the trending parameters) such that the biased k{sub eff}, when increased for the uncertainty is less than 0.95. (3) Establish bounding conditions for the isotopic concentration and criticality calculations. Three bounding axial profiles have been established to assure the ''end effect'' is accounted for conservatively. (4) Use the validated codes and bounding conditions to generate package loading criteria (burnup credit loading curves). Burnup credit loading curves show the minimum burnup required for a given initial enrichment. The utility burnup record is compared to this requirement after the utility accounts for the uncertainty in its record. Separate curves may be generated for each assembly design, various minimum cooling times and burnable absorber histories. (5) Verify that SNF assemblies meet the package loading criteria and confirm proper assembly selection prior to loading. A measurement of the average assembly burnup is required and that measurement must be within 10% of the utility burnup record for the assembly to be accepted. The measurement device must be accurate to within 10%. Each step is described in detail for use with any computer code system and is then demonstrated with the SCALE 4.2 computer code package using 27BURNUPLIB cross sections.

DOE

1997-04-01T23:59:59.000Z

172

DUSCOBS - a depleted-uranium silicate backfill for transport, storage, and disposal of spent nuclear fuel  

SciTech Connect (OSTI)

A Depleted Uranium Silicate COntainer Backfill System (DUSCOBS) is proposed that would use small, isotopically-depleted uranium silicate glass beads as a backfill material inside storage, transport, and repository waste packages containing spent nuclear fuel (SNF). The uranium silicate glass beads would fill all void space inside the package including the coolant channels inside SNF assemblies. Based on preliminary analysis, the following benefits have been identified. DUSCOBS improves repository waste package performance by three mechanisms. First, it reduces the radionuclide releases from SNF when water enters the waste package by creating a local uranium silicate saturated groundwater environment that suppresses (1) the dissolution and/or transformation of uranium dioxide fuel pellets and, hence, (2) the release of radionuclides incorporated into the SNF pellets. Second, the potential for long-term nuclear criticality is reduced by isotopic exchange of enriched uranium in SNF with the depleted uranium (DU) in the glass. Third, the backfill reduces radiation interactions between SNF and the local environment (package and local geology) and thus reduces generation of hydrogen, acids, and other chemicals that degrade the waste package system. In addition, the DUSCOBS improves the integrity of the package by acting as a packing material and ensures criticality control for the package during SNF storage and transport. Finally, DUSCOBS provides a potential method to dispose of significant quantities of excess DU from uranium enrichment plants at potential economic savings. DUSCOBS is a new concept. Consequently, the concept has not been optimized or demonstrated in laboratory experiments.

Forsberg, C.W.; Pope, R.B.; Ashline, R.C.; DeHart, M.D.; Childs, K.W.; Tang, J.S.

1995-11-30T23:59:59.000Z

173

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

SciTech Connect (OSTI)

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

NONE

1997-12-01T23:59:59.000Z

174

Present experience of NRI REZ with preparation of spent nuclear fuel shipment to Russian Federation  

SciTech Connect (OSTI)

The Nuclear Research Institute Rez plc (NRI) jointed the Russian Research Reactor Fuel Return (RRRFR) programme under the US-Russian Global Threat Reduction Initiative (GTRI) initiative and started the preparation of the spent nuclear fuel (SNF) shipment from the LVR-15 research reactor back to the Russian Federation (RF). The transport of 16 SKODA VPVR/M casks with EK-10, IRT-2M 80 %, and IRT-2M 36% fuel types is planned for the autumn of 2007. The paper describes the experience gained so far during the preparatory works for the SNF shipment (facility equipment modification, cask licenses) and the actual preparation of the SNF for transport, in particular its checking, repacking in a hot cell, loading into the VPVR/M casks, drying, manipulation, completion of the transport documentation, etc., including its transport to the SNF storage facility at the NRI before it is shipped to the RF. The paper also briefly describes a regulatory framework for these activities with a focus on legislative and methodological aspects of the return of vitrified waste back to the Czech Republic. (author)

Svitak, F.; Broz, V.; Hrehor, M.; Marek, M.; Novosad, P.; Podlaha, J.; Rychecky, J. [Nuclear Research Institute Rez plc, Husinec 130, CZ-25068 (Czech Republic)

2008-07-15T23:59:59.000Z

175

FY13 Summary Report on the Augmentation of the Spent Fuel Composition Dataset for Nuclear Forensics: SFCOMPO/NF  

SciTech Connect (OSTI)

This report documents the FY13 efforts to enhance a dataset of spent nuclear fuel isotopic composition data for use in developing intrinsic signatures for nuclear forensics. A review and collection of data from the open literature was performed in FY10. In FY11, the Spent Fuel COMPOsition (SFCOMPO) excel-based dataset for nuclear forensics (NF), SFCOMPO/NF was established and measured data for graphite production reactors, Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs) were added to the dataset and expanded to include a consistent set of data simulated by calculations. A test was performed to determine whether the SFCOMPO/NF dataset will be useful for the analysis and identification of reactor types from isotopic ratios observed in interdicted samples.

Brady Raap, Michaele C.; Lyons, Jennifer A.; Collins, Brian A.; Livingston, James V.

2014-03-31T23:59:59.000Z

176

Spent Nuclear Fuel Project path forward: nuclear safety equivalency to comparable NRC-licensed facilities  

SciTech Connect (OSTI)

This document includes the Technical requirements which meet the nuclear safety objectives of the NRC regulations for fuel treatment and storage facilities. These include requirements regarding radiation exposure limits, safety analysis, design and construction. This document also includes administrative requirements which meet the objectives of the major elements of the NRC licensing process. These include formally documented design and safety analysis, independent technical review, and oppportunity for public involvement.

Garvin, L.J.

1995-11-01T23:59:59.000Z

177

Initial evaluation of dry storage issues for spent nuclear fuels in wet storage at the Idaho Chemical Processing Plant  

SciTech Connect (OSTI)

The Pacific Northwest Laboratory has evaluated the basis for moving selected spent nuclear fuels in the CPP-603 and CPP-666 storage pools at the Idaho Chemical Processing Plant from wet to dry interim storage. This work is being conducted for the Lockheed Idaho Technologies Company as part of the effort to determine appropriate conditioning and dry storage requirements for these fuels. These spent fuels are from 22 test reactors and include elements clad with aluminum or stainless steel and a wide variety of fuel materials: UAl{sub x}, UAl{sub x}-Al and U{sub 3}O{sub 8}-Al cermets, U-5% fissium, UMo, UZrH{sub x}, UErZrH, UO{sub 2}-stainless steel cermet, and U{sub 3}O{sub 8}-stainless steel cermet. The study also included declad uranium-zirconium hydride spent fuel stored in the CPP-603 storage pools. The current condition and potential failure mechanisms for these spent fuels were evaluated to determine the impact on conditioning and dry storage requirements. Initial recommendations for conditioning and dry storage requirements are made based on the potential degradation mechanisms and their impacts on moving the spent fuel from wet to dry storage. Areas needing further evaluation are identified.

Guenther, R.J.; Johnson, A.B. Jr.; Lund, A.L.; Gilbert, E.R. [and others

1996-07-01T23:59:59.000Z

178

Containment and Analysis Capability Insights Gained from Drop Testing Representative Spent Nuclear Fuel Containers  

SciTech Connect (OSTI)

The National Spent Nuclear Fuel Program (NSNFP), operating from the Idaho National Engineering and Environmental Laboratory (INEEL), developed the standardized Department of Energy (DOE) spent nuclear fuel (SNF) canister. This canister is designed to be loaded with DOE SNF (including other radioactive materials) and then be used during interim storage, during transportation to the nations repository, and for final disposal at the repository without having to be reopened. The canister has been fully designed and has completed significant testing that clearly demonstrates that it can safely achieve its intended design goals. During 1999, nine 457-mm diameter test canisters were fabricated at the INEEL to represent the standardized DOE SNF canister design. Various "worst case" internals were incorporated. Seven of the test canisters were 4.57 m long and weighed approximately 2721 kg, while two were 3.00 m long and weighed approximately 1360 kg and 1725 kg. Seven of the test canisters were dropped from 9 m onto an essentially unyielding flat surface and one of the test canisters was dropped from 1 m onto a 15-cm diameter puncture post. The final test canister was dropped from 61 cm onto a 50.8 mm thick vertically oriented steel plate, and then fell over to impact another 50.8 mm thick vertically oriented steel plate. This last test represented a canister dropping onto another larger container such as a repository disposal container or waste package. The 1999 drop testing was performed at Sandia National Laboratories (SNL). The nine test canisters experienced varying degrees of damage to their skirts, lifting rings, and pressure boundary components (heads and main body). However, all of the canisters were shown to have maintained their pressure boundary (through pressure testing). Four heavily damaged canisters were also shown to be leaktight via helium leak testing. Pre- and post-drop finite element (FE) analyses were also performed. The results clearly indicated that accurate predictions of canister responses to the drop tests were achieved. The results achieved for the standardized canister can also be applicable to other well-constructed containers (canisters, casks, cans, vessels, etc.) subjected to similar loads. Properly designed containers can maintain a containment system after being subjected to dynamically induced high strains and FE computer analyses can accurately predict the resulting responses.

Morton, Dana Keith; Snow, Spencer David; Rahl, Tommy Ervin; Ware, Arthur Gates

2001-08-01T23:59:59.000Z

179

Technology status in support of refined technical baseline for the Spent Nuclear Fuel project. Revision 1  

SciTech Connect (OSTI)

The Spent Nuclear Fuel Project (SNFP) has undertaken technology acquisition activities focused on supporting the technical basis for the removal of the N Reactor fuel from the K Basins to an interim storage facility. The purpose of these technology acquisition activities has been to identify technology issues impacting design or safety approval, to establish the strategy for obtaining the necessary information through either existing project activities, or the assignment of new work. A set of specific path options has been identified for each major action proposed for placing the N Reactor fuel into a ``stabilized`` form for interim storage as part of this refined technical basis. This report summarizes the status of technology information acquisition as it relates to key decisions impacting the selection of specific path options. The following specific categories were chosen to characterize and partition the technology information status: hydride issues and ignition, corrosion, hydrogen generation, drying and conditioning, thermal performance, criticality and materials accountability, canister/fuel particulate behavior, and MCO integrity. This report represents a preliminary assessment of the technology information supporting the SNFP. As our understanding of the N Reactor fuel performance develops the technology information supporting the SNFP will be updated and documented in later revisions to this report. Revision 1 represents the incorporation of peer review comments into the original document. The substantive evolution in our understanding of the technical status for the SNFP (except section 3) since July 1995 have not been incorporated into this revision.

Puigh, R.J.; Toffer, H.; Heard, F.J.; Irvin, J.J.; Cooper, T.D.

1995-10-20T23:59:59.000Z

180

DEVELOPING AN INTEGRATED NATIONAL STRATEGY FOR THE DISPOSITION OF SPENT NUCLEAR FUEL  

SciTech Connect (OSTI)

This paper summarizes the Department of Energy's (DOE's) current efforts to strengthen its activities for the management and disposition of DOE-owned spent nuclear fuel (SNF). In August 2002 an integrated, ''corporate project'' was initiated by the Office of Environmental Management (EM) to develop a fully integrated strategy for disposition of the approximately {approx}250,000 DOE SNF assemblies currently managed by EM. Through the course of preliminary design, the focus of this project rapidly evolved to become DOE-wide. It is supported by all DOE organizations involved in SNF management, and represents a marked change in the way DOE conducts its business. This paper provides an overview of the Corporate Project for Integrated/Risk-Driven Disposition of SNF (Corporate SNF Project), including a description of its purpose, scope and deliverables. It also summarizes the results of the integrated project team's (IPT's) conceptual design efforts, including the identification of project/system requirements and alternatives. Finally, this paper highlights the schedule of the corporate project, and its progress towards development of a DOE corporate strategy for SNF disposition.

Gelles, C.M.

2003-02-27T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" 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

Spent Nuclear Fuel Project document control and Records Management Program Description  

SciTech Connect (OSTI)

The Spent Nuclear Fuel (SNF) Project document control and records management program, as defined within this document, is based on a broad spectrum of regulatory requirements, Department of Energy (DOE) and Project Hanford and SNF Project-specific direction and guidance. The SNF Project Execution Plan, HNF-3552, requires the control of documents and management of records under the auspices of configuration control, conduct of operations, training, quality assurance, work control, records management, data management, engineering and design control, operational readiness review, and project management and turnover. Implementation of the controls, systems, and processes necessary to ensure compliance with applicable requirements is facilitated through plans, directives, and procedures within the Project Hanford Management System (PHMS) and the SNF Project internal technical and administrative procedures systems. The documents cited within this document are those which directly establish or define the SNF Project document control and records management program. There are many peripheral documents that establish requirements and provide direction pertinent to managing specific types of documents that, for the sake of brevity and clarity, are not cited within this document.

MARTIN, B.M.

2000-05-18T23:59:59.000Z

182

FATE Unified Modeling Method for Spent Nuclear Fuel and Sludge Processing, Shipping and Storage - 13405  

SciTech Connect (OSTI)

A unified modeling method applicable to the processing, shipping, and storage of spent nuclear fuel and sludge has been incrementally developed, validated, and applied over a period of about 15 years at the US DOE Hanford site. The software, FATE{sup TM}, provides a consistent framework for a wide dynamic range of common DOE and commercial fuel and waste applications. It has been used during the design phase, for safety and licensing calculations, and offers a graded approach to complex modeling problems encountered at DOE facilities and abroad (e.g., Sellafield). FATE has also been used for commercial power plant evaluations including reactor building fire modeling for fire PRA, evaluation of hydrogen release, transport, and flammability for post-Fukushima vulnerability assessment, and drying of commercial oxide fuel. FATE comprises an integrated set of models for fluid flow, aerosol and contamination release, transport, and deposition, thermal response including chemical reactions, and evaluation of fire and explosion hazards. It is one of few software tools that combine both source term and thermal-hydraulic capability. Practical examples are described below, with consideration of appropriate model complexity and validation. (authors)

Plys, Martin; Burelbach, James; Lee, Sung Jin; Apthorpe, Robert [Fauske and Associates, LLC, 16W070 83rd St., Burr Ridge, IL, 60527 (United States)] [Fauske and Associates, LLC, 16W070 83rd St., Burr Ridge, IL, 60527 (United States)

2013-07-01T23:59:59.000Z

183

Damaged Spent Nuclear Fuel at U.S. DOE Facilities Experience and Lessons Learned  

SciTech Connect (OSTI)

From a handling perspective, any spent nuclear fuel (SNF) that has lost its original technical and functional design capabilities with regard to handling and confinement can be considered as damaged. Some SNF was damaged as a result of experimental activities and destructive examinations; incidents during packaging, handling, and transportation; or degradation that has occurred during storage. Some SNF was mechanically destroyed to protect proprietary SNF designs. Examples of damage to the SNF include failed cladding, failed fuel meat, sectioned test specimens, partially reprocessed SNFs, over-heated elements, dismantled assemblies, and assemblies with lifting fixtures removed. In spite of the challenges involved with handling and storage of damaged SNF, the SNF has been safely handled and stored for many years at DOE storage facilities. This report summarizes a variety of challenges encountered at DOE facilities during interim storage and handling operations along with strategies and solutions that are planned or were implemented to ameliorate those challenges. A discussion of proposed paths forward for moving damaged and nondamaged SNF from interim storage to final disposition in the geologic repository is also presented.

Brett W. Carlsen; Eric Woolstenhulme; Roger McCormack

2005-11-01T23:59:59.000Z

184

DEMONSTRATION OF LONG-TERM STORAGE CAPABILITY FOR SPENT NUCLEAR FUEL IN L BASIN  

SciTech Connect (OSTI)

The U.S. Department of Energy decisions for the ultimate disposition of its inventory of used nuclear fuel presently in, and to be received and stored in, the L Basin at the Savannah River Site, and schedule for project execution have not been established. A logical decision timeframe for the DOE is following the review of the overall options for fuel management and disposition by the Blue Ribbon Commission on America's Nuclear Future (BRC). The focus of the BRC review is commercial fuel; however, the BRC has included the DOE fuel inventory in their review. Even though the final report by the BRC to the U.S. Department of Energy is expected in January 2012, no timetable has been established for decisions by the U.S. Department of Energy on alternatives selection. Furthermore, with the imminent lay-up and potential closure of H-canyon, no ready path for fuel disposition would be available, and new technologies and/or facilities would need to be established. The fuel inventory in wet storage in the 3.375 million gallon L Basin is primarily aluminum-clad, aluminum-based fuel of the Materials Test Reactor equivalent design. An inventory of non-aluminum-clad fuel of various designs is also stored in L Basin. Safe storage of fuel in wet storage mandates several high-level 'safety functions' that would be provided by the Structures, Systems, and Components (SSCs) of the storage system. A large inventory of aluminum-clad, aluminum-based spent nuclear fuel, and other nonaluminum fuel owned by the U.S. Department of Energy is in wet storage in L Basin at the Savannah River Site. An evaluation of the present condition of the fuel, and the Structures, Systems, or Components (SSCs) necessary for its wet storage, and the present programs and storage practices for fuel management have been performed. Activities necessary to validate the technical bases for, and verify the condition of the fuel and the SSCs under long-term wet storage have also been identified. The overall conclusion is that the fuel can be stored in L Basin, meeting general safety functions for fuel storage, for an additional 50 years and possibly beyond contingent upon continuation of existing fuel management activities and several augmented program activities. It is concluded that the technical bases and well-founded technologies have been established to store spent nuclear fuel in the L Basin. Methodologies to evaluate the fuel condition and characteristics, and systems to prepare fuel, isolate damaged fuel, and maintain water quality storage conditions have been established. Basin structural analyses have been performed against present NPH criteria. The aluminum fuel storage experience to date, supported by the understanding of the effects of environmental variables on materials performance, demonstrates that storage systems that minimize degradation and provide full retrievability of the fuel up to and greater than 50 additional years will require maintaining the present management programs, and with the recommended augmented/additional activities in this report.

Sindelar, R.; Deible, R.

2011-04-27T23:59:59.000Z

185

Characteristics of potential repository wastes: Volume 4, Appendix 4A, Nuclear reactors at educational institutions of the United States; Appendix 4B, Data sheets for nuclear reactors at educational institutions; Appendix 4C, Supplemental data for Fort St. Vrain spent fuel; Appendix 4D, Supplemental data for Peach Bottom 1 spent fuel; Appendix 4E, Supplemental data for Fast Flux Test Facility  

SciTech Connect (OSTI)

Volume 4 contains the following appendices: nuclear reactors at educational institutions in the United States; data sheets for nuclear reactors at educational institutions in the United States(operational reactors and shut-down reactors); supplemental data for Fort St. Vrain spent fuel; supplemental data for Peach Bottom 1 spent fuel; and supplemental data for Fast Flux Test Facility.

Not Available

1992-07-01T23:59:59.000Z

186

NRC Technical Research Program to Evaluate Extended Storage and Transportation of Spent Nuclear Fuel - 12547  

SciTech Connect (OSTI)

Any new direction proposed for the back-end of spent nuclear fuel (SNF) cycle will require storage of SNF beyond the current licensing periods. The Nuclear Regulatory Commission (NRC) has established a technical research program to determine if any changes in the 10 CFR part 71, and 72 requirements, and associated guidance might be necessary to regulate the safety of anticipated extended storage, and subsequent transport of SNF. This three part program of: 1) analysis of knowledge gaps in the potential degradation of materials, 2) short-term research and modeling, and 3) long-term demonstration of systems, will allow the NRC to make informed regulatory changes, and determine when and if additional monitoring and inspection of the systems is necessary. The NRC has started a research program to obtain data necessary to determine if the current regulatory guidance is sufficient if interim dry storage has to be extended beyond the currently approved licensing periods. The three-phased approach consists of: - the identification and prioritization of potential degradation of the components related to the safe operation of a dry cask storage system, - short-term research to determine if the initial analysis was correct, and - a long-term prototypic demonstration project to confirm the models and results obtained in the short-term research. The gap analysis has identified issues with the SCC of the stainless steel canisters, and SNF behavior. Issues impacting the SNF and canister internal performance such as high and low temperature distributions, and drying have also been identified. Research to evaluate these issues is underway. Evaluations have been conducted to determine the relative values that various types of long-term demonstration projects might provide. These projects or follow-on work is expected to continue over the next five years. (authors)

Einziger, R.E.; Compton, K.; Gordon, M.; Ahn, T.; Gonzales, H. [United States Nuclear Regulatory Commission, Rockville, Maryland 20852 (United States); Pan, Y. [Center for Nuclear Waste Regulatory Analyses, San Antonio, TX 78238 (United States)

2012-07-01T23:59:59.000Z

187

Naval Engineering A National Naval Obligation  

E-Print Network [OSTI]

As part of its national obligations, ONR must ensure US world leadership in those unique technology areas that insure naval superiority. ONR accomplishes this mission through research, recruitment and education, maintaining ...

Chryssostomidis, Chryssostomos

2000-05-16T23:59:59.000Z

188

Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579  

SciTech Connect (OSTI)

General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2 discharge reuse. The EM2 waste disposal profile is effectively only fission products, which reduces the mass (about 3% vs LWR), average half life, heat and long term radio-toxicity of the disposal. Widespread implementation of EM2 fuel cycle is highly significant as it would increase world energy reserves; the remaining energy in U.S. LWR SNF alone exceeds that in the U.S. natural gas reserves. Unlike many LWR SNF disposition concepts, the EM2 fuel cycle conversion of SNF produces energy and associated revenue such that the overall project is cost effective. By providing conversion of SNF to fission products the fuel cycle is closed and a non-repository LWR SNF disposition path is created and overall repository requirements are significantly reduced. (authors)

Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)] [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

2013-07-01T23:59:59.000Z

189

Hanford spent nuclear fuel project recommended path forward, volume III: Alternatives and path forward evaluation supporting documentation  

SciTech Connect (OSTI)

Volume I of the Hanford Spent Nuclear Fuel Project - Recommended Path Forward constitutes an aggressive series of projects to construct and operate systems and facilities to safely retrieve, package, transport, process, and store K Basins fuel and sludge. Volume II provided a comparative evaluation of four Alternatives for the Path Forward and an evaluation for the Recommended Path Forward. Although Volume II contained extensive appendices, six supporting documents have been compiled in Volume III to provide additional background for Volume II.

Fulton, J.C.

1994-10-01T23:59:59.000Z

190

Spent nuclear fuels project: FY 1995 multi-year program plan, WBS {number_sign}1.4  

SciTech Connect (OSTI)

The mission of the Spent Nuclear Fuel (SNF) program is to safely, reliably, and efficiently manage, condition, transport, and store Department of Energy (DOE)-owned SNF, so that it meets acceptance criteria for disposal in a permanent repository. The Hanford Site Spent Nuclear Fuel strategic plan for accomplishing the project mission is: Establish near-term safe storage in the 105-K Basins; Complete national Environmental Policy Act (NEPA) process to obtain a decision on how and where spent nuclear fuel will be managed on the site; Define and establish alternative interim storage on site or transport off site to support implementation of the NEPA decision; and Define and establish a waste package qualified for final disposition. This report contains descriptions of the following: Work Breakdown Structure; WBS Dictionary; Responsibility Assignment Matrix; Program Logic Diagrams; Program Master Baseline Schedule; Program Performance Baseline Schedule; Milestone List; Milestone Description Sheets; Cost Baseline Summary by Year; Basis of Estimate; Waste Type Data; Planned Staffing; and Fiscal Year Work Plan.

Denning, J.L.

1994-09-01T23:59:59.000Z

191

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

Broader source: Energy.gov [DOE]

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

192

Comparison and Analysis of Regulatory and Derived Requirements for Certain DOE Spent Nuclear Fuel Shipments; Lessons Learned for Future Spent Fuel Transportation Campaigns  

SciTech Connect (OSTI)

Radioactive materials transportation is stringently regulated by the Department of Transportation and the Nuclear Regulatory Commission to protect the public and the environment. As a Federal agency, however, the U.S. Department of Energy (DOE) must seek State, Tribal and local input on safety issues for certain transportation activities. This interaction has invariably resulted in the imposition of extra-regulatory requirements, greatly increasing transportation costs and delaying schedules while not significantly enhancing the level of safety. This paper discusses the results an analysis of the regulatory and negotiated requirements established for a July 1998 shipment of spent nuclear fuel from foreign countries through the west coast to the Idaho National Engineering and Environmental Laboratory (INEEL). Staff from the INEEL Nuclear Materials Engineering and Disposition Department undertook the analysis in partnership with HMTC, to discover if there were instances where requirements derived from stakeholder interactions duplicate, contradict, or otherwise overlap with regulatory requirements. The study exhaustively lists and classifies applicable Department of Transportation (DOT) and Nuclear Regulatory Commission (NRC) regulations. These are then compared with a similarly classified list of requirements from the Environmental Impact Statements (EIS) and those developed during stakeholder negotiations. Comparison and analysis reveals numerous attempts to reduce transportation risk by imposing more stringent safety measures than those required by DOT and NRC. These usually took the form of additional inspection, notification and planning requirements. There are also many instances of overlap with, and duplication of regulations. Participants will gain a greater appreciation for the need to understand the risk-oriented basis of the radioactive materials regulations and their effectiveness in ensuring safety when negotiating extra-regulatory requirements.

Kramer, George L., Ph.D.; Fawcett, Rick L.; Rieke, Philip C.

2003-02-27T23:59:59.000Z

193

National spent fuel program preliminary report RCRA characteristics of DOE-owned spent nuclear fuel DOE-SNF-REP-002. Revision 3  

SciTech Connect (OSTI)

This report presents information on the preliminary process knowledge to be used in characterizing all Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) types that potentially exhibit a Resource Conservation and Recovery Act (RCRA) characteristic. This report also includes the process knowledge, analyses, and rationale used to preliminarily exclude certain SNF types from RCRA regulation under 40 CFR {section}261.4(a)(4), ``Identification and Listing of Hazardous Waste,`` as special nuclear and byproduct material. The evaluations and analyses detailed herein have been undertaken as a proactive approach. In the event that DOE-owned SNF is determined to be a RCRA solid waste, this report provides general direction for each site regarding further characterization efforts. The intent of this report is also to define the path forward to be taken for further evaluation of specific SNF types and a recommended position to be negotiated and established with regional and state regulators throughout the DOE Complex regarding the RCRA-related policy issues.

NONE

1995-07-01T23:59:59.000Z

194

RISKIND: A computer program for calculating radiological consequences and health risks from transportation of spent nuclear fuel  

SciTech Connect (OSTI)

This report presents the technical details of RISKIND, a computer code designed to estimate potential radiological consequences and health risks to individuals and the collective population from exposures associated with the transportation of spent nuclear fuel. RISKIND is a user-friendly, interactive program that can be run on an IBM or equivalent personal computer under the Windows{trademark} environment. Several models are included in RISKIND that have been tailored to calculate the exposure to individuals under various incident-free and accident conditions. The incident-free models assess exposures from both gamma and neutron radiation and can account for different cask designs. The accident models include accidental release, atmospheric transport, and the environmental pathways of radionuclides from spent fuels; these models also assess health risks to individuals and the collective population. The models are supported by databases that are specific to spent nuclear fuels and include a radionuclide inventory and dose conversion factors. In addition, the flexibility of the models allows them to be used for assessing any accidental release involving radioactive materials. The RISKIND code allows for user-specified accident scenarios as well as receptor locations under various exposure conditions, thereby facilitating the estimation of radiological consequences and health risks for individuals. Median (50% probability) and typical worst-case (less than 5% probability of being exceeded) doses and health consequences from potential accidental releases can be calculated by constructing a cumulative dose/probability distribution curve for a complete matrix of site joint-wind-frequency data. These consequence results, together with the estimated probability of the entire spectrum of potential accidents, form a comprehensive, probabilistic risk assessment of a spent nuclear fuel transportation accident.

Yuan, Y.C. [Square Y Consultants, Orchard Park, NY (US); Chen, S.Y.; Biwer, B.M.; LePoire, D.J. [Argonne National Lab., IL (US)

1995-11-01T23:59:59.000Z

195

Summary engineering description of underwater fuel storage facility for foreign research reactor spent nuclear fuel  

SciTech Connect (OSTI)

This document is a summary description for an Underwater Fuel Storage Facility (UFSF) for foreign research reactor (FRR) spent nuclear fuel (SNF). A FRR SNF environmental Impact Statement (EIS) is being prepared and will include both wet and dry storage facilities as storage alternatives. For the UFSF presented in this document, a specific site is not chosen. This facility can be sited at any one of the five locations under consideration in the EIS. These locations are the Idaho National Engineering Laboratory, Savannah River Site, Hanford, Oak Ridge National Laboratory, and Nevada Test Site. Generic facility environmental impacts and emissions are provided in this report. A baseline fuel element is defined in Section 2.2, and the results of a fission product analysis are presented. Requirements for a storage facility have been researched and are summarized in Section 3. Section 4 describes three facility options: (1) the Centralized-UFSF, which would store the entire fuel element quantity in a single facility at a single location, (2) the Regionalized Large-UFSF, which would store 75% of the fuel element quantity in some region of the country, and (3) the Regionalized Small-UFSF, which would store 25% of the fuel element quantity, with the possibility of a number of these facilities in various regions throughout the country. The operational philosophy is presented in Section 5, and Section 6 contains a description of the equipment. Section 7 defines the utilities required for the facility. Cost estimates are discussed in Section 8, and detailed cost estimates are included. Impacts to worker safety, public safety, and the environment are discussed in Section 9. Accidental releases are presented in Section 10. Standard Environmental Impact Forms are included in Section 11.

Dahlke, H.J.; Johnson, D.A.; Rawlins, J.K.; Searle, D.K.; Wachs, G.W.

1994-10-01T23:59:59.000Z

196

Best Practices for Finite Element Analysis of Spent Nuclear Fuel Transfer, Storage, and Transportation Systems  

SciTech Connect (OSTI)

Storage casks and transportation packages for spent nuclear fuel (SNF) are designed to confine SNF in sealed canisters or casks, provide structural integrity during accidents, and remove decay through a storage or transportation overpack. The transfer, storage, and transportation of SNF in dry storage casks and transport packages is regulated under 10 CFR Part 72 and 10 CFR Part 71, respectively. Finite Element Analysis (FEA) is used with increasing frequency in Safety Analysis Reports and other regulatory technical evaluations related to SNF casks and packages and their associated systems. Advances in computing power have made increasingly sophisticated FEA models more feasible, and as a result, the need for careful review of such models has also increased. This paper identifies best practice recommendations that stem from recent NRC review experience. The scope covers issues common to all commercially available FEA software, and the recommendations are applicable to any FEA software package. Three specific topics are addressed: general FEA practices, issues specific to thermal analyses, and issues specific to structural analyses. General FEA practices covers appropriate documentation of the model and results, which is important for an efficient review process. The thermal analysis best practices are related to cask analysis for steady state conditions and transient scenarios. The structural analysis best practices are related to the analysis of casks and associated payload during standard handling and drop scenarios. The best practices described in this paper are intended to identify FEA modeling issues and provide insights that can help minimize associated uncertainties and errors, in order to facilitate the NRC licensing review process.

Bajwa, Christopher S.; Piotter, Jason; Cuta, Judith M.; Adkins, Harold E.; Klymyshyn, Nicholas A.; Fort, James A.; Suffield, Sarah R.

2010-08-11T23:59:59.000Z

197

Application of Diagnostic/Prognostic Methods to Critical Equipment for the Spent Nuclear Fuel Cleanup Program  

SciTech Connect (OSTI)

The management of the Spent Nuclear Fuel (SNF) project at the Hanford K-Basin in the 100 N Area has successfully restructured the preventive maintenance, spare parts inventory requirements, and the operator rounds data requirements. In this investigation, they continue to examine the different facets of the operations and maintenance (O&M) of the K-Basin cleanup project in search of additional reliability and cost savings. This report focuses on the initial findings of a team of PNNL engineers engaged to identify potential opportunities for reducing the cost of O&M through the application of advanced diagnostics (fault determination) and prognostics (residual life/reliability determination). The objective is to introduce predictive technologies to eliminate or reduce high impact equipment failures. The PNNL team in conjunction with the SNF engineers found the following major opportunities for cost reduction and/or enhancing reliability: (1) Provide data routing and automated analysis from existing detection systems to a display center that will engage the operations and engineering team. This display will be operator intuitive with system alarms and integrated diagnostic capability. (2) Change operating methods to reduce major transients induced in critical equipment. This would reduce stress levels on critical equipment. (3) Install a limited sensor set on failure prone critical equipment to allow degradation or stressor levels to be monitored and alarmed. This would provide operators and engineers with advance guidance and warning of failure events. Specific methods for implementation of the above improvement opportunities are provided in the recommendations. They include an Integrated Water Treatment System (IWTS) decision support system, introduction of variable frequency drives on certain pump motors, and the addition of limited diagnostic instrumentation on specified critical equipment.

Casazza, Lawrence O.; Jarrell, Donald B.; Koehler, Theresa M.; Meador, Richard J.; Wallace, Dale E.

2002-02-28T23:59:59.000Z

198

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

SciTech Connect (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 (if licensed) 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. The objective of this lessons learned study was to identify successful, best-in-class trends and commonalities from past shipping campaigns, which OCRWM could consider when planning for the development and operation of a repository transportation system. Note: this paper is for analytical and discussion purposes only, and is not an endorsement of, or commitment by, OCRWM to follow any of the comments or trends. If OCRWM elects to make such commitments at a future time, they will be appropriately documented in formal programmatic policy statements, plans and procedures. Reviewers examined an extensive study completed in 2003 by DOE's National Transportation Program (NTP), Office of Environmental Management (EM), as well as plans and documents related to SNF shipments since issuance of the NTP report. OCRWM examined specific planning, business, institutional and operating practices that have been identified by DOE, its transportation contractors, and stakeholders as important issues that arise repeatedly. In addition, the review identifies lessons learned or activities/actions which were found not to be productive to the planning and conduct of SNF shipments (i.e., negative impacts). This paper is a 'looking back' summary of lessons learned across multiple transportation campaigns. Not all lessons learned are captured here, and participants in some of the campaigns have divergent opinions and perspectives about which lessons are most critical. This analysis is part of a larger OCRWM benchmarking effort to identify best practices to consider in future transportation of radioactive materials ('looking forward'). Initial findings from this comprehensive benchmarking analysis are expected to be available in late fall 2006.

M. Keister; K, McBride

2006-08-28T23:59:59.000Z

199

Standard guide for pyrophoricity/combustibility testing in support of pyrophoricity analyses of metallic uranium spent nuclear fuel  

E-Print Network [OSTI]

1.1 This guide covers testing protocols for testing the pyrophoricity/combustibility characteristics of metallic uranium-based spent nuclear fuel (SNF). The testing will provide basic data for input into more detailed computer codes or analyses of thermal, chemical, and mechanical SNF responses. These analyses would support the engineered barrier system (EBS) design bases and safety assessment of extended interim storage facilities and final disposal in a geologic repository. The testing also could provide data related to licensing requirements for the design and operation of a monitored retrievable storage facility (MRS) or independent spent fuel storage installation (ISFSI). 1.2 This guide describes testing of metallic uranium and metallic uranium-based SNF in support of transportation (in accordance with the requirements of 10CFR71), interim storage (in accordance with the requirements of 10CFR72), and geologic repository disposal (in accordance with the requirements of 10CFR60/63). The testing described ...

American Society for Testing and Materials. Philadelphia

2007-01-01T23:59:59.000Z

200

E-Print Network 3.0 - aluminum-clad spent nuclear Sample Search...  

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

of Texas... Stephen Bolt Ben Lail Adam Davis 12;Nuclear History 1942 - First self sustaining nuclear... 1946 - Atomic Energy Act of 1946 was passed to control nuclear energy...

Note: This page contains sample records for the topic "naval spent nuclear" 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

E-Print Network 3.0 - aluminum-based spent nuclear Sample Search...  

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

of Texas... Stephen Bolt Ben Lail Adam Davis 12;Nuclear History 1942 - First self sustaining nuclear... 1946 - Atomic Energy Act of 1946 was passed to control nuclear energy...

202

LETTER REPORT. INDEPENDENT CONFIRMATORY SURVEY RESULTS OF SOILS ASSOCIATED WITH THE ARGYLE STREET SEWER LINE AT THE UNITED NUCLEAR CORPORATION NAVAL PRODUCTS SITE, NEW HAVEN, CONNECTICUT  

SciTech Connect (OSTI)

Oak Ridge Institute for Science and Education (ORISE) personnel visited the United Nuclear Corporation (UNC) Naval Products site on three separate occasions during the months of October and November 2011. The purpose of these visits was to conduct confirmatory surveys of soils associated with the Argyle Street sewer line that was being removed. Soil samples were collected from six different, judgmentally determined locations in the Argyle Street sewer trench. In addition to the six soil samples collected by ORISE, four replicate soil samples were collected by Cabrera Services, Inc. (CSI) for analysis by the ORISE laboratory. Replicate samples S0010 and S0011 were final status survey (FSS) bias samples; S0012 was an FSS systematic sample; and S0015 was a waste characterization sample. Six soil samples were also collected for background determination. Uranium-235 and uranium-238 concentrations were determined via gamma spectroscopy; the spectra were also reviewed for other identifiable photopeaks. Radionuclide concentrations for these soil samples are provided. In addition to the replicate samples and the samples collected by ORISE, CSI submitted three soil samples for inter-laboratory comparison analyses. One sample was from the background reference area, one was from waste characterization efforts (material inside the sewer line), and one was a FSS sample. The inter-laboratory comparison analyses results between ORISE and CSI were in agreement, except for one sample collected in the reference area. Smear results For Argyle Street sewer pipes are tabulated.

Adams, Wade C.

2012-01-24T23:59:59.000Z

203

Spent nuclear fuel characterization for a bounding reference assembly for the receiving basin for off-site fuel  

SciTech Connect (OSTI)

The Basis for Interim Operation (BIO) for the Receiving Basin for Off-Site Fuel (RBOF) facility at the Department of Energy (DOE) Savannah River Site (SRS) nuclear materials production complex, developed in accordance with draft DOE-STD-0019-93, required a hazard categorization for the safety analysis section as outlined in DOE-STD-1027-92. The RBOF facility was thus established as a Category-2 facility (having potential for significant on-site consequences from a radiological release) as defined in DOE 5480.23. Given the wide diversity of spent nuclear fuel stored in the RBOF facility, which made a detailed assessment of the total nuclear inventory virtually impossible, the categorization required a conservative calculation based on the concept of a hypothetical, bounding reference fuel assembly integrated over the total capacity of the facility. This scheme not only was simple but also precluded a potential delay in the completion of the BIO.

Kahook, S.D.; Garrett, R.L.; Canas, L.R.; Beckum, M.J. [Westinghouse Savannah River, Aiken, SC (United States)

1995-07-01T23:59:59.000Z

204

Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying (CVD) Facility Operations Manual  

SciTech Connect (OSTI)

The mission of the Spent Nuclear Fuel (SNF) Project Cold Vacuum Drying Facility (CVDF) is to achieve the earliest possible removal of free water from Multi-Canister Overpacks (MCOs). The MCOs contain metallic uranium SNF that have been removed from the 100K Area fuel storage water basins (i.e., the K East and K West Basins) at the US. Department of Energy Hanford Site in Southeastern Washington state. Removal of free water is necessary to halt water-induced corrosion of exposed uranium surfaces and to allow the MCOs and their SNF payloads to be safely transported to the Hanford Site 200 East Area and stored within the SNF Project Canister Storage Building (CSB). The CVDF is located within a few hundred yards of the basins, southwest of the 165KW Power Control Building and the 105KW Reactor Building. The site area required for the facility and vehicle circulation is approximately 2 acres. Access and egress is provided by the main entrance to the 100K inner area using existing roadways. The CVDF will remove free. water from the MCOs to reduce the potential for continued fuel-water corrosion reactions. The cold vacuum drying process involves the draining of bulk water from the MCO and subsequent vacuum drying. The MCO will be evacuated to a pressure of 8 torr or less and backfilled with an inert gas (helium). The MCO will be sealed, leak tested, and then transported to the CSB within a sealed shipping cask. (The MCO remains within the same shipping Cask from the time it enters the basin to receive its SNF payload until it is removed from the Cask by the CSB MCO handling machine.) The CVDF subproject acquired the required process systems, supporting equipment, and facilities. The cold vacuum drying operations result in an MCO containing dried fuel that is prepared for shipment to the CSB by the Cask transportation system. The CVDF subproject also provides equipment to dispose of solid wastes generated by the cold vacuum drying process and transfer process water removed from the MCO back to the K Basins.

IRWIN, J.J.

2000-11-18T23:59:59.000Z

205

Criticality safety considerations in the geologic disposal of spent nuclear fuel assemblies  

SciTech Connect (OSTI)

Features of geologic disposal which hamper the demonstration that criticality cannot occur therein include possible changes of shape and form, intrusion of water as a neutron moderator, and selective leaching of spent fuel constituents. If the criticality safety of spent fuel disposal depends on burnup, independent measurements verifying the burnup should be performed prior to disposal. The status of nondestructive analysis method which might provide such verification is discussed. Calculations were performed to assess the potential for increasing the allowed size of a spent fuel disposal canister if potential water intrusion were limited by close-packing the enclosed rods. Several factors were identified which severely limited the potential of this application. The theoretical limit of hexagonal close-packing cannot be achieved due to fuel rod bowing. It is concluded that disposal canisters should be sized on the basis of assumed optimum moderation. Several topics for additional research were identified during this limited study.

Gore, B.F.; McNair, G.W.; Heaberlin, S.W.

1980-05-01T23:59:59.000Z

206

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

SciTech Connect (OSTI)

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

Not Available

1994-12-01T23:59:59.000Z

207

Corrosion property of 9Cr-ODS steel in nitric acid solution for spent nuclear fuel reprocessing  

SciTech Connect (OSTI)

Corrosion tests of oxide dispersion strengthened with 9% Cr (9Cr-ODS) steel, which is one of the desirable materials for cladding tube of sodium-cooled fast reactors, in pure nitric acid solution, spent FBR fuel solution, and its simulated solution were performed to understand the corrosion behavior in a spent nuclear fuel reprocessing. In this study, the 9Cr-ODS steel with lower effective chromium content was evaluated to understand the corrosion behavior conservatively. As results, the tube-type specimens of the 9Cr-ODS steels suffered severe weight loss owing to active dissolution at the beginning of the immersion test in pure nitric acid solution in the range from 1 to 3.5 M. In contrast, the weight loss was decreased and they showed a stable corrosion in the higher nitric acid concentration, the dissolved FBR fuel solution, and its simulated solution by passivation. The corrosion rates of the 9Cr-ODS steel in the dissolved FBR fuel solution and its simulated solution were 1-2 mm/y and showed good agreement with each other. The passivation was caused by the shift of corrosion potential to noble side owing to increase in nitric acid concentration or oxidative ions in the dissolved FBR fuel solution and the simulated spent fuel solution. (authors)

Takeuchi, M.; Koizumi, T. [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Inoue, M.; Koyama, S.I. [Japan Atomic Energy Agency, Oarai-machi, Higashi-ibaraki-gun, Ibaraki 311-1393 (Japan)

2013-07-01T23:59:59.000Z

208

Parameter Selection for Department of Energy Spent Nuclear Fuel to be Used in the Yucca Mountain License Application  

SciTech Connect (OSTI)

This report contains the chemical, physical, and radiological parameters that were chosen to represent the U.S. Department of Energy spent nuclear fuel in the Yucca Mountain license application. It also contains the selected packaging requirements for the various fuel types and the criticality controls that were used. The data are reported for representative fuels and bounding fuels in groups of fuels that were selected for the analysis. The justification for the selection of each parameter is given. The data reported were not generated under any quality assurance program.

D. L. Fillmore

2003-10-01T23:59:59.000Z

209

Spent Nuclear Fuel Self-Induced XRF to Predict Pu to U Content  

E-Print Network [OSTI]

) in July 2008 and January 2009. These measurements successfully showed that it is possible to measure the Pu x-ray peak at 103.7 keV in PWR spent fuel (~1 percent Pu) using a planar HPGe detector. Prior to these measurement campaigns, the Pu peak has only...

Stafford, Alissa Sarah

2010-10-12T23:59:59.000Z

210

FIELD-DEPLOYABLE SAMPLING TOOLS FOR SPENT NUCLEAR FUEL INTERROGATION IN LIQUID STORAGE  

SciTech Connect (OSTI)

Methodology and field deployable tools (test kits) to analyze the chemical and microbiological condition of aqueous spent fuel storage basins and determine the oxide thickness on the spent fuel basin materials were developed to assess the corrosion potential of a basin. this assessment can then be used to determine the amount of time fuel has spent in a storage basin to ascertain if the operation of the reactor and storage basin is consistent with safeguard declarations or expectations and assist in evaluating general storage basin operations. The test kit was developed based on the identification of key physical, chemical and microbiological parameters identified using a review of the scientific and basin operations literature. The parameters were used to design bench scale test cells for additional corrosion analyses, and then tools were purchased to analyze the key parameters. The tools were used to characterize an active spent fuel basin, the Savannah River Site (SRS) L-Area basin. The sampling kit consisted of a total organic carbon analyzer, an YSI multiprobe, and a thickness probe. The tools were field tested to determine their ease of use, reliability, and determine the quality of data that each tool could provide. Characterization confirmed that the L Area basin is a well operated facility with low corrosion potential.

Berry, T.; Milliken, C.; Martinez-Rodriguez, M.; Hathcock, D.; Heitkamp, M.

2012-09-12T23:59:59.000Z

211

Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 3, Site team reports  

SciTech Connect (OSTI)

A self assessment was conducted of those Hanford facilities that are utilized to store Reactor Irradiated Nuclear Material, (RINM). The objective of the assessment is to identify the Hanford inventories of RINM and the ES & H concerns associated with such storage. The assessment was performed as proscribed by the Project Plan issued by the DOE Spent Fuel Working Group. The Project Plan is the plan of execution intended to complete the Secretary`s request for information relevant to the inventories and vulnerabilities of DOE storage of spent nuclear fuel. The Hanford RINM inventory, the facilities involved and the nature of the fuel stored are summarized. This table succinctly reveals the variety of the Hanford facilities involved, the variety of the types of RINM involved, and the wide range of the quantities of material involved in Hanford`s RINM storage circumstances. ES & H concerns are defined as those circumstances that have the potential, now or in the future, to lead to a criticality event, to a worker radiation exposure event, to an environmental release event, or to public announcements of such circumstances and the sensationalized reporting of the inherent risks.

Not Available

1993-11-01T23:59:59.000Z

212

Preliminary Microstructural Characterization of Gadolinium-Enriched Stainless Steels for Spent Nuclear Fuel Baskets (title change from A)  

SciTech Connect (OSTI)

Gadolinium (Gd) is a very potent neutron absorber that can potentially provide the nuclear criticality safety required for interim storage, transport, and final disposal of spent nuclear fuel. Gd could be incorporated into an alloy that can be fabricated into baskets to provide structural support, corrosion resistance, and nuclear criticality control. In particular, Gd alloyed with stainless steel has been identified as a material that may fulfill these functional requirements. However, no information is available in the open literature that describes the influence of Gd on the microstructure and resultant mechanical properties of stainless steels alloyed with Gd. Such information is vital for determination of the suitability of these types of alloys for the intended application. Characterization of Gd-stainless steel (Gd-SS) alloys is also necessary for an American Society for Testing and Materials (ASTM) material specification, subsequent code approval by the American Society of Mechanical Engineers (ASME), and regulatory approval by the Nuclear Regulatory Commission for subsequent use by the nuclear industry. The Department of Energy National Spent Nuclear Fuel Program at Idaho National Engineering and Environmental Laboratory has commissioned Lehigh University and Sandia National Laboratories to characterize the properties of a series of Gd-SS alloys to assess their suitability for the spent fuel basket application. Preliminary microstructural characterization results are presented on Gd stainless steels. Small gas tungsten arc buttons were prepared by melting 316L stainless steel with 0.1 to 10 wt.% Gd. These samples were characterized by light optical and electron optical microscopy to determine the distribution of alloying elements and volume fraction of Gd-rich phase. The results acquired to date indicate that no Gd is dissolved in the austenite matrix. Instead, the Gd was present as an interdendritic constituent, and the amount of the Gd-rich constituent increased with nominal Gd concentration. The microstructure were similar to berated stainless steels in that each alloy system contains a hard secondary constituent dispersed in a ductile austenitic matrix. Microstructure-mechanical property correlations were therefore developed from previous work on berated stainless steels in order to guide selection of compositions of larger scale Gd-alloyed heats. In turn, these large-scale heats will form the basis for further investigations in which detailed microstructure, mechanical property, and corrosion resistance relationships will be developed.

DUPONT,J.N.; ROBINO,CHARLES V.; STEPHENS JR.,JOHN J.; MCCONNELL,PAUL E.; MIZIA,R.; BRANAGAN,D.

2000-07-24T23:59:59.000Z

213

Hanford Spent Nuclear Fuel Project evaluation of multi-canister overpack venting and monitoring options during staging of K basins fuel  

SciTech Connect (OSTI)

This engineering study recommends whether multi-canister overpacks containing spent nuclear fuel from the Hanford K Basins should be staged in vented or a sealed, but ventable, condition during staging at the Canister Storage Building prior to hot vacuum conditioning and interim storage. The integrally related issues of MCO monitoring, end point criteria, and assessing the practicality of avoiding venting and Hot Vacuum Conditioning for a portion of the spent fuel are also considered.

Wiborg, J.C.

1995-12-01T23:59:59.000Z

214

A National Tracking Center for Monitoring Shipments of HEU, MOX, and Spent Nuclear Fuel: How do we implement?  

SciTech Connect (OSTI)

Nuclear material safeguards specialists and instrument developers at US Department of Energy (USDOE) National Laboratories in the United States, sponsored by the National Nuclear Security Administration (NNSA) Office of NA-24, have been developing devices to monitor shipments of UF6 cylinders and other radioactive materials , . Tracking devices are being developed that are capable of monitoring shipments of valuable radioactive materials in real time, using the Global Positioning System (GPS). We envision that such devices will be extremely useful, if not essential, for monitoring the shipment of these important cargoes of nuclear material, including highly-enriched uranium (HEU), mixed plutonium/uranium oxide (MOX), spent nuclear fuel, and, potentially, other large radioactive sources. To ensure nuclear material security and safeguards, it is extremely important to track these materials because they contain so-called direct-use material which is material that if diverted and processed could potentially be used to develop clandestine nuclear weapons . Large sources could be used for a dirty bomb also known as a radioactive dispersal device (RDD). For that matter, any interdiction by an adversary regardless of intent demands a rapid response. To make the fullest use of such tracking devices, we propose a National Tracking Center. This paper describes what the attributes of such a center would be and how it could ultimately be the prototype for an International Tracking Center, possibly to be based in Vienna, at the International Atomic Energy Agency (IAEA).

Mark Schanfein

2009-07-01T23:59:59.000Z

215

Nuclear Dynamics Consequence Analysis (NDCA) for the Disposal of Spent Nuclear Fuel in an Underground Geologic Repository--Volume 2: Methodology and Results  

SciTech Connect (OSTI)

The US Department of Energy Office of Environmental Management's (DOE/EM's) National Spent Nuclear Fuel Program (NSNFP), through a collaboration between Sandia National Laboratories (SNL) and Idaho National Engineering and Environmental Laboratory (INEEL), is conducting a systematic Nuclear Dynamics Consequence Analysis (NDCA) of the disposal of SNFs in an underground geologic repository sited in unsaturated tuff. This analysis is intended to provide interim guidance to the DOE for the management of the SNF while they prepare for final compliance evaluation. This report presents results from a Nuclear Dynamics Consequence Analysis (NDCA) that examined the potential consequences and risks of criticality during the long-term disposal of spent nuclear fuel owned by DOE-EM. This analysis investigated the potential of post-closure criticality, the consequences of a criticality excursion, and the probability frequency for post-closure criticality. The results of the NDCA are intended to provide the DOE-EM with a technical basis for measuring risk which can be used for screening arguments to eliminate post-closure criticality FEPs (features, events and processes) from consideration in the compliance assessment because of either low probability or low consequences. This report is composed of an executive summary (Volume 1), the methodology and results of the NDCA (Volume 2), and the applicable appendices (Volume 3).

Taylor, L.L.; Wilson, J.R.; Sanchez, L.C.; Aguilar, R.; Trellue, H.R.; Cochrane, K.; Rath, J.S.

1998-10-01T23:59:59.000Z

216

Risk Insights Associated with Incident-Free Transportation of Spent Nuclear Fuel To Yucca Mountain Using RADTRAN 5.5  

SciTech Connect (OSTI)

The Yucca Mountain Final Environmental Impact Statement (YM EIS)[1] included an analysis of the environmental impacts associated with the transport of spent nuclear fuel (SNF) from multiple locations across the US to Yucca Mountain for incident-free and accident conditions. While the radiological risks contained in the YM EIS were calculated to be small, it is important to recognize the many conservatisms that were utilized to calculate these risks. This paper identifies conservative assumptions associated with the YM EIS calculation of incident free transportation risk, and provides an estimate of incident free transportation risk using more realistic assumptions. While it is important to use conservative assumptions in the evaluation of the environmental impacts associated with the proposed repository, it is equally important that the public and decision makers understand the conservative nature of the results presented. This paper will provide that perspective regarding the incident free transportation impacts and summarizes the results of a more detailed EPRI report on this subject, 'Assessment of Incident Free Transport Risk for Transport of Spent Nuclear Fuel to Yucca Mountain Using RADTRAN 5.5'. [2] (authors)

Supko, E.M. [Energy Resources International, Inc., 101518 St., NW, Suite 650, Washington, DC 20036 (United States); Kessler, J.H. [Electric Power Research Institute, 1300 West W.T. Harris Blvd., Charlotte NC 28262 (United States)

2006-07-01T23:59:59.000Z

217

Standard guide for evaluation of materials used in extended service of interim spent nuclear fuel dry storage systems  

E-Print Network [OSTI]

1.1 Part of the total inventory of commercial spent nuclear fuel (SNF) is stored in dry cask storage systems (DCSS) under licenses granted by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this guide is to provide information to assist in supporting the renewal of these licenses, safely and without removal of the SNF from its licensed confinement, for periods beyond those governed by the term of the original license. This guide provides information on materials behavior under conditions that may be important to safety evaluations for the extended service of the renewal period. This guide is written for DCSS containing light water reactor (LWR) fuel that is clad in zirconium alloy material and stored in accordance with the Code of Federal Regulations (CFR), at an independent spent-fuel storage installation (ISFSI). The components of an ISFSI, addressed in this document, include the commercial SNF, canister, cask, and all parts of the storage installation including the ISFSI pad. The language of t...

American Society for Testing and Materials. Philadelphia

2010-01-01T23:59:59.000Z

218

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs draft environmental impact statement. Summary  

SciTech Connect (OSTI)

This document analyzes at a programmatic level the potential environmental consequences over the next 40 years of alternatives related to the transportation, receipt, processing, and storage of spent nuclear fuel under the responsibility of the US Department of Energy. It also analyzes the site-specific consequences of the Idaho National Engineering Laboratory sitewide actions anticipated over the next 10 years for waste and spent nuclear fuel management and environmental restoration. For programmatic spent nuclear fuel management, this document analyzes alternatives of no action, decentralization, regionalization, centralization and the use of the plans that existed in 1992/1993 for the management of these materials. For the Idaho National Engineering Laboratory, this document analyzes alternatives of no action, ten-year plan, minimum and maximum treatment, storage, and disposal of US Department of Energy wastes.

Not Available

1994-06-01T23:59:59.000Z

219

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

SciTech Connect (OSTI)

This document analyzes at a programmatic level the potential environmental consequences over the next 40 years of alternatives related to the transportation, receipt, processing, and storage of spent nuclear fuel under the responsibility of the US Department of Energy. It also analyzes the site-specific consequences of the Idaho National Engineering Laboratory sitewide actions anticipated over the next 10 years for waste and spent nuclear fuel management and environmental restoration. For programmatic spent nuclear fuel management this document analyzes alternatives of no action, decentralization, regionalization, centralization and the use of the plans that existed in 1992/1993 for the management of these materials. For the Idaho National Engineering Laboratory, this document analyzes alternatives of no action, ten-year plan, minimum and maximum and maximum treatment, storage, and disposal of US Department of Energy wastes.

Not Available

1994-06-01T23:59:59.000Z

220

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement. Volume 1  

SciTech Connect (OSTI)

This document analyzes at a pregrammatic level the potential environmental consequences over the next 40 years of alternatives related to the transportation, receipt, processing, and storage of spent nuclear fuel under the responsibility of the US Department of Energy. It also analyzes the site-specific consequences of the Idaho National Engineering Laboratory sitewide actions anticipated over the next 10 years for waste and spent nuclear fuel management and environmental restoration. For pregrammatic spent nuclear fuel management, this document analyzes alternatives of no action, decentralization, regionalization, centralization and the use of the plans that existed in 1992/1993 for the management of these materials. For the Idaho National Engineering Laboratory, this document analyzes alternatives of no action, ten-year plan, minimum and maximum treatment, storage, and disposal of US Department of Energy wastes.

Not Available

1994-06-01T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" 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

Acceptance of spent nuclear fuel in multiple element sealed canisters by the Federal Waste Management System  

SciTech Connect (OSTI)

This report is one of a series of eight prepared by E.R. Johnson Associates, Inc. (JAI) under ORNL's contract with DOE's OCRWM Systems Integration Program and in support of the Annual Capacity Report (ACR) Issue Resolution Process. The report topics relate specifically to the list of high priority technical waste acceptance issues developed jointly by DOE and a utility-working group. JAI performed various analyses and studies on each topic to serve as starting points for further discussion and analysis leading eventually to finalizing the process by which DOE will accept spent fuel and waste into its waste management system. The eight reports are concerned with the conditions under which spent fuel and high level waste will be accepted in the following categories: (1) failed fuel; (2) consolidated fuel and associated structural parts; (3) non-fuel-assembly hardware; (4) fuel in metal storage casks; (5) fuel in multi-element sealed canisters; (6) inspection and testing requirements for wastes; (7) canister criteria; (8) spent fuel selection for delivery; and (9) defense and commercial high-level waste packages. 14 refs., 27 figs.

Not Available

1990-03-01T23:59:59.000Z

222

Nuclear Forensics Attributing the Source of Spent Fuel Used in an RDD Event  

SciTech Connect (OSTI)

An RDD attack against the U.S. is something America needs to prepare against. If such an event occurs the ability to quickly identify the source of the radiological material used in an RDD would aid investigators in identifying the perpetrators. Spent fuel is one of the most dangerous possible radiological sources for an RDD. In this work, a forensics methodology was developed and implemented to attribute spent fuel to a source reactor. The specific attributes determined are the spent fuel burnup, age from discharge, reactor type, and initial fuel enrichment. It is shown that by analyzing the post-event material, these attributes can be determined with enough accuracy to be useful for investigators. The burnup can be found within a 5% accuracy, enrichment with a 2% accuracy, and age with a 10% accuracy. Reactor type can be determined if specific nuclides are measured. The methodology developed was implemented into a code call NEMASYS. NEMASYS is easy to use and it takes a minimum amount of time to learn its basic functions. It will process data within a few minutes and provide detailed information about the results and conclusions.

M.R. Scott

2005-06-01T23:59:59.000Z

223

Reduction of the Radiotoxicity of Spent Nuclear Fuel Using a Two-Tiered System Comprising Light Water Reactors and Accelerator-Driven Systems  

SciTech Connect (OSTI)

Two main issues regarding the disposal of spent nuclear fuel from nuclear reactors in the United States in the geological repository Yucca Mountain are: (1) Yucca Mountain is not designed to hold the amount of fuel that has been and is proposed to be generated in the next few decades, and (2) the radiotoxicity (i.e., biological hazard) of the waste (particularly the actinides) does not decrease below that of natural uranium ore for hundreds of thousands of years. One solution to these problems may be to use transmutation to convert the nuclides in spent nuclear fuel to ones with shorter half-lives. Both reactor and accelerator-based systems have been examined in the past for transmutation; there are advantages and disadvantages associated with each. By using existing Light Water Reactors (LWRs) to burn a majority of the plutonium in spent nuclear fuel and Accelerator-Driven Systems (ADSs) to transmute the remainder of the actinides, the benefits of each type of system can be realized. The transmutation process then becomes more efficient and less expensive. This research searched for the best combination of LWRs with multiple recycling of plutonium and ADSs to transmute spent nuclear fuel from past and projected nuclear activities (assuming little growth of nuclear energy). The neutronic design of each system is examined in detail although thermal hydraulic performance would have to be considered before a final system is designed. The results are obtained using the Monte Carlo burnup code Monteburns, which has been successfully benchmarked for MOX fuel irradiation and compared to other codes for ADS calculations. The best combination of systems found in this research includes 41 LWRs burning mixed oxide fuel with two recycles of plutonium ({approx}40 years operation each) and 53 ADSs to transmute the remainder of the actinides from spent nuclear fuel over the course of 60 years of operation.

H.R. Trellue

2003-06-01T23:59:59.000Z

224

Applications of nuclear data covariances to criticality safety and spent fuel characterization  

SciTech Connect (OSTI)

Covariance data computational methods and data used for sensitivity and uncertainty analysis within the SCALE nuclear analysis code system are presented. Applications in criticality safety calculations and used nuclear fuel analysis are discussed.

Williams, Mark L [ORNL] [ORNL; Ilas, Germina [ORNL] [ORNL; Marshall, William BJ J [ORNL] [ORNL; Rearden, Bradley T [ORNL] [ORNL

2014-01-01T23:59:59.000Z

225

On selection and operation of an international interim storage facility for spent nuclear fuel  

E-Print Network [OSTI]

Disposal of post-irradiation fuel from nuclear reactors has been an issue for the nuclear industry for many years. Most countries currently have no long-term disposal strategy in place. Therefore, the concept of an ...

Burns, Joe, 1966-

2004-01-01T23:59:59.000Z

226

Measurements on spent-fuel assemblies at Arkansas Nuclear One using the Fork system. Final report, January 1995  

SciTech Connect (OSTI)

The Fork measurement system has been used to examine spent-fuel assemblies at the two reactors of Arkansas Nuclear One, operated by Entergy Operations, Inc. The Unit 1 reactor is a Babcock and Wilcox (B and W) design, and the Unit 2 reactor is a Combustion Engineering (CE) design. The neutron and gamma-ray emissions from individual spent-fuel assemblies were measured in the storage pools by raising each assembly pathway out of the storage rack and performing a measurement near the center of the assembly. The overall accuracy of the measurements after corrections is about 2%. Thirty-four assemblies were examined at Unit 1, and forty-one assemblies at Unit 2. The average deviation of the burnup measurements from the calibration was 3.0% at Unit 1 and 3.5% at Unit 2, indicating 2 to 3% random variation among the reactor records. There was no indication of clearly anomalous assemblies. Axial Scans of the variation in neutron and gamma ray emission were obtained by collecting data at several locations along the length of three assemblies at Unit 2. Two of these assemblies were nonstandard in that each contained a small neutron source. The sources were detected by the axial scans. The test program was a cooperative effort involving Sandia National Laboratories, Los Alamos National Laboratory, Entergy Operations, Inc., the Electric Power Research Institute, and the Office of Civilian Radioactive Waste Management of the US Department of Energy.

Ewing, R.I.; Bronowski, D.R. [Sandia National Labs., Albuquerque, NM (United States); Bosler, G.E.; Siebelist, R. [Los Alamos National Lab., NM (United States); Priore, J.; Hansford, C.H.; Sullivan, S. [Entergy Operations, Inc., Russellville, AR (United States). Arkansas Nuclear One

1997-03-01T23:59:59.000Z

227

Spent Fuel Disposal Trust Fund (Maine)  

Broader source: Energy.gov [DOE]

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

228

Consideration of critically when directly disposing highly enriched spent nuclear fuel in unsaturated tuff: Bounding estimates  

SciTech Connect (OSTI)

This report presents one of 2 approaches (bounding calculations) which were used in a 1994 study to examine the possibility of a criticality in a repository. Bounding probabilities, although rough, point to the difficulty of creating conditions under which a critical mass could be assembled (container corrosion, separation of neutron absorbers from fissile material, collapse or precipitation of fissile material) and how significant the geochemical and hydrologic phenomena are. The study could not conceive of a mechanism consistent with conditions under which an atomic explosion could occur. Should a criticality occur in or near a container in the future, boundary consequence calculations showed that fissions from one critical event (<10{sup 20} fissions, if similar to aqueous and metal accidents and experiments) are quite small compared to the amount of fissions represented by the spent fuel itself. If it is assumed that the containers necessary to hold the highly enriched spent fuel went critical once per day for 1 million years, creating an energy release of about 10{sup 20} fissions, the number of fissions equals about 10{sup 28}, which corresponds to only 1% of the fission inventory in a repository containing 70,000 metric tons of heavy metal, the expected size for the proposed repository at Yucca Mountain, Nevada.

Rechard, R.P.; Tierney, M.S.; Sanchez, L.C.; Martell, M.-A.

1996-05-01T23:59:59.000Z

229

Room at the Mountain: Estimated Maximum Amounts of Commercial Spent Nuclear Fuel Capable of Disposal in a Yucca Mountain Repository  

SciTech Connect (OSTI)

The purpose of this paper is to present an initial analysis of the maximum amount of commercial spent nuclear fuel (CSNF) that could be emplaced into a geological repository at Yucca Mountain. This analysis identifies and uses programmatic, material, and geological constraints and factors that affect this estimation of maximum amount of CSNF for disposal. The conclusion of this initial analysis is that the current legislative limit on Yucca Mountain disposal capacity, 63,000 MTHM of CSNF, is a small fraction of the available physical capacity of the Yucca Mountain system assuming the current high-temperature operating mode (HTOM) design. EPRI is confident that at least four times the legislative limit for CSNF ({approx}260,000 MTHM) can be emplaced in the Yucca Mountain system. It is possible that with additional site characterization, upwards of nine times the legislative limit ({approx}570,000 MTHM) could be emplaced. (authors)

Kessler, John H. [Electric Power Research Institute - EPRI, 3420 Hillview Avenue, Palo Alto, California 94304 (United States); Kemeny, John [University of Arizona, Tucson AZ 85721 (United States); King, Fraser [Integrity Corrosion Consulting, Ltd., 6732 Silverview Drive NW, Calgary, Alberta (Canada); Ross, Alan M. [Alan M. Ross and Associates, 1061 Gray Fox Circle Pleasanton, CA 94566 (Canada); Ross, Benjamen [Disposal Safety, Inc., Bethesda, MD 20814 (United States)

2006-07-01T23:59:59.000Z

230

Spent nuclear fuel project multi-year work plan WBS {number_sign}1.4.1  

SciTech Connect (OSTI)

The Spent Nuclear Fuel (SNF) Project Multi-Year Work Plan (MYWP) is a controlled living document that contains the current SNF Project Technical, Schedule and Cost Baselines. These baselines reflect the current Project execution strategies and are controlled via the change control process. Other changes to the MYWP document will be controlled using the document control process. These changes will be processed as they are approved to keep the MYWP a living document. The MYWP will be maintained continuously as the project baseline through the life of the project and not revised annually. The MYWP is the one document which summarizes and links these three baselines in one place. Supporting documentation for each baseline referred to herein may be impacted by changes to the MYWP, and must also be revised through change control to maintain consistency.

Wells, J.L.

1997-03-01T23:59:59.000Z

231

Development of Technologies for the Simultaneous Separation of Cesium and Strontium from Spent Nuclear Fuel as Part of an Advanced Fuel Cycle  

SciTech Connect (OSTI)

As part of the Advanced Fuel Cycle Initiative, two solvent extraction technologies are being developed to simultaneously separate cesium and strontium from dissolved spent nuclear fuel. The first process utilizes a solvent consisting of chlorinated cobalt dicarbollide and polyethylene glycol extractants in a phenyltrifluoromethyl sulfone diluent. Recent improvements to the process include development of a new, non-nitroaromatic diluent and development of new stripping reagents, including a regenerable strip reagent that can be recovered and recycled. Countercurrent flowsheets have been designed and tested on simulated and actual spent nuclear fuel feed streams with both cesium and strontium removal efficiencies of greater than 99 %. The second process developed to simultaneously separate cesium and strontium from spent nuclear fuel is based on two highly-specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. A solvent composition has been developed that enables both elements to be removed together and, in fact, a synergistic effect was observed with strontium distributions in the combined solvent that are much higher that in the strontium extraction (SREX) process. Initial laboratory test results of the new combined cesium and strontium extraction process indicate good extraction and stripping performance. A flowsheet for treatment of spent nuclear fuel is currently being developed.

Jack D. Law; R. Scott HErbst; David H. Meikrantz; Dean R. Peterman; Catherine L. Riddle; Richard D. Tillotson; Terry A. Todd

2005-04-01T23:59:59.000Z

232

Spent Nuclear Fuel (SNF) Project Cask and MCO Helium Purge System Design Review Completion Report Project A.5 and A.6  

SciTech Connect (OSTI)

This report documents the results of the design verification performed on the Cask and Multiple Canister Over-pack (MCO) Helium Purge System. The helium purge system is part of the Spent Nuclear Fuel (SNF) Project Cask Loadout System (CLS) at 100K area. The design verification employed the ''Independent Review Method'' in accordance with Administrative Procedure (AP) EN-6-027-01.

ARD, K.E.

2000-04-19T23:59:59.000Z

233

Successful Completion of the Largest Shipment of Russian Research Reactor High-Enriched Uranium Spent Nuclear Fuel from Czech Republic to Russian Federation  

SciTech Connect (OSTI)

On December 8, 2007, the largest shipment of high-enriched uranium spent nuclear fuel was successfully made from a Russian-designed nuclear research reactor in the Czech Republic to the Russian Federation. This accomplishment is the culmination of years of planning, negotiations, and hard work. The United States, Russian Federation, and the International Atomic Energy Agency have been working together on the Russian Research Reactor Fuel Return (RRRFR) Program in support of the Global Threat Reduction Initiative. In February 2003, RRRFR Program representatives met with the Nuclear Research Institute in Re, Czech Republic, and discussed the return of their high-enriched uranium spent nuclear fuel to the Russian Federation for reprocessing. Nearly 5 years later, the shipment was made. This paper discusses the planning, preparations, coordination, and cooperation required to make this important international shipment.

Michael Tyacke; Dr. Igor Bolshinsky; Jeff Chamberlin

2008-07-01T23:59:59.000Z

234

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

SciTech Connect (OSTI)

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

Rechard, R.P. [ed.

1995-03-01T23:59:59.000Z

235

Hazard Evaluation for Storage of Spent Nuclear Fuel at the Solid Waste Treatment Facility  

SciTech Connect (OSTI)

This report is prepared as the initial step in the safety assurance process described in 10 CFR 830 Subpart B, Nuclear Safety Management, and HNF-PRO-700, Safety Basis Development.

ERPENBECK, E.G.

2002-11-12T23:59:59.000Z

236

Mass-spectrometric determination of americium and curium in spent nuclear fuel  

SciTech Connect (OSTI)

It was shown that it is possible to carry out isotopic analysis of americium and curium isolated in one fraction from a sample of nuclear fuel without chemical separation of the elements. This became possible as a result of the use in the mass-spectrometric measurements of a highly efficient ion source with surface ionization in a partially closed cavity. With this ion source, the content of americium and curium in nuclear fuel was determined by isotope dilution.

Kalygin, V.V.; Gabeskiriya, V.Ya.

1987-01-01T23:59:59.000Z

237

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

SciTech Connect (OSTI)

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

Rechard, R.P. [ed.

1993-12-01T23:59:59.000Z

238

Assessment of the impacts of spent fuel disassembly alternatives on the Nuclear Waste Isolation System. [Preparing and packaging spent fuel assemblies for geologic disposal  

SciTech Connect (OSTI)

The objective of this report was to evaluate four possible alternative methods of preparing and packaging spent fuel assemblies for geologic disposal against the Reference Process of unmodified spent fuel. The four alternative processes were: (1) End fitting removal, (2) Fission gas venting and resealing, (3) Fuel bundle disassembly and close packing of fuel pins, and (4) Fuel shearing and immobilization. Systems analysis was used to develop a basis of comparison of the alternatives. Conceptual processes and facility layouts were devised for each of the alternatives, based on technology deemed feasible for the purpose. Assessments were made of 15 principal attributes from the technical, operational, safety/risk, and economic considerations related to each of the alternatives, including both the surface packaging and underground repository operations. Specific attributes of the alternative processes were evaluated by assigning a number for each that expressed its merit relative to the corresponding attribute of the Reference Process. Each alternative process was then ranked by summing the numbers for attributes in each of the four assessment areas and collectively. Fuel bundle disassembly and close packing of fuel pins was ranked the preferred method of disposal of spent fuel. 63 references, 46 figures, 46 tables.

Not Available

1984-07-01T23:59:59.000Z

239

Vertical Drop of the Naval SNF Long Waste Package On Unyielding Surface  

SciTech Connect (OSTI)

The purpose of this calculation is to determine the structural response of a Naval SNF (Spent Nuclear Fuel) Long Waste Package (WP) subjected to 2 m-vertical drop on unyielding surface (US). The scope of this document is limited to reporting the calculation results in terms of maximum stress intensities. This calculation is associated with the waste package design; calculation is performed by the Waste Package Design group. AP-3.12Q, Revision 0, ICN 0, Calculations, is used to perform the calculation and develop the document. The finite element calculation is performed by using the commercially available ANSYS Version (V) 5.4 finite element code. The result of this calculation is provided in terms of maximum stress intensities.

S. Mastilovic

2006-07-21T23:59:59.000Z

240

Spent Fuel Working Group report on inventory and storage of the Department`s spent nuclear fuel and other reactor irradiated nuclear materials and their environmental, safety and health vulnerabilities. Volume 2, Working Group Assessment Team reports; Vulnerability development forms; Working group documents  

SciTech Connect (OSTI)

The Secretary of Energy`s memorandum of August 19, 1993, established an initiative for a Department-wide assessment of the vulnerabilities of stored spent nuclear fuel and other reactor irradiated nuclear materials. A Project Plan to accomplish this study was issued on September 20, 1993 by US Department of Energy, Office of Environment, Health and Safety (EH) which established responsibilities for personnel essential to the study. The DOE Spent Fuel Working Group, which was formed for this purpose and produced the Project Plan, will manage the assessment and produce a report for the Secretary by November 20, 1993. This report was prepared by the Working Group Assessment Team assigned to the Hanford Site facilities. Results contained in this report will be reviewed, along with similar reports from all other selected DOE storage sites, by a working group review panel which will assemble the final summary report to the Secretary on spent nuclear fuel storage inventory and vulnerability.

Not Available

1993-11-01T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" 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

Application of Neutron-Absorbing Structural-Amorphous Metal (SAM) Coatings for Spent Nuclear Fuel (SNF) Container to Enhance Criticality Safety Control  

SciTech Connect (OSTI)

This report describes the analysis and modeling approaches used in the evaluation for criticality-control applications of the neutron-absorbing structural-amorphous metal (SAM) coatings. The applications of boron-containing high-performance corrosion-resistant material (HPCRM)--amorphous metal as the neutron-absorbing coatings to the metallic support structure can enhance criticality safety controls for spent nuclear fuel in baskets inside storage containers, transportation casks, and disposal containers. The use of these advanced iron-based, corrosion-resistant materials to prevent nuclear criticality in transportation, aging, and disposal containers would be extremely beneficial to the nuclear waste management programs.

Choi, J

2007-01-12T23:59:59.000Z

242

BOARD OF ADVISORS TO THE PRESIDENTS OF THE NAVAL POSTGRADUATE SCHOOL & NAVAL WAR COLLEGE  

E-Print Network [OSTI]

BOARD OF ADVISORS TO THE PRESIDENTS OF THE NAVAL POSTGRADUATE SCHOOL & NAVAL WAR COLLEGE October 17 attendance (choose all that apply): Naval Postgraduate School Subcommittee, 17 October 2012 Naval War College

243

Operational Challenges of Extended Dry Storage of Spent Nuclear Fuel - 12550  

SciTech Connect (OSTI)

As a result of the termination of the Yucca Mountain used fuel repository program and a continuing climate of uncertainty in the national policy for nuclear fuel disposition, the likelihood has increased that extended storage, defined as more than 60 years, and subsequent transportation of used nuclear fuel after periods of extended storage may become necessary. Whether at the nation's 104 nuclear energy facilities, or at one or more consolidated interim storage facilities, the operational challenges of extended storage and transportation will depend upon the future US policy for Used Fuel Management and the future Regulatory Framework for EST, both of which should be developed with consideration of their operational impacts. Risk insights into the regulatory framework may conclude that dry storage and transportation operations should focus primarily on ensuring canister integrity. Assurance of cladding integrity may not be beneficial from an overall risk perspective. If assurance of canister integrity becomes more important, then mitigation techniques for potential canister degradation mechanisms will be the primary source of operational focus. If cladding integrity remains as an important focus, then operational challenges to assure it would require much more effort. Fundamental shifts in the approach to design a repository and optimize the back-end of the fuel cycle will need to occur in order to address the realities of the changes that have taken place over the last 30 years. Direct disposal of existing dual purpose storage and transportation casks will be essential to optimizing the back end of the fuel cycle. The federal used fuel management should focus on siting and designing a repository that meets this objective along with the development of CIS, and possibly recycling. An integrated approach to developing US policy and the regulatory framework must consider the potential operational challenges that they would create. Therefore, it should be integral to these efforts to redefine retrievability to apply to the dual purpose cask, and not to apply to individual assemblies. (authors)

Nichol, M. [Nuclear Energy Institute, Washington DC (United States)

2012-07-01T23:59:59.000Z

244

Discussing spent nuclear fuel in high school classrooms: addressing public fears through early education  

SciTech Connect (OSTI)

The Inreach program combines the Deep River Science Academy (DRSA) 'learning through research' approach with state of the art communication technology to bring scientific research to high school classrooms. The Inreach program follows the DRSA teaching model where a university student tutor works on a research project with scientific staff at AECL's Chalk River Laboratories. Participating high school classes are located across Canada. The high school students learn about the ongoing research activities via weekly web conferences. In order to engage the students and encourage participation in the conferences, themed exercises linked to the research project are provided to the students. The DRSA's Inreach program uses a cost-effective internet technology to reach a wide audience, in an interactive setting, without anyone leaving their desks or offices. An example Inreach research project is presented here: an investigation of the potential of the Canadian supercritical water cooled reactor (SCWR) concept to burn transuranic elements (Np, Pu, Am, Cm) to reduce the impact of used nuclear fuel. During this project a university student worked with AECL (Atomic Energy of Canada Limited) researchers on technical aspects of the project, and high school students followed their progress and learned about the composition, hazards, and disposition options for used nuclear fuel. Previous projects included the effects of tritium on cellular viability and neutron diffraction measurement of residual stresses in automobile engines.

Winkel, S. [Deep River Science Academy, 20 Forest Ave. P.O. Box 600, Deep River, Ontario K0J 1P0 (Canada); Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Sullivan, J.; Jones, S.; Sullivan, K. [Deep River Science Academy, 20 Forest Ave. P.O. Box 600, Deep River, Ontario K0J 1P0 (Canada); Hyland, B.; Pencer, J.; Colton, A. [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada)

2013-07-01T23:59:59.000Z

245

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

SciTech Connect (OSTI)

In the U.S. Department of Energy's ''Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada,'' the Department states that certain broad transportation-related decisions can be made. These include the choice of a mode of transportation nationally (mostly legal-weight truck or mostly rail) and in Nevada (mostly rail, mostly legal-weight truck, or mostly heavy-haul truck with use of an associated intermodal transfer station), as well as the choice among alternative rail corridors or heavy-haul truck routes with use of an associated intermodal transfer station in Nevada. Although a rail line does not service the Yucca Mountain site, the Department has identified mostly rail as its preferred mode of transportation, both nationally and in the State of Nevada. If mostly rail is selected for Nevada, the Department would then identify a preference for one of the rail corridors in consultation with affected stakeholders, particularly the State of Nevada. DOE would then select the rail corridor and initiate a process to select a specific rail alignment within the corridor for the construction of a rail line. Five proposed rail corridors were analyzed in the Final Environmental Impact Statement. The assessment considered the impacts of constructing a branch rail line in the five 400-meter (0.25mile) wide corridors. Each corridor connects the Yucca Mountain site with an existing mainline railroad in Nevada.

Sweeney, Robin L,; Lechel, David J.

2003-02-25T23:59:59.000Z

246

Analysis of dust samples collected from spent nuclear fuel interim storage containers at Hope Creek, Delaware, and Diablo Canyon, California.  

SciTech Connect (OSTI)

Potentially corrosive environments may form on the surface of spent nuclear fuel dry storage canisters by deliquescence of deposited dusts. To assess this, samples of dust were collected from in-service dry storage canisters at two near-marine sites, the Hope Creek and Diablo Canyon storage installations, and have been characterized with respect to mineralogy, chemistry, and texture. At both sites, terrestrially-derived silicate minerals, including quartz, feldspars, micas, and clays, comprise the largest fraction of the dust. Also significant at both sites were particles of iron and iron-chromium metal and oxides generated by the manufacturing process. Soluble salt phases were minor component of the Hope Creek dusts, and were compositionally similar to inland salt aerosols, rich in calcium, sulfate, and nitrate. At Diablo Canyon, however, sea-salt aerosols, occurring as aggregates of NaCl and Mg-sulfate, were a major component of the dust samples. The seasalt aerosols commonly occurred as hollow spheres, which may have formed by evaporation of suspended aerosol seawater droplets, possibly while rising through the heated annulus between the canister and the overpack. The differences in salt composition and abundance for the two sites are attributed to differences in proximity to the open ocean and wave action. The Diablo Canyon facility is on the shores of the Pacific Ocean, while the Hope Creek facility is on the shores of the Delaware River, several miles from the open ocean.

Bryan, Charles R.; Enos, David George

2014-07-01T23:59:59.000Z

247

Draft Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada  

SciTech Connect (OSTI)

The Proposed Action addressed in this EIS is to construct, operate and monitor, and eventually close a geologic repository at Yucca Mountain in southern Nevada for the disposal of spent nuclear fuel and high-level radioactive waste currently in storage at 72 commercial and 5 DOE sites across the United States. The EIS evaluates (1) projected impacts on the Yucca Mountain environment of the construction, operation and monitoring, and eventual closure of the geologic repository; (2) the potential long-term impacts of repository disposal of spent nuclear fuel and high-level radioactive waste; (3) the potential impacts of transporting these materials nationally and in the State of Nevada; and (4) the potential impacts of not proceeding with the Proposed Action.

N /A

1999-08-13T23:59:59.000Z

248

A practical strategy for reducing the future security risk of United States spent nuclear fuel  

SciTech Connect (OSTI)

Depletion calculations show that advanced oxide (AOX) fuels can be used in existing light water reactors (LWRs) to achieve and maintain virtually any desired level of US (US) reactor-grade plutonium (R-Pu) inventory. AOX fuels are composed of a neutronically inert matrix loaded with R-Pu and erbium. A 1/2 core load of 100% nonfertile, 7w% R-Pu AOX and 3.9 w% UO{sub 2} has a net total plutonium ({sup TOT}Pu) destruction rate of 310 kg/yr. The 20% residual {sup TOT}Pu in discharged AOX contains > 55% {sup 242}Pu making it unattractive for nuclear explosive use. A three-phase fuel-cycle development program sequentially loading 60 LWRs with 100% mixed oxide, 50% AOX with a nonfertile component displacing only some of the {sup 238}U, and 50% AOX, which is 100% nonfertile, could reduce the US plutonium inventory to near zero by 2050.

Chodak, P. III; Buksa, J.J. [Los Alamos National Lab., NM (United States). Nuclear Systems Design and Analysis Group

1997-06-01T23:59:59.000Z

249

Radiological survey of the Mare Island Naval Shipyard, Alameda Naval Air Station, and Hunters Point Shipyard  

SciTech Connect (OSTI)

Since 1963, the Eastern Environmental Radiation Facility (EERF), US Environmental Protection Agency (USEPA), in cooperation with the US Naval Sea Systems Command (NAVSEA) has surveyed facilities serving nuclear-powered warships on the Atlantic and Pacific coasts and the Gulf of Mexico. These surveys assess whether the operation of nuclear-powered warships, during construction, maintenance, overhaul, or refueling, have created elevated levels of radioactivity. The surveys emphasize sampling those areas and pathways that could expose the public. In 1984, NAVSEA requested that EPA survey all active facilities serving nuclear-powered warships over the next three years. This report contains the results of surveys conducted at Naval facilities located at Mare Island, Alameda, and Hunters Point in the San Francisco region. The locations of these facilities are shown. 3 refs., 4 figs., 3 tabs.

Semler, M.O.; Blanchard, R.L. (Environmental Protection Agency, Montgomery, AL (USA). Eastern Environmental Radiation Facility)

1989-06-01T23:59:59.000Z

250

Preliminary Simulations for Geometric Optimization of a High-Energy Delayed Gamma Spectrometer for Direct Assay of Pu in Spent Nuclear Fuel  

SciTech Connect (OSTI)

High-energy, beta-delayed gamma-ray spectroscopy is under investigation as part of the Next Generation Safeguard Initiative effort to develop non-destructive assay instruments for plutonium mass quantification in spent nuclear fuel assemblies. Results obtained to date indicate that individual isotope-specific signatures contained in the delayed gamma-ray spectra can potentially be used to quantify the total fissile content and individual weight fractions of fissile and fertile nuclides present in spent fuel. Adequate assay precision for inventory analysis can be obtained using a neutron generator of sufficient strength and currently available detection technology. In an attempt to optimize the geometric configuration and material composition for a delayed gamma measurement on spent fuel, the current study applies MCNPX, a Monte Carlo radiation transport code, in order to obtain the best signal-to-noise ratio. Results are presented for optimizing the neutron spectrum tailoring material, geometries to maximize thermal or fast fissions from a given neutron source, and detector location to allow an acceptable delayed gamma-ray signal while achieving a reasonable detector lifetime while operating in a high-energy neutron field. This work is supported in part by the Next Generation Safeguards Initiative, Office of Nuclear Safeguards and Security, National Nuclear Security Administration.

Kulisek, Jonathan A.; Campbell, Luke W.; Rodriguez, Douglas C.

2012-06-07T23:59:59.000Z

251

Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Draft Environmental Impact Statement; Volume 1, Appendix F, Nevada Test Site and Oak Ridge Reservation Spent Nuclear Fuel Management Programs  

SciTech Connect (OSTI)

This volume addresses the interim storage of spent nuclear fuel (SNF) at two US Department of Energy sites, the Nevada Test Site (NTS) and the Oak Ridge Reservation (ORR). These sites are being considered to provide a reasonable range of alternative settings at which future SNF management activities could be conducted. These locations are not currently involved in management of large quantities of SNF; NTS has none, and ORR has only small quantities. But NTS and ORR do offer experience and infrastructure for the handling, processing and storage of radioactive materials, and they do exemplify a broad spectrum of environmental parameters. This broad spectrum of environmental parameters will provide, a perspective on whether and how such location attributes may relate to potential environmental impacts. Consideration of these two sites will permit a programmatic decision to be based upon an assessment of the feasible options without bias, to the current storage sites. This volume is divided into four parts. Part One is the volume introduction. Part Two contains chapters one through five for the NTS, as well as references contained in chapter six. Part Three contains chapters one through five for the ORR, as well as references contained in chapter six. Part Four is summary information including the list of preparers, organizations contacted, acronyms, and abbreviations for both the NTS and the ORR. A Table of Contents, List of Figures, and List of Tables are included in parts Two, Three, and Four. This approach permitted the inclusion of both sites in one volume while maintaining consistent chapter numbering.

NONE

1994-06-01T23:59:59.000Z

252

Critical Analysis of Dry Storage Temperature Limits for Zircaloy-Clad Spent Nuclear Fuel Based on Diffusion Controlled Cavity Growth  

SciTech Connect (OSTI)

Interim dry storage of spent nuclear fuel (SNF) rods is of critical concern because a shortage of existing SNF wet storage capacity combined with delays in the availability of a permanent disposal repository has led to an increasing number of SNF rods being placed into interim dry storage. Safe interim dry storage must be maintained for a minimum of twenty years according to the Standard Review Plan for Dry Cask Storage Systems [1] and the Code of Federal Regulations, 10 CFR Part 72 [2]. Interim dry storage licensees must meet certain safety conditions when storing SNF rods to ensure that there is a ''very low probability (e.g. 0.5%) of cladding breach during long-term storage'' [1]. Commercial SNF typically consists of uranium oxide pellets surrounded by a thin cladding. The cladding is usually an {alpha}-zirconium based alloy know as ''Zircaloy''. In dry storage, the SNF rods are confined in one of several types of cask systems approved by the Nuclear Regulatory Commission (NRC). ''The cask system must be designed to prevent degradation of fuel cladding that results in a type of cladding breach, such as axial-splits or ductile fracture, where irradiated UO{sub 2} particles may be released. In addition, the fuel cladding should not degrade to the point where more than one percent of the fuel rods suffer pinhole or hairline crack type failure under normal storage conditions [1].'' The NRC has approved two models [3,4] for use by proposed dry storage licensees to determine the maximum initial temperature limit for nuclear fuel rods in dry storage that supposedly meet the above criteria and yield consistent temperature limits. Though these two models are based on the same fundamental failure theory, different assumptions have been made including the choice of values for material constants in the failure equation. This report will examine and compare the similarities and inconsistencies of these two models. It will illustrate some of the shortcomings of the current models and suggest modifications as well as some experiments that should be started in the near future. This report will also discuss changes in the current NRC standards with regard to the adoption of a strain-based model to be used to determine maximum allowable temperatures of the SNF.

Hayes, T.A.; Rosen, R.S.; Kassner, M.E.

1999-12-01T23:59:59.000Z

253

EQ6 Calculations for Chemical Degradation Of N Reactor (U-Metal) Spent Nuclear Fuel Waste Packages  

SciTech Connect (OSTI)

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating Contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the N Reactor, a graphite moderated reactor at the Department of Energy's (DOE) Hanford Site (ref. 1). The N Reactor core was fueled with slightly enriched (0.947 wt% and 0.947 to 1.25 wt% {sup 235}U in Mark IV and Mark IA fuels, respectively) U-metal clad in Zircaloy-2 (Ref. 1, Sec. 3). Both types of N Reactor SNF have been considered for disposal at the proposed Yucca Mountain site. For some WPs, the outer shell and inner shell may breach (Ref. 3) allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing two multi-canister overpacks (MCO) with either six baskets of Mark IA or five baskets of Mark IV intact N Reactor SNF rods (Ref. 1, Sec. 4) and two high-level waste (HLW) glass pour canisters (GPCs) arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which fissile uranium will remain in the WP after corrosion/dissolution of the initial WP configuration (2) The extent to which fissile uranium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations) of the simulations, is limited to time periods up to 6.35 x 10{sup 5} years. This longer time frame is closer to the one million year time horizon recently recommended by the National Academy of Sciences to the Environmental Protection Agency for performance assessment related to a nuclear repository (Ref. 5). However, it is important to note that after 100,000 years, most of the materials of interest (fissile materials) will have either been removed from the WP, reached a steady state, or been transmuted.

P. Bernot

2001-02-27T23:59:59.000Z

254

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

SciTech Connect (OSTI)

This performance assessment characterized plausible treatment options conceived by the Idaho National Engineering Laboratory (INEL) for its spent fuel and high-level radioactive waste and then modeled the performance of the resulting waste forms in two hypothetical, deep, geologic repositories: one in bedded salt and the other in granite. The results of the performance assessment are intended to help guide INEL in its study of how to prepare wastes and spent fuel for eventual permanent disposal. This assessment was part of the Waste Management Technology Development Program designed to help the US Department of Energy develop and demonstrate the capability to dispose of its nuclear waste. Although numerous caveats must be placed on the results, the general findings were as follows: Though the waste form behavior depended upon the repository type, all current and proposed waste forms provided acceptable behavior in the salt and granite repositories.

Rechard, R.P. [ed.

1993-12-01T23:59:59.000Z

255

Development of a novel solvent for the simultaneous separation of strontium and cesium from dissolved Spent Nuclear Fuel solutions  

SciTech Connect (OSTI)

The recovery of Cs and Sr from acidic solutions by solvent extraction has been investigated. The goal of this project was to develop an extraction process to remove Cs and Sr from high-level waste in an effort to reduce the heat loading in storage. Solvents for the extraction of Cs and Sr separately have been used on both caustic and acidic spent nuclear fuel waste in the past. The objective of this research was to find a suitable solvent for the extraction of both Cs and Sr simultaneously from acidic nitrate media. The solvents selected for this research possess good stability and extraction behavior when mixed together. The extraction experiments were performed with 4 ,4,(5 )-Di-(tbutyldicyclohexano)- 18-crown-6 {DtBuCH18C6}, Calix[4]arene-bis-(tert-octylbenzocrown-6) {BOBCalixC6} and 1-(2,2,3,3-tetrafluoropropoxy)-3-(4-sec-butylphenoxy)-2-propanol {Cs-7SB modifier} in a branched aliphatic kerosene {Isopar L}. The BOBCalixC6 and Cs-7SB modifier were developed at Oak Ridge National Laboratory (ORNL) by Bonnesen et al. [1]. The values obtained from the SREX solvent for DSr in 1 M nitric acid ranged from 0.7 to 2.2 at 25oC and 10oC respectively. The values for DCs in 1 M nitric acid with the CSSX solvent ranged from 8.0 to 46.0 at 25oC and 10oC respectively. A new mixed solvent, developed at the Idaho National Engineering and Environmental Laboratory (INEEL) by Riddle et al. [2], showed distributions for Sr ranging from 8.8 to 17.4 in 1 M nitric acid at 25oC and 10oC respectively. The DCs for the mixed solvent ranged from 7.7 to 20.2 in 1 M nitric acid at 25oC to 10oC respectively. The unexpectedly high distributions for Sr at both 25oC and 10oC show a synergy in the mixed solvent. The DCs, although lower than with CSSX solvent, still showed good extraction behavior.

Catherine L. Riddle; John D. Baker; Jack D. Law; Christopher A. McGrath; David H. Meikrantz; Bruce J. Mincher; Dean R. Peterman; Terry A. Todd

2004-10-01T23:59:59.000Z

256

Measurement and analysis of gamma-rays emitted from spent nuclear fuel above 3 MeV  

SciTech Connect (OSTI)

The Next Generation Safeguard Initiative (NGSI) includes an effort to determine the mass content of fissile isotopes contained within spent fuel through the spectroscopy of high-energy delayed gamma rays. Studies being performed indicate the primary difficulty is the ability to detect the desired signal in the presence of the intense background associated with spent fuel fission products. An enabling technology for this application is high-resolution high-purity germanium (HPGe) detectors capable of operating efficiently in at extremely high count rates. This presentation will describe the prospects of high-rate germanium detectors and delayed-gamma techniques, primarily discussing the efforts to merge these into a unique and viable system for measuring spent fuel.

Rodriguez, Douglas C. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Anderson, Elaina R. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); State Univ. of New York at Stony Brook, NY (United States); Anderson, Kevin K. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Campbell, Luke W. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Fast, James E. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Jarman, Kenneth D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Kulisek, Jonathan A. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Orton, Christopher R. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Runkle, Robert C. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Stave, Sean [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

2013-08-28T23:59:59.000Z

257

Naval Spent Fuel Rail Shipment Accident Exercise Objectives  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGY TAXBalanced Scorecard Federal2Energy Second Quarter Report1-93 JulyDepartment

258

Naval Spent Fuel Rail Shipment Accident Exercise Objectives | Department of  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNG |September 15, 2010 Printing andNancyFairbanksNaturalDecreeCouncilEnergy

259

Fracture of aluminum naval structures  

E-Print Network [OSTI]

Structural catastrophic failure of naval vessels due to extreme loads such as underwater or air explosion, high velocity impact (torpedoes), or hydrodynamic loads (high speed vessels) is primarily caused by fracture. ...

Galanis, Konstantinos, 1970-

2007-01-01T23:59:59.000Z

260

Final Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada  

SciTech Connect (OSTI)

The purpose of this environmental impact statement (EIS) is to provide information on potential environmental impacts that could result from a Proposed Action to construct, operate and monitor, and eventually close a geologic repository for the disposal of spent nuclear fuel and high-level radioactive waste at the Yucca Mountain site in Nye County, Nevada. The EIS also provides information on potential environmental impacts from an alternative referred to as the No-Action Alternative, under which there would be no development of a geologic repository at Yucca Mountain.

N /A

2002-10-25T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" 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

EQ6 Calculation for Chemical Degradation of Shippingport LWBR (TH/U Oxide) Spent Nuclear Fuel Waste Packages  

SciTech Connect (OSTI)

The Monitored Geologic Repository (MGR) Waste Package Department of the Civilian Radioactive Waste Management System Management & Operating contractor (CRWMS M&O) performed calculations to provide input for disposal of spent nuclear fuel (SNF) from the Shippingport Light Water Breeder Reactor (LWBR) (Ref. 1). The Shippingport LWBR SNF has been considered for disposal at the potential Yucca Mountain site. Because of the high content of fissile material in the SNF, the waste package (WP) design requires special consideration of the amount and placement of neutron absorbers and the possible loss of absorbers and SNF materials over geologic time. For some WPs, the outer shell corrosion-resistant material (CRM) and the corrosion-allowance inner shell may breach (Refs. 2 and 3), allowing the influx of water. Water in the WP will moderate neutrons, increasing the likelihood of a criticality event within the WP; and the water may, in time, gradually leach the fissile components and neutron absorbers from the WP, further affecting the neutronics of the system. This study presents calculations of the long-term geochemical behavior of WPs containing a Shippingport LWBR SNF seed assembly, and high-level waste (HLW) glass canisters arranged according to the codisposal concept (Ref. 4). The specific study objectives were to determine: (1) The extent to which criticality control material, suggested for this WP design, will remain in the WP after corrosion/dissolution of the initial WP configuration (such that it can be effective in preventing criticality); (2) The extent to which fissile uranium and fertile thorium will be carried out of the degraded WP by infiltrating water (such that internal criticality is no longer possible, but the possibility of external criticality may be enhanced); and (3) The nominal chemical composition for the criticality evaluations of the WP design, and to suggest the range of parametric variations for additional evaluations. The scope of this calculation, the chemical compositions (and subsequent criticality evaluations), of the simulations are limited to time periods up to 3.17 x 10{sup 5} years. This longer time frame is closer to the one million year time horizon recently recommended by the National Academy of Sciences to the Environmental Protection Agency for performance assessment related to a nuclear repository (Ref. 5). However, it is important to note that after 100,000 years, most of the materials of interest (fissile and absorber materials) will have either been removed from the WP, reached a steady state, or been transmuted. The calculation included elements with high neutron-absorption cross sections, notably gadolinium (Gd), as well as the fissile materials. The results of this analysis will be used to ensure that the type and amount of criticality control material used in the WP design will prevent criticality.

S. Arthur

2000-09-14T23:59:59.000Z

262

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

SciTech Connect (OSTI)

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

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

1984-04-01T23:59:59.000Z

263

LWR spent fuel reduction by the removal of U and the compact storage of Pu with FP for long-term nuclear sustainability  

SciTech Connect (OSTI)

Fast breeder reactors (FBR) nuclear fuel cycle is needed for long-term nuclear sustainability while preventing global warming and maximum utilizing the limited uranium (U) resources. The 'Framework for Nuclear Energy Policy' by the Japanese government on October 2005 stated that commercial FBR deployment will start around 2050 under its suitable conditions by the successive replacement of light water reactors (LWR) to FBR. Even after Fukushima Daiichi Nuclear Power Plant accident which made Japanese tendency slow down the nuclear power generation activities, Japan should have various options for energy resources including nuclear, and also consider the delay of FBR deployment and increase of LWR spent fuel (LWR-SF) storage amounts. As plutonium (Pu) for FBR deployment will be supplied from LWR-SF reprocessing and Japan will not possess surplus Pu, the authors have developed the flexible fuel cycle initiative (FFCI) for the transition from LWR to FBR. The FFCI system is based on the possibility to stored recycled materials (U, Pu)temporarily for a suitable period according to the FBR deployment rate to control the Pu demand/supply balance. This FFCI system is also effective after the Fukushima accident for the reduction of LWR-SF and future LWR-to-FBR transition. (authors)

Fukasawa, T.; Hoshino, K. [Hitachi-GE Nuclear Energy, Ltd, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Takano, M. [Japan Atomic Energy Agency, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Sato, S. [Hokkaido University, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan); Shimazu, Y. [Fukui University, 3-1-1 Saiwai, Hitachi, Ibaraki, 317-0073 (Japan)

2013-07-01T23:59:59.000Z

264

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

SciTech Connect (OSTI)

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

Not Available

1994-06-01T23:59:59.000Z

265

State of Washington Department of Health Radioactive air emissions notice of construction phase 1 for spent nuclear fuel project - cold vacuum drying facility, project W-441  

SciTech Connect (OSTI)

This notice of construction (NOC) provides information regarding the source and the estimated annual possession quantity resulting from operation of the Cold Vacuum Drying Facility (CVDF). Additional details on emissions generated by the operation of the CVDF will be discussed again in the Phase 11 NOC. This document serves as a NOC pursuant to the requirements of WAC 246-247-060 for the completion of Phase I NOC, defined as the pouring of concrete for the foundation flooring, construction of external walls, and construction of the building excluding the installation of CVDF process equipment. A Phase 11 NOC will be submitted for approval prior to installing and is defined as the completion of the CVDF, which consisted installation of process equipment, air emissions control, and emission monitoring equipment. About 80 percent of the U.S. Department of Energy`s spent nuclear fuel (SNF) inventory is stored under water in the Hanford Site K Basins. Spent nuclear fuel in the K West Basin is contained in closed canisters while the SNF in the K East Basin is in open canisters, which allow free release of corrosion products to the K East Basin water.

Turnbaugh, J.E.

1996-08-15T23:59:59.000Z

266

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

SciTech Connect (OSTI)

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

Dilger, Fred [Black Mountain Research, Henderson, NV 81012 (United States)] [Black Mountain Research, Henderson, NV 81012 (United States); Halstead, Robert J. [State of Nevada Agency for Nuclear Projects, Carson City, NV 80906 (United States)] [State of Nevada Agency for Nuclear Projects, Carson City, NV 80906 (United States); Ballard, James D. [Department of Sociology, California State University, Northridge, CA 91330 (United States)] [Department of Sociology, California State University, Northridge, CA 91330 (United States)

2013-07-01T23:59:59.000Z

267

Progress Letter Report on Bending Fatigue Test System Development for Spent Nuclear Fuel Vibration Integrity Study (Out-of-cell fatigue testing development - Task 2.4)  

SciTech Connect (OSTI)

Vibration integrity of high burn-up spent nuclear fuel in transportation remains to be a critical component of US nuclear waste management system. The structural evaluation of package for spent fuel transportation eventually will need to see if the content or spent fuel is in a subcritical condition. However, a system for testing and characterizing such spent fuel is still lacking mainly due to the complication involved with dealing radioactive specimens in a hot cell environment. Apparently, the current state-of-the-art in spent fuel research and development is quite far away from the delivery of reliable mechanical property data for the assessment of spent fuels in the transport package evaluation. Under the sponsorship of US NRC, ORNL has taken the challenge in developing a robust testing system for spent fuel in hot cell. An extensive literature survey was carried out and unique requirements of such testing system were identified. The U-frame setup has come to the top among various designs examined for reverse bending fatigue test of spent fuel rod. The U-frame has many features that deserve mentioned here: Easy to install spent fuel rod in test; Less linkages than in conventional bending test setup such as three-point or four-point bending; Target the failure mode relevant to the fracture of spent fuel rod in transportation by focusing on pure bending; The continuous calibrations and modifications resulted in the third generation (3G) U-frame testing setup. Rigid arms are split along the LBB axis at rod sample ends. For each arm, this results in a large arm body and an end piece. Mating halves of bushings were modified into two V-shaped surfaces on which linear roller bearings (LRB) are embedded. The rod specimen is installed into the test fixture through opening and closing slide end-pieces. The 3G apparently has addressed major issues of setup identified in the previous stage and been proven to be eligible to be further pursued in this project. On the other hand, the purchase of universal testing machine or Bose dual LM2 TB was completed and the testing system was delivered to ORNL in August 2012. The preliminary confirmation of the system and on-site training were given by Bose field engineer and regional manager on 8/1-8/2/2012. The calibration of Bose testing system has been performed by ORNL because the integration of ORNL setup into the Bose TestBench occurred after the installation. Major challenge with this process arose from two aspects: 1) the load control involves two load cells, and 2) U-frame setup itself is a non-standard specimen. ORNL has been able to implement the load control through Cycle Indirect along with pinning the U-frame setup. Two meetings with ORNL hot-cell group (November 2012 and January 2013) were held to discuss the potential issues with both epoxy mounting of rigid sleeve and U-frame setup. Many suggestions were provided to make the procedure friendlier to the manipulator in hot cell. Addressing of these suggestions resulted in another cycle of modifications of both vise mold and setup. The initial meeting with ORNL I&C group occurred in November 2012 with regard to the Bose cable modification and design of central panel to integrate the cables and wires. The first round of cable modification and central panel fabrication was completed in February 2012. The testing with the modified cables exhibited substantial noises and the testing system was not shown to be stable. It was believed the cross talk was responsible to the noise, and a central panel with a better grounding and shielding was highly recommended. The central panel has been re-designed and fabricated in March 2013. In the subsequent period, the ORNL made substantial effort to debug the noises with the load cell channel, and to resolve the noises and nonlinearity with RDP LVDTs related to the integration of RDP LVDTs to Bose system. At the same time, ORNL has completed the verification tests of Bose test system, including cycle tests under reversal bending in load control, bending tests under monotonic load, and cycle test

Wang, Jy-An John [ORNL; Wang, Hong [ORNL; Cox, Thomas S [ORNL; Baldwin, Charles A [ORNL; Bevard, Bruce Balkcom [ORNL

2013-08-01T23:59:59.000Z

268

NAVAL POSTGRADUAm SCHOOL Monterey, California  

E-Print Network [OSTI]

NAVAL POSTGRADUAm SCHOOL Monterey, California A WHOLESALE LEVEL CONSUMABLE ITEM DEMAND PATI TYPE AND DATES COVERED Master's Thesis 4. TITLE AND SUBTITLE A WHOLESALE LEVEL CONSUMABLE DEMAND is unlimited. A Wholesale Level Consumable Item Inventory Model for Non-Stationary Demand Patterns Glenn C

269

HANSF 1.3 Users Manual FAI/98-40-R2 Hanford Spent Nuclear Fuel (SNF) Safety Analysis Model [SEC 1 and 2  

SciTech Connect (OSTI)

The HANSF analysis tool is an integrated model considering phenomena inside a multi-canister overpack (MCO) spent nuclear fuel container such as fuel oxidation, convective and radiative heat transfer, and the potential for fission product release. This manual reflects the HANSF version 1.3.2, a revised version of 1.3.1. HANSF 1.3.2 was written to correct minor errors and to allow modeling of condensate flow on the MCO inner surface. HANSF 1.3.2 is intended for use on personal computers such as IBM-compatible machines with Intel processors running under Lahey TI or digital Visual FORTRAN, Version 6.0, but this does not preclude operation in other environments.

DUNCAN, D.R.

1999-10-07T23:59:59.000Z

270

U.S. Environmental Protection Agency Clear Air Act notice of construction for the spent nuclear fuel project - Cold Vaccum Drying Facility, project W-441  

SciTech Connect (OSTI)

This document provides information regarding the source and the estimated quantity of potential airborne radionuclide emissions resulting from the operation of the Cold Vacuum Drying (CVD) Facility. The construction of the CVD Facility is scheduled to commence on or about December 1996, and will be completed when the process begins operation. This document serves as a Notice of Construction (NOC) pursuant to the requirements of 40 Code of Federal Regulations (CFR) 61 for the CVD Facility. About 80 percent of the U.S. Department of Energy`s spent nuclear fuel (SNF) inventory is stored under water in the Hanford Site K Basins. Spent nuclear fuel in the K West Basin is contained in closed canisters, while the SNF in the K East Basin is in open canisters, which allow release of corrosion products to the K East Basin water. Storage of the current inventory in the K Basins was originally intended to be on an as-needed basis to sustain operation of the N Reactor while the Plutonium-Uranium Extraction (PUREX) Plant was refurbished and restarted. The decision in December 1992 to deactivate the PURF-X Plant left approximately 2,100 MT (2,300 tons) of uranium as part of the N Reactor SNF in the K Basins with no means for near-term removal and processing. The CVD Facility will be constructed in the 100 Area northwest of the 190 K West Building, which is in close proximity to the K East and K West Basins (Figures 1 and 08572). The CVD Facility will consist of five processing bays, with four of the bays fully equipped with processing equipment and the fifth bay configured as an open spare bay. The CVD Facility will have a support area consisting of a control room, change rooms, and other functions required to support operations.

Turnbaugh, J.E.

1996-11-25T23:59:59.000Z

271

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

Broader source: Energy.gov [DOE]

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

272

Probability of spent fuel transportation accidents  

SciTech Connect (OSTI)

The transported volume of spent fuel, incident/accident experience and accident environment probabilities were reviewed in order to provide an estimate of spent fuel accident probabilities. In particular, the accident review assessed the accident experience for large casks of the type that could transport spent (irradiated) nuclear fuel. This review determined that since 1971, the beginning of official US Department of Transportation record keeping for accidents/incidents, there has been one spent fuel transportation accident. This information, coupled with estimated annual shipping volumes for spent fuel, indicated an estimated annual probability of a spent fuel transport accident of 5 x 10/sup -7/ spent fuel accidents per mile. This is consistent with ordinary truck accident rates. A comparison of accident environments and regulatory test environments suggests that the probability of truck accidents exceeding regulatory test for impact is approximately 10/sup -9//mile.

McClure, J. D.

1981-07-01T23:59:59.000Z

273

Intermodal transportation of spent fuel  

SciTech Connect (OSTI)

Concepts for transportation of spent fuel in rail casks from nuclear power plant sites with no rail service are under consideration by the US Department of Energy in the Commercial Spent Fuel Management program at the Pacific Northwest Laboratory. This report identifies and evaluates three alternative systems for intermodal transfer of spent fuel: heavy-haul truck to rail, barge to rail, and barge to heavy-haul truck. This report concludes that, with some modifications and provisions for new equipment, existing rail and marine systems can provide a transportation base for the intermodal transfer of spent fuel to federal interim storage facilities. Some needed land transportation support and loading and unloading equipment does not currently exist. There are insufficient shipping casks available at this time, but the industrial capability to meet projected needs appears adequate.

Elder, H.K.

1983-09-01T23:59:59.000Z

274

Successful Deployment of System for the Storage and Retrieval of Spent/Used Nuclear Fuel from Hanford K-West Fuel Storage Basin-13051  

SciTech Connect (OSTI)

In 2012, a system was deployed to remove, transport, and interim store chemically reactive and highly radioactive sludge material from the Hanford Site's 105-K West Fuel Storage Basin that will be managed as spent/used nuclear fuel. The Knockout Pot (KOP) sludge in the 105-K West Basin was a legacy issue resulting from the spent nuclear fuel (SNF) washing process applied to 2200 metric tons of highly degraded fuel elements following long-term underwater storage. The washing process removed uranium metal and other non-uranium constituents that could pass through a screen with 0.25-inch openings; larger pieces are, by definition, SNF or fuel scrap. When originally retrieved, KOP sludge contained pieces of degraded uranium fuel ranging from 600 microns (?m) to 6350 ?m mixed with inert material such as aluminum hydroxide, aluminum wire, and graphite in the same size range. In 2011, a system was developed, tested, successfully deployed and operated to pre-treat KOP sludge as part of 105-K West Basin cleanup. The pretreatment process successfully removed the vast majority of inert material from the KOP sludge stream and reduced the remaining volume of material by approximately 65 percent, down to approximately 50 liters of material requiring management as used fuel. The removal of inert material resulted in significant waste minimization and project cost savings because of the reduced number of transportation/storage containers and improvement in worker safety. The improvement in worker safety is a result of shorter operating times and reduced number of remote handled shipments to the site fuel storage facility. Additionally in 2011, technology development, final design, and cold testing was completed on the system to be used in processing and packaging the remaining KOP material for removal from the basin in much the same manner spent fuel was removed. This system was deployed and successfully operated from June through September 2012, to remove and package the last of the SNF fragments from the 105-K West Basin, ending the long and complex history of the KOP sludge materials. The planning and execution of this project demonstrated how the graded application of DOE O 413.3B, Program and Project Management for the Acquisition of Capital Assets[1], DOE-STD-1189-2008, Integration of Safety into the Design Process[2], and DOE G 413.3-4, U.S. Department of Energy Technology Readiness Assessment Guide[3], can positively affect the outcome of project implementation in the DOE Complex. Provided herein are relevant information, ideas, and tools for use by other projects facing similar issues in managing high risk waste streams in a complex regulatory environment. Positive aspects are also provided of appropriate time spent in detailed planning, design, and testing in non-hazardous environments to reduce project risks in both cost and safety performance, as well as improving confidence in meeting project goals through predictable and reliable performance. (authors)

Quintero, Roger; Smith, Sahid [U.S. Department of Energy, Richland Operations Office, Richland, Washington 99352 (United States)] [U.S. Department of Energy, Richland Operations Office, Richland, Washington 99352 (United States); Blackford, Leonard Ty; Johnson, Mike W.; Raymond, Richard; Sullivan, Neal; Sloughter, Jim [CH2M HILL Plateau Remediation Company, Richland, Washington 99352 (United States)] [CH2M HILL Plateau Remediation Company, Richland, Washington 99352 (United States)

2013-07-01T23:59:59.000Z

275

CORROSION OF ALUMINUM CLAD SPENT NUCLEAR FUEL IN THE 70 TON CASK DURING TRANSFER FROM L AREA TO H-CANYON  

SciTech Connect (OSTI)

Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 260 {degrees}C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 {degrees}C. These losses are not expected to impact the overall confinement function of the aluminum cladding.

Mickalonis, J.

2014-06-01T23:59:59.000Z

276

THE MULTI-ISOTOPE PROCESS (MIP) MONITOR: A NEAR-REAL-TIME, NON-DESTRUCTIVE, INDICATOR OF SPENT NUCLEAR FUEL REPROCESSING CONDITIONS  

SciTech Connect (OSTI)

Researchers from Pacific Northwest National Laboratory and The Ohio State University are working to develop a system for monitoring spent nuclear fuel reprocessing facilities on-line, non-destructively, and in near-real-time. This method, known as the Multi-Isotope Process (MIP) Monitor, is based upon the measurement of distribution patterns of a suite of indicator (radioactive) isotopes present within product and waste streams of a nuclear reprocessing facility. Signatures from these indicator isotopes are monitored on-line by gamma spectrometry and compared, in near-real-time, to patterns representing "normal" process conditions using multivariate pattern recognition software. By targeting gamma-emitting indicator isotopes, the MIP Monitor approach is compatible with the use of small, portable, high-resolution gamma detectors that may be easily deployed throughout an existing facility. In addition, utilization of a suite of radio-elements, including ones with multiple oxidation states, increases the likelihood that attempts to divert material via process manipulation would be detected. Proof-of-principle modeling exercises simulating changes in acid strength have been completed and the results are promising. Laboratory validation is currently under way and significant results are available. The latest experimental results, along with an overview of the method will be presented.

Schwantes, Jon M.; Orton, Christopher R.; Fraga, Carlos G.; Douglas, Matthew; Christensen, Richard

2010-05-07T23:59:59.000Z

277

MONITORING SPENT NUCLEAR FUEL REPROCESSING CONDITIONS NON-DESTRUCTIVELY AND IN NEAR-REAL-TIME USING THE MULTI-ISOTOPE PROCESS (MIP) MONITOR  

SciTech Connect (OSTI)

Researchers from Pacific Northwest National Laboratory and The Ohio State University are working to develop a system for monitoring spent nuclear fuel reprocessing facilities on-line, nondestructively, and in near-real-time. This method, known as the Multi-Isotope Process (MIP) Monitor, is based upon the measurement of distribution patterns of a suite of indicator (radioactive) isotopes present within product and waste streams of a nuclear reprocessing facility. Signatures from these indicator isotopes are monitored on-line by gamma spectrometry and compared, in near-real-time, to patterns representing "normal" process conditions using multivariate pattern recognition software. By targeting gamma-emitting indicator isotopes, the MIP Monitor approach is compatible with the use of small, portable, high-resolution gamma detectors that may be easily deployed throughout an existing facility. In addition, utilization of a suite of radio-elements, including ones with multiple oxidation states, increases the likelihood that attempts to divert material via process manipulation would be detected. Proof-of-principle modeling exercises simulating changes in acid strength have been completed and the results are promising. Laboratory testing is currently under way and significant results are available. Recent experimental results, along with an overview of the method are presented.

Orton, Christopher R.; Fraga, Carlos G.; Douglas, Matthew; Christensen, Richard; Schwantes, Jon M.

2010-05-07T23:59:59.000Z

278

Neptunium - Uranium - Plutonium Co-Extraction in TBP-based Solvent Extraction Processes for Spent Nuclear Fuel Recycling  

SciTech Connect (OSTI)

The US, through the Global Nuclear Energy Partnership, is currently engaged in efforts aimed at closing the nuclear fuel cycle. Neptunium behavior is important to understand for transuranic recycling because of its complex oxidation chemistry. The Pacific Northwest National Laboratory is investigating neptunium oxidation chemistry in the context of the PUREX process. Neptunium extraction in the PUREX process relies on maintaining either IV or V oxidation states. Qualitative conversion of neptunium(V) to neptunium(VI) was achieved within 5 hours in 6 M nitric acid at 95 deg. C. However, the VI state was not maintained during a batch contact test simulating the PUREX process and neptunium reduced to the V state, rendering it inextractable. Vanadium(V) was found to be effective in maintaining neptunium(VI) by adding it to a simulated irradiated nuclear fuel feed in 6 M nitric acid and to the scrub acid in the batch contact simulation of the PUREX process. Computer simulations of the PUREX process with a typical irradiated nuclear fuel in 6 M nitric acid as feed indicated little impact of the higher acid concentration on the behavior of fission products of moderate extractability. We plan to perform countercurrent tests of this modified PUREX process in the near future. (authors)

Arm, S.T.; Abrefah, J.; Lumetta, G.J.; Sinkov, S.I. [Battelle PNWD, Pacific Northwest National Laboratory, 902 Battelle Boulevard, PO Box 999, Richland, Washington, 99352 (United States)

2007-07-01T23:59:59.000Z

279

The Planning, Licensing, Modifications, and Use of a Russian Vessel for Shipping Spent Nuclear Fuel by Sea in Support of the DOE RRRFR Program  

SciTech Connect (OSTI)

The Russian Research Reactor Fuel Return (RRRFR) Program, under the U.S. Department of Energys Global Threat Reduction Initiative, began returning Russian-supplied high-enriched uranium (HEU) spent nuclear fuel (SNF), stored at Russian-designed research reactors throughout the world, to Russia in January 2006. During the first years of making HEU SNF shipments, it became clear that the modes of transportation needed to be expanded from highway and railroad to include sea and air to meet the extremely aggressive commitment of completing the first series of shipments by the end of 2010. The first shipment using sea transport was made in October 2008 and used a non-Russian flagged vessel. The Russian government reluctantly allowed a one-time use of the foreign-owned vessel into their highly secured seaport, with the understanding that any future shipments would be made using a vessel owned and operated by a Russian company. ASPOL-Baltic of St. Petersburg, Russia, owns and operates a small fleet of vessels and has a history of shipping nuclear materials. ASPOL-Baltics vessels were licensed for shipping nuclear materials; however, they were not licensed to transport SNF materials. After a thorough review of ASPOL Baltics capabilities and detailed negotiations, it was agreed that a contract would be let with ASPOL-Baltic to license and refit their MCL Trader vessel for hauling SNF in support of the RRRFR Program. This effort was funded through a contract between the RRRFR Program, Idaho National Laboratory, and Radioactive Waste Management Plant of Swierk, Poland. This paper discusses planning, Russian and international maritime regulations and requirements, Russian authorities reviews and approvals, licensing, design, and modifications made to the vessel in preparation for SNF shipments. A brief summary of actual shipments using this vessel, experiences, and lessons learned also are described.

Michael Tyacke; Dr. Igor Bolshinsky; Wlodzimierz Tomczak; Sergey Naletov; Oleg Pichugin

2001-10-01T23:59:59.000Z

280

INL Site Portion of the April 1995 Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Mamagement Programmatic Final Environmental Impact Statement  

SciTech Connect (OSTI)

In April 1995, the Department of Energy (DOE) and the Department of the Navy, as a cooperating agency, issued the Programmatic Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental Restoration and Waste Management Programs Final Environmental Impact Statement (1995 EIS). The 1995 EIS analyzed alternatives for managing The Department's existing and reasonably foreseeable inventories of spent nuclear fuel through the year 2035. It also included a detailed analysis of environmental restoration and waste management activities at the Idaho National Engineering and Environmental Laboratory (INEEL). The analysis supported facility-specific decisions regarding new, continued, or planned environmental restoration and waste management operations. The Record of Decision (ROD) was signed in June 1995 and amended in February 1996. It documented a number of projects or activities that would be implemented as a result of decisions regarding INL Site operations. In addition to the decisions that were made, decisions on a number of projects were deferred or projects have been canceled. DOE National Environmental Policy Act (NEPA) implementing procedures (found in 10 CFR Part 1 021.330(d)) require that a Supplement Analysis of site-wide EISs be done every five years to determine whether the site-wide EIS remains adequate. While the 1995 EIS was not a true site-wide EIS in that several programs were not included, most notably reactor operations, this method was used to evaluate the adequacy of the 1995 EIS. The decision to perform a Supplement Analysis was supported by the multi-program aspect of the 1995 EIS in conjunction with the spirit of the requirement for periodic review. The purpose of the SA is to determine if there have been changes in the basis upon which an EIS was prepared. This provides input for an evaluation of the continued adequacy of the EIS in light of those changes (i.e., whether there are substantial changes in the proposed action, significant new circumstances, or new information relevant to environmental concerns). This is not to question the previous analysis or decisions based on that analysis, but whether the environmental impact analyses are still adequate in light of programmatic changes. In addition, the information for each of the projects for which decisions were deferred in the ROD needs to be reviewed to determine if decisions can be made or if any additional NEP A analysis needs to be completed. The Supplement Analysis is required to contain sufficient information for DOE to determine whether (1) an existing EIS should be supplemented, (2) a new EIS should be prepared, or (3) no further NEP A documentation is required.

N /A

2005-06-30T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" from the National Library of EnergyBeta (NLEBeta).
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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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281

Optimization of combined delayed neutron and differential die-away prompt neutron signal detection for characterization of spent nuclear fuel assemblies  

SciTech Connect (OSTI)

The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy (DOE) has funded multiple laboratories and universities to develop a means to accurately quantify the Plutonium (Pu) mass in spent nuclear fuel assemblies and ways to also detect potential diversion of fuel pins. Delayed Neutron (DN) counting provides a signature somewhat more sensitive to {sup 235}U than Pu while Differential Die-Away (DDA) is complementary in that it has greater sensitivity to Pu. The two methods can, with care, be combined into a single instrument which also provides passive neutron information. Individually the techniques cannot robustly quantify the Pu content but coupled together the information content in the signatures enables Pu quantification separate to the total fissile content. The challenge of merging DN and DDA, prompt neutron (PN) signal, capabilities in the same design is the focus of this paper. Other possibilities also suggest themselves, such as a direct measurement of the reactivity (multiplication) by either the boost in signal obtained during the active interrogation itself or by the extension of the die-away profile. In an early study, conceptual designs have been modeled using a neutron detector comprising fission chambers or 3He proportional counters and a {approx}14 MeV neutron Deuterium-Tritium (DT) generator as the interrogation source. Modeling was performed using the radiation transport code Monte Carlo N-Particles eXtended (MCNPX). Building on this foundation, the present paper quantifies the capability of a new design using an array of {sup 3}He detectors together with fission chambers to optimize both DN and PN detections and active characterization, respectively. This new design was created in order to minimize fission in {sup 238}U (a nuisance DN emitter), to use a realistic neutron generator, to reduce the cost and to achieve near spatial interrogation and detection of the DN and PN, important for detection of diversion, all within the constraints of a single practical instrument. Both DN and PN detections are active techniques using the signal from the most prominent fissile isotopes of spent nuclear fuel that respond the best to a slow neutron interrogation, {sup 235}U, {sup 239}U and {sup 241}PU. The performance is characterized against a library of 64 assemblies and 40 diversion scenarios at different burnup (BU), cooling-time (CT) and initial enrichment (IE) in fresh water.

Blanc, Pauline [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Croft, Stephen [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Swinhoe, M [Los Alamos National Laboratory; Lee, T [NON LANL

2010-12-02T23:59:59.000Z

282

A Transportation Risk Assessment Tool for Analyzing the Transport of Spent Nuclear Fuel and High-Level Radioactive Waste to the Proposed Yucca Mountain Repository  

SciTech Connect (OSTI)

The Yucca Mountain Transportation Database was developed as a data management tool for assembling and integrating data from multiple sources to compile the potential transportation impacts presented in the Draft Environmental Impact Statement for a Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive Waste at Yucca Mountain, Nye County, Nevada (DEIS). The database uses the results from existing models and codes such as RADTRAN, RISKIND, INTERLINE, and HIGHWAY to estimate transportation-related impacts of transporting spent nuclear fuel and high-level radioactive waste from commercial reactors and U. S. Department of Energy (DOE) facilities to Yucca Mountain. The source tables in the database are compendiums of information from many diverse sources including: radionuclide quantities for each waste type; route and route characteristics for rail, legal-weight truck, heavy haul. truck, and barge transport options; state-specific accident and fatality rates for routes selected for analysis; packaging and shipment data by waste type; unit risk factors; the complex behavior of the packaged waste forms in severe transport accidents; and the effects of exposure to radiation or the isotopic specific effects of radionclides should they be released in severe transportation accidents. The database works together with the codes RADTRAN (Neuhauser, et al, 1994) and RISKlND (Yuan, et al, 1995) to calculate incident-free dose and accident risk. For the incident-free transportation scenario, the database uses RADTRAN and RISKIND-generated data to calculate doses to offlink populations, onlink populations, people at stops, crews, inspectors, workers at intermodal transfer stations, guards at overnight stops, and escorts, as well as non-radioactive pollution health effects. For accident scenarios, the database uses RADTRAN-generated data to calculate dose risks based on ingestion, inhalation, resuspension, immersion (cloudshine), and groundshine as well as non-radioactive traffic fatalities. The Yucca Mountain EIS Transportation Database was developed using Microsoft Access 97{trademark} software and the Microsoft Windows NT{trademark} operating system. The database consists of tables for storing data, forms for selecting data for querying, and queries for retrieving the data in a predefined format. Database queries retrieve records based on input parameters and are used to calculate incident-free and accident doses using unit risk factors obtained from RADTRAN results. The next section briefly provides some background that led to the development of the database approach used in preparing the Yucca Mountain DEIS. Subsequent sections provide additional details on the database structure and types of impacts calculated using the database.

Ralph Best; T. Winnard; S. Ross; R. Best

2001-08-17T23:59:59.000Z

283

Shielding and criticality analyses of phase I reference truck and rail cask designs for spent nuclear fuel  

SciTech Connect (OSTI)

Results are presented herein to determine the adequacy with respect to shielding regulations of reference designs for a truck cask containing 2 PWR or 5 BWR assemblies of standard burnup (45 GWd/MTU for PWR, 40 GWd/MTU for BWR) and 1 PWR assembly with extended burnup (55 GWd/MTU). The study also includes reference and modified rail cask designs with projected payloads of 8, 10, or 12 PWR assemblies. The burnup/age trends are analyzed in one dimension for both Pb and depleted uranium (DU) gamma-ray shields. The results of the two-dimensional shielding analysis uphold the one-dimensional results as being an appropriate means of studying the burnup/age trends for the truck cask. These results show that the reference design for the Pb-shield truck cask is inadequate for all cases considered, while the DU-shield truck cask is capable of carrying the desired payloads. The one-dimensional shielding analysis results for the reference Pb and DU rail casks indicate substantial margins exist in the side doses for reasonable burnup/age combinations. For a Pb-cask configuration, margins exist primarily for long-cooled (15 years) fuel. For the modified Pb and DU rail casks, the 2-m dose rates offer substantial margins below the regulatory limits for all burnup values considered provided the spent fuel has cooled for {>=}10 years. The modified Pb and DU casks yield essentially identical results and, hence, could be considered equivalent from a shielding perspective. The criticality analyses that were performed indicate that a truck basket can be designed to provide an adequate subcritical margin for 2 PWR assemblies enriched to 5 wt%. While the 10- and 12- assembly rail cask designs are very close to the regulatory limit of 0.95 for k{sub eff}, after accounting for a 0.01 {Delta}k bias and 2 standard deviations, the limit is exceeded by about 3%. It is believed that a combination of decreased enrichments and/or increased water gaps should allow these baskets to be acceptable.

Broadhead, B.L.; Childs, R.L.; Parks, C.V.

1996-03-01T23:59:59.000Z

284

The Environmental Protection Agency's Safety Standards for Disposal of Spent Nuclear Fuel: Potential Path Forward in Response to the Report of the Blue Ribbon Commission on America's Nuclear Future - 13388  

SciTech Connect (OSTI)

Following the decision to withdraw the Yucca Mountain license application, the Department of Energy created a Blue Ribbon Commission (BRC) on America's Nuclear Future, tasked with recommending a national strategy to manage the back end of the nuclear fuel cycle. The BRC issued its final report in January 2012, with recommendations covering transportation, storage and disposal of spent nuclear fuel (SNF); potential reprocessing; and supporting institutional measures. The BRC recommendations on disposal of SNF and high-level waste (HLW) are relevant to the U.S. Environmental Protection Agency (EPA), which shares regulatory responsibility with the Nuclear Regulatory Commission (NRC): EPA issues 'generally applicable' performance standards for disposal repositories, which are then implemented in licensing. For disposal, the BRC endorses developing one or more geological repositories, with siting based on an approach that is adaptive, staged and consent-based. The BRC recommends that EPA and NRC work cooperatively to issue generic disposal standards-applying equally to all sites-early in any siting process. EPA previously issued generic disposal standards that apply to all sites other than Yucca Mountain. However, the BRC concluded that the existing regulations should be revisited and revised. The BRC proposes a number of general principles to guide the development of future regulations. EPA continues to review the BRC report and to assess the implications for Agency action, including potential regulatory issues and considerations if EPA develops new or revised generic disposal standards. This review also involves preparatory activities to define potential process and public engagement approaches. (authors)

Forinash, Betsy; Schultheisz, Daniel; Peake, Tom [U.S. Environmental Protection Agency, Radiation Protection Division (United States)] [U.S. Environmental Protection Agency, Radiation Protection Division (United States)

2013-07-01T23:59:59.000Z

285

U.S. Naval war College Application | Department of Energy  

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

College Application U.S. Naval war College Application U.S. Naval War College Civilian Student Application. Naval War College Application More Documents & Publications DOE F 471.1...

286

THE TESTING OF COMMERCIALLY AVAILABLE ENGINEERING AND PLANT SCALE ANNULAR CENTRIFUGAL CONTACTORS FOR THE PROCESSING OF SPENT NUCLEAR FUEL  

SciTech Connect (OSTI)

Annular centrifugal contactors are being evaluated for process scale solvent extraction operations in support of United State Advanced Fuel Cycle Initiative goals. These contactors have the potential for high stage efficiency if properly employed and optimized for the application. Commercially available centrifugal contactors are being tested at the Idaho National Laboratory to support this program. Hydraulic performance and mass transfer efficiency have been measured for portions of an advanced nuclear fuel cycle using 5-cm diameter annular centrifugal contactors. Advanced features, including low mix sleeves and clean-in-place rotors, have also been evaluated in 5-cm and 12.5-cm contactors.

Jack D. Law; David Meikrantz; Troy Garn; Nick Mann; Scott Herbst

2006-10-01T23:59:59.000Z

287

Improvement of operational safety of dual-purpose transport packaging set for naval SNF in storage  

SciTech Connect (OSTI)

Available in abstract form only. Full text of publication follows: In recent ten years a new technology of management of irradiated nuclear fuel (SNF) at the final stage of fuel cycle has been intensely developing on a basis of a new type of casks used for interim storage of SNF and subsequent transportation therein to the place of processing, further storage or final disposal. This technology stems from the concept of a protective cask which provides preservation of its content (SNF) and fulfillment of all other safety requirements for storage and transportation of SNF. Radiation protection against emissions and non-distribution of activity outside the cask is ensured by physical barriers, i.e. all-metal or composite body, shells, inner cavities for irradiated fuel assemblies (SFA), lids with sealing systems. Residual heat release of SFA is discharged to the environment by natural way: through emission and convection of surrounding air. By now more than 100 dual purpose packaging sets TUK-108/1 are in operation in the mode of interim storage and transportation of SNF from decommissioned nuclear powered submarines (NPS). In accordance with certificate, spent fuel is stored in TUK-108/1 on the premises of plants involved in NPS dismantlement for 2 years, whereupon it is transported for processing to PO Mayak. At one Far Eastern plant Zvezda involved in NPS dismantlement there arose a complicated situation due to necessity to extend period of storage of SNF in TUK- 108/1. To ensure safety over a longer period of storage of SNF in TUK-108/1 it is essential to modify conditions of storage by removing of residual water and filling the inner cavity of the cask with an inert gas. Within implementation of the international 1.1- 2 project Development of drying technology for the cask TUK-108/1 intended for naval SNF under the Program, there has been developed the technology of preparation of the cask for long-term storage of SNF in TUK-108/1, the design of a mobile TUK-108/1 drying facility; a pilot facility has been manufactured. This report describes key issues of cask drying technology, justification of terms of dry storage of naval SNF in no-108/1, design features of the mobile drying facility, results of tests of the pilot facility at the Far Eastern plant Zvezda. (authors)

Guskov, Vladimir; Korotkov, Gennady [JSC 'KBSM' (Russian Federation); Barnes, Ella [US Environmental Protection Agency - EPA (United States); Snipes, Randy [Oak Ridge National Laboratory - ORNL, 1 Bethel Valley Rd, Oak Ridge, TN 37830 (United States)

2007-07-01T23:59:59.000Z

288

Categorical Exclusion Determinations: Naval Nuclear Propulsion Program |  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious Rank EERE:YearRound-Up fromDepartmentTie Ltd:JuneNovember 26,Energy Service Center

289

About Naval Reactors | National Nuclear Security Administration  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc Documentation RUCProductstwrmrAre the Effects of GlobalASCRAbigailAboutquestions from ourNNSA

290

naval reactors | National Nuclear Security Administration  

National Nuclear Security Administration (NNSA)

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

291

Evaluation of Options for Permanent Geologic Disposal of Spent...  

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

policy decisions regarding strategies for the management and permanent disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) in the United States requiring...

292

President Reagan Calls for a National Spent Fuel Storage Facility...  

National Nuclear Security Administration (NNSA)

Spent Fuel Storage Facility Washington, DC The Reagan Administration announces a nuclear energy policy that anticipates the establishment of a facility for the storage of...

293

2013 Federal Energy and Water Management Award Winner Naval Sea...  

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

Naval Sea Systems Command 2013 Federal Energy and Water Management Award Winner Naval Sea Systems Command fewm13nswcphiladelphiahighres.pdf fewm13nswcphiladelphia.pdf More...

294

Naval Petroleum Reserve No. 3 Disposition Decision Analysis and...  

Energy Savers [EERE]

Naval Petroleum Reserve No. 3 Disposition Decision Analysis and Timeline Naval Petroleum Reserve No. 3 Disposition Decision Analysis and Timeline This Report to Congress provides a...

295

Final LDRD report : nanoscale mechanisms in advanced aging of materials during storage of spent %22high burnup%22 nuclear fuel.  

SciTech Connect (OSTI)

We present the results of a three-year LDRD project focused on understanding microstructural evolution and related property changes in Zr-based nuclear cladding materials towards the development of high fidelity predictive simulations for long term dry storage. Experiments and modeling efforts have focused on the effects of hydride formation and accumulation of irradiation defects. Key results include: determination of the influence of composition and defect structures on hydride formation; measurement of the electrochemical property differences between hydride and parent material for understanding and predicting corrosion resistance; in situ environmental transmission electron microscope observation of hydride formation; development of a predictive simulation for mechanical property changes as a function of irradiation dose; novel test method development for microtensile testing of ionirradiated material to simulate the effect of neutron irradiation on mechanical properties; and successful demonstration of an Idaho National Labs-based sample preparation and shipping method for subsequent Sandia-based analysis of post-reactor cladding.

Clark, Blythe G.; Rajasekhara, Shreyas; Enos, David George; Dingreville, Remi Philippe Michel; Doyle, Barney Lee; Hattar, Khalid Mikhiel; Weiner, Ruth F.

2013-09-01T23:59:59.000Z

296

A methodology to quantify the release of spent nuclear fuel from dry casks during security-related scenarios.  

SciTech Connect (OSTI)

Assessing the risk to the public and the environment from a release of radioactive material produced by accidental or purposeful forces/environments is an important aspect of the regulatory process in many facets of the nuclear industry. In particular, the transport and storage of radioactive materials is of particular concern to the public, especially with regard to potential sabotage acts that might be undertaken by terror groups to cause injuries, panic, and/or economic consequences to a nation. For many such postulated attacks, no breach in the robust cask or storage module containment is expected to occur. However, there exists evidence that some hypothetical attack modes can penetrate and cause a release of radioactive material. This report is intended as an unclassified overview of the methodology for release estimation as well as a guide to useful resource data from unclassified sources and relevant analysis methods for the estimation process.

Durbin, Samuel G.; Luna, Robert Earl

2013-11-01T23:59:59.000Z

297

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

Broader source: Energy.gov [DOE]

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

298

A TRANSPORTATION RISK ASSESSMENT TOOL FOR ANALYZING THE TRANSPORT OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE WASTE TO THE PROPOSED YUCCA MOUNTAIN REPOSITORY  

SciTech Connect (OSTI)

The Yucca Mountain Draft Environmental Impact Statement (DEIS) analysis addressed the potential for transporting spent nuclear fuel and high-level radioactive waste from 77 origins for 34 types of spent fuel and high-level radioactive waste, 49,914 legal weight truck shipments, and 10,911 rail shipments. The analysis evaluated transportation over 59,250 unique shipment links for travel outside Nevada (shipment segments in urban, suburban or rural zones by state), and 22,611 links in Nevada. In addition, the analysis modeled the behavior of 41 isotopes, 1091 source terms, and used 8850 food transfer factors (distinct factors by isotope for each state). The analysis also used mode-specific accident rates for legal weight truck, rail, and heavy haul truck by state, and barge by waterway. This complex mix of data and information required an innovative approach to assess the transportation impacts. The approach employed a Microsoft{reg_sign} Access database tool that incorporated data from many sources, including unit risk factors calculated using the RADTRAN IV transportation risk assessment computer program. Using Microsoft{reg_sign} Access, the analysts organized data (such as state-specific accident and fatality rates) into tables and developed queries to obtain the overall transportation impacts. Queries are instructions to the database describing how to use data contained in the database tables. While a query might be applied to thousands of table entries, there is only one sequence of queries that is used to calculate a particular transportation impact. For example, the incident-free dose to off-link populations in a state is calculated by a query that uses route segment lengths for each route in a state that could be used by shipments, populations for each segment, number of shipments on each segment, and an incident-free unit risk factor calculated using RADTRAN IV. In addition to providing a method for using large volumes of data in the calculations, the queries provide a straight-forward means used to verify results. Another advantage of using the MS Access database was the ability to develop query hierarchies using nested queries. Calculations were broken into a series of steps, each step represented by a query. For example, the first query might calculate the number of shipment kilometers traveled through urban, rural and suburban zones for all states. Subsequent queries could join the shipment kilometers query results with another table containing unit risk factors calculated using RADTRAN IV to produce radiological impacts. Through the use of queries, impacts by origin, mode, fuel type or many other parameters can be obtained. The paper will show both the flexibility of the assessment tool and the ease it provides for verifying results.

NA

2001-02-15T23:59:59.000Z

299

Management of Legacy Spent Nuclear Fuel Wastes at the Chalk River Laboratories: The Challenges and Innovative Solutions Implemented - 13301  

SciTech Connect (OSTI)

AECL's Fuel Packaging and Storage (FPS) Project was initiated in 2004 to retrieve, transfer, and stabilize an identified inventory of degraded research reactor fuel that had been emplaced within in-ground 'Tile Hole' structures in Chalk River Laboratories' Waste Management Area in the 1950's and 60's. Ongoing monitoring of the legacy fuel storage conditions had identified that moisture present in the storage structures had contributed to corrosion of both the fuel and the storage containers. This prompted the initiation of the FPS Project which has as its objective to design, construct, and commission equipment and systems that would allow for the ongoing safe storage of this fuel until a final long-term management, or disposition, pathway was available. The FPS Project provides systems and technologies to retrieve and transfer the fuel from the Waste Management Area to a new facility that will repackage, dry, safely store and monitor the fuel for a period of 50 years. All equipment and the new storage facility are designed and constructed to meet the requirements for Class 1 Nuclear Facilities in Canada. (authors)

Schruder, Kristan [Atomic Energy of Canada Limited - Chalk River Laboratories, Chalk River, Ontario (Canada)] [Atomic Energy of Canada Limited - Chalk River Laboratories, Chalk River, Ontario (Canada); Goodwin, Derek [Rolls-Royce Civil Nuclear Canada Limited, 678 Neal Dr., Peterborough, Ontario (Canada)] [Rolls-Royce Civil Nuclear Canada Limited, 678 Neal Dr., Peterborough, Ontario (Canada)

2013-07-01T23:59:59.000Z

300

Spent fuel storage requirements 1993--2040  

SciTech Connect (OSTI)

Historical inventories of spent fuel are combined with U.S. Department of Energy (DOE) projections of future discharges from commercial nuclear reactors in the United States to provide estimates of spent fuel storage requirements through the year 2040. The needs are estimated for storage capacity beyond that presently available in the reactor storage pools. These estimates incorporate the maximum capacities within current and planned in-pool storage facilities and any planned transshipments of spent fuel to other reactors or facilities. Existing and future dry storage facilities are also discussed. The nuclear utilities provide historical data through December 1992 on the end of reactor life are based on the DOE/Energy Information Administration (EIA) estimates of future nuclear capacity, generation, and spent fuel discharges.

Not Available

1994-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" 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

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

SciTech Connect (OSTI)

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

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

1998-09-01T23:59:59.000Z

302

Volatilization of Fission Products from Metallic Melts in the Melt-Dilute Treatment Technology Development for Al-Based DOE Spent Nuclear Fuels  

SciTech Connect (OSTI)

The melt-dilute treatment technology is being developed to facilitate the ultimate disposition of highly enriched Al-Base DOE spent nuclear fuels in a geologic repository such as that proposed for Yucca Mountain. Currently, approximately 28 MTHM is expected to be returned to the Savannah River Site from domestic and foreign research reactors. The melt-dilute treatment technology will melt the fuel assemblies to reduce their volume and alloys them with depleted uranium to isotopically dilute the 235U concentration. The resulting alloy is cast into a form for long term geologic repository storage. Benefits accrued from the melt-dilute process include the potential for significant volume reduction; reduced criticality potential, and proliferation concerns. A critical technology element in the development of the melt-dilute process is the development of offgas system requirements. The volatilization of radioactive species during the melting stage of the process primarily constitutes the offgas in this process. Several of the species present following irradiation of a fuel assembly have been shown to be volatile or semi-volatile under reactor core melt-down conditions. Some of the key species that have previously been studied are krypton, iodine, and cesium. All of these species have been shown to volatilize during melting experiments however, the degree to which they are released is highly dependent upon atmosphere, fuel burnup, temperature, and fuel composition. With this in mind an analytical and experimental program has been undertaken to assess the volatility and capture of species under the melt-dilute operating conditions.

Adams, T.

1999-11-18T23:59:59.000Z

303

Analysis of a Nuclear Accident: Fission and Activation Product Releases from the Fukushima Daiichi Nuclear Facility as Remote Indicators of Source Identification, Extent of Release, and State of Damaged Spent Nuclear Fuel  

SciTech Connect (OSTI)

Measurements of several radionuclides within environmental samples taken from the Fukushima Daiichi nuclear facility and reported on the Tokyo Electric Power Company website following the recent tsunami-initiated catastrophe were evaluated for the purpose of identifying the source term, reconstructing the release mechanisms, and estimating the extent of the release. 136Cs/137Cs and 134Cs/137Cs ratios identified Units 1-3 as the major source of radioactive contamination to the surface soil close to the facility. A trend was observed between the fraction of the total core inventory released for a number of fission product isotopes and their corresponding Gibbs Free Energy of formation for the primary oxide form of the isotope, suggesting that release was dictated primarily by chemical volatility driven by temperature and reduction potential within the primary containment vessels of the vented reactors. The absence of any major fractionation beyond volatilization suggested all coolant had evaporated by the time of venting. High estimates for the fraction of the total inventory released of more volatile species (Te, Cs, I) indicated the damage to fuel bundles was likely extensive, minimizing any potential containment due to physical migration of these species through the fuel matrix and across the cladding wall. 238Pu/239,240Pu ratios close-in and at 30 km from the facility indicated that the damaged reactors were the major contributor of Pu to surface soil at the source but that this contribution likely decreased rapidly with distance from the facility. The fraction of the total Pu inventory released to the environment from venting units 1 and 3 was estimated to be ~0.003% based upon Pu/Cs isotope ratios relative to the within-reactor modeled inventory prior to venting and was consistent with an independent model evaluation that considered chemical volatility based upon measured fission product release trends. Significant volatile radionuclides within the spent fuel at the time of venting but not as yet observed and reported within environmental samples are suggested as potential analytes of concern for future environmental surveys around the site.

Schwantes, Jon M.; Orton, Christopher R.; Clark, Richard A.

2012-09-10T23:59:59.000Z

304

Sandia National Laboratories: Carderock Naval Surface War-fare Center  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the1 -the Mid-Infrared0Energy AdvancedEnergy Commission Linde,CapabilitiesCarderock Naval

305

A Technical Review of Non-Destructive Assay Research for the Characterization of Spent Nuclear Fuel Assemblies Being Conducted Under the US DOE NGSI - 11544  

E-Print Network [OSTI]

03715, Institute of Nuclear Materials Management 50th Annual04602, Institute of Nuclear Materials Management 51st AnnualInstitute of Nuclear Materials Management 51st Annual

Croft, S.

2012-01-01T23:59:59.000Z

306

Comparative naval architecture analysis of diesel submarines  

E-Print Network [OSTI]

Many comparative naval architecture analyses of surface ships have been performed, but few published comparative analyses of submarines exist. Of the several design concept papers, reports and studies that have been written ...

Torkelson, Kai Oscar

2005-01-01T23:59:59.000Z

307

Mitch S. Daugherty Nuclear Engineering and Planning Manager  

E-Print Network [OSTI]

Mitch S. Daugherty Nuclear Engineering and Planning Manager Naval Sea Systems Command, Norfolk Naval Shipyard Mitch Daugherty is the Nuclear Engineering and Planning Manager and the senior civilian. Daugherty also headed the Nuclear Refueling Division, the Nuclear Test Engineering Division, and the Nuclear

308

Shielding analysis for the 300 area light water reactor spent nuclear fuel within a modified multi-canister overpack canister in a modified multi-canister overpack cask  

SciTech Connect (OSTI)

Spent light water reactor fuel is to be moved out of the 324 Building. It is anticipated that intact fuel assemblies will be loaded in a modified Multi-Canister Overpack Canister, which in turn will be placed in an Overpack Transportation Cask. An estimate of gamma ray dose rates from a transportation cask is desired.

Gedeon, S.R.

1997-04-11T23:59:59.000Z

309

A Technical Review of Non-Destructive Assay Research for the Characterization of Spent Nuclear Fuel Assemblies Being Conducted Under the US DOE NGSI - 11544  

E-Print Network [OSTI]

NA-24), National Nuclear Security Administration, U.S.nuclear energy together with the emergence of alternative fuel cycles. Introduction Energy security

Croft, S.

2012-01-01T23:59:59.000Z

310

U.S. NAVAL ACADEMY COMPUTER SCIENCE DEPARTMENT  

E-Print Network [OSTI]

U.S. NAVAL ACADEMY COMPUTER SCIENCE DEPARTMENT TECHNICAL REPORT Mobile Vehicle Teleoperated Over Akin and Frederick L. Crabbe U.S. Naval Academy Computer Science Department 572M Holloway Rd, Stop 9F

Crabbe, Frederick

311

Inspection and Gamma-Ray Dose Rate Measurements of the Annulus of the VSC-17 Concrete Spent Nuclear Fuel Storage Cask  

SciTech Connect (OSTI)

The air cooling annulus of the Ventilated Storage Cask (VSC)-17 spent fuel storage cask was inspected using a Toshiba 7 mm (1/4) CCD video camera. The dose rates observed in the annular space were measured to provide a reference for the activity to which the camera(s) being tested were being exposed. No gross degradation, pitting, or general corrosion was observed.

P. L. Winston

2007-09-01T23:59:59.000Z

312

Recycling spent sandblasting grit and similar wastes as aggregate in asphaltic concrete. Technical data sheet  

SciTech Connect (OSTI)

The Naval Facilities Engineering Service Center (NFESC), Port Hueneme, California, and Naval Facilities Engineering Command Engineering Field Activity, West, San Bruno, California, took overall leadership in identifying and testing methods to manage waste generated from a machine shop located at Hunters Point Annex, at Naval Station, Treasure Island, California. Ship cleaning and equipment maintenance resulted in the accumulation of 4,665 tons of spent sandblasting grit at the site. The spent grit, consisting of silica sand plus a small amount of slag-derived grit, had the physical characteristics of coarse-grained beach sand and also contained fragments of coatings. The spent grit had the potential for exhibiting hazardous characteristics since the coatings included lead-based primers, copper, and butyltin-containing antifouling topcoats. The most beneficial application of reusing the spent grit was to use it as a replacement for some of the fine aggregate in asphaltic concrete. A test program was established that included characterization, bench-scale testing, long-term pilot scale testing, and a full-scale demonstration. Full-scale asphalt production provided samples which proved both the chemical leaching resistance and physical performance characteristics were acceptable.

Heath, J.C.; Nelson, B.

1998-12-01T23:59:59.000Z

313

Analysis of a Nuclear Accident: Fission and Activation Product Releases from the Fukushima Daiichi Nuclear Facility as Remote Indicators of Source Identification, Extent of Release, and State of Damaged Spent Nuclear Fuel  

SciTech Connect (OSTI)

Evidence of the release Pu from the Fukushima Daiichi nuclear power station to the local environment and surrounding communities and estimates on fraction of total fuel inventory released

Schwantes, Jon M.; Orton, Christopher R.; Clark, Richard A.

2011-12-05T23:59:59.000Z

314

Yangtze Patrol: American Naval Forces in China  

E-Print Network [OSTI]

Yangtze Patrol: American Naval Forces in China A Selected, Partially-Annotated Bibliography literature of the United States Navy in China. mvh #12;"Like Chimneys in Summer" The thousands of men who served on the China Station before World War II have been all but forgotten, except in the mythology

315

Spent fuel management fee methodology and computer code user's manual.  

SciTech Connect (OSTI)

The methodology and computer model described here were developed to analyze the cash flows for the federal government taking title to and managing spent nuclear fuel. The methodology has been used by the US Department of Energy (DOE) to estimate the spent fuel disposal fee that will provide full cost recovery. Although the methodology was designed to analyze interim storage followed by spent fuel disposal, it could be used to calculate a fee for reprocessing spent fuel and disposing of the waste. The methodology consists of two phases. The first phase estimates government expenditures for spent fuel management. The second phase determines the fees that will result in revenues such that the government attains full cost recovery assuming various revenue collection philosophies. These two phases are discussed in detail in subsequent sections of this report. Each of the two phases constitute a computer module, called SPADE (SPent fuel Analysis and Disposal Economics) and FEAN (FEe ANalysis), respectively.

Engel, R.L.; White, M.K.

1982-01-01T23:59:59.000Z

316

Investigation of the condition of spent-fuel pool components  

SciTech Connect (OSTI)

It is currently projected that spent nuclear fuel, which is discharged from the reactor and then stored in water pools, may remain in those pools for several decades. Other studies have addressed the expected integrity of the spent fuel during extended water storage; this study assesses the integrity of metallic spent fuel pool components. Results from metallurgical examinations of specimens taken from stainless steel and aluminum components exposed in spent fuel pools are presented. Licensee Event Reports (LERs) relating to problems with spent fuel components were assessed and are summarized to define the types of operational problems that have occurred. The major conclusions of this study are: aluminum and stainless steel spent fuel pool components have a good history of performance in both deionized and borated water pools. Although some operational problems involving pool components have occurred, these problems have had minimal impacts.

Kustas, F.M.; Bates, S.O.; Opitz, B.E.; Johnson, A.B. Jr.; Perez, J.M. Jr.; Farnsworth, R.K.

1981-09-01T23:59:59.000Z

317

AIAA 94-4688 Topaz II Nuclear Powered  

E-Print Network [OSTI]

Y Y AIAA 94-4688 Topaz II Nuclear Powered SAR Satellite M. Feuerstein and Dr. 9. Agrawal Naval Astronautics 370 L'EnfantPromenade, S.W., Washington, D.C. 20024 #12;AIAA-94-4688 TOPAZ 11NUCLEAR POWERED SAR at the Naval Postgraduate School. Thc design team integrated a Topaz I1 nuclear power system with an EOS

318

James F. Argue Nuclear Engineering and Planning Manager  

E-Print Network [OSTI]

James F. Argue Nuclear Engineering and Planning Manager Portsmouth Naval Shipyard Naval Sea Systems Command In June of 2003 Mr. Argue was assigned as the Nuclear Engineering and Planning Manager, a Senior Executive position. Mr. Argue is the head of the Nuclear Engineering and Planning Department

319

13.400 Introduction to Naval Architecture, Fall 2004  

E-Print Network [OSTI]

Introduction to principles of naval architecture, ship geometry, hydrostatics, calculation and drawing of curves of form, intact and damaged stability, hull structure strength calculations and ship resistance. Projects ...

Herbein, David

320

Naval Construction Battalion Center Gulfport- Mississippi Power Partnership Success Story  

Broader source: Energy.gov [DOE]

Presentation covers the Naval Construction Battalion Center Gulfport - Mississippi Power Partnership success storygiven at the Spring 2009 Federal Utility Partnership Working Group (FUPWG) meeting...

Note: This page contains sample records for the topic "naval spent nuclear" 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

Department of Energy, Office of Naval Petroleum & Oil Shale Reserves  

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

Items that may be marked "disposrtron not Office of Naval Petroleum & Oil Shale Reserves approved" or "withdrawn" In column 10 4 Nameof Personwith whom to confer 5...

322

The united kingdom's changing requirements for spent fuel storage  

SciTech Connect (OSTI)

The UK is adopting an open fuel cycle, and is necessarily moving to a regime of long term storage of spent fuel, followed by geological disposal once a geological disposal facility (GDF) is available. The earliest GDF receipt date for legacy spent fuel is assumed to be 2075. The UK is set to embark on a programme of new nuclear build to maintain a nuclear energy contribution of 16 GW. Additionally, the UK are considering a significant expansion of nuclear energy in order to meet carbon reduction targets and it is plausible to foresee a scenario where up to 75 GW from nuclear power production could be deployed in the UK by the mid 21. century. Such an expansion, could lead to spent fuel storage and its disposal being a dominant issue for the UK Government, the utilities and the public. If the UK were to transition a closed fuel cycle, then spent fuel storage should become less onerous depending on the timescales. The UK has demonstrated a preference for wet storage of spent fuel on an interim basis. The UK has adopted an approach of centralised storage, but a 16 GW new build programme and any significant expansion of this may push the UK towards distributed spent fuel storage at a number of reactors station sites across the UK.

Hodgson, Z.; Hambley, D.I.; Gregg, R.; Ross, D.N. [National Nuclear Laboratory, Chadwick House, Birchwood Park, Warrington, Cheshire WA3 6AE (United Kingdom)

2013-07-01T23:59:59.000Z

323

Naval Reactors | Y-12 National Security Complex  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary)morphinanInformation Desert Southwest Regionat Cornell Batteries &NSTCurrent Issues & Trends SeeNaval

324

324 Building spent fuel segments pieces and fragments removal summary report  

SciTech Connect (OSTI)

As part of the 324 Building Deactivation Project, all Spent Nuclear Fuel (SNF) and Special Nuclear Material were removed. The removal entailed packaging the material into a GNS-12 cask and shipping it to the Central Waste Complex (CWC).

SMITH, C L

2003-01-09T23:59:59.000Z

325

Supplement Analysis of Environmental Effects of Changes in DOE's Preferred Alternative for Management of Spent Nuclear Fuel from the K basins at the Hanford Site, Richland, Washington  

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

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

326

Mission Need Statement: Idaho Spent Fuel Facility Project  

SciTech Connect (OSTI)

Approval is requested based on the information in this Mission Need Statement for The Department of Energy, Idaho Operations Office (DOE-ID) to develop a project in support of the mission established by the Office of Environmental Management to "complete the safe cleanup of the environmental legacy brought about from five decades of nuclear weapons development and government-sponsored nuclear energy research". DOE-ID requests approval to develop the Idaho Spent Fuel Facility Project that is required to implement the Department of Energy's decision for final disposition of spent nuclear fuel in the Geologic Repository at Yucca Mountain. The capability that is required to prepare Spent Nuclear Fuel for transportation and disposal outside the State of Idaho includes characterization, conditioning, packaging, onsite interim storage, and shipping cask loading to complete shipments by January 1,2035. These capabilities do not currently exist in Idaho.

Barbara Beller

2007-09-01T23:59:59.000Z

327

Extraction of uranium from spent fuels using liquefied gases  

SciTech Connect (OSTI)

For reprocessing of spent nuclear fuels, a novel method to extract actinides from spent fuel using highly compressed gases, nitrogen dioxide and carbon dioxide was proposed. As a fundamental study, the nitrate conversion with liquefied nitrogen dioxide and the nitrate extraction with supercritical carbon dioxide were demonstrated by using uranium dioxide powder, uranyl nitrate and tri-n-butylphosphate complex in the present study. (authors)

Sawada, Kayo; Hirabayashi, Daisuke; Enokida, Youichi [EcoTopia Science Institute, Nagoya University, Furo-cho, Chikusa-ku, Nagoya, 464-8603 (Japan)

2007-07-01T23:59:59.000Z

328

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

SciTech Connect (OSTI)

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

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

1993-12-31T23:59:59.000Z

329

Scintillator spent fuel monitor  

SciTech Connect (OSTI)

A monitor for rapidly measuring the gross gamma-ray flux immediately above spent fuel assemblies in underwater storage racks has been developed. It consists of a plastic scintillator, photomultiplier, collimator, and a small battery-powered electronics package. The crosstalk from an isolated fuel assembly to an adjacent void is only about 2%. The mean difference between the measured gamma-ray flux and the flux estimated from the declared burnup and cooling time with a simple formula is 22%.

Moss, C.E.; Nixon, K.V.; Bernard, W.

1980-01-01T23:59:59.000Z

330

THE NAVAL RESEARCH ENTERPRISE AND PLASMA PHYSICS RESEARCH AT THE NAVAL RESEARCH LAB  

E-Print Network [OSTI]

, to create Xray simulators for testing nuclear weapons effects, and to understand high altitude nuclear ex

Shyy, Wei

331

Spent Fuel Working Group Report. Volume 1  

SciTech Connect (OSTI)

The Department of Energy is storing large amounts of spent nuclear fuel and other reactor irradiated nuclear materials (herein referred to as RINM). In the past, the Department reprocessed RINM to recover plutonium, tritium, and other isotopes. However, the Department has ceased or is phasing out reprocessing operations. As a consequence, Department facilities designed, constructed, and operated to store RINM for relatively short periods of time now store RINM, pending decisions on the disposition of these materials. The extended use of the facilities, combined with their known degradation and that of their stored materials, has led to uncertainties about safety. To ensure that extended storage is safe (i.e., that protection exists for workers, the public, and the environment), the conditions of these storage facilities had to be assessed. The compelling need for such an assessment led to the Secretary`s initiative on spent fuel, which is the subject of this report. This report comprises three volumes: Volume I; Summary Results of the Spent Fuel Working Group Evaluation; Volume II, Working Group Assessment Team Reports and Protocol; Volume III; Operating Contractor Site Team Reports. This volume presents the overall results of the Working Group`s Evaluation. The group assessed 66 facilities spread across 11 sites. It identified: (1) facilities that should be considered for priority attention. (2) programmatic issues to be considered in decision making about interim storage plans and (3) specific vulnerabilities for some of these facilities.

O`Toole, T.

1993-11-01T23:59:59.000Z

332

Status of Proposed Repository for Latin-American Spent Fuel  

SciTech Connect (OSTI)

This report compiles preliminary information that supports the premise that a repository is needed in Latin America and analyzes the nuclear situation (mainly in Argentina and Brazil) in terms of nuclear capabilities, inventories, and regional spent-fuel repositories. The report is based on several sources and summarizes (1) the nuclear capabilities in Latin America and establishes the framework for the need of a permanent repository, (2) the International Atomic Energy Agency (IAEA) approach for a regional spent-fuel repository and describes the support that international institutions are lending to this issue, (3) the current situation in Argentina in order to analyze the Argentinean willingness to find a location for a deep geological repository, and (4) the issues involved in selecting a location for the repository and identifies a potential location. This report then draws conclusions based on an analysis of this information. The focus of this report is mainly on spent fuel and does not elaborate on other radiological waste sources.

Ferrada, J.J.

2004-10-04T23:59:59.000Z

333

NAC-1 cask dose rate calculations for LWR spent fuel  

SciTech Connect (OSTI)

A Nuclear Assurance Corporation nuclear fuel transport cask, NAC-1, is being considered as a transport and storage option for spent nuclear fuel located in the B-Cell of the 324 Building. The loaded casks will be shipped to the 200 East Area Interim Storage Area for dry interim storage. Several calculations were performed to assess the photon and neutron dose rates. This report describes the analytical methods, models, and results of this investigation.

CARLSON, A.B.

1999-02-24T23:59:59.000Z

334

Reactor-specific spent fuel discharge projections: 1985 to 2020  

SciTech Connect (OSTI)

The creation of four spent-fuel data bases that contain information on the projected amounts of spent fuel to be discharged from US commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent-fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water reactors (BWR). The projections are based on individual reactor information supplied by the US reactor owners. The basic information is adjusted to conform to Energy Information Agency (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: (1) No New Orders with Extended Burnup, (2) No New Orders with Constant Burnup, (3) Middle Case with Extended Burnup, and (4) Middle Case with Constant Burnup. Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum-at-reactor storage, and for storage requirements assuming maximum-at-reactor plus intra-utility transshipment of spent fuel.

Heeb, C.M.; Libby, R.A.; Walling, R.C.; Purcell, W.L.

1986-09-01T23:59:59.000Z

335

Reactor-specific spent fuel discharge projections, 1987-2020  

SciTech Connect (OSTI)

The creation of five reactor-specific spent fuel data bases that contain information on the projected amounts of spent fuel to be discharged from U.S. commercial nuclear reactors through the year 2020 is described. The data bases contain detailed spent fuel information from existing, planned, and projected pressurized water reactors (PWR) and boiling water eactors (BWR), and one existing high temperature gas reactor (HTGR). The projections are based on individual reactor information supplied by the U.S. reactor owners. The basic information is adjusted to conform to Energy Information Administration (EIA) forecasts for nuclear installed capacity, generation, and spent fuel discharged. The EIA cases considered are: No New Orders (assumes increasing burnup), No New Orders with No Increased Burnup, Upper Reference (assumes increasing burnup), Upper Reference with No Increased Burnup, and Lower Reference (assumes increasing burnup). Detailed, by-reactor tables are provided for annual discharged amounts of spent fuel, for storage requirements assuming maximum at-reactor storage, and for storage requirements assuming maximum at-reactor storage plus intra-utility transshipment of spent fuel. 8 refs., 8 figs., 10 tabs.

Walling, R.C.; Heeb, C.M.; Purcell, W.L.

1988-03-01T23:59:59.000Z

336

NUHOMS modular spent-fuel storage system: Performance testing  

SciTech Connect (OSTI)

This report documents the results of a heat transfer and shielding performance evaluation of the NUTECH HOrizontal MOdular Storage (NUHOMS{reg sign}) System utilized by the Carolina Power and Light Co. (CP L) in an Independent Spent Fuel Storage Installation (ISFSI) licensed by the US Nuclear Regulatory Commission (NRC). The ISFSI is located at CP L's H. B. Robinson Nuclear Plant (HBR) near Hartsville, South Carolina. The demonstration included testing of three modules, first with electric heaters and then with spent fuel. The results indicated that the system was conservatively designed, with all heat transfer and shielding design criteria easily met. 5 refs., 45 figs., 9 tabs.

Strope, L.A.; McKinnon, M.A. (Pacific Northwest Lab., Richland, WA (USA)); Dyksterhouse, D.J.; McLean, J.C. (Carolina Power and Light Co., Raleigh, NC (USA))

1990-09-01T23:59:59.000Z

337

International Atomic Energy Agency (IAEA) activities on spent fuel management options  

SciTech Connect (OSTI)

Many countries have in the past several decades opted for storage of spent fuel for undefined periods of time. They have adopted the 'wait and see' strategy for spent fuel management. A relatively small number of countries have adopted reprocessing and use of MOX fuel as part of their strategy in spent fuel management. From the 10, 000 tonnes of heavy metal that is removed annually from nuclear reactors throughout the world, only approximately 30 % is currently being reprocessed. Continuous re-evaluation of world energy resources, announcement of the Global Nuclear Energy Partnership (GNEP) and the Russian initiative to form international nuclear centers, including reprocessing, are changing the stage for future development of nuclear energy. World energy demand is expected to more than double by 2050, and expansion of nuclear energy is a key to meeting this demand while reducing pollution and greenhouse gases. Since its foundation, the International Atomic Energy Agency (IAEA) has served as an interface between countries in exchanging information on the peaceful development of nuclear energy and at the same time guarding against proliferation of materials that could be used for nuclear weapons. The IAEA's Department of Nuclear Energy has been generating technical documents, holding meetings and conferences, and supporting technical cooperation projects to facilitate this exchange of information. This paper focuses on the current status of IAEA activities in the field of spent fuel management being carried out by the Division of Nuclear Fuel Cycle and Waste Technology. Information on those activities could be found on the web site link www.iaea.org/OurWork/ST/NE/NEFW/nfcms. To date, the IAEA has given priority in its spent fuel management activities to supporting Member States in their efforts to deal with growing accumulations of spent power reactor fuel. There is technical consensus that the present technologies for spent fuel storage, wet and dry, provide adequate protection to people and environment. As storage durations grow, the IAEA has expanded its work related to the implications of extended storage periods. Operation and maintenance of containers for storage and transport have also been investigated related to long term storage periods. In addition, as international interest in reprocessing of spent fuel increases, the IAEA continues to serve as a crossroads for sharing the latest developments in spent fuel treatment options. A Coordinated Research Project is currently addressing spent fuel performance assessment and research to evaluate long term effects of storage on spent fuel. The effect of increased burnup and mixed oxide fuels on spent fuel management is also the focus of interest as it follows the trend in optimizing the use of nuclear fuel. Implications of damaged fuel on storage and transport as well as burnup credit in spent fuel applications are areas that the IAEA is also investigating. Since spent fuel management considerations require social stability and institutional control, those aspects are taken into account in most IAEA activities. Data requirements and records management as storage durations extend were also investigated as well as the potential for regional spent fuel storage facilities. Spent fuel management activities continue to be coordinated with others in the IAEA to ensure compliance and consistency with efforts in the Department of Safety and Security and the Department of Safeguards, as well as with activities related to geologic disposal. Either disposal of radioactive waste or spent fuel will be an ultimate consideration in all spent fuel management options. Updated information on spent fuel treatment options that include fuel reprocessing as well as transmutation of minor actinides are investigated to optimize the use of nuclear fuel and minimize impact on environment. Tools for spent fuel management economics are also investigated to facilitate assessment of industrial applicability for these options. Most IAEA spent fuel management activities will ultimately be reported in o

Lovasic, Z.; Danker, W. [International Atomic Energy Agency (IAEA) Vienna (Austria)

2007-07-01T23:59:59.000Z

338

Congressional Delegation visits Naval Reactors Facility | National Nuclear  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series to User Group andCompositional VariationCluster)Photosynthetic

339

More About NNSA's Naval Reactors Office | National Nuclear Security  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas Conchas recovery challenge fundProject8MistakesAdministration About Us / Our Programs

340

More About NNSA's Naval Reactors Office | National Nuclear Security  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsrucLas Conchas recovery challenge fundProject8MistakesAdministration About Us / Our

Note: This page contains sample records for the topic "naval spent nuclear" 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

2013 Annual Planning Summary for the Naval Nuclear Propulsion Program |  

Energy Savers [EERE]

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home Page on Delicious RankCombustionImprovement Awardflash2007-42attachment1.pdfmodule(EE)2012

342

NA 30 - Deputy Administrator for Naval Reactors | National Nuclear Security  

National Nuclear Security Administration (NNSA)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "ofEarlyEnergyDepartmentNationalRestart of the Review of the Yucca0 NationalJ Page i PARTAdministration

343

Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities  

National Nuclear Security Administration (NNSA)

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

344

FY 2012 Budget Hearing Testimony on Nuclear Nonproliferation and Naval  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May JunDatastreamsmmcrcalgovInstrumentsruc DocumentationP-Series toESnet4:Epitaxial ThinFOR IMMEDIATE RELEASE17 Budget

345

Facts and issues of direct disposal of spent fuel; Revision 1  

SciTech Connect (OSTI)

This report reviews those facts and issues that affect the direct disposal of spent reactor fuels. It is intended as a resource document for those impacted by the current Department of Energy (DOE) guidance that calls for the cessation of fuel reprocessing. It is not intended as a study of the specific impacts (schedules and costs) to the Savannah River Site (SRS) alone. Commercial fuels, other low enriched fuels, highly enriched defense-production, research, and naval reactor fuels are included in this survey, except as prevented by rules on classification.

Parks, P.B.

1993-10-01T23:59:59.000Z

346

TEPP - Spent Nuclear Fuel | Department of Energy  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of Energy Power.pdf11-161-LNGInternational EnergyCommitteeRenewable Energy,Section 180(c)CHARTER

347

RADIATION ANALYSIS OF A SPENT-FUEL STORAGE CASK  

E-Print Network [OSTI]

RADIATION ANALYSIS OF A SPENT-FUEL STORAGE CASK by J.K. Shultis Department of Mechanical;Radiation Analysis of a Spect-Fuel Storage Cask by J.K.Shultis Dept. Mechanical and Nuclear Engineering compositions for the air, soil, berm and concrete were taken from the references indicated in Table 1. 1.3 Cask

Shultis, J. Kenneth

348

West Valley facility spent fuel handling, storage, and shipping experience  

SciTech Connect (OSTI)

The result of a study on handling and shipping experience with spent fuel are described in this report. The study was performed by Pacific Northwest Laboratory (PNL) and was jointly sponsored by the US Department of Energy (DOE) and the Electric Power Research Institute (EPRI). The purpose of the study was to document the experience with handling and shipping of relatively old light-water reactor (LWR) fuel that has been in pool storage at the West Valley facility, which is at the Western New York Nuclear Service Center at West Valley, New York and operated by DOE. A subject of particular interest in the study was the behavior of corrosion product deposits (i.e., crud) deposits on spent LWR fuel after long-term pool storage; some evidence of crud loosening has been observed with fuel that was stored for extended periods at the West Valley facility and at other sites. Conclusions associated with the experience to date with old spent fuel that has been stored at the West Valley facility are presented. The conclusions are drawn from these subject areas: a general overview of the West Valley experience, handling of spent fuel, storing of spent fuel, rod consolidation, shipping of spent fuel, crud loosening, and visual inspection. A list of recommendations is provided. 61 refs., 4 figs., 5 tabs.

Bailey, W.J.

1990-11-01T23:59:59.000Z

349

NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944  

E-Print Network [OSTI]

#12;#12;11 #12;2 NUCLEAR POWER AND RESEARCH REACTORS 1939 1942 1943 1944 Nuclear fission discovered 430 nuclear power reactors are operating in the world, and 103 nuclear power plants produce 20, naval reactors, and nuclear power plants. Oak Ridge experiments byArt Snell in 1944 showed that 10 tons

Pennycook, Steve

350

Carl Vinson and pre-war naval legislation 1932-1940  

E-Print Network [OSTI]

, House and Senate debates, and behind the scenes conferences between Vinson and officials of the Navy Department and Roosevelt Administration which helped decide the course of naval expansion. It shows that while many people contributed passing naval...

Svonavec, Stephen Charles

1994-01-01T23:59:59.000Z

351

HFIR spent fuel management alternatives  

SciTech Connect (OSTI)

The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems' Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere.

Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

1992-10-15T23:59:59.000Z

352

HFIR spent fuel management alternatives  

SciTech Connect (OSTI)

The High Flux Isotope Reactor (HFIR) at Martin Marietta Energy Systems` Oak Ridge National Laboratory (ORNL) has been unable to ship its spent fuel to Savannah River Site (SRS) for reprocessing since 1985. The HFIR storage pools are expected to fill up in the February 1994 to February 1995 time frame. If a management altemative to existing HFIR pool storage is not identified and implemented before the HFIR pools are full, the HFIR will be forced to shut down. This study investigated several alternatives for managing the HFIR spent fuel, attempting to identify options that could be implemented before the HFIR pools are full. The options investigated were: installing a dedicated dry cask storage facility at ORNL, increasing HFIR pool storage capacity by clearing the HFIR pools of debris and either close-packing or stacking the spent fuel elements, storing the spent fuel at another ORNL pool, storing the spent fuel in one or more hot cells at ORNL, and shipping the spent fuel offsite for reprocessing or storage elsewhere.

Begovich, J.M.; Green, V.M.; Shappert, L.B.; Lotts, A.L.

1992-10-15T23:59:59.000Z

353

Nuclear Energy Policy University of Nevada ? Reno 27 March...  

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

* Safe, effective disposition of spent nuclear fuel not yet demonstrated What's driving nuclear expansion * Rapid increase in global energy demand * Rising importance of carbon...

354

DOE Selects Savannah River Nuclear Solutions, LLC to Manage and...  

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

activities. Environmental cleanup activities include management of spent nuclear fuel, nuclear materials, and non high-level radioactive waste; deactivation and decommissioning...

355

Radiological consequences of ship collisions that might occur in U.S. Ports during the shipment of foreign research reactor spent nuclear fuel to the United States in break-bulk freighters  

SciTech Connect (OSTI)

Accident source terms, source term probabilities, consequences, and risks are developed for ship collisions that might occur in U.S. ports during the shipment of spent fuel from foreign research reactors to the United States in break-bulk freighters.

Sprung, J.L.; Bespalko, S.J.; Massey, C.D.; Yoshimura, R. [Sandia National Laboratory, Albuquerque, NM (United States); Johnson, J.D. [GRAM Inc., Albuquerque, NM (United States); Reardon, P.C. [PCRT Technologies, Albuquerque, NM (United States); Ebert, M.W.; Gallagher D.W. [Science Applications International Corp., Reston, VA (United States)

1996-08-01T23:59:59.000Z

356

Spent fuel utilization in a compact traveling wave reactor  

SciTech Connect (OSTI)

In recent years, several innovative designs of nuclear reactors are proposed. One of them is Traveling Wave Reactor (TWR). The unique characteristic of a TWR is the capability of breeding its own fuel in the reactor. The reactor is fueled by mostly depleted, natural uranium or spent nuclear fuel and a small amount of enriched uranium to initiate the fission process. Later on in the core, the reactor gradually converts the non-fissile material into the fissile in a process like a traveling wave. In this work, a TWR with spent nuclear fuel blanket was studied. Several parameters such as reactivity coefficients, delayed neutron fraction, prompt neutron generation lifetime, and fission power, were analyzed. The discharge burnup composition was also analyzed. The calculation is performed by a continuous energy Monte Carlo code McCARD.

Hartanto, Donny; Kim, Yonghee [Korea Advanced Institute of Science and Technology 373-1 Kusong-dong, Yuseong-gu, Daejeon, 305-701 (Korea, Republic of)

2012-06-06T23:59:59.000Z

357

INSTRUCTIONS FOR SUBMITTING NUCLEAR  

E-Print Network [OSTI]

. Decommissioning 13 I. Performance 13 J. Nuclear Fuel 14 K. Nuclear Insurance 14 L. Relicensing or Plant RetirementCALIFORNIA ENERGY COMMISSION INSTRUCTIONS FOR SUBMITTING NUCLEAR POWER PLANT-RELATED DATA of Submitted Data 3 NUCLEAR POWER PLANT DATA REQUESTS 6 A. Environmental Impacts 6 B. Spent Fuel Generation 8 C

358

Information handbook on independent spent fuel storage installations  

SciTech Connect (OSTI)

In this information handbook, the staff of the U.S. Nuclear Regulatory Commission describes (1) background information regarding the licensing and history of independent spent fuel storage installations (ISFSIs), (2) a discussion of the licensing process, (3) a description of all currently approved or certified models of dry cask storage systems (DCSSs), and (4) a description of sites currently storing spent fuel in an ISFSI. Storage of spent fuel at ISFSIs must be in accordance with the provisions of 10 CFR Part 72. The staff has provided this handbook for information purposes only. The accuracy of any information herein is not guaranteed. For verification or for more details, the reader should refer to the respective docket files for each DCSS and ISFSI site. The information in this handbook is current as of September 1, 1996.

Raddatz, M.G.; Waters, M.D.

1996-12-01T23:59:59.000Z

359

Spent fuel and fuel pool component integrity. Annual report, FY 1980  

SciTech Connect (OSTI)

During program FY 1980 staff members of the Spent Fuel and Fuel Pool Component Integrity Program at Pacific Northwest Laboratory (PNL) completed the following major tasks: represented DOE on the international Behavior of Fuel Assemblies in Storage (BEFAST) Committee; the program manager, A.B. Johnson, Jr., participated in an International Survey of Water Reactor Spent Fuel Storage Experience, which was conducted jointly by the International Atomic Energy Agency (Vienna) and the Nuclear Energy Agency (Paris); provided written testimony and cross statement for the Proposed Rulemaking on Storage and Disposal of Nuclear Waste; acquired and began examination of the world's oldest pool-stored Zircaloy-clad fuel from the Shippingport reactor, stored approx. 21 years in deionized water; acquired and began examination of stainless-clad spent fuel from the Connecticut Yankee Reactor (PWR); negotiated for specimens from components stored in spent fuel pools at fuel storage facilities from the Savannah River Plant, Aiken, South Carolina, Zion (PWR) spent fuel pool, Zion, Illinois, and La Crosse (BWR) spent fuel pool, La Crosse, Wisconsin; planned for examinations in FY 81 of specimens from the three spent fuel pools; investigated a low-temperature stress corrosion cracking mechanism that developed in piping at a few PWR spent fuel pools. This report summarizes the results of these activities and investigations. Details are provided in the presentationsand publications generated under this program and summarized in Appendix A.

Johnson, A.B. Jr.; Bailey, W.J.; Bradley, E.R.; Bruemmer, S.M.; Langstaff, D.C.

1981-09-01T23:59:59.000Z

360

362 SDSU General Catalog 2012-2013 Naval Science  

E-Print Network [OSTI]

turbine, and diesel propulsion engines receive in-depth study. Leadership topics in an engineering setting Power and Maritime Affairs (3) Sea power and maritime affairs; general concept of sea power including Merchant Marine; role of naval warfare components used to support the Navy's mission; sea power

Gallo, Linda C.

Note: This page contains sample records for the topic "naval spent nuclear" 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

368 SDSU General Catalog 2013-2014 Naval Science  

E-Print Network [OSTI]

turbine, and diesel propulsion engines receive in-depth study. Leadership topics in an engineering setting Power and Maritime Affairs (3) Sea power and maritime affairs; general concept of sea power including Merchant Marine; role of naval warfare components used to support the Navy's mission; sea power

Gallo, Linda C.

362

Colleagues and Friends of NPS: Recently the Naval Postgraduate  

E-Print Network [OSTI]

@nps.edu Journalists Barbara Honegger MC2 (SW) Corey Truax MC3 Kellie Arakawa Photographers Javier Chagoya MC2 (SW) Corey Truax MC2 Kellie Arakawa Naval Postgraduate School President Daniel T. Oliver Provost Executive. Full versions of all articles are available at www.nps.edu #12;July 2008 Contents 7 New GSEAS Dean Dr

363

A Monte Carlo based spent fuel analysis safeguards strategy assessment  

SciTech Connect (OSTI)

Safeguarding nuclear material involves the detection of diversions of significant quantities of nuclear materials, and the deterrence of such diversions by the risk of early detection. There are a variety of motivations for quantifying plutonium in spent fuel assemblies by means of nondestructive assay (NDA) including the following: strengthening the capabilities of the International Atomic Energy Agencies ability to safeguards nuclear facilities, shipper/receiver difference, input accountability at reprocessing facilities and burnup credit at repositories. Many NDA techniques exist for measuring signatures from spent fuel; however, no single NDA technique can, in isolation, quantify elemental plutonium and other actinides of interest in spent fuel. A study has been undertaken to determine the best integrated combination of cost effective techniques for quantifying plutonium mass in spent fuel for nuclear safeguards. A standardized assessment process was developed to compare the effective merits and faults of 12 different detection techniques in order to integrate a few techniques and to down-select among the techniques in preparation for experiments. The process involves generating a basis burnup/enrichment/cooling time dependent spent fuel assembly library, creating diversion scenarios, developing detector models and quantifying the capability of each NDA technique. Because hundreds of input and output files must be managed in the couplings of data transitions for the different facets of the assessment process, a graphical user interface (GUI) was development that automates the process. This GUI allows users to visually create diversion scenarios with varied replacement materials, and generate a MCNPX fixed source detector assessment input file. The end result of the assembly library assessment is to select a set of common source terms and diversion scenarios for quantifying the capability of each of the 12 NDA techniques. We present here the generalized assessment process, the techniques employed to automate the coupled facets of the assessment process, and the standard burnup/enrichment/cooling time dependent spent fuel assembly library. We also clearly define the diversion scenarios that will be analyzed during the standardized assessments. Though this study is currently limited to generic PWR assemblies, it is expected that the results of the assessment will yield an adequate spent fuel analysis strategy knowledge that will help the down-select process for other reactor types.

Fensin, Michael L [Los Alamos National Laboratory; Tobin, Stephen J [Los Alamos National Laboratory; Swinhoe, Martyn T [Los Alamos National Laboratory; Menlove, Howard O [Los Alamos National Laboratory; Sandoval, Nathan P [Los Alamos National Laboratory

2009-01-01T23:59:59.000Z

364

Spent-fuel-storage alternatives  

SciTech Connect (OSTI)

The Spent Fuel Storage Alternatives meeting was a technical forum in which 37 experts from 12 states discussed storage alternatives that are available or are under development. The subject matter was divided into the following five areas: techniques for increasing fuel storage density; dry storage of spent fuel; fuel characterization and conditioning; fuel storage operating experience; and storage and transport economics. Nineteen of the 21 papers which were presented at this meeting are included in this Proceedings. These have been abstracted and indexed. (ATT)

Not Available

1980-01-01T23:59:59.000Z

365

Storage of LWR spent fuel in air. Volume 3, Results from exposure of spent fuel to fluorine-contaminated air  

SciTech Connect (OSTI)

The Behavior of Spent Fuel in Storage (BSFS) Project has conducted research to develop data on spent nuclear fuel (irradiated U0{sub 2}) that could be used to support design, licensing, and operation of dry storage installations. Test Series B conducted by the BSFS Project was designed as a long-term study of the oxidation of spent fuel exposed to air. It was discovered after the exposures were completed in September 1990 that the test specimens had been exposed to an atmosphere of bottled air contaminated with an unknown quantity of fluorine. This exposure resulted in the test specimens reacting with both the oxygen and the fluorine in the oven atmospheres. The apparent source of the fluorine was gamma radiation-induced chemical decomposition of the fluoro-elastomer gaskets used to seal the oven doors. This chemical decomposition apparently released hydrofluoric acid (HF) vapor into the oven atmospheres. Because the Test Series B specimens were exposed to a fluorine-contaminated oven atmosphere and reacted with the fluorine, it is recommended that the Test Series B data not be used to develop time-temperature limits for exposure of spent nuclear fuel to air. This report has been prepared to document Test Series B and present the collected data and observations.

Cunningham, M.E.; Thomas, L.E.

1995-06-01T23:59:59.000Z

366

Naval Research Laboratory Technology Marketing Summaries - Energy  

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

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr MayAtmosphericNuclear Security Administration the Contributions andDataNational Library of1, 2007 (nextNauru Island

367

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

E-Print Network [OSTI]

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

368

Method For Processing Spent (Trn,Zr)N Fuel  

DOE Patents [OSTI]

A new process for recycling spent nuclear fuels, in particular, mixed nitrides of transuranic elements and zirconium. The process consists of two electrorefiner cells in series configuration. A transuranic element such as plutonium is reduced at the cathode in the first cell, zirconium at the cathode in the second cell, and nitrogen-15 is released and captured for reuse to make transuranic and zirconium nitrides.

Miller, William E. (Naperville, IL); Richmann, Michael K. (Woodridge, IL)

2004-07-27T23:59:59.000Z

369

Energy use baselining study for the National Naval Medical Center  

SciTech Connect (OSTI)

This report provides an energy consumption profile for fourteen buildings at the National Naval Medical Center (NNMC) in Bethesda, Maryland. Recommendations are also made for viable energy efficiency projects funded with assistance from the servicing utility (Potomic Electric Power Company) in the form of rebates and incentives available in their Demand Side Management (DSM) program and through Shared Energy Savings (SES) projects. This report also provides estimates of costs and potential energy savings of the recommended projects.

Parker, G.B.; Halverson, M.A.

1992-04-01T23:59:59.000Z

370

Test plan for reactions between spent fuel and J-13 well water under unsaturated conditions  

SciTech Connect (OSTI)

The Yucca Mountain Site Characterization Project is evaluating the long-term performance of a high-level nuclear waste form, spent fuel from commercial reactors. Permanent disposal of the spent fuel is possible in a potential repository to be located in the volcanic tuff beds near Yucca Mountain, Nevada. During the post-containment period the spent fuel could be exposed to water condensation since of the cladding is assumed to fail during this time. Spent fuel leach (SFL) tests are designed to simulate and monitor the release of radionuclides from the spent fuel under this condition. This Test Plan addresses the anticipated conditions whereby spent fuel is contacted by small amounts of water that trickle through the spent fuel container. Two complentary test plans are presented, one to examine the reaction of spent fuel and J-13 well water under unsaturated conditions and the second to examine the reaction of unirradiated UO{sub 2} pellets and J-13 well water under unsaturated conditions. The former test plan examines the importance of the water content, the oxygen content as affected by radiolysis, the fuel burnup, fuel surface area, and temperature. The latter test plant examines the effect of the non-presence of Teflon in the test vessel.

Finn, P.A.; Wronkiewicz, D.J.; Hoh, J.C.; Emery, J.W.; Hafenrichter, L.D.; Bates, J.K.

1993-01-01T23:59:59.000Z

371

The French national program for spent fuel and high-level waste management  

SciTech Connect (OSTI)

From its very beginning, the French national program for spent fuel and HLW management is aimed at the recycling of energetic materials and the safe disposal of nuclear waste. Spent fuel reprocessing is the cornerstone of this program, since it directly opens the way to energetic material recycling, waste minimization and safe conditioning. It is complemented by the HLW management program which is defined by the HLW disposal regulation and the Waste Act issued in 1991.

Giraud, J.P.; Demontalembert, J.A. [COGEMA, Velizy-Villacoublay (France)

1993-12-31T23:59:59.000Z

372

Actinide removal from spent salts  

DOE Patents [OSTI]

A method for removing actinide contaminants (uranium and thorium) from the spent salt of a molten salt oxidation (MSO) reactor is described. Spent salt is removed from the reactor and analyzed to determine the contaminants present and the carbonate concentration. The salt is dissolved in water, and one or more reagents are added to precipitate the thorium as thorium oxide and/or the uranium as either uranium oxide or as a diuranate salt. The precipitated materials are filtered, dried and packaged for disposal as radioactive waste. About 90% of the thorium and/or uranium present is removed by filtration. After filtration, salt solutions having a carbonate concentration >20% can be dried and returned to the reactor for re-use. Salt solutions containing a carbonate concentration <20% require further clean-up using an ion exchange column, which yields salt solutions that contain less than 0.1 ppm of thorium or uranium.

Hsu, Peter C. (Pleasanton, CA); von Holtz, Erica H. (Livermore, CA); Hipple, David L. (Livermore, CA); Summers, Leslie J. (Livermore, CA); Adamson, Martyn G. (Danville, CA)

2002-01-01T23:59:59.000Z

373

Naval Petroleum and Oil Shale Reserves annual report of operations for fiscal year 1996  

SciTech Connect (OSTI)

During fiscal year 1996, the Department of Energy continued to operate Naval Petroleum Reserve No. 1 in California and Naval Petroleum Reserve No. 3 in Wyoming through its contractors. In addition, natural gas operations were conducted at Naval Petroleum Reserve No. 3. All productive acreage owned by the Government at Naval Petroleum Reserve No. 2 in California was produced under lease to private companies. The locations of all six Naval Petroleum and Oil Shale Reserves are shown in a figure. Under the Naval Petroleum Reserves Production Act of 1976, production was originally authorized for six years, and based on findings of national interest, the President was authorized to extend production in three-year increments. President Reagan exercised this authority three times (in 1981, 1984, and 1987) and President Bush authorized extended production once (in 1990). President Clinton exercised this authority in 1993 and again in October 1996; production is presently authorized through April 5, 2000. 4 figs. 30 tabs.

NONE

1996-12-31T23:59:59.000Z

374

Spent fuel integrity during transportation  

SciTech Connect (OSTI)

The conditions of recent shipments of light water reactor spent fuel were surveyed. The radioactivity level of cask coolant was examined in an attempt to find the effects of transportation on LWR fuel assemblies. Discussion included potential cladding integrity loss mechanisms, canning requirements, changes of radioactivity levels, and comparison of transportation in wet or dry media. Although integrity loss or degradation has not been identified, radioactivity levels usually increase during transportation, especially for leaking assemblies.

Funk, C.W.; Jacobson, L.D.

1980-01-01T23:59:59.000Z

375

E-Print Network 3.0 - andreyev bay naval Sample Search Results  

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

at the exact location (inside the bay) where the naval operations take place... . Armstrong, 2004. Prediction of instantaneous currents in San Diego ... Source: Chu, Peter C. -...

376

FEMP ESPC Success Story- U.S. Naval Station, Guantanamo Bay, Cuba  

Broader source: Energy.gov [DOE]

Fact sheet describes the Energy Savings Performance Contract (ESPC) success story on environmental stewardship and cost savings at the U.S. Naval Station at Guantanamo Bay, Cuba.

377

Safety Aspects of Dry Spent Fuel Storage and Spent Fuel Management - 13559  

SciTech Connect (OSTI)

Dry storage systems are characterized by passive and inherent safety systems ensuring safety even in case of severe incidents or accidents. After the events of Fukushima, the advantages of such passively and inherently safe dry storage systems have become more and more obvious. As with the storage of all radioactive materials, the storage of spent nuclear fuel (SF) and high-level radioactive waste (HLW) must conform to safety requirements. Following safety aspects must be achieved throughout the storage period: - safe enclosure of radioactive materials, - safe removal of decay heat, - securing nuclear criticality safety, - avoidance of unnecessary radiation exposure. The implementation of these safety requirements can be achieved by dry storage of SF and HLW in casks as well as in other systems such as dry vault storage systems or spent fuel pools, where the latter is neither a dry nor a passive system. Furthermore, transport capability must be guaranteed during and after storage as well as limitation and control of radiation exposure. The safe enclosure of radioactive materials in dry storage casks can be achieved by a double-lid sealing system with surveillance of the sealing system. The safe removal of decay heat must be ensured by the design of the storage containers and the storage facility. The safe confinement of radioactive inventory has to be ensured by mechanical integrity of fuel assembly structures. This is guaranteed, e.g. by maintaining the mechanical integrity of the fuel rods or by additional safety measures for defective fuel rods. In order to ensure nuclear critically safety, possible effects of accidents have also to be taken into consideration. In case of dry storage it might be necessary to exclude the re-positioning of fissile material inside the container and/or neutron moderator exclusion might be taken into account. Unnecessary radiation exposure can be avoided by the cask or canister vault system itself. In Germany dry storage of SF in casks fulfills both transport and storage requirements. Mostly, storage facilities are designed as concrete buildings above the ground, but due to regional constraints, one storage facility has also been built as a rock tunnel. The decay heat is always removed by natural air flow; further technical equipment is not needed. The removal of decay heat and shielding had been modeled and calculated by state-of-the-art computer codes before such a facility has been built. TueV and BAM present their long experience in the licensing process for sites and casks and inform about spent nuclear fuel management and issues concerning dry storage of spent nuclear fuel. Different storage systems and facilities in Germany, Europe and world-wide are compared with respect to the safety aspects mentioned above. Initial points are the safety issues of wet storage of SF, and it is shown how dry storage systems can ensure the compliance with the mentioned safety criteria over a long storage period. The German storage concept for dry storage of SF and HLW is presented and discussed. Exemplarily, the process of licensing, erection and operation of selected German dry storage facilities is presented. (authors)

Botsch, W.; Smalian, S.; Hinterding, P. [TUV NORD Nuclear c/o TUV NORD EnSys Hannover GmbH and Co.KG, Dept. Radiation Protection and Waste Disposal, Am TueV 1, 30519 Hannover (Germany)] [TUV NORD Nuclear c/o TUV NORD EnSys Hannover GmbH and Co.KG, Dept. Radiation Protection and Waste Disposal, Am TueV 1, 30519 Hannover (Germany); Voelzke, H.; Wolff, D.; Kasparek, E. [BAM Federal Institute for Materials Research and Testing Division 3.4 Safety of Storage Containers Unter den Eichen 44-46, 12203 Berlin (Germany)] [BAM Federal Institute for Materials Research and Testing Division 3.4 Safety of Storage Containers Unter den Eichen 44-46, 12203 Berlin (Germany)

2013-07-01T23:59:59.000Z

378

The Universal Canister Strategy in Spent Fuel Reprocessing: UC-C a Real Industrial Improvement  

SciTech Connect (OSTI)

In commercial nuclear activities, spent fuel back end management is a key issue for nuclear countries as spent fuel represent most of national civil nuclear waste legacy. Ensuring public safety and protection of the environment, now and in the future has been and still remains a major commitment, it is still the subject of thorough development efforts and active public debates. Considerable benefits can be obtained from the Universal Canister strategy as implemented in France in spent fuel treatment and waste conditioning based on reprocessing. COGEMA developed sophisticated waste conditioning processes to simplify High Level and Long Lived Intermediate Level Waste storage and final disposal. Main benefits are: waste stabilization by immobilization and encapsulation; ultimate waste toxicity reduction; drastic ultimate waste volume reduction; and ultimate waste packages standardization.

Thomasson, J.; Barithel, S.; Cocaud, A.; Derycke, P.; Pierre, P.

2003-02-25T23:59:59.000Z

379

Thermomechanical modeling of the Spent Fuel Test-Climax  

SciTech Connect (OSTI)

The Spent Fuel Test-Climax (SFT-C) was conducted to evaluate the feasibility of retrievable deep geologic storage of commercially generated spent nuclear-reactor fuel assemblies. One of the primary aspects of the test was to measure the thermomechanical response of the rock mass to the extensive heating of a large volume of rock. Instrumentation was emplaced to measure stress changes, relative motion of the rock mass, and tunnel closures during three years of heating from thermally decaying heat sources, followed by a six-month cooldown period. The calculations reported here were performed using the best available input parameters, thermal and mechanical properties, and power levels which were directly measured or inferred from measurements made during the test. This report documents the results of these calculations and compares the results with selected measurements made during heating and cooling of the SFT-C.

Butkovich, T.R.; Patrick, W.C.

1986-02-01T23:59:59.000Z

380

Nuclear Waste Technical Review Board Performance Plan  

E-Print Network [OSTI]

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

Note: This page contains sample records for the topic "naval spent nuclear" 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

Maintenance practices for emergency diesel generator engines onboard United States Navy Los Angeles class nuclear submarines  

E-Print Network [OSTI]

The United States Navy has recognized the rising age of its nuclear reactors. With this increasing age comes increasing importance of backup generators. In addition to the need for decay heat removal common to all (naval ...

Hawks, Matthew Arthur

2006-01-01T23:59:59.000Z

382

2012 Annual Planning Summary for Naval Reactors | Department of Energy  

Energy Savers [EERE]

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

383

Fallon Naval Air Station Geothermal Area | Open Energy Information  

Open Energy Info (EERE)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Home Page on Google Bookmark EERE: Alternative Fuels Data Center Home5b9fcbce19 NoPublic Utilities Address:011-DNA Jump37.California: Energy Resources Jump4748456°,Fallon Naval Air Station

384

Spent Fuel Transportation Risk Assessment  

Office of Environmental Management (EM)

AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE:1 First Use of Energy for All Purposes (Fuel and Nonfuel),Feet) Year Jan Feb Mar Apr May Jun Jul(Summary) "of EnergyEnergyENERGYWomen Owned Small Business Webinar June 20,Department ofSpecialEnergy SpelmanSpent

385

The characteristic assessment of spent ion exchange resin from PUSPATI TRIGA REACTOR (RTP) for immobilization process  

SciTech Connect (OSTI)

In this paper, spent ion exchange resin generated from PUSPATI TRIGA reactor (RTP) in Malaysian Nuclear Agency were characterized based on the water content, radionuclide content and radionuclide leachability. The result revealed that the water content in the spent resin is 48%. Gamma spectrometry analysis indicated the presence of {sup 134}Cs, {sup 137}Cs, {sup 152}Eu, {sup 54}Mn, {sup 58}Co, {sup 60}Co and {sup 65}Zn. The leachability test shows a small concentrations (<1 Bq/l) of {sup 152}Eu and {sup 134}Cs were leached out from the spent resin while {sup 60}Co activity concentrations slightly exceeded the limit generally used for industrial wastewater i.e. 1 Bq/l. Characterization of spent ion exchange resin sampled from RTP show that this characterization is important as a basis to immobilize this radioactive waste using geopolymer technology.

Wahida, Nurul [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor, Malaysia and Malaysian Nuclear Agency, Bangi 43000 Kajang, Selangor (Malaysia); Yasir, Muhamad Samudi; Majid, Amran Ab; Irwan, M. N. [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor (Malaysia); Wahab, Mohd Abd; Marzukee, Nik; Paulus, Wilfred; Phillip, Esther; Thanaletchumy [Malaysian Nuclear Agency, Bangi 43000 Kajang, Selangor (Malaysia)

2014-09-03T23:59:59.000Z

386

Coastal Inundation due to Tide, Surge, Waves, and Sea Level Rise at Naval Station Norfolk  

E-Print Network [OSTI]

Coastal Inundation due to Tide, Surge, Waves, and Sea Level Rise at Naval Station Norfolk Honghai of future sea level rise (SLR) scenarios and to evaluate the potential coastal inundation at Naval Station and sea level rise threats to coastal residents and coastal military facilities, the US Strategic

US Army Corps of Engineers

387

a contaminant in decline: long-term tbt monitoring at a naval base in Western australia  

E-Print Network [OSTI]

a contaminant in decline: long-term tbt monitoring at a naval base in Western australia john a planulatus) in and around the RAN naval base in Cockburn Sound, WesternAustralia, was initiated and continued, Australia. 2 Current address: ES Link Services Pty Ltd, PO Box 10, Castlemaine, VIC 3450, Australia. 3

Burgman, Mark

388

An approach to determine a defensible spent fuel ratio.  

SciTech Connect (OSTI)

Sabotage of spent nuclear fuel casks remains a concern nearly forty years after attacks against shipment casks were first analyzed and has a renewed relevance in the post-9/11 environment. A limited number of full-scale tests and supporting efforts using surrogate materials, typically depleted uranium dioxide (DUO2), have been conducted in the interim to more definitively determine the source term from these postulated events. In all the previous studies, the postulated attack of greatest interest was by a conical shape charge (CSC) that focuses the explosive energy much more efficiently than bulk explosives. However, the validity of these large-scale results remain in question due to the lack of a defensible Spent Fuel Ratio (SFR), defined as the amount of respirable aerosol generated by an attack on a mass of spent fuel compared to that of an otherwise identical DUO2 surrogate. Previous attempts to define the SFR have resulted in estimates ranging from 0.42 to 12 and include suboptimal experimental techniques and data comparisons. Different researchers have suggested using SFR values of 3 to 5.6. Sound technical arguments exist that the SFR does not exceed a value of unity. A defensible determination of the SFR in this lower range would greatly reduce the calculated risk associated with the transport and dry storage of spent nuclear fuel. Currently, Oak Ridge National Laboratory (ORNL) is in possession of several samples of spent nuclear fuel (SNF) that were used in the original SFR studies in the 1980's and were intended for use in a modern effort at Sandia National Laboratories (SNL) in the 2000's. A portion of these samples are being used for a variety of research efforts. However, the entirety of SNF samples at ORNL is scheduled for disposition at the Waste Isolation Pilot Plant (WIPP) by approximately the end of 2015. If a defensible SFR is to be determined for use in storage and transportation security analyses, the need to begin this effort is urgent in order to secure the only known available SNF samples with a clearly defined path to disposal.

Durbin, Samuel G.; Lindgren, Eric Richard

2014-03-01T23:59:59.000Z

389

Office of Naval Petroleum and Oil Shale Reserves  

E-Print Network [OSTI]

Worldwide supplies of conventional oil will soon reach a peak production rate and begin an irreversible long-term decline. Options to augment liquid fuel supplies in the United States have once again begun to focus on oil shale as long-term source of reliable, affordable, and secure oil. The United States government has long recognized the strategic potential of the nations vast oil shale resources to support national security. President Taft in 1912 established an Office of Naval Petroleum and Oil Shale Reserves and charged that office with ensuring oil for naval military operations. This office continues to oversee the United States strategic interest in oil shale. Americas 2 trillion barrel oil shale resource is recognized as having the same production potential as Canadas tar sands. Tar sand production, initiated in the 1960s, has increased steadily to more than 1 million per barrels/day and is moving toward a near-term goal of 2.5 million barrels per day by 2017. This amount of oil is equivalent to the volume of oil currently imported by the United States from Middle East countries. Tar sands production has enabled Canada to add 174 billion barrels to its recoverable oil reserves, making Canadas proved reserves second only to those of Saudi Arabia.

unknown authors

390

Licensing of spent fuel dry storage and consolidated rod storage: A Review of Issues and Experiences  

SciTech Connect (OSTI)

The results of this study, performed by Pacific Northwest Laboratory (PNL) and sponsored by the US Department of Energy (DOE), respond to the nuclear industry's recommendation that a report be prepared that collects and describes the licensing issues (and their resolutions) that confront a new applicant requesting approval from the US Nuclear Regulatory Commission (NRC) for dry storage of spent fuel or for large-scale storage of consolidated spent fuel rods in pools. The issues are identified in comments, questions, and requests from the NRC during its review of applicants' submittals. Included in the report are discussions of (1) the 18 topical reports on cask and module designs for dry storage fuel that have been submitted to the NRC, (2) the three license applications for dry storage of spent fuel at independent spent fuel storage installations (ISFSIs) that have been submitted to the NRC, and (3) the three applications (one of which was later withdrawn) for large-scale storage of consolidated fuel rods in existing spent fuel storage pools at reactors that were submitted tot he NRC. For each of the applications submitted, examples of some of the issues (and suggestions for their resolutions) are described. The issues and their resolutions are also covered in detail in an example in each of the three subject areas: (1) the application for the CASTOR V/21 dry spent fuel storage cask, (2) the application for the ISFSI for dry storage of spent fuel at Surry, and (3) the application for full-scale wet storage of consolidated spent fuel at Millstone-2. The conclusions in the report include examples of major issues that applicants have encountered. Recommendations for future applicants to follow are listed. 401 refs., 26 tabs.

Bailey, W.J.

1990-02-01T23:59:59.000Z

391

Naval Petroleum Reserve No. 1 (Elk Hills): Supplemental environmental impact statement. Record of decision  

SciTech Connect (OSTI)

Pursuant to the Council on Environmental Quality regulations, which implement the procedural provisions of the National Environmental Policy Act, and the US Department of Energy National Environmental Policy Act regulations, the Department of Energy, Office of Fossil Energy, is issuing a Record of Decision on the continued operation of Naval Petroleum Reserve No. 1, Kern County, California. The Department of Energy has decided to continue current operations at Naval Petroleum Reserve No. 1 and implement additional well drilling, facility development projects and other activities necessary for continued production of Naval Petroleum Reserve No. 1 in accordance with the requirements of the Naval Petroleum Reserves Production Act of 1976. The final Supplemental Environmental Impact Statement, entitled ``Petroleum Production at Maximum Efficient Rate, Naval Petroleum Reserve No. 1 (Elk Hills), Kern County, California (DOE/SEIS-0158),`` was released on September 3, 1993.

Not Available

1994-02-01T23:59:59.000Z

392

EIS-0218: Proposed Nuclear Weapons Nonproliferation Policy Concerning...  

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

the potential environmental impacts of adopting a policy to manage foreign research reactor spent nuclear fuel containing uranium enriched in the United States. In...

393

Principal Media Contact: DT Townsend Savannah River Nuclear Solutions...  

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

Nuclear Science Week, dozens of students and teachers recently spent time with Savannah River National Laboratory (SRNL) scientists, touring the Department of Energy's (DOE)...

394

apex nuclear fuel: Topics by E-print Network  

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

fuel fabrication, in-core physics and fuel management of uranium, thorium and other fuel types, ... Kazimi, Mujid S. 19 Nuclear Waste Imaging and Spent Fuel Verification by...

395

Characterization of spent fuel approved testing material: ATM-103  

SciTech Connect (OSTI)

The characterization data obtained to date are described for Approved Testing Material (ATM)-103, which is spent fuel from Assembly D101 of pressurized-water reactor Calvert Cliffs, No. 1. This report is one in a series being written by the Materials Characterization Center (MCC) at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US nuclear waste repository program. ATM-103 consists of 176 full-length irradiated fuel rods with rod-average burnups of about 2600 GJ/kgM (30 MWd/kgM) and less than 1% fission gas release. Characterization data include 1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; 2) isotopic gamma scans; 3) fission gas analyses; 4) ceramography of the fuel and metallography of the cladding; 5) special fuels studies involving analytical transmission electron microscopy (AEM); 6) calculated nuclide inventories and radioactivities in the fuel and cladding; and 7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report. 10 refs., 103 figs., 63 tabs.

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

1988-04-01T23:59:59.000Z

396

Characterization of spent fuel approved testing material---ATM-105  

SciTech Connect (OSTI)

The characterization data obtained to data are described for Approved Testing Material 105 (ATM-105), which is spent fuel from Bundles CZ346 and CZ348 of the Cooper Nuclear Power Plant, a boiling-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-105 consists of 88 full-length irradiated fuel rods with rod-average burnups of about 2400 GJ/kgM (28 MWd/kgM) and expected fission gas release of about 1%. Characterization data include (1) descriptions of as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel are being conducted and will be included in planned revisions of this report.

Guenther, R.J.; Blahnik, D.E.; Campbell, T.K.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

1991-12-01T23:59:59.000Z

397

Characterization of spent fuel approved testing material--ATM-104  

SciTech Connect (OSTI)

The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding.

Guenther, R.J.; Blahnik, D.E.; Jenquin, U.P.; Mendel, J.E.; Thomas, L.E.; Thornhill, C.K.

1991-12-01T23:59:59.000Z

398

Thermoelectric powered wireless sensors for spent fuel monitoring  

SciTech Connect (OSTI)

This paper describes using thermoelectric generators to power wireless sensors to monitor spent nuclear fuel during dry-cask storage. OrigenArp was used to determine the decay heat of the spent fuel at different times during the service life of the dry-cask. The Engineering Equation Solver computer program modeled the temperatures inside the spent fuel storage facility during its service life. The temperature distribution in a thermoelectric generator and heat sink was calculated using the computer program Finite Element Heat Transfer. From these temperature distributions the power produced by the thermoelectric generator was determined as a function of the service life of the dry-cask. In addition, an estimation of the path loss experienced by the wireless signal can be made based on materials and thickness of the structure. Once the path loss is known, the transmission power and thermoelectric generator power requirements can be determined. This analysis estimates that a thermoelectric generator can produce enough power for a sensor to function and transmit data from inside the dry-cask throughout its service life. (authors)

Carstens, T.; Corradini, M.; Blanchard, J. [Dept. of Engineering Physics, Univ. of Wisconsin-Madison, Madison, WI 53706 (United States); Ma, Z. [Dept. of Electrical and Computer Engineering, Univ. of Wisconsin-Madison, Madison, WI 53706 (United States)

2011-07-01T23:59:59.000Z

399

Determining Reactor Flux from Xenon-136 and Cesium-135 in Spent Fuel  

E-Print Network [OSTI]

The ability to infer the reactor flux from spent fuel or seized fissile material would enhance the tools of nuclear forensics and nuclear nonproliferation significantly. We show that reactor flux can be inferred from the ratios of xenon-136 to xenon-134 and cesium-135 to cesium-137. If the average flux of a reactor is known, the flux inferred from measurements of spent fuel could help determine whether that spent fuel was loaded as a blanket or close to the mid-plane of the reactor. The cesium ratio also provides information on reactor shutdowns during the irradiation of fuel, which could prove valuable for identifying the reactor in question through comparisons with satellite reactor heat monitoring data. We derive analytic expressions for these correlations and compare them to experimental data and to detailed reactor burn simulations. The enrichment of the original uranium fuel affects the correlations by up to 3 percent, but only at high flux.

A. C. Hayes; Gerard Jungman

2012-05-30T23:59:59.000Z

400

Naval Reactors Facility environmental monitoring report, calendar year 2001  

SciTech Connect (OSTI)

The results of the radiological and nonradiological environmental monitoring programs for 2001 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U. S. Environmental Protection Agency and the U. S. Department of Energy.

NONE

2002-12-31T23:59:59.000Z

Note: This page contains sample records for the topic "naval spent nuclear" from the National Library of EnergyBeta (NLEBeta).
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they are not comprehensive nor are they the most current set.
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401

Naval Reactors Facility Environmental Monitoring Report, Calendar Year 2003  

SciTech Connect (OSTI)

The results of the radiological and nonradiological environmental monitoring programs for 2003 at the Naval Reactors Facility are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with Federal and State regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the U.S. Environmental Protection Agency and the U.S. Department of Energy.

None

2003-12-31T23:59:59.000Z

402

1997 environmental monitoring report for the Naval Reactors Facility  

SciTech Connect (OSTI)

The results of the radiological and nonradiological environmental monitoring programs for 1997 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE).

NONE

1997-12-31T23:59:59.000Z

403

National Report Joint Convention on the Safety of Spent Fuel...  

Office of Environmental Management (EM)

National Report Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management National Report Joint Convention on the Safety of Spent...

404

Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition...  

Office of Environmental Management (EM)

Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition Strategy Lessons Learned Report, NNSA, Feb 2010 Idaho Spent Fuel Facility (ISFF) Project, Appropriate Acquisition...

405

Operating experience feedback report: Assessment of spent fuel cooling. Volume 12  

SciTech Connect (OSTI)

This report documents the results of an independent assessment by a team from the Office of Analysis and Evaluation of Operational Data of spent-fuel-pool (SFP) cooling in operating nuclear power plants. The team assessed the likelihood and consequences of an extended loss of SFP cooling and suggested corrective actions, based on their findings.

Ibarra, J.G.; Jones, W.R.; Lanik, G.F.; Ornstein, H.L.; Pullani, S.V.

1997-02-01T23:59:59.000Z

406

Nuclear Regulatory Commission issuances, January 1997. Volume 45, Number 1  

SciTech Connect (OSTI)

This book contains issuances of the Atomic Safety and Licensing Board, Nuclear Regulatory Commission and Director`s Decision for January 1997. The issuances concern Sequoyah Fuels Corporation and General Atomics Gore, Oklahoma Site decontamination and decommissioning funding; Louisiana Energy Services, Claiborne Enrichment Center denies appeal to review emergency planning; General Public Utilities Nuclear Corporation, Oyster Creek Nuclear Generating station, challenges to technical specifications concerning spent fuel pool; and Consumers Power Company, Palisades Nuclear Plant dry cask storage of spent nuclear fuel.

NONE

1997-01-01T23:59:59.000Z

407

The Evolution of Dry Spent Fuel Storage in the United States  

SciTech Connect (OSTI)

This paper reviews the evolution of Dry Spent Fuel storage technology and application in the United States. Dating back to the legislation signed by Jimmy Carter on April 7, 1977, to outlaw spent fuel reprocessing, the nations spent fuel pools are gradually becoming filled to capacity. This has necessitated the development of new technologies to store spent fuel in dry casks, predominantly at nuclear power plant sites, awaiting the availability of the federal repository at Yucca Mountain. Site-specific conditions and changes in types of fuel being discharged from reactors have driven a constant evolution of technologies to support this critical need. This paper provides an overview of those changes, which have influenced the evolution of dry storage technology. Focus is provided more towards current technology and cask loading practices, as opposed to those technologies, which are no longer in heavy use. Detailed pictorial material is presented showing the loading sequences of various systems in current use. This paper provides a critical primer on Dry Spent Fuel Storage technology. It provides anyone who is new to dry storage, or who is contemplating initiating dry storage at a nuclear plant site, with useful background and history upon which to build programmatic decisions. (authors)

McGough, M.S. [Duratek Inc., 695 Bamesley Lane, Alpharetta, GA 30022 (United States); Bland, D.W. [TriVis, Inc., 1001 Yeager Parkway, Pelham, AL 35124 (United States)

2006-07-01T23:59:59.000Z

408

End-to-end calculation of the radiation characteristics of VVER-1000 spent fuel assemblies  

SciTech Connect (OSTI)

The results of end-to-end calculation of the radiation characteristics of VVER-1000 spent nuclear fuel are presented. Details of formation of neutron and gamma-radiation sources are analyzed. Distributed sources of different types of radiation are considered. A comparative analysis of calculated radiation characteristics is performed with the use of nuclear data from different ENDF/B and EAF files and ANSI/ANS and ICRP standards.

Linge, I. I.; Mitenkova, E. F., E-mail: mit@ibrae.ac.ru; Novikov, N. V. [Russian Academy of Sciences, Nuclear Safety Institute (Russian Federation)

2012-12-15T23:59:59.000Z

409

Fluoride removal from water with spent catalyst  

SciTech Connect (OSTI)

The adsorption of fluoride from water with spent catalyst was studied. Adsorption density of fluoride decreased with increasing pH. Linear adsorption isotherm was utilized to describe the adsorption reaction. The adsorption was a first-order reaction, and the rate constant increased with decreasing surface loading. Adsorption reaction of fluoride onto spent catalyst was endothermic, and the reaction rate increased slightly with increasing temperature. Fluoro-alumino complex and free fluoride ion were involved in the adsorption reaction. It is proposed that both the silica and alumina fractions of spent catalyst contribute to the removal of fluoride from aqueous solution. Coulombic interaction is proposed as the major driving force of the adsorption reaction of fluoride onto spent catalyst.

Lai, Y.D.; Liu, J.C. [National Taiwan Institute of Technology, Taipei (Taiwan, Province of China)

1996-12-01T23:59:59.000Z

410

HYDRAULIC CEMENT PREPARATION FROM LURGI SPENT SHALE  

E-Print Network [OSTI]

cement from spent oil shale," Vol. 10, No. 4, p. 54S,Colorado's primary oil shale resource for vertical modifiedSimulated effects of oil-shale development on the hydrology

Mehta, P.K.

2013-01-01T23:59:59.000Z

411

HYDRAULIC CEMENT PREPARATION FROM LURGI SPENT SHALE  

E-Print Network [OSTI]

hydraulic cement from spent oil shale," Vol. 10, No. 4, p.J. W. , "Colorado's primary oil shale resource for verticalSimulated effects of oil-shale development on the hydrology

Mehta, P.K.

2013-01-01T23:59:59.000Z

412

Application and analysis of stiffened side shell panel failure for naval patrol craft  

E-Print Network [OSTI]

Over their lifetime, naval patrol craft are subjected to many different types of loading scenarios, most of which are perfectly safe. In rare instances, through a variety of different reasons, these craft are loaded beyond ...

Mothander, Matthew K. A., Lieutenant (Matthew Kristian Alden)

2009-01-01T23:59:59.000Z

413

Naval ship propulsion and electric power systems selection for optimal fuel consumption  

E-Print Network [OSTI]

Although propulsion and electric power systems selection is an important part of naval ship design, respective decisions often have to be made without detailed ship knowledge (resistance, propulsors, etc.). Propulsion and ...

Sarris, Emmanouil

2011-01-01T23:59:59.000Z

414

Design and analysis of a permanent magnet generator for naval applications  

E-Print Network [OSTI]

This paper discusses the electrical and magnetic design and analysis of a permanent magnet generation module for naval applications. Numerous design issues are addressed and several issues are raised about the potential ...

Rucker, Jonathan E. (Jonathan Estill)

2005-01-01T23:59:59.000Z

415

Modular machinery arrangement and its impact in early-stage naval electric ship design  

E-Print Network [OSTI]

Electrical power demands for naval surface combatants are projected to rise with the development of increasingly complex and power intensive combat systems. This trend also coincides with the need of achieving maximum fuel ...

Jurkiewicz, David J. (David James)

2012-01-01T23:59:59.000Z

416

A study of the Naval Construction Force project material supply chain  

E-Print Network [OSTI]

The Naval Construction Force (NCF) performs construction projects in all areas of the world during both peacetime and war. While some of these projects occur in populated areas where project materials are readily available, ...

Stasick, Steven J. (Steven James), 1970-

2004-01-01T23:59:59.000Z

417

Analysis of harmonic distortion in an Integrated Power System for naval applications  

E-Print Network [OSTI]

This research quantifies the voltage distortion over the broad range of operating conditions experienced by a Naval warship. A steady state model of an Integrated Power System (IPS) was developed in a commercially available ...

West, Edward G., S.M. Massachusetts Institute of Technology

2005-01-01T23:59:59.000Z