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1

EA-1889: Disposal of Decommissioned, Defueled Naval Reactor Plants from USS  

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

89: Disposal of Decommissioned, Defueled Naval Reactor Plants 89: Disposal of Decommissioned, Defueled Naval Reactor Plants from USS Enterprise (CVN 65) at the Hanford Site, Richland, Washington EA-1889: Disposal of Decommissioned, Defueled Naval Reactor Plants from USS Enterprise (CVN 65) at the Hanford Site, Richland, Washington Summary This EA, prepared by the Department of the Navy, evaluates the environmental impacts of the disposal of decommissioned, defueled, naval reactor plants from the USS Enterprise at DOE's Hanford Site, Richland, Washington. DOE participated as a cooperating agency in the preparation of this EA. The Department of the Navy issued its FONSI on August 23, 2012. Public Comment Opportunities No public comment opportunities available at this time. Documents Available for Download August 23, 2012

2

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

National Nuclear Security Administration (NNSA)

Naval Reactors Office The Naval Nuclear Propulsion Program provides militarily effective nuclear propulsion plants and ensures their safe, reliable and long-lived operation. This...

3

Naval Nuclear Propulsion Plants | National Nuclear Security Administra...  

National Nuclear Security Administration (NNSA)

Naval Nuclear Propulsion Plants In naval nuclear propulsion plants, fissioning of uranium atoms in the reactor core produces heat. Because the fission process also produces...

4

Management of Naval Reactors' Cyber Security Program, OIG-0884  

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

Naval Reactors' Naval Reactors' Cyber Security Program DOE/IG-0884 April 2013 U.S. Department of Energy Office of Inspector General Office of Audits and Inspections Department of Energy Washington, DC 20585 April 12, 2013 MEMORANDUM FOR THE SECRETARY FROM: Gregory H. Friedman Inspector General SUBJECT: INFORMATION: Audit Report on "Management of Naval Reactors' Cyber Security Program" INTRODUCTION AND OBJECTIVE The Naval Reactors Program (Naval Reactors), an organization within the National Nuclear Security Administration, provides the military with safe and reliable nuclear propulsion plants to power warships and submarines. Naval Reactors maintains responsibility for activities supporting the United States Naval fleet nuclear propulsion systems, including research and

5

Audit Report - Naval Reactors Information Technology System Development Efforts, IG-0879  

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

Naval Reactors Information Naval Reactors Information Technology System Development Efforts DOE/IG-0879 December 2012 U.S. Department of Energy Office of Inspector General Office of Audits & Inspections Department of Energy Washington, DC 20585 December 21, 2012 MEMORANDUM FOR THE ADMINISTRATOR, NATIONAL NUCLEAR SECURITY ADMINISTRATION FROM: Gregory H. Friedman Inspector General SUBJECT: INFORMATION: Audit Report on the "Naval Reactors Information Technology System Development Efforts" INTRODUCTION AND OBJECTIVE The Naval Reactors Program (Naval Reactors), an organization within the National Nuclear Security Administration, was established to provide the military with safe and reliable nuclear propulsion plants to power warships and submarines. Naval Reactors maintains responsibility

6

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

National Nuclear Security Administration (NNSA)

to skip to the main content Facebook Flickr RSS Twitter YouTube More About NNSA's Naval Reactors Office | National Nuclear Security Administration Our Mission Managing the...

7

1996 environmental monitoring report for the Naval Reactors Facility  

Science Conference Proceedings (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

8

Request for Naval Reactors Comment on Proposed Prometheus Space Flight Nuclear Reactor High Tier Reactor Safety Requirements and for Naval Reactors Approval to Transmit These Requirements to JPL  

SciTech Connect

The purpose of this letter is to request Naval Reactors comments on the nuclear reactor high tier requirements for the PROMETHEUS space flight reactor design, pre-launch operations, launch, ascent, operation, and disposal, and to request Naval Reactors approval to transmit these requirements to Jet Propulsion Laboratory to ensure consistency between the reactor safety requirements and the spacecraft safety requirements. The proposed PROMETHEUS nuclear reactor high tier safety requirements are consistent with the long standing safety culture of the Naval Reactors Program and its commitment to protecting the health and safety of the public and the environment. In addition, the philosophy on which these requirements are based is consistent with the Nuclear Safety Policy Working Group recommendations on space nuclear propulsion safety (Reference 1), DOE Nuclear Safety Criteria and Specifications for Space Nuclear Reactors (Reference 2), the Nuclear Space Power Safety and Facility Guidelines Study of the Applied Physics Laboratory.

D. Kokkinos

2005-04-28T23:59:59.000Z

9

Fuel Cell Power Plant Experience Naval Applications  

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

clean clean Fuel Cell Power Plant Experience Naval Applications US Department of Energy/ Office of Naval Research Shipboard Fuel Cell Workshop Washington, DC March 29, 2011 FuelCell Energy, the FuelCell Energy logo, Direct FuelCell and "DFC" are all registered trademarks (®) of FuelCell Energy, Inc. *FuelCell Energy, Inc. *Renewable and Liquid Fuels Experience *HTPEM Fuel Cell Stack for Shipboard APU *Solid Oxide Experience and Applications DOE-ONR Workshop FuelCell Energy, the FuelCell Energy logo, Direct FuelCell and "DFC" are all registered trademarks (®) of FuelCell Energy, Inc. FuelCell Energy, Inc. * Premier developer of fuel cell technology - founded in 1969 * Over 50 power installations in North America, Europe, and Asia * Industrial, commercial, utility

10

NAVAL REACTORS PHYSICS HANDBOOK. VOLUME I. SELECTED BASIC TECHNIQUES  

SciTech Connect

The purpose of this work is to present the most pertinent parts of the body of physics knowledge which has been built up in the course of the Naval and Shippingport (PWR) Reactor Programs, with the aim of providing a background of understanding for those interested in nuclear core design. Volume 1 of this handbook was planned to bring together topics in the basic theoretical and experimental material which are of especially wide interest, including those common to both thermal and intermediate neutron energy reactor types. The physics design of light water-moderated and -cooled reactors is covered in Volume 2 (classified), and that of intermediate neutron-energy power reactors in Volume 3. The emphasis in Volume 1 is thus on light water reactor systems, and as many recent advances in reactor physics of the Naval and Shippingport Reactor Programs as possible have been included.

Radkowsky, A. ed.

1964-01-01T23:59:59.000Z

11

Tag: Naval Reactors | Y-12 National Security Complex  

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

Naval Reactors Naval Reactors Tag: Naval Reactors Displaying 1 - 7 of 7... Category: Employees & Retirees "Cook"ing at Y-12 for 70 years We have an enduring mission. Y-12 plays a key role in it. And a nuclear deterrent remains the ultimate insurance policy for America. More... Category: News Y-12 Knows Uranium Y-12 produces many forms of uranium. More... Category: News A Rich Resource Requires Recovery Given the value and scarcity of enriched uranium, Y-12 recycles and reuses as much of it as possible. More... Category: News Seawolf Manufacturing Challenge For decades, attack submarines were either fast or quiet - but never both. The fast subs were so loud that an enemy could hear them long before they were within striking distance. More... Category: News Reliable fuel source

12

Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities  

National Nuclear Security Administration (NNSA)

Defense Nuclear Nonproliferation and Naval Reactors Activities Defense Nuclear Nonproliferation and Naval Reactors Activities before the House Committee on Appropriations Subcommittee on Energy & Water Development | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Congressional Testimony > Statement on Defense Nuclear

13

Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities  

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

Defense Nuclear Nonproliferation and Naval Reactors Activities Defense Nuclear Nonproliferation and Naval Reactors Activities before the House Committee on Appropriations Subcommittee on Energy & Water Development | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Congressional Testimony > Statement on Defense Nuclear

14

U.S. Department Of Energy Naval Reactors Laboratory Field Office  

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

Of Energy Naval Reactors Laboratory Field Office Knolls Laboratory National Environmental Policy Act (NEPA) Categorical Exclusion (CX) Determination Summary Form BUILDING A10...

15

Naval Reactors Facility environmental monitoring report, calendar year 2001  

Science Conference Proceedings (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

16

Naval Reactors Facility environmental monitoring report, calendar year 1999  

Science Conference Proceedings (OSTI)

The results of the radiological and nonradiological environmental monitoring programs for 1999 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 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 (EPA) and the U.S. Department of Energy (DOE).

None

2000-12-01T23:59:59.000Z

17

Naval Reactors Facility environmental monitoring report, calendar year 2000  

Science Conference Proceedings (OSTI)

The results of the radiological and nonradiological environmental monitoring programs for 2000 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 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 (EPA) and the U.S. Department of Energy (DOE).

None

2001-12-01T23:59:59.000Z

18

Naval Reactors Facility Environmental Monitoring Report, Calendar Year 2003  

Science Conference Proceedings (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

19

EIS-0259 Final Environmental Impact Statement On The Disposal Of Decommissioned, Defueled Cruiser, Ohio Class, And Los Angeles Class Naval Reactor Plants  

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

-. -. 1. PURPOSE AND NEED U.S. Navy nuclear ships are decommissioned and defieled at the end of their usefi Metime, when the cost of continued operation is not justified by their fitary capabfity, or when the ship is no longer needed. The Navy needs to disposition the reactor compartments born defieled and decommissioned ctisers, and OHIO Class and LOS ANGELES Class submarines. The number of reactor compartments under consideration by this Environmental Impact Statement is about 100. These reactor compartments are in addition to the pre-LOS ANGELES Class submarines tieady being disposed of under the Na&s 1984 Find Environmental hpact Statement (USN, 1984a). Newer types of U.S. Navy nuclear-powered stips that are not expected to be decommissioned in the next 20 years (e.g., aircraft carriers, SEAWOLF Class submarines) are not included in this fid Enviromentd Impact Statement.

20

EIS-0259 Final Environmental Impact Statement On The Disposal Of Decommissioned, Defueled Cruiser, Ohio Class, And Los Angeles Class Naval Reactor Plants  

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

... -. FEASIBILITY STUDY FOR LEAD REMOVAL FROM AND STRUCTURAL RESTORATION OF CRUISER, OHIO, AND LOS ANGELES CLASS REACTOR COMPARTMENT DISPOSAL PACWGES Appendix A A-i . Table of Contents E~CUT~ S~Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..A.3 1. ~TRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .." . . . . . . . . . . . . . . . ..A.3 2. DESCR~TION OF S~ELD~G LEAD CONTmD ~ REACTOR commmm PAcmGEs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..A.4 2.1 PermanentSMeldingLead . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..A-4 2.2 MisceUaneousLead . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..A-5 2.3 Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................"A-5 2.4 Assumptions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .................A-5 3. SHIELD~G LEAD REMOVQ

Note: This page contains sample records for the topic "naval reactor plants" 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

September 10, 2003, Board Public Meeting - Naval Reactors Approach to Oversight  

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

4 4 25 22 As I previously mentioned this morning, we will receive testimony from experienced representatives from other organizations. First, I would like to welcome representatives from the U.S. Naval Sea Systems Command, Mr. Thomas Beckett and Mr. Storm Kauffman. If you would be kind enough to give your names and titles so the stenographer can identify you for the record. MR. BECKETT: Thank you, Mr. Chairman. Thomas H. Beckett. I ' m the Deputy Director for Naval Reactors, a joint Department of the Navy/Department of Energy Program. MR. KAUFFMAN: Storm Kauffman. I ' m the Director of Reactor Safety and Analysis for the Naval Reactors Program CHAIRMAN CONWAY: Mr. Beckett. MR. BECKETT: Thank you, Mr. Chairman, and let me thank you and the other Board Members f

22

Naval Nuclear Propulsion Program Directorate, Washington ...  

Science Conference Proceedings (OSTI)

Naval Nuclear Propulsion Program Directorate, Washington, DC. NVLAP Lab Code: 100565-10. Address and Contact Information: Naval Reactors ...

2013-07-26T23:59:59.000Z

23

NEUTRONIC REACTOR POWER PLANT  

DOE Patents (OSTI)

This patent relates to a nuclear reactor power plant incorporating an air-cooled, beryllium oxide-moderated, pebble bed reactor. According to the invention means are provided for circulating a flow of air through tubes in the reactor to a turbine and for directing a sidestream of the circu1ating air through the pebble bed to remove fission products therefrom as well as assist in cooling the reactor. (AEC)

Metcalf, H.E.

1962-12-25T23:59:59.000Z

24

DOE - Office of Legacy Management -- Naval Ordnance Plant - MI 0-03  

Office of Legacy Management (LM)

Plant - MI 0-03 Plant - MI 0-03 FUSRAP Considered Sites Site: NAVAL ORDNANCE PLANT (MI.0-03) Eliminated from further consideration under FUSRAP - Referred to DoD for action Designated Name: Not Designated Alternate Name: None Location: Centerline , Michigan MI.0-03-1 Evaluation Year: 1987 MI.0-03-1 Site Operations: Assembled bomb components. MI.0-03-1 Site Disposition: Eliminated - No Authority - Referred to DoD MI.0-03-1 Radioactive Materials Handled: None Indicated Primary Radioactive Materials Handled: None Radiological Survey(s): None Indicated Site Status: Eliminated from further consideration under FUSRAP - Referred to DoD for action MI.0-03-1 Also see Documents Related to NAVAL ORDNANCE PLANT MI.0-03-1 - DOE Letter; J.Fiore to C.Shafer; Subject: Information on

25

Tag: Naval Reactors  

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

9/all en "Cook"ing at Y-12 for 70 9/all en "Cook"ing at Y-12 for 70 years http://www.y12.doe.gov/employees-retirees/y-12-times/cooking-y-12-70-years

We have an enduring mission. Y-12 plays a key role in it. And a nuclear deterrent remains the ultimate insurance policy for America.

Plant maintenance and advanced reactors, 2005  

SciTech Connect

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: First U.S. EPRs in 2015, by Ray Ganthner, Framatome ANP; Pursuing several opportunities, by William E. (Ed) Cummins, Westinghouse Electric Company; Vigorous plans to develop advanced reactors, by Yuliang Sun, Tsinghua University, China; Multiple designs, small and large, by Kumiaki Moriya, Hitachi Ltd., Japan; Sealed and embedded for safety and security, by Handa Norihiko, Toshiba Corporation, Japan; Scheduled online in 2010, by Johan Slabber, PMBR (Pty) Ltd., South Africa; Multi-application reactors, by Nikolay G. Kodochigov, OKBM, Russia; Six projects under budget and on schedule, by David F. Togerson, AECL, Canada; Creating a positive image, by Scott Peterson, Nuclear Energy Institute (NEI); Advanced plans for nuclear power's renaissance, by John Cleveland, International Atomic Energy Agency, Austria; and, Plant profile: last five outages in less than 20 days, by Beth Rapczynski, Exelon Nuclear.

Agnihotri, Newal (ed.)

2005-09-15T23:59:59.000Z

27

Plant maintenance and advanced reactors, 2007  

Science Conference Proceedings (OSTI)

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: A new day for energy in America; Committed to success more than ever, by Andy White, GE--Hitachi Nuclear Energy; Competitive technology for decades, by Steve Tritch, Westinghouse Electric Company; Pioneers of positive community relationship, by Exelon Nuclear; A robust design for 60-years, by Ray Ganthner, Areva; Aiming at no evacuation plants, by Kumiaki Moriya, Hitachi-GE Nuclear Energy, Ltd.; and, Desalination and hydrogen economy, by Dr. I. Khamis, International Atomic Energy Agency. Industry innovation articles in this issue are: Reactor vessel closure head project, by Jeff LeClair, Prairie Island Nuclear Generating Plant; and Submersible remote-operated vehicle, by Michael S. Rose, Entergy's Fitzpatrick Nuclear Station.

Agnihotri, Newal (ed.)

2007-09-15T23:59:59.000Z

28

Generic small modular reactor plant design.  

SciTech Connect

This report gives an overview of expected design characteristics, concepts, and procedures for small modular reactors. The purpose of this report is to provide those who are interested in reducing the cost and improving the safety of advanced nuclear power plants with a generic design that possesses enough detail in a non-sensitive manner to give merit to their conclusions. The report is focused on light water reactor technology, but does add details on what could be different in a more advanced design (see Appendix). Numerous reactor and facility concepts were used for inspiration (documented in the bibliography). The final design described here is conceptual and does not reflect any proposed concept or sub-systems, thus any details given here are only relevant within this report. This report does not include any design or engineering calculations.

Lewis, Tom Goslee,; Cipiti, Benjamin B.; Jordan, Sabina Erteza; Baum, Gregory A.

2012-12-01T23:59:59.000Z

29

Plant maintenance and advanced reactors issue, 2004  

SciTech Connect

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Optimism about the future of nuclear power, by Ruth G. Shaw, Duke Power Company; Licensed in three countries, by GE Energy; Enhancing public acceptance, by Westinghouse Electric Company; Standardized MOV program, by Ted Neckowicz, Exelon; Inservice testing, by Steven Unikewicz, U.S. Nuclear Regulatory Commission; Asian network for education, Fatimah Mohd Amin, Malaysian Institute for Nuclear Technology Research; and, Cooling water intake optimization, by Jeffrey M. Jones and Bert Mayer, P.E., Framatome ANP.

Agnihotri, Newal (ed.)

2004-09-15T23:59:59.000Z

30

Plant maintenance and advanced reactors issue, 2008  

Science Conference Proceedings (OSTI)

The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Technologies of national importance, by Tsutomu Ohkubo, Japan Atomic Energy Agency, Japan; Modeling and simulation advances brighten future nuclear power, by Hussein Khalil, Argonne National Laboratory, Energy and desalination projects, by Ratan Kumar Sinha, Bhabha Atomic Research Centre, India; A plant with simplified design, by John Higgins, GE Hitachi Nuclear Energy; A forward thinking design, by Ray Ganthner, AREVA; A passively safe design, by Ed Cummins, Westinghouse Electric Company; A market-ready design, by Ken Petrunik, Atomic Energy of Canada Limited, Canada; Generation IV Advanced Nuclear Energy Systems, by Jacques Bouchard, French Commissariat a l'Energie Atomique, France, and Ralph Bennett, Idaho National Laboratory; Innovative reactor designs, a report by IAEA, Vienna, Austria; Guidance for new vendors, by John Nakoski, U.S. Nuclear Regulatory Commission; Road map for future energy, by John Cleveland, International Atomic Energy Agency, Vienna, Austria; and, Vermont's largest source of electricity, by Tyler Lamberts, Entergy Nuclear Operations, Inc. The Industry Innovation article is titled Intelligent monitoring technology, by Chris Demars, Exelon Nuclear.

Agnihotri, Newal (ed.)

2009-09-15T23:59:59.000Z

31

Wind Power Plant Evaluation Naval Auxiliary Landing Field, San Clemente Island, California: Period of Performance 24 September 1999--15 December 2000  

DOE Green Energy (OSTI)

The purpose of this report is to evaluate the wind power benefits and impacts to the San Clement Island wind power system, including energy savings, emissions reduction, system stability, and decreased naval dependence on fossil fuel at the island. The primary goal of the SCI wind power system has been to operate with the existing diesel power plant and provide equivalent or better power quality and system reliability than the existing diesel system. The wind system is intended to reduce, as far as possible, the use of diesel fuel and the inherent generation of nitrogen oxide emissions and other pollutants.

Olsen, T.L.; Gulman, P.J.; McKenna, E.

2000-12-11T23:59:59.000Z

32

Liquid metal cooled nuclear reactor plant system  

DOE Patents (OSTI)

A liquid metal cooled nuclear reactor having a passive cooling system for removing residual heat resulting for fuel decay during reactor shutdown, or heat produced during a mishap. The reactor system is enhanced with sealing means for excluding external air from contact with the liquid metal coolant leaking from the reactor vessel during an accident. The invention also includes a silo structure which resists attack by leaking liquid metal coolant, and an added unique cooling means.

Hunsbedt, Anstein (Los Gatos, CA); Boardman, Charles E. (Saratoga, CA)

1993-01-01T23:59:59.000Z

33

Liquid Metal Fast Breeder Reactor plant maintenance and equipment design  

Science Conference Proceedings (OSTI)

This paper provides a summary of maintenance equipment considerations and actual plant handling experiences from operation of a sodium-cooled reactor, the Fast Flux Test Facility (FFTF). Equipment areas relating to design, repair techniques, in-cell handling, logistics and facility services are discussed. Plant design must make provisions for handling and replacement of components within containment or allow for transport to an ex-containment area for repair. The modular cask assemblies and transporter systems developed for FFTF can service major plant components as well as smaller units. The plant and equipment designs for the Clinch River Breeder Reactor (CRBR) plant have been patterned after successful FFTF equipment.

Swannack, D.L.

1982-06-07T23:59:59.000Z

34

Multiple Irradiation Capsule Experiment (MICE)-3B Irradiation Test of Space Fuel Specimens in the Advanced Test Reactor (ATR) - Close Out Documentation for Naval Reactors (NR) Information  

SciTech Connect

Few data exist for UO{sub 2} or UN within the notional design space for the Prometheus-1 reactor (low fission rate, high temperature, long duration). As such, basic testing is required to validate predictions (and in some cases determine) performance aspects of these fuels. Therefore, the MICE-3B test of UO{sub 2} pellets was designed to provide data on gas release, unrestrained swelling, and restrained swelling at the upper range of fission rates expected for a space reactor. These data would be compared with model predictions and used to determine adequacy of a space reactor design basis relative to fission gas release and swelling of UO{sub 2} fuel and to assess potential pellet-clad interactions. A primary goal of an irradiation test for UN fuel was to assess performance issues currently associated with this fuel type such as gas release, swelling and transient performance. Information learned from this effort may have enabled use of UN fuel for future applications.

M. Chen; CM Regan; D. Noe

2006-01-09T23:59:59.000Z

35

Naval applications study areas  

SciTech Connect

This memorandum discusses study areas and items that will require attention for the naval studies of the utilization of nuclear propulsion in a submarine-based missile system.

Hadley, J. W.

1962-06-20T23:59:59.000Z

36

Clinch River Breeder Reactor Plant Project: construction schedule  

SciTech Connect

The construction schedule for the Clinch River Breeder Reactor Plant and its evolution are described. The initial schedule basis, changes necessitated by the evaluation of the overall plant design, and constructability improvements that have been effected to assure adherence to the schedule are presented. The schedule structure and hierarchy are discussed, as are tools used to define, develop, and evaluate the schedule.

Purcell, W.J.; Martin, E.M.; Shivley, J.M.

1982-01-01T23:59:59.000Z

37

A FUEL REPROCESSING PLANT FOR FAST CERAMIC REACTORS  

SciTech Connect

A study was made of the adaptation of the HAPO anion exchange process to the reprocessing of Fast Ceramic Reactor (FCR) fuel using the Idaho Small Plant Concept. It is shown that the anion exchange flowsheet meets the reprocessing objectives of the FCR case and can be adequately accommodated in the Small Plant Concept. Capacities of up to 1550 Md(e) are feasible in the Small Plant and unit reprocessing costs range from 0.14 to 0.28 mills/kwh depending on the number of reactors to be processed. (auth)

Alter, H.W.

1962-02-01T23:59:59.000Z

38

THE SODIUM GRAPHITE REACTOR: TOMMORROW'S POWER PLANT  

SciTech Connect

A description is given of the Advanced Sodium Graphite Reactor Power Plant, including the reactor, heat transfer systems, generatirg plant, control systems, and the economics of producing 256 Mw(e). The safety of this design is due to its unusually low operating pressure, absence of chemically incompatible materials in the core, and excellent stability under atatic and dynamic conditions. The reactor is being constructed at Hallam, Nebraska, at a probable cost of 0 to 0/kw, exclusive of the first core costs. The 151 fuel elements of uranium carbide are enriched to 2.75 at.% U/sup 235/ and clad in stainless steel. The average thermal neutron flux in the fuel is 8 x 10/sup 13/ n/cm/sup 2/sec. (B.O.G.)

Beeley, R.J.; Lowell, E.G.; Polak, H.; Renard, J.

1960-04-25T23:59:59.000Z

39

Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants  

E-Print Network (OSTI)

Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants ANL-AFCI-168 of Nuclear Reactor and Chemical Separation Plants ANL-AFCI-168 by G. Palmiotti, J. Cahalan, P. Pfeiffer, T;2 ANL-AFCI-168 Requirements for Advanced Simulation of Nuclear Reactor and Chemical Separation Plants G

Anitescu, Mihai

40

AN ADVANCED SODIUM-GRAPHITE REACTOR NUCLEAR POWER PLANT  

SciTech Connect

An advanced sodium-cooled, graphite-moderated nuclear power plant is described which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency. Steam is generated at 2400 psig, superheated to 1050 deg F and, after partial expansion in the turbine, reheated to 1000 deg F. Net thermal efficiency of the plant is 42.3%. In a plant sized to produce a net electrical output of 256 Mw, the estimated cost is 8232/kw. Estimated cost of power generation is 6.7 mills/kwh. In a similar plant with a net electrical output of 530 Mw, the estimated power generating cost is 5.4 mills/ kwh. Most of the components of the plant are within the capability of current technology. The major exception is the fuel material, uranium carbide. Preliminary results of the development work now in progress indicate that uranium carbide would be an excellent fuel for high-temperature reactors, but temperature and burnup limitation have yet to be firmly established. Additional development work is also required on the steam generators. These are the single-barrier type similar to those which will be used in the Enrico Fernri Fast Breeder Reactor plant but produce steam at higher pressure and temperature. Questions also remain regarding the use of nitrogen as a cover gas over sodium at 1200 deg F and compatibility of the materials used in the primary neutron shield. All of these questions are currently under investigation. (auth)

Churchill, J.R.; Renard, J.

1960-03-15T23:59:59.000Z

Note: This page contains sample records for the topic "naval reactor plants" 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

Overall plant concept for a tank-type fast reactor  

SciTech Connect

Japanese nuclear industries are expressing interest in the merits of the tank-type FBR as a large plant (demonstration) after JOYO (experimental, in operation) and MONJU (prototype, under construction). In response to this growing interest in a tank-type FBR demonstration plant, Hitachi has initiated a conceptual study of a 1000 MWe tank plant concept in collaboration with GE and Bechtel. Key objectives of this study have been: to select reliable and competitive tank plant concepts, with emphases on a seismic-resistant and compact tank reactor system;to select reliable shutdown heat removal system;and to identify R and D items needed for early 1990s construction. Design goals were defined as follows: capital costs must be less than twice, and as close as practical to 1.5 those of equivalent LWR plants;earthquake resistant structures to meet stringent Japanese seismic conditions must be as simple and reliable as practical;safety must be maintained at LWR-equivalent risks;and R and D needs must be limited to minimum cost for the limited time allowed. This paper summarizes the overall plant concepts with some selected topics, whereas detailed descriptions of the reactor assembly and the layout design are found in separate papers.

Yamaki, Hideo; Davies, S.M.; Goodman, L.

1984-01-01T23:59:59.000Z

42

L-Reactor Operation Savannah River Plant Aiken, SC  

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

51371 (F.R.) 51371 (F.R.) NOTICES DEPARTMENT OF ENERGY L-Reactor Operation, Savannah River Plant Aiken, South Carolina; Finding of No Significant Impact Monday, August 23, 1982 *36691 The Department of Energy (DOE) proposes to resume operation of L- Reactor at its Savannah River Plant at Aiken, South Carolina, as soon as it is ready for operation, scheduled for October 1983. The environmental impacts of the resumption of operation have been evaluated in an environmental assessment (DOE/EA-0195), prepared in accordance with the National Environmental Policy Act of 1969 (NEPA) as implemented by regulations promulgated by the Council on Environmental Quality (CEQ) (40 CFR Parts 1500 -1508, November 1978) and DOE implementing guidelines (45 FR 20694, March 28, 1980). Based on the analysis in the assessment, DOE has

43

Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.  

SciTech Connect

This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

2006-12-11T23:59:59.000Z

44

Trojan Nuclear Power Plant Reactor Vessel and Internals Removal: Trojan Nuclear Plant Decommissioning Experience  

Science Conference Proceedings (OSTI)

One goal of the EPRI Decommissioning Technology Program is to capture the growing utility experience in nuclear plant decommissioning activities for the benefit of other utilities facing similar challenges in the future. This report provides historical information on the background, scope, organization, schedule, cost, contracts, and support activities associated with the Trojan Nuclear Plant Reactor Vessel and Internals Removal (RVAIR) Project. Also discussed are problems, successes, and lessons learned...

2000-10-16T23:59:59.000Z

45

Dual-phase reactor plant with partitioned isolation condenser  

DOE Patents (OSTI)

A nuclear energy plant housing a boiling-water reactor utilizes an isolation condenser in which a single chamber is partitioned into a distributor plenum and a collector plenum. Steam accumulates in the distributor plenum and is conveyed to the collector plenum through an annular manifold that includes tubes extending through a condenser pool. The tubes provide for a transfer of heat from the steam, forming a condensate. The chamber has a disk-shaped base, a cylindrical sidewall, and a semispherical top. This geometry results in a compact design that exhibits significant performance and cost advantages over prior designs.

Hui, Marvin M. (Cupertino, CA)

1992-01-01T23:59:59.000Z

46

Fast reactor power plant design having heat pipe heat exchanger  

DOE Patents (OSTI)

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Huebotter, P.R.; McLennan, G.A.

1984-08-30T23:59:59.000Z

47

Fast reactor power plant design having heat pipe heat exchanger  

DOE Patents (OSTI)

The invention relates to a pool-type fission reactor power plant design having a reactor vessel containing a primary coolant (such as liquid sodium), and a steam expansion device powered by a pressurized water/steam coolant system. Heat pipe means are disposed between the primary and water coolants to complete the heat transfer therebetween. The heat pipes are vertically oriented, penetrating the reactor deck and being directly submerged in the primary coolant. A U-tube or line passes through each heat pipe, extended over most of the length of the heat pipe and having its walls spaced from but closely proximate to and generally facing the surrounding walls of the heat pipe. The water/steam coolant loop includes each U-tube and the steam expansion device. A heat transfer medium (such as mercury) fills each of the heat pipes. The thermal energy from the primary coolant is transferred to the water coolant by isothermal evaporation-condensation of the heat transfer medium between the heat pipe and U-tube walls, the heat transfer medium moving within the heat pipe primarily transversely between these walls.

Huebotter, Paul R. (Western Springs, IL); McLennan, George A. (Downers Grove, IL)

1985-01-01T23:59:59.000Z

48

EPRI NMAC Maintainability Review of the Pebble Bed Modular Reactor Demonstration Plant  

Science Conference Proceedings (OSTI)

This report provides information to the designers of pebble bed reactor helium-driven gas turbine plants and to others who are considering the purchase of this type of plant.

2002-05-13T23:59:59.000Z

49

100-MW NUCLEAR POWER PLANT UTILIZING A SODIUM COOLED, GRAPHITE MODERATED REACTOR  

SciTech Connect

The conceptual design of a 100 Mw(e) nuclear power plant is described. The plant utilized a sodium-cooled graphite-moderated reactor with stainless- steel clad. slightiy enriched UO/sub 2/ fuel. The reactor is provided with three main coolant circuits, and the steam cycle has three stages of regenerative heating. The plant control system allows automatic operation over the range of 20 to 100% load, or manual operation at all loads. The site, reactor, sodium systems, reactor auxiliaries, fuel handling, instrumentation, turbine-generator, buildings. and safety measures are described. Engineering drawings are included. (W.D.M.)

1958-02-28T23:59:59.000Z

50

Life of Plant Activity Estimates for a Nominal 1000 MWe Pressurized Water Reactor and Boiling Water Reactor  

Science Conference Proceedings (OSTI)

Decommissioning nuclear power plant and disposal site managers must understand the radioactive source term of a nuclear power plant to effectively manage disposition of these materials. This study estimates the radioactive source term from nominal 1000 MWe pressurized water and boiling water reactors to support decisions related to radioactive waste storage, processing, and disposal through decommissioning.BackgroundThis study examines the radionuclide ...

2012-12-05T23:59:59.000Z

51

THE SODIUM GRAPHITE REACTOR POWER PLANT FOR CPPD  

SciTech Connect

The plant arrangement, component design, and the functions of various systems are described and illustrated. Relative estimated costs of the systems and major components are indicated. The reactor core is designed around requiremouts for 254 thermal megawatts, 950 deg F maximum sodium temperature, stainless steel clad graphite moderator blocks, and low enrichment (0.015 to 0.04 U/sup 235/) uranium fuel elements. The fuel cycle is described for the possible fuel elements. The fuel cost factors are discussed. Burn-up limitations encountered for metallic fuel in the SGR temperature range indicate UO/sub 2/ the more desirable choice. The estimated cost of electrical energy associated with the UO/sub 2/ fuel is given. (auth)

Olson, R.L.; Gerber, R.C.; Gordon, R.B.; Ross-Clunis, H.A.; Stolz, J.F.

1958-10-31T23:59:59.000Z

52

REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT  

DOE Patents (OSTI)

A power generator in the form of a flywheel with four reactors positioned about its rim is described. The reactors are so positioned that steam, produced in the reactor, exists tangentially to the flywheel, giving it a rotation. The reactors are incompletely moderated without water. The water enters the flywheel at its axis, under sufficient pressure to force it through the reactors, where it is converted to steam. The fuel consists of parallel twisted ribbons assembled to approximate a cylinder.

Loeb, E.

1961-01-17T23:59:59.000Z

53

FY 2012 Budget Hearing Testimony on Nuclear Nonproliferation and Naval  

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

on Nuclear Nonproliferation and Naval on Nuclear Nonproliferation and Naval Reactor Programs before the House Appropriations Committee, Energy and Water Development Subcommittee | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Congressional Testimony > FY 2012 Budget Hearing Testimony on Nuclear ...

54

FY 2012 Budget Hearing Testimony on Nuclear Nonproliferation and Naval  

National Nuclear Security Administration (NNSA)

on Nuclear Nonproliferation and Naval on Nuclear Nonproliferation and Naval Reactor Programs before the House Appropriations Committee, Energy and Water Development Subcommittee | National Nuclear Security Administration Our Mission Managing the Stockpile Preventing Proliferation Powering the Nuclear Navy Emergency Response Recapitalizing Our Infrastructure Continuing Management Reform Countering Nuclear Terrorism About Us Our Programs Our History Who We Are Our Leadership Our Locations Budget Our Operations Media Room Congressional Testimony Fact Sheets Newsletters Press Releases Speeches Events Social Media Video Gallery Photo Gallery NNSA Archive Federal Employment Apply for Our Jobs Our Jobs Working at NNSA Blog Home > Media Room > Congressional Testimony > FY 2012 Budget Hearing Testimony on Nuclear ...

55

NAVAL POSTGRADUATE SCHOOL MONTEREY, CALIFORNIA  

E-Print Network (OSTI)

), Stockholm Intl Peace Research Inst, Oxford University Oress 337 p., 1997. Arkin, WM and J Handler, Naval Geochemistry (eds. AC Sigleo and A Hattori), Lewis Publishers, Inc., Chelsea, MI, 97-119, 1985. Brungot, AL

56

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

57

Naval Spent Fuel Rail Shipment Accident Exercise Objectives ...  

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

Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives Naval Spent Fuel Rail Shipment Accident Exercise Objectives...

58

2013 Annual Planning Summary for the Naval Nuclear Propulsion...  

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

Naval Nuclear Propulsion Program 2013 Annual Planning Summary for the Naval Nuclear Propulsion Program 2013 Annual Planning Summary for the Naval Nuclear Propulsion Program The...

59

Assessment of light water reactor power plant cost and ultra-acceleration depreciation financing  

E-Print Network (OSTI)

Although in many regions of the U.S. the least expensive electricity is generated from light-water reactor (LWR) plants, the fixed (capital plus operation and maintenance) cost has increased to the level where the cost ...

El-Magboub, Sadek Abdulhafid.

60

COST STUDY OF A 100-Mw(e) DIRECT-CYCLE BOILING WATER REACTOR PLANT  

SciTech Connect

A technical and economic evaluation is presented of a direct-cycle light- water boiling reactor designed for natural circulation and internal steam-water separation. The reference lOO-Mw(e) reactor power plant design evolved from the study should have the best chance (compared to similar plants) of approaching the 8 to 9 mill/kwh total power-cost level. (W.D.M.)

Bullinger, C.F.; Harrer, J.M.

1960-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "naval reactor plants" 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

PRISM; The plant design concept for the U. S. advanced liquid metal reactor program  

SciTech Connect

The US program for development of an advanced liquid metal reactor (ALMR) is proceeding into a new phase of focused design development. This new phase started at the beginning of 1989; its objective is to complete the conceptual design of the US ALMR, with supporting key feature tests, sufficiently to enter a more detailed design phase and subsequent construction of a prototype reactor plant. A project goal is to demonstrate by actual performance of the reactor its passive, inherent safety features and thereby provide the technical basis for certification of the design by the Nuclear Regulatory Commission (NRC). This paper reports on the PRISM (power reactor inherently safe module) reactor concept which in combination with the IFR (integral fast reactor) metal fuel cycle being developed by Argonne National Laboratory, was selected by DOE in 1988 as the reference design for the US ALMR program.

Berglund, R.C.; Tippets, F.E. (GE Nuclear Energy, Advance Nuclear Technology, San Jose, CA (US))

1989-01-01T23:59:59.000Z

62

Assessement of Codes and Standards Applicable to a Hydrogen Production Plant Coupled to a Nuclear Reactor  

DOE Green Energy (OSTI)

This is an assessment of codes and standards applicable to a hydrogen production plant to be coupled to a nuclear reactor. The result of the assessment is a list of codes and standards that are expected to be applicable to the plant during its design and construction.

M. J. Russell

2006-06-01T23:59:59.000Z

63

Naval Petroleum Reserves | Department of Energy  

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

Naval Petroleum Reserves For much of the 20th century, the Naval Petroleum and Oil Shale Reserves served as a contingency source of fuel for the Nation's military. All that...

64

Space Nuclear Power Plant Pre-Conceptual Design Report, For Information  

Science Conference Proceedings (OSTI)

This letter transmits, for information, the Project Prometheus Space Nuclear Power Plant (SNPP) Pre-Conceptual Design Report completed by the Naval Reactors Prime Contractor Team (NRPCT). This report documents the work pertaining to the Reactor Module, which includes integration of the space nuclear reactor with the reactor radiation shield, energy conversion, and instrumentation and control segments. This document also describes integration of the Reactor Module with the Heat Rejection segment, the Power Conditioning and Distribution subsystem (which comprise the SNPP), and the remainder of the Prometheus spaceship.

B. Levine

2006-01-27T23:59:59.000Z

65

Naval Nuclear Propulsion Program Directorate, Washington ...  

Science Conference Proceedings (OSTI)

Naval Nuclear Propulsion Program Directorate, Washington, DC. NVLAP Lab Code: 100565-0. Address and Contact Information: ...

2013-08-23T23:59:59.000Z

66

Naval Nuclear Propulsion Program Directorate, Washington ...  

Science Conference Proceedings (OSTI)

Naval Nuclear Propulsion Program Directorate, Washington, DC. NVLAP Lab Code: 100565-2. Address and Contact Information: Point Loma, Bldg. ...

2013-08-23T23:59:59.000Z

67

NAVAL RESEARCH LABORATORY Information Technology Solutions  

power (CHP) or emergency backup power Small, High Efficiency, Recuperated Ceramic Turboshaft Engine NAVAL RESEARCH LABORATORY TECHNOLOGY T RANSFER ...

68

Study of Pu consumption in Advanced Light Water Reactors. Evaluation of GE Advanced Boiling Water Reactor plants  

SciTech Connect

Timely disposal of the weapons plutonium is of paramount importance to permanently safeguarding this material. GE`s 1300 MWe Advanced Boiling Water Reactor (ABWR) has been designed to utilize fill] core loading of mixed uranium-plutonium oxide fuel. Because of its large core size, a single ABWR reactor is capable of disposing 100 metric tons of plutonium within 15 years of project inception in the spiking mode. The same amount of material could be disposed of in 25 years after the start of the project as spent fuel, again using a single reactor, while operating at 75 percent capacity factor. In either case, the design permits reuse of the stored spent fuel assemblies for electrical energy generation for the remaining life of the plant for another 40 years. Up to 40 percent of the initial plutonium can also be completely destroyed using ABWRS, without reprocessing, either by utilizing six ABWRs over 25 years or by expanding the disposition time to 60 years, the design life of the plants and using two ABWRS. More complete destruction would require the development and testing of a plutonium-base fuel with a non-fertile matrix for an ABWR or use of an Advanced Liquid Metal Reactor (ALMR). The ABWR, in addition, is fully capable of meeting the tritium target production goals with already developed target technology.

Not Available

1993-05-13T23:59:59.000Z

69

Condensate Polishing Guidelines for Pressurized Water Reactor and Boiling Water Reactor Plants - 2004 Revision  

Science Conference Proceedings (OSTI)

Successful condensate polishing allows more reliable operation of nuclear units by maintaining control of ionic and particulate impurity transport to the pressurized water reactor (PWR) steam generators and the boiling water reactor (BWR) and recirculation system. This report presents revisions of EPRI's 1997 nuclear industry consensus guidelines for the design and operation of deep bed and filter demineralizer condensate polishers. These guidelines are consistent with the 2000 revisions of EPRI's "BWR W...

2004-03-16T23:59:59.000Z

70

POWER PLANT USING A STEAM-COOLED NUCLEAR REACTOR  

SciTech Connect

A method of providing efficient and economic means for obtaining reheat from nuclear heat is described. A steamcooled steam-moderated reactor produces high-pressure, high-temperature steam. A multi-stage steam turbine partially expands the high-pressure steam, which is then withdrawn and reheated, and then further expanded for producing useful power. The saturated steam is superheated by leading it through tubular passages provided in the fuel assemblies of a nuclear reactor, leading the useful part of the superheated steam into a steam turbine in which it expands to a predetermined intermediate pressure, leading the steam at that reduced pressure from the turbine back into the reactor where it is reheated by flowing through other tubular passages in the fuel assemblies, and returning the reheated steam to the turbine for further expansion. (M.C.G.)

Nettel, F.; Nakanishi, T.

1963-10-29T23:59:59.000Z

71

Explosions and other uncontrolled chemical reactions at non-reactor nuclear facilities of the Savannah River Plant  

Science Conference Proceedings (OSTI)

This paper describes five energetic reactions (explosions) that have occurred at the Savannah River Plant non-reactor nuclear facilities. 1 fig.

Durant, W.S.; Gray, L.W.; Wallace, R.M.; Yau, W.W.F.

1988-01-01T23:59:59.000Z

72

Materials Reliability Program: Inspection and Evaluation Guidelines for Reactor Vessel Bottom-Mounted Nozzles in U.S. PWR Plants (MR P-206)  

Science Conference Proceedings (OSTI)

This report presents inspection and evaluation guidelines for reactor vessel bottom-mounted nozzles in U.S. pressurized water reactor (PWR) plants.

2009-03-23T23:59:59.000Z

73

Applicability of base-isolation R D in non-reactor facilities to a nuclear reactor plant  

SciTech Connect

Seismic isolation is gaining increased attention worldwide for use in a wide spectrum of critical facilities, ranging from hospitals and computing centers to nuclear power plants. While the fundamental principles and technology are applicable to all of these facilities, the degree of assurance that the actual behavior of the isolation systems is as specified varies with the nature of the facility involved. Obviously, the level of effort to provide such assurance for a nuclear power plant will be much greater than that required for, say, a critical computer facility. The question, therefore, is to what extent can research and development (R D) for non-nuclear use be used to provide technological data needed for seismic isolation of a nuclear power plant. This question, of course is not unique to seismic isolation. Virtually every structural component, system, or piece of equipment used in nuclear power plants is also used in non- nuclear facilities. Experience shows that considerable effort is needed to adapt conventional technology into a nuclear power plant. Usually, more thorough analysis is required, material and fabrication quality-control requirements are more stringent as are controls on field installation. In addition, increased emphasis on maintainability and inservice inspection throughout the life of the plant is generally required to gain acceptance in nuclear power plant application. This paper reviews the R D programs ongoing for seismic isolation in non-nuclear facilities and related experience and makes a preliminary assessment of the extent to which such R D and experience can be used for nuclear power plant application. Ways are suggested to improve the usefulness of such non-nuclear R D in providing the high level of confidence required for the use of seismic isolation in a nuclear reactor plant. 2 refs.

Seidensticker, R.W.; Chang, Y.W.

1990-01-01T23:59:59.000Z

74

NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions  

SciTech Connect

This document provides key definitions, plant capabilities, and inputs and assumptions related to the Next Generation Nuclear Plant to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor. These definitions, capabilities, and assumptions were extracted from a number of NGNP Project sources such as licensing related white papers, previously issued requirement documents, and preapplication interactions with the Nuclear Regulatory Commission (NRC).

Wayne Moe

2013-05-01T23:59:59.000Z

75

Knowledge and abilities catalog for nuclear power plant operators: Boiling water reactors, Revision 1  

SciTech Connect

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWRs) (NUREG-1123, Revision 1) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog along with the Operator Licensing Examiner Standards (NUREG-1021) and the Examiner`s Handbook for Developing Operator Licensing Written Examinations (NUREG/BR-0122), will cover the topics listed under Title 10, Code of Federal Regulations, Part 55 (10 CFR 55). The BWR Catalog contains approximately 7,000 knowledge and ability (K/A) statements for ROs and SROs at BWRs. The catalog is organized into six major sections: Organization of the Catalog, Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Functions, Emergency and Abnormal Plant Evolutions, Components, and Theory. Revision 1 to the BWR Catalog represents a modification in form and content of the original catalog. The K/As were linked to their applicable 10 CFR 55 item numbers. SRO level K/As were identified by 10 CFR 55.43 item numbers. The plant-wide generic and system generic K/As were combined in one section with approximately one hundred new K/As. Component Cooling Water and Instrument Air Systems were added to the Systems Section. Finally, High Containment Hydrogen Concentration and Plant Fire On Site evolutions added to the Emergency and Abnormal Plant Evolutions section.

NONE

1995-08-01T23:59:59.000Z

76

Selection of power plant elements for future reactor space electric power systems  

DOE Green Energy (OSTI)

Various types of reactor designs, electric power conversion equipment, and reject-heat systems to be used in nuclear reactor power plants for future space missions were studied. The designs included gas-cooled, liquid-cooled, and heat-pipe reactors. For the power converters, passive types such as thermoelectric and thermionic converters and dynamic types such as Brayton, potassium Rankine, and Stirling cycles were considered. For the radiators, heat pipes for transfer and radiating surface, pumped fluid for heat transfer with fins as the radiating surface, and pumped fluid for heat transfer with heat pipes as the radiating surface were considered. After careful consideration of weights, sizes, reliabilities, safety, and development cost and time, a heat-pipe reactor design, thermoelectric converters, and a heat-pipe radiator for an experimental program were selected.

Buden, D.; Bennett, G.A.; Copper, K.

1979-09-01T23:59:59.000Z

77

Environmental impacts of nonfusion power systems. [Data on environmental effects of all power sources that may be competitive with fusion reactor power plants  

DOE Green Energy (OSTI)

Data were collected on the environmental effects of power sources that may be competitive with future fusion reactor power plants. Data are included on nuclear power plants using HTGR, LMBR, GCFR, LMFBR, and molten salt reactors; fossil-fuel electric power plants; geothermal power plants; solar energy power plants, including satellite-based solar systems; wind energy power plants; ocean thermal gradient power plants; tidal energy power plants; and power plants using hydrogen and other synthetic fuels as energy sources.

Brouns, R.J.

1976-09-01T23:59:59.000Z

78

Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants  

Science Conference Proceedings (OSTI)

Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design.

Woo, H.H.; Lu, S.C.

1981-09-15T23:59:59.000Z

79

ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT  

DOE Green Energy (OSTI)

An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322C and 750C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

M. G. McKellar; E. A. Harvego; A. M. Gandrik

2010-11-01T23:59:59.000Z

80

Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant  

DOE Green Energy (OSTI)

The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood.

Chang H. Oh; Eung Soo Kim; Steven Sherman

2008-04-01T23:59:59.000Z

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


81

Thermal Hydraulic Analyses for Coupling High Temperature Gas-Cooled Reactor to Hydrogen Plant  

DOE Green Energy (OSTI)

The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. A number of possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermal-hydraulic and cycle-efficiency evaluations of the different configurations and coolants. The thermal-hydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermalhydraulic and efficiency points of view.

C.H. Oh; R. Barner; C. B. Davis; S. Sherman; P. Pickard

2006-08-01T23:59:59.000Z

82

Considerations Associated with Reactor Technology Selection for the Next Generation Nuclear Plant Project  

Science Conference Proceedings (OSTI)

At the inception of the Next Generation Nuclear Plant Project and during predecessor activities, alternative reactor technologies have been evaluated to determine the technology that best fulfills the functional and performance requirements of the targeted energy applications and market. Unlike the case of electric power generation where the reactor performance is primarily expressed in terms of economics, the targeted energy applications involve industrial applications that have specific needs in terms of acceptable heat transport fluids and the associated thermodynamic conditions. Hence, to be of interest to these industrial energy applications, the alternative reactor technologies are weighed in terms of the reactor coolant/heat transport fluid, achievable reactor outlet temperature, and practicality of operations to achieve the very high reliability demands associated with the petrochemical, petroleum, metals and related industries. These evaluations have concluded that the high temperature gas-cooled reactor (HTGR) can uniquely provide the required ranges of energy needs for these target applications, do so with promising economics, and can be commercialized with reasonable development risk in the time frames of current industry interest i.e., within the next 10-15 years.

L.E. Demick

2010-09-01T23:59:59.000Z

83

J. Plasma Fusion Res. SERIES, Vol. 8 (2009) Alternative Fusion Reactors as Future Commercial Power Plants  

E-Print Network (OSTI)

Alternative reactor based on a field-reversed configuration (FRC) has advantages of the cylindrical geometry, the open field line geometry (direct energy conversion (DEC) of the charged-particle flow), and high ? (plasma pressure/magnetic-field pressure). This paper aims to evaluate the attractiveness of a low radioactive FRC fusion core. Analysis of a conceptual deuterium- helium-3 (D- 3 He) fusion power reactor is presented and reference point is defined. Principal parameters of the D- 3 He plasma reference case (RC) and comparison with conceptual D- 3 He tokamak and FRC power plants are shown.

Sergei V. Ryzhkov

2008-01-01T23:59:59.000Z

84

DESIGN AND FEASIBILITY STUDY OF A PEBBLE BED REACTOR-STEAM POWER PLANT  

SciTech Connect

Originally issued as S and P 1963A, Parts I and II. A design and feasibility study of a pebble bed reactorsteam power plant is presented, The reactor design which evolved from this study is a 125 Mwe heliumcooled two-region thermal breeder, operating on the uranium-thorium cycle, in which all core structural materials are graphite. Fuel is in the form of unclad spherical elements of graphite, containing fissile and fertile material. The primary loop consists of the reactor plus three steam generators and blowers in parallel. Nuclear characteristics, costs, etc., are given. (W.D.M.)

1958-05-01T23:59:59.000Z

85

Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.  

DOE Green Energy (OSTI)

In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for both the PBMR and prismatic design. The main focus of this report is the RPV for both design concepts with emphasis on material selection.

Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

2007-03-21T23:59:59.000Z

86

Geothermal energy at Long Beach Naval Shipyard and Naval Station and at Seal Beach Naval Weapons Station, California. Final report  

DOE Green Energy (OSTI)

The purpose of this project was to determine and evaluate sources of geothermal energy at two military bases in southern California, the Long Beach Naval Shipyard and Naval Station and the Seal Beach Naval Weapons Station. One part of the project focused on the natural geothermal characteristics beneath the naval bases. Another part focused on the geothermal energy produced by oilfield operations on and adjacent to each base. Results of the study are presented here for the US Department of the Navy to use in its program to reduce its reliance on petrolem by the development of different sources of energy. The study was accomplished under a cooperative agreement between the US Department of Energy's San Francisco Operations Office and the Department of the Navy's Naval Weapons Center, China Lake, California, for joint research and development of geothermal energy at military installations.

Higgins, C.T.; Chapman, R.H.

1984-01-01T23:59:59.000Z

87

EA-1008: Continued Development of Naval Petroleum Reserve No...  

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

8: Continued Development of Naval Petroleum Reserve No. 3 (Sitewide), Natrona County, Wyoming EA-1008: Continued Development of Naval Petroleum Reserve No. 3 (Sitewide), Natrona...

88

EA-1236: Preparation for Transfer of Ownership of Naval Petroleum...  

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

236: Preparation for Transfer of Ownership of Naval Petroleum Reserve No. 3, Natrona County, WY EA-1236: Preparation for Transfer of Ownership of Naval Petroleum Reserve No. 3,...

89

Distributed energy resources at naval base ventura county building 1512  

E-Print Network (OSTI)

Naval Base Ventura County Standby Generator OptimizationC&H Engineering performed a standby generator optimizationOn Naval Base Ventura County Standby Generator Optimization

Bailey, Owen C.; Marnay, Chris

2004-01-01T23:59:59.000Z

90

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

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

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

91

A Review of OLED Research at Naval Research Laboratory  

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

Division at Naval Research Laboratory. Her research is focused on organic light emitting diode (OLED) material and devices. She will discuss the research activities at Naval...

92

The Next Generation Nuclear Plant/Advanced Gas Reactor Fuel Irradiation Experiments in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energys Next Generation Nuclear Plant (NGNP) Program will be irradiating eight separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energys lead laboratory for nuclear energy development. The ATR is one of the worlds premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next ten years to demonstrate and qualify new particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006, and the second experiment (AGR-2) is currently in the design phase. The design of test trains, as well as the support systems and fission product monitoring system that will monitor and control the experiment during irradiation will be discussed. In addition, the purpose and differences between the two experiments will be compared and the irradiation results to date on the first experiment will be presented.

S. Blaine Grover

2009-09-01T23:59:59.000Z

93

Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants  

Science Conference Proceedings (OSTI)

This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

Wood, RT

2004-09-27T23:59:59.000Z

94

Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen Production Plant  

DOE Green Energy (OSTI)

A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540C and 900C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating-current, AC, to direct-current, DC, conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.12% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

E. A. Harvego; M. G. McKellar; M. S. Sohal; J. E. O'Brien; J. S. Herring

2008-08-01T23:59:59.000Z

95

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have concluded, however, that with adequate engineered cooling of the vessel, the A508/533 steels are also acceptable.

J. K. Wright; R. N. Wright

2008-04-01T23:59:59.000Z

96

Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan (PLN-2803)  

Science Conference Proceedings (OSTI)

The U.S. Department of Energy (DOE) has selected the High-Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production, with an outlet gas temperature in the range of 750C, and a design service life of 60 years. The reactor design will be a graphite-moderated, helium-cooled, prismatic, or pebble bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. This technology development plan details the additional research and development (R&D) required to design and license the NGNP RPV, assuming that A 508/A 533 is the material of construction. The majority of additional information that is required is related to long-term aging behavior at NGNP vessel temperatures, which are somewhat above those commonly encountered in the existing database from LWR experience. Additional data are also required for the anticipated NGNP environment. An assessment of required R&D for a Grade 91 vessel has been retained from the first revision of the R&D plan in Appendix B in somewhat less detail. Considerably more development is required for this steel compared to A 508/A 533 including additional irradiation testing for expected NGNP operating temperatures, high-temperature mechanical properties, and extensive studies of long-term microstructural stability.

J. K. Wright; R. N. Wright

2010-07-01T23:59:59.000Z

97

Fuel Summary Report: Shippingport Light Water Breeder Reactor  

SciTech Connect

The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749.

Illum, D.B.; Olson, G.L.; McCardell, R.K.

1999-01-01T23:59:59.000Z

98

NGNP: High Temperature Gas-Cooled Reactor Key Definitions, Plant Capabilities, and Assumptions  

SciTech Connect

This document is intended to provide a Next Generation Nuclear Plant (NGNP) Project tool in which to collect and identify key definitions, plant capabilities, and inputs and assumptions to be used in ongoing efforts related to the licensing and deployment of a high temperature gas-cooled reactor (HTGR). These definitions, capabilities, and assumptions are extracted from a number of sources, including NGNP Project documents such as licensing related white papers [References 1-11] and previously issued requirement documents [References 13-15]. Also included is information agreed upon by the NGNP Regulatory Affairs group's Licensing Working Group and Configuration Council. The NGNP Project approach to licensing an HTGR plant via a combined license (COL) is defined within the referenced white papers and reference [12], and is not duplicated here.

Phillip Mills

2012-02-01T23:59:59.000Z

99

DOE - Office of Legacy Management -- Shippingport Atomic Power Plant - PA  

Office of Legacy Management (LM)

Shippingport Atomic Power Plant - Shippingport Atomic Power Plant - PA 13 FUSRAP Considered Sites Site: SHIPPINGPORT ATOMIC POWER PLANT (PA.13 ) Eliminated from further consideration under FUSRAP. Designated Name: Not Designated Alternate Name: Duquesne Light Company PA.13-1 Location: 25 miles west of Pittsburgh in Beaver County , Shippingport , Pennsylvania PA.13-2 Evaluation Year: circa 1987 PA.13-3 Site Operations: First commercially operated nuclear power reactor. Joint project (Federal Government an Duquesne Light Company) to demonstrate pressurized water reactor technology and to generate electricity. Plant operated by Duquesne Light Company under supervision of the Office of the DOE Deputy Assistant Secretary for Naval Reactors -- 1957 to October 1982. PA.13-2 Site Disposition: Eliminated - No Authority. DOE chartered Major Project #118, Shippingport Station Decommissioning Project completed cleanup in 1989. PA.13-1

100

DOE - Office of Legacy Management -- Naval Ordnance Test Station - CA 06  

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

Ordnance Test Station - CA 06 Ordnance Test Station - CA 06 FUSRAP Considered Sites Site: NAVAL ORDNANCE TEST STATION (CA.06) Eliminated from consideration under FUSRAP Designated Name: Not Designated Alternate Name: China Lake Naval Weapons Center Salt Wells Pilot Plant CA.06-1 Location: Inyokern , California CA.06-1 Evaluation Year: 1987 CA.06-1 Site Operations: Naval facility; experimental development work on shape charges and quality castings on a pilot plant scale. CA.06-1 Site Disposition: Eliminated - No indication that radioactive materials were handled at the site CA.06-1 Radioactive Materials Handled: None Indicated CA.06-1 Primary Radioactive Materials Handled: None CA.06-1 Radiological Survey(s): No Site Status: Eliminated from consideration under FUSRAP Also see

Note: This page contains sample records for the topic "naval reactor plants" 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

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

102

The dynamics of naval resource allocation  

Science Conference Proceedings (OSTI)

Naval policy planning requires viewing the entire spectrum of annual resource allocation, over a long period. Emerging research into the trade offs between force asset levels and fund flows to support those assets is described. The method underlying ...

Rolf Clark

1981-01-01T23:59:59.000Z

103

Study of Pu consumption in advanced light water reactors: Evaluation of GE advanced boiling water reactor plants - compilation of Phase 1B task reports  

SciTech Connect

This report contains an extensive evaluation of GE advanced boiling water reactor plants prepared for United State Department of Energy. The general areas covered in this report are: core and system performance; fuel cycle; infrastructure and deployment; and safety and environmental approval.

NONE

1993-09-15T23:59:59.000Z

104

A STEAM POWER INSTALLATION FOR NUCLEAR POWER PLANT WITH GAS-COOLED REACTORS  

SciTech Connect

A steam power plant is designed for use with gas-cooled power reactors. In this plant, the turbine is divided into two sections, one high pressure and the other low pressure, the low-pressure turbine being the condensing turbine. The feed water from the condensing turbine is divided into two streams, one of which is brought to a higher pressure than the other. The high-pressure feed water is evaporated and superheated in the heat exchanger and then supplied to the high-pressure turbine, while the low-pressure feed water is evaporated and mixed with the exhaust steam of the highpressure turbine before superhenting and then passing to the low-pressure condensing turbine. Circulation of the reactor coolant is effected by a blower driven by a series turbine with no regulating devices and arranged in the steam plant circuit upstream of the low-pressure turbine; such a turbine works with constant efficiency over its whole load range. (D.L.C.)

1961-03-01T23:59:59.000Z

105

Modular pebble-bed reactor reforming plant design for process heat  

Science Conference Proceedings (OSTI)

This report describes a preliminary design study of a Modular Pebble-Bed Reactor System Reforming (MPB-R) Plant. The system uses one pressure vessel for the reactor and a second pressure vessel for the components, i.e., reformer, steam generator and coolant circulator. The two vessels are connected by coaxial pipes in an arrangement known as the side-by-side (SBS). The goal of the study is to gain an understanding of this particular system and to identify any technical issues that must be resolved for its application to a modular reformer plant. The basic conditions for the MPB-R were selected in common with those of the current study of the MRS-R in-line prismatic fuel concept, specifically, the module core power of 250 MWt, average core power density of 4.1 w/cc, low enriched uranium (LEU) fuel with a /sup 235/U content of 20% homogeneously mixed with thorium, and a target burnup of 80,000 MWD/MT. Study results include the pebble-bed core neutronics and thermal-hydraulic calculations. Core characteristics for both the once-through-then-out (OTTO) and recirculation of fuel sphere refueling schemes were developed. The plant heat balance was calculated with 55% of core power allotted to the reformer.

Lutz, D.E.; Cowan, C.L.; Davis, C.R.; El Sheikh, K.A.; Hui, M.M.; Lipps, A.J.; Wu, T.

1982-09-01T23:59:59.000Z

106

REACTOR  

DOE Patents (OSTI)

A pressurized water reactor in which automatic control is achieved by varying the average density of the liquid moderator-cooiant is patented. Density is controlled by the temperature and power level of the reactor ftself. This control can be effected by the use of either plate, pellet, or tubular fuel elements. The fuel elements are disposed between upper and lower coolant plenum chambers and are designed to permit unrestricted coolant flow. The control chamber has an inlet opening communicating with the lower coolant plenum chamber and a restricted vapor vent communicating with the upper coolant plenum chamber. Thus, a variation in temperature of the fuel elements will cause a variation in the average moderator density in the chamber which directly affects the power level of the reactor.

Roman, W.G.

1961-06-27T23:59:59.000Z

107

Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2  

SciTech Connect

This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.

NONE

1997-12-01T23:59:59.000Z

108

A DESIGN STUDY OF A LOW POWER AQUEOUS HOMOGENEOUS BOILING REACTOR POWER PLANT  

SciTech Connect

This design study describes a reactor and associated power plant that has been designed to produce 100 kv of net electric power and 400 kv of hot water space heating at a total thermal output of 1300 kw. The fuel consists of a solution of UO/sub 2/SO/sub 4/ in light water. Power is removed from the core by boiling the fuel solution and transferring the heat to the secondary steam system by condensing primary water on the external surface of a bayonet type boiler and boiling secondary water within the tubes. Saturated steam, produced in the boiler at 225 psia (Full Power) is used to drive a turbo generator, Extraction steam from the turbine is used, at a reduced pressure, for space heating. The initial loading of the reactor is approximately 4.8 kg of U/sub 235/ and operation based on an average load factor of 80% will require fuel addition at the rate of about 580 grams per year. It may be desirable to replace the fuel in the core after a period of 5 years operation due to the accumulation of corrosion products. The reactor control is affected automatically by power demand. The major objective has been to design a reactor that is reliable and simple, requiring little if any operating personnel and routine maintenance only which can be performed by one man. The design should stress simplicity of the system, ease of erection at the site, initial transportability, reliability and ease of operation; these characteristics are then expected to result in greatly reduced effort and manpower support over a conventional system. (auth)

Mong, B.A.; Colgan, J.E.; D' Elia, R.A.; Mooradian, J.S.; Rhode, G.K.; Wood, P.M.

1955-06-01T23:59:59.000Z

109

SINGLE-FLUID TWO-REGION AQUEOUS HOMOGENEOUS REACTOR POWER PLANT CONCEPTUAL DESIGN AND FEASIBILITY STUDY. Final Report  

SciTech Connect

The feasibility of a 150 Mwe aqueous homogeneous nuclear power plant was investigated by a joint study team of the Nuclear Power Group and The Babcock & Wilcox Company. In this concept, the reactor is a single-fluid two-region design in which the fuel solution circulates through the thoria pellet blanket as the coolant. Componeats and plant arrangement were designed to provide maximum overhead accessibility for maintenance. All components in contact with reactor fuel at high pressure are themselves enclosed in close-fitting high-pressure containment envelopes. (auth)

1957-07-01T23:59:59.000Z

110

Commercial-Scale Performance Predictions for High-Temperature Electrolysis Plants Coupled to Three Advanced Reactor Types  

DOE Green Energy (OSTI)

This report presents results of system analyses that have been developed to assess the hydrogen production performance of commercial-scale high-temperature electrolysis (HTE) plants driven by three different advanced reactor power-cycle combinations: a high-temperature helium cooled reactor coupled to a direct Brayton power cycle, a supercritical CO2-cooled reactor coupled to a direct recompression cycle, and a sodium-cooled fast reactor coupled to a Rankine cycle. The system analyses were performed using UniSim software. The work described in this report represents a refinement of previous analyses in that the process flow diagrams include realistic representations of the three advanced reactors directly coupled to the power cycles and integrated with the high-temperature electrolysis process loops. In addition, this report includes parametric studies in which the performance of each HTE concept is determined over a wide range of operating conditions. Results of the study indicate that overall thermal-to- hydrogen production efficiencies (based on the low heating value of the produced hydrogen) in the 45 - 50% range can be achieved at reasonable production rates with the high-temperature helium cooled reactor concept, 42 - 44% with the supercritical CO2-cooled reactor and about 33 - 34% with the sodium-cooled reactor.

M. G. McKellar; J. E. O'Brien; J. S. Herring

2007-09-01T23:59:59.000Z

111

EA-0531: Proposed Natural Gas Protection Program for Naval Oil...  

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

1: Proposed Natural Gas Protection Program for Naval Oil Shale Reserves Nos. 1 and 3, Garfield County, Colorado EA-0531: Proposed Natural Gas Protection Program for Naval Oil Shale...

112

The Naval Petroleum and Oil Shale Reserves | Department of Energy  

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

The Naval Petroleum and Oil Shale Reserves The Naval Petroleum and Oil Shale Reserves To ensure sufficient fuel for the fleet, the Government began withdrawing probable oil-bearing...

113

Title 10, Chapter 641 Pertaining to Naval Petroleum Reserves in U.S.C. |  

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

Title 10, Chapter 641 Pertaining to Naval Petroleum Reserves in Title 10, Chapter 641 Pertaining to Naval Petroleum Reserves in U.S.C. Title 10, Chapter 641 Pertaining to Naval Petroleum Reserves in U.S.C. CITE: 10USC7420 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7421 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7422 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7423 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7424 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7425 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7427 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7428 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7429 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7430 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7431 CHAPTER 641--NAVAL PETROLEUM RESERVES CITE: 10USC7432 CHAPTER 641--NAVAL PETROLEUM RESERVES

114

Lunar Nuclear Power Plant With Solid Core Reactor, Heatpipes and Thermoelectric Conversion  

Science Conference Proceedings (OSTI)

This is a lunar nuclear power plant with the advantages of minimum mass, with no moving parts, no pumped liquid coolant, a solid metal rugged core, with no single point of failure. The electrical output is 100 kilowatts with a 500 kilowatt thermal reactor. The thermoelectric converters surround the potassium heatpipes from the core and water heatpipes surround the converter and connect to the radiator. The solid core reactor is made from HT9 alloy. The fuel is uranium oxide with 90% enrichment. The thermoelectric converter is bonded to the outside of the 1.10 inch ID heat pipe and is 30 inches long. The thermoelectric couple is Si/SiGe-Si/SiC Quantum Well with over 20% efficiency with an 890 K hot side and a 490 K cold side and produces 625 Watts. 176 converters produce 110 kWe. With less than 10% loss in controls this yields 100 kWe for use. The cylindrical thermoelectric converter is designed and fabricated by HIPing to keep brittle materials in compression and to ensure conductivity. The solid core is fabricated by machining the heatpipe tubes with 6 grooves that are diffusion bonded together by HIPing to form the fuel tubes. The maximum temperature of the heat pipes is 940 K and the return flow temperature is 890 K. The reactor core is hexagonal shaped, 61 cm. wide and 76.2 cm high with 12 rotating control drums surrounding it. There is shielding to protect components and human habitation. The radiator is daisy shaped at 45 degrees with each petal 5.5 meters long. The design life is ten years.

Sayre, Edwin D. [Engineering Consultant, 218 Brooke Acres Drive, Los Gatos, CA 95032 (United States); Ring, Peter J. [Advanced Methods and Materials, 1190 Mountain View-Alviso Rd. Suite P, Sunnyvale, CA 94089 (United States); Brown, Neil [Engineering Consultant, 5134 Cordoy Lane, San Jose, CA 95124 (United States); Elsner, Norbert B.; Bass, John C. [Hi-Z Technology, Inc., 7606 Miramar Rd. Suite 7400, San Diego, CA 92126 (United States)

2008-01-21T23:59:59.000Z

115

Induced Radioactivity and Waste Classification of Reactor Zone Components of the Chernobyl Nuclear Power Plant Unit 1 After Final Shutdown  

SciTech Connect

The dismantlement of the reactor core materials and surrounding structural components is a major technical concern for those planning closure and decontamination and decommissioning of the Chernobyl Nuclear Power Plant (NPP). Specific issues include when and how dismantlement should be accomplished and what the radwaste classification of the dismantled system would be at the time it is disassembled. Whereas radiation levels and residual radiological characteristics of the majority of the plant systems are directly measured using standard radiation survey and radiochemical analysis techniques, actual measurements of reactor zone materials are not practical due to high radiation levels and inaccessibility. For these reasons, neutron transport analysis was used to estimate induced radioactivity and radiation levels in the Chernobyl NPP Unit 1 reactor core materials and structures.Analysis results suggest that the optimum period of safe storage is 90 to 100 yr for the Unit 1 reactor. For all of the reactor components except the fuel channel pipes (or pressure tubes), this will provide sufficient decay time to allow unlimited worker access during dismantlement, minimize the need for expensive remote dismantlement, and allow for the dismantled reactor components to be classified as low- or medium-level radioactive waste. The fuel channel pipes will remain classified as high-activity waste requiring remote dismantlement for hundreds of years due to the high concentration of induced {sup 63}Ni in the Zircaloy pipes.

Bylkin, Boris K. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Davydova, Galina B. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Zverkov, Yuri A. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Krayushkin, Alexander V. [Russian Research Center 'Kurchatov Institute' (Russian Federation); Neretin, Yuri A. [Chernobyl Nuclear Power Plant (Ukraine); Nosovsky, Anatoly V. [Slavutych Division of the International Chernobyl Center (Ukraine); Seyda, Valery A. [Chernobyl Nuclear Power Plant (Ukraine); Short, Steven M. [Pacific Northwest National Laboratory (United States)

2001-10-15T23:59:59.000Z

116

High Temperature Gas-Cooled Reactors Lessons Learned Applicable to the Next Generation Nuclear Plant  

SciTech Connect

The purpose of this report is to identify possible issues highlighted by these lessons learned that could apply to the NGNP in reducing technical risks commensurate with the current phase of design. Some of the lessons learned have been applied to the NGNP and documented in the Preconceptual Design Report. These are addressed in the background section of this document and include, for example, the decision to use TRISO fuel rather than BISO fuel used in the Peach Bottom reactor; the use of a reactor pressure vessel rather than prestressed concrete found in Fort St. Vrain; and the use of helium as a primary coolant rather than CO2. Other lessons learned, 68 in total, are documented in Sections 2 through 6 and will be applied, as appropriate, in advancing phases of design. The lessons learned are derived from both negative and positive outcomes from prior HTGR experiences. Lessons learned are grouped according to the plant, areas, systems, subsystems, and components defined in the NGNP Preconceptual Design Report, and subsequent NGNP project documents.

J. M. Beck; L. F. Pincock

2011-04-01T23:59:59.000Z

117

Protected air-cooled condenser for the Clinch River Breeder Reactor Plant  

SciTech Connect

The long term residual heat removal for the Clinch River Breeder Reactor Plant (CRBRP) is accomplished through the use of three protected air-cooled condensers (PACC's) each rated at 15M/sub t/ following a normal or emergency shutdown of the reactor. Steam is condensed by forcing air over the finned and coiled condenser tubes located above the steam drums. The steam flow is by natural convection. It is drawn to the PACC tube bundle for the steam drum by the lower pressure region in the tube bundle created from the condensing action. The concept of the tube bundle employs a unique patented configuration which has been commercially available through CONSECO Inc. of Medfore, Wisconsin. The concept provides semi-parallel flow that minimizes subcooling and reduces steam/condensate flow instabilities that have been observed on other similar heat transfer equipment such as moisture separator reheaters (MSRS). The improved flow stability will reduce temperature cycling and associated mechanical fatigue. The PACC is being designed to operate during and following the design basis earthquake, depressurization from the design basis tornado and is housed in protective building enclosure which is also designed to withstand the above mentioned events.

Louison, R.; Boardman, C.E.

1981-05-29T23:59:59.000Z

118

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

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

119

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

120

Separation Requirements for a Hydrogen Production Plant and High-Temperature Nuclear Reactor  

DOE Green Energy (OSTI)

This report provides the methods, models, and results of an evaluation for locating a hydrogen production facility near a nuclear power plant. In order to answer the risk-related questions for this combined nuclear and chemical facility, we utilized standard probabilistic safety assessment methodologies to answer three questions: what can happen, how likely is it, and what are the consequences? As part of answering these questions, we developed a model suitable to determine separation distances for hydrogen process structures and the nuclear plant structures. Our objective of the model-development and analysis is to answer key safety questions related to the placement of one or more hydrogen production plants in the vicinity of a high-temperature nuclear reactor. From a thermal-hydraulic standpoint we would like the two facilities to be quite close. However, safety and regulatory implications force the separation distance to be increased, perhaps substantially. Without answering these safety questions, the likelihood for obtaining a permit to construct and build such as facility in the U.S. would be questionable. The quantitative analysis performed for this report provides us with a scoping mechanism to determine key parameters related to the development of a nuclear-based hydrogen production facility. From our calculations, we estimate that when the separation distance is less than 100m, the core damage frequency is large enough (greater than 1E-6/yr) to become problematic in a risk-informed environment. However, a variety of design modifications, for example blast-deflection barriers, were explored to determine the impact of potential mitigating strategies. We found that these mitigating cases may significantly reduce risk and should be explored as the design for the hydrogen production facility evolves.

Curtis Smith; Scott Beck; Bill Galyean

2005-09-01T23:59:59.000Z

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121

Separation Requirements for a Hydrogen Production Plant and High-Temperature Nuclear Reactor  

Science Conference Proceedings (OSTI)

This paper presents an overview of the engineering methods, models, and results used in an evaluation for locating a hydrogen production facility near a proposed next-generation nuclear power plant. Standard probabilistic safety assessment methodologies were used to answer the risk-related questions for a combined nuclear and chemical facility: what can go wrong? how likely is it to happen? and what are the consequences of it happening? As part of answering these questions, a model was developed suitable for determining the distances separating a hydrogen-production process and nuclear plant structures. The objective of the model-development and analysis is to answer key safety questions relating to the placement of one or more hydrogen production plants in the vicinity of a high-temperature nuclear reactor. From a thermal-hydraulic efficiency standpoint, close proximity of the two facilities is beneficial. Safety and regulatory implications, however, force the separation to be increased, perhaps substantially. The likelihood of obtaining a permit to construct and build such as facility in the United States without answering these safety questions is uncertain. The quantitative analysis performed and described in this paper offers a scoping mechanism to determine key parameters relating to the development of a nuclear-based hydrogen production facility. The calculations indicate that when the facilities are less than 100 m apart, the core damage frequency is large enough (greater than 1E-6/yr) to become problematic in a risk-informed environment. However, a variety of design modifications (blast-deflection barriers, for example) could significantly reduce risk and should be further explored as design of the hydrogen production facility evolves.

Curtis Smith; Scott Beck; William Galyean

2006-06-01T23:59:59.000Z

122

Plant Modernization with Digital Reactor Protection System Safety System Upgrades at US Nuclear Power Stations  

SciTech Connect

As the current fleet of nuclear power plants in the US reaches 25+ years of operation, obsolescence is driving many utilities to implement upgrades to both their safety and non-safety-related Instrumentation and Control (I and C) Systems. Digital technology is the predominant replacement technology for these upgrades. Within the last 15 years, digital control systems have been deployed in non-safety- related control applications at many utilities. In addition, a few utilities have replaced small safety-related systems utilizing digital technology. These systems have shown digital technology to be robust, reliable and simpler to maintain. Based upon this success, acceptance of digital technology has gained momentum with both utilities and regulatory agencies. Today, in an effort to extend the operating lives of their nuclear stations and resolve obsolescence of critical components, utilities are now pursuing digital technology for replacement of their primary safety systems. AREVA is leading this effort in the United States with the first significant digital upgrade of a major safety system. AREVA has previously completed upgrades to safety-related control systems emergency diesel engine controls and governor control systems for a hydro station which serves as the emergency power source for a nuclear station. Currently, AREVA is implementing the replacement of both the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS) on all three units at a US PWR site. (authors)

Heckle, Wm. Lloyd; Bolian, Tricia W. [AREVA NP, an AREVA and Siemens Company, 1345 Ridgeland Parkway, Suite 130 (United States)

2006-07-01T23:59:59.000Z

123

Study of Pu consumption in light water reactors: Evaluation of GE advanced boiling water reactor plants, compilation of Phase 1C task reports  

Science Conference Proceedings (OSTI)

This report summarizes the evaluations conducted during Phase 1C of the Pu Disposition Study have provided further results which reinforce the conclusions reached during Phase 1A & 1B: These conclusions clearly establish the benefits of the fission option and the use of the ABWR as a reliable, proven, well-defined and cost-effective means available to disposition the weapons Pu. This project could be implemented in the near-term at a cost and on a schedule being validated by reactor plants currently under construction in Japan and by cost and schedule history and validated plans for MOX plants in Europe. Evaluations conducted during this phase have established that (1) the MOX fuel is licensable based on existing criteria for new fuel with limited lead fuel rod testing, (2) that the applicable requirements for transport, handling and repository storage can be met, and (3) that all the applicable safeguards criteria can be met.

Not Available

1994-01-15T23:59:59.000Z

124

Removal plan for Shippingport pressurized water reactor core 2 blanket fuel assemblies form T plant to the canister storage building  

Science Conference Proceedings (OSTI)

This document presents the current strategy and path forward for removal of the Shippingport Pressurized Water Reactor Core 2 blanket fuel assemblies from their existing storage configuration (wet storage within the T Plant canyon) and transport to the Canister Storage Building (designed and managed by the Spent Nuclear Fuel. Division). The removal plan identifies all processes, equipment, facility interfaces, and documentation (safety, permitting, procedures, etc.) required to facilitate the PWR Core 2 assembly removal (from T Plant), transport (to the Canister storage Building), and storage to the Canister Storage Building. The plan also provides schedules, associated milestones, and cost estimates for all handling activities.

Lata

1996-09-26T23:59:59.000Z

125

Design of the Next Generation Nuclear Plant Graphite Creep Experiments for Irradiation in the Advanced Test Reactor  

SciTech Connect

The United States Department of Energys Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energys lead laboratory for nuclear energy development. The ATR is one of the worlds premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of the experiment. The final design phase for the first experiment was completed in September 2008, and the fabrication and assembly of the experiment test train as well as installation and testing of the control and support systems that will monitor and control the experiment during irradiation are being completed in early calendar 2009. The first experiment is scheduled to be ready for insertion in the ATR by April 30, 2009. This paper will discuss the design of the experiment including the test train and the temperature and compressive load monitoring, control, and data collection systems.

S. Blaine Grover

2009-05-01T23:59:59.000Z

126

Distributed energy resources at naval base ventura county building 1512  

E-Print Network (OSTI)

up by a DER system. Distributed Energy Resources at NavalFebruary 2003. Distributed Energy Resources in Practice: ARyan. January 2004. Distributed Energy Resources Customer

Bailey, Owen C.; Marnay, Chris

2004-01-01T23:59:59.000Z

127

DOE - Office of Legacy Management -- Naval Oil Shale Reserves...  

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

Oil Shale Reserves Site - 013 FUSRAP Considered Sites Site: Naval Oil Shale Reserves Site (013 ) Designated Name: Alternate Name: Location: Evaluation Year: Site Operations: Site...

128

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

129

Feasibility of Reduced Boron Concentration Operation in Pressurized Water Reactor Plants  

Science Conference Proceedings (OSTI)

Technical Paper / Second Seminar on Accelerated Testing of Materials in Spent Nuclear Fuel and High-Level Waste Storage Systems / Fission Reactors

Aung Tharn Daing; Myung-Hyun Kim

130

Analysis of operating alternatives for the Naval Computer and Telecommunications Station Cogeneration Facility at Naval Air Station North Island, San Diego, California  

SciTech Connect

The Naval Facilities Engineering Command Southwestern Division commissioned Pacific Northwest Laboratory (PNL), in support of the US Department of Energy (DOE) Federal Energy Management Program (FEMP), to determine the most cost-effective approach to the operation of the cogeneration facility in the Naval Computer and Telecommunications Station (NCTS) at the Naval Air Station North Island (NASNI). Nineteen alternative scenarios were analyzed by PNL on a life-cycle cost basis to determine whether to continue operating the cogeneration facility or convert the plant to emergency-generator status. This report provides the results of the analysis performed by PNL for the 19 alternative scenarios. A narrative description of each scenario is provided, including information on the prime mover, electrical generating efficiency, thermal recovery efficiency, operational labor, and backup energy strategy. Descriptions of the energy and energy cost analysis, operations and maintenance (O&M) costs, emissions and related costs, and implementation costs are also provided for each alternative. A summary table presents the operational cost of each scenario and presents the result of the life-cycle cost analysis.

Parker, S.A.; Carroll, D.M.; McMordie, K.L.; Brown, D.R.; Daellenbach, K.K.; Shankle, S.A.; Stucky, D.J.

1993-12-01T23:59:59.000Z

131

Implementation plan for operating alternatives for the Naval Computer and Telecommunications Station cogeneration facility at Naval Air Station North Island, San Diego, California  

SciTech Connect

The goal of the US Department of Energy (DOE) Federal Energy Management Program (FEMP) is to facilitate energy efficiency improvements at federal facilities. This is accomplished by a balanced program of technology development, facility assessment, and use of cost-sharing procurement mechanisms. Technology development focuses upon the tools, software, and procedures used to identify and evaluate energy efficiency technologies and improvements. For facility assessment, FEMP provides metering equipment and trained analysts to federal agencies exhibiting a commitment to improve energy use efficiency. To assist in procurement of energy efficiency measures, FEMP helps federal agencies devise and implement performance contracting and utility demand-side management strategies. Pacific Northwest Laboratory (PNL) supports the FEMP mission of energy systems modernization. Under this charter, the Laboratory and its contractors work with federal facility energy managers to assess and implement energy efficiency improvements at federal facilities nationwide. The SouthWestern Division of the Naval Facilities Engineering Command, in cooperation with FEMP, has tasked PNL with developing a plan for implementing recommended modifications to the Naval Computer and Telecommunications Station (NCTS) cogeneration plant at the Naval Air Station North Island (NASNI) in San Diego. That plan is detailed in this report.

Carroll, D.M.; Parker, S.A.; Stucky, D.J.

1994-04-01T23:59:59.000Z

132

Thermal Design of an Ultrahigh Temperature Vapor Core Reactor Combined Cycle Nuclear Power Plant  

SciTech Connect

Current work modeling high temperature compact heat exchangers may demonstrate the design feasibility of a Vapor Core Reactor (VCR) driven combined cycle power plant. For solid nuclear fuel designs, the cycle efficiency is typically limited by a metallurgical temperature limit which is dictated by fuel and structural melting points. In a vapor core, the gas/vapor phase nuclear fuel is uniformly mixed with the topping cycle working fluid. Heat is generated homogeneously throughout the working fluid thus extending the metallurgical temperature limit. Because of the high temperature, magnetohydrodynamic (MHD) generation is employed for topping cycle power extraction. MHD rejected heat is transported via compact heat exchanger to a conventional Brayton gas turbine bottoming cycle. High bottoming cycle mass flow rates are required to remove the waste heat because of low heat capacities for the bottoming cycle gas. High mass flow is also necessary to balance the high Uranium Tetrafluoride (UF{sub 4}) mass flow rate in the topping cycle. Heat exchanger design is critical due to the high temperatures and corrosive influence of fluoride compounds and fission products existing in VCR/MHD exhaust. Working fluid compositions for the topping cycle include variations of Uranium Tetrafluoride, Helium and various electrical conductivity seeds for the MHD. Bottoming cycle working fluid compositions include variations of Helium and Xenon. Some thought has been given to include liquid metal vapor in the bottoming cycle for a Cheng or evaporative cooled design enhancement. The NASA Glenn Lewis Research Center code Chemical Equilibrium with Applications (CEA) is utilized for evaluating chemical species existing in the gas stream. Work being conducted demonstrates the compact heat exchanger design, utilization of the CEA code, and assessment of different topping and bottoming working fluid compositions. (authors)

Bays, Samuel E.; Anghaie, Samim; Smith, Blair; Knight, Travis [Innovative Space Power and Propulsion Institute, University of Florida, 202 Nuclear Science Building, Gainesville, FL 32611 (United States)

2004-07-01T23:59:59.000Z

133

Naval Reserve Force : cost and benefit analysis of reducing the number of Naval Surface Reserve Force operating budget holders ; .  

E-Print Network (OSTI)

??The Quadrennial Defense Review 1997 recommended reductions of civilian and military personnel associated with infrastructure. The Naval Reserve Force is aggressively pursuing options to reduce (more)

Young, Eric Coy

1997-01-01T23:59:59.000Z

134

Geothermal energy at Long Beach Naval Shipyard and Naval Station and at Seal Beach Naval Weapons Station, California. Final Report 1  

DOE Green Energy (OSTI)

The purpose of this project was to determine and evaluate sources of geothermal energy at two military bases in southern California, the Long Beach Naval Shipyard and Naval Station and the Seal Beach Naval Weapons Station. One part of the project focused on the natural geothermal characteristics beneath the naval bases. Another part focused on the geothermal energy produced by oilfield operations on and adjacent to each base. Results of the study are presented here for the US Department of the Navy to use in its program to reduce its reliance on petroleum by the development of different sources of energy. The project required research of various reports and data, both published and unpublished, particularly those of the California Department of Conservation, Division of Oil and Gas and of oil companies with leases on or adjacent to the naval bases. Important field investigations included the measurement of well-head temperatures of fluids produced from selected oil wells at each naval base and a detailed gravity survey of the Seal Beach naval base and vicinity. The well-head temperatures were needed to evaluate individual wells as sources of geothermal energy, while the gravity survey attempted to discover subsurface geologic structures that might contain geothermal fluids of temperatures higher than those predicted by the regional geothermal conditions.

Higgins, C.T.; Chapman, R.H.

1984-01-01T23:59:59.000Z

135

Success Story: Naval Medical Center San Diego Co-Generation Project...  

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

Success Story: Naval Medical Center San Diego Co-Generation Project Success Story: Naval Medical Center San Diego Co-Generation Project Presentation covers the FUPWG Fall Meeting,...

136

Advanced Light Water Reactor Plants System 80+{trademark} Design Certification Program. Annual progress report, October 1, 1992--September 30, 1993  

SciTech Connect

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1993 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW{sub t} (1350 MWe) Pressurized Water Reactor (PWR). The design consists of an essentially complete plant. It is based on evolutionary improvements to the Standardized System 80 nuclear steam supply system in operation at Palo Verde Units 1, 2, and 3, and the Duke Power Company P-81 balance-of-plant (BOP) that was designed and partially constructed at the Cherokee plant site. The System 80/P-81 original design has been substantially enhanced to increase conformance with the EPRI ALWR Utility Requirements Document (URD). Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units form the basis of the Korean standardization program. The full System 80+ standard design has been offered to the Republic of China, in response to their recent bid specification. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was submitted to the NRC and a Draft Safety Evaluation Report was issued by the NRC in October 1992. CESSAR-DC contains the technical basis for compliance with the EPRI URD for simplified emergency planning. The Nuclear Steam Supply System (NSSS) is the standard ABB-Combustion Engineering two-loop arrangement with two steam generators, two hot legs and four cold legs each with a reactor coolant pump. The System 80+ standard plant includes a sperical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual containment.

Not Available

1993-12-31T23:59:59.000Z

137

CX-008819: Categorical Exclusion Determination | Department of...  

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

Plant Project CX(s) Applied: B1.15 Date: 06202012 Location(s): Idaho Offices(s): Naval Nuclear Propulsion Program, Naval Reactors The scope of the Naval Reactors Facility (NRF)...

138

A Tabu Search Heuristic for Resource Management in Naval Warfare  

Science Conference Proceedings (OSTI)

Effective utilization of scarce resources, in particular weapon resources, is a prominent issue in naval anti-air warfare. In this paper, defence plans are constructed to guide the allocation and scheduling of different types of defence weapons against ... Keywords: defence plan, naval warfare, resource management, tabu search, weapon

Dale E. Blodgett; Michel Gendreau; Franois Guertin; Jean-Yves Potvin; Ren Sguin

2003-03-01T23:59:59.000Z

139

Mirror Advanced Reactor Study (MARS). Final report. Volume 1-B. Commercial fusion electric plant  

SciTech Connect

Volume 1-B contains the following chapters: (1) blanket and reflector; (2) central cell shield; (3) central cell structure; (4) heat transport and energy conversion; (5) tritium systems; (6) cryogenics; (7) maintenance; (8) safety; (9) radioactivity, activation, and waste disposal; (10) instrumentation and control; (11) balance of plant; (12) plant startup and operation; (13) plant availability; (14) plant construction; and (15) economic analysis.

Donohue, M.L.; Price, M.E. (eds.)

1984-07-01T23:59:59.000Z

140

Interim report VII, production test IP-549-A half-plant low alum feed water treatment at F Reactor  

SciTech Connect

A half-plant low alum water treatment test began at F Reactor on January 16, 1963. The test, which had been prompted by the analysis of ledge corrosion attack on fuel elements, will demonstrate whether or not high alum feed is responsible for increasing the frequency of ledge and groove corrosion attack on fuel element surfaces. The effect will be evaluated by comparing visual examination results obtained from the normal production fuel irradiated in process water treated with two different alum feed rates. Six 20-column fuel discharges, ten columns from each side of the reactor, have been taken during the test as follows: (1) One discharge prior to the start of the test. (2) One discharge such that the test side was exposed to coolant treated with both high and low alum feed. (3) Four discharges under test conditions. This report discusses the results obtained from the fifth discharge under test conditions.

Geier, R.G.

1964-03-18T23:59:59.000Z

Note: This page contains sample records for the topic "naval reactor plants" 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

Integrated Surface Geophysical Methods for Characterization of the Naval  

Open Energy Info (EERE)

Integrated Surface Geophysical Methods for Characterization of the Naval Integrated Surface Geophysical Methods for Characterization of the Naval Air Warfare Center, New Jersey Jump to: navigation, search OpenEI Reference LibraryAdd to library Web Site: Integrated Surface Geophysical Methods for Characterization of the Naval Air Warfare Center, New Jersey Author USGS Published Publisher Not Provided, 2013 DOI Not Provided Check for DOI availability: http://crossref.org Online Internet link for Integrated Surface Geophysical Methods for Characterization of the Naval Air Warfare Center, New Jersey Citation USGS. Integrated Surface Geophysical Methods for Characterization of the Naval Air Warfare Center, New Jersey [Internet]. 2013. [updated 2013/01/03;cited 2013/11/22]. Available from: http://water.usgs.gov/ogw/bgas/toxics/NAWC-surface.html

142

Naval Spent Fuel Rail Shipment Accident Exercise Objectives  

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

NAVAL SPENT FUEL RAIL SHIPMENT NAVAL SPENT FUEL RAIL SHIPMENT ACCIDENT EXERCISE OBJECTIVES * Familiarize stakeholders with the Naval spent fuel ACCIDENT EXERCISE OBJECTIVES Familiarize stakeholders with the Naval spent fuel shipping container characteristics and shipping practices * Gain understanding of how the NNPP escorts who accompany the spent fuel shipments will interact with civilian emergency services representatives g y p * Allow civilian emergency services agencies the opportunity to evaluate their response to a pp y p simulated accident * Gain understanding of how the communications links that would be activated in an accident involving a Naval spent fuel shipment would work 1 NTSF May 11 ACCIDENT EXERCISE TYPICAL TIMELINE * Conceptual/Organizational Meeting - April 6 E R T i d it t t d TYPICAL TIMELINE

143

Anti-proliferation safeguard system for General Electric's PRISM reactor plant  

E-Print Network (OSTI)

The proliferation resistance of a nuclear power plant has become an increasingly important issue due to the political climate of nuclear power at the present. Any new power plant that is constructed must be proliferation ...

Tenorio, Luis E

2008-01-01T23:59:59.000Z

144

SUPPLEMENT ANALYSIS OF FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL  

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

FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL FOREIGN RESEARCH REACTOR srENT NUCLEAR FUEL TRANSPORTATION ALONG OTHER THAN~. PRESENTATIVE ROUTE FROM CONCORD NAVAL WEAPO~~ STATION TO IDAHO NATIONAL ENGINEERING AND ENVIRONMENTAL LADORA TORY Introduction The Department of Energy is planning to transport foreign research reactor spent nuclear fuel by rail from the Concord Naval Weapons Station (CNWS), Concord, California, to the Idaho National Engineering and Environmental Laboratory (INEEL). The environmental analysis supporting the decision to transport, by rail or truck, foreign research reactor spent nuclear fuel from CNWS to the INEEL is contained in +he Final Environmental Impact Statement on a Proposed Nuclear Weapons Nonproliftration Policy Concerning Foreign Research Reactor

145

Nuclear Reactor Technologies | Department of Energy  

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

Reactor Technologies Nuclear Reactor Technologies TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority TVA Watts Bar Nuclear Power Plant | Photo...

146

Generic Requirements Specification for Upgrading the Safety-Related Reactor Trip and Engineered Safety Features Actuation Systems in Westinghouse PWR Nuclear Power Plants  

Science Conference Proceedings (OSTI)

To address obsolescence concerns, a generic requirements specification for digital upgrades to existing reactor trip systems and engineered safety features actuation systems in a Westinghouse pressurized water reactor (PWR) was developed. System requirements are based on a 4-loop PWR with a solid-state protection system since this typifies the most advanced capability level. However, the specification is applicable to relay-based 2- and 3-loop plants where some or all of the advances in the newest solid-...

2001-10-19T23:59:59.000Z

147

Advanced light water reactor plants system 80+{trademark} design certification program. Annual progress report, October 1, 1993--September 30, 1994  

SciTech Connect

The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80{sup +}{trademark} during the U.S. government`s 1994 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2 and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems. Some design enhancements incorporated in the System 80+ design are included in the four units currently under construction in the Republic of Korea. These units and the System 80+ design form the basis of the Korean standardization program. The Nuclear Island portion of the System 80+ standard design has also been offered to the Republic of China, in response to their bid specification for an ALWR. The ABB-CE Standard Safety Analysis Report (CESSAR-DC) was docketed by the Nuclear Regulatory Commission (NRC) in May 1991 and a Draft Safety Evaluation Report (DSER) was issued in October 1992.

Not Available

1995-01-01T23:59:59.000Z

148

Naval Petroleum Reserve No. 3 Disposition Decision Analysis and Timeline  

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

This Report to Congress provides a summary of the analysis supporting DOE's determination to dispose of the Naval Petroleum Reserve No. 3 through sale of all right, title, interest on the open market.

149

Distributed energy resources at naval base ventura county building 1512  

E-Print Network (OSTI)

the prices of electricity and gas, that might make PV costprices increase by 10% Distributed Energy Resources at Naval Base Ventura Country Building 1512 over current Public Works levels, then PV

Bailey, Owen C.; Marnay, Chris

2004-01-01T23:59:59.000Z

150

Distributed energy resources at naval base ventura county building 1512  

E-Print Network (OSTI)

the affects of energy prices and tariff structures on energythe default SCE tariff, total energy bills for Building 1512$0.1097. This tariff Distributed Energy Resources at Naval

Bailey, Owen C.; Marnay, Chris

2004-01-01T23:59:59.000Z

151

Nuclear power plants: structure and function  

SciTech Connect

Topics discussed include: steam electric plants; BWR type reactors; PWR type reactors; thermal efficiency of light water reactors; other types of nuclear power plants; the fission process and nuclear fuel; fission products and reactor afterheat; and reactor safety.

Hendrie, J.M.

1983-01-01T23:59:59.000Z

152

Mirror Advanced Reactor Study (MARS). Final report. Volume 2. Commercial fusion synfuels plant  

DOE Green Energy (OSTI)

Volume 2 contains the following chapters: (1) synfuels; (2) physics base and parameters for TMR; (3) high-temperature two-temperature-zone blanket system for synfuel application; (4) thermochemical hydrogen processes; (5) interfacing the sulfur-iodine cycle; (6) interfacing the reactor with the thermochemical process; (7) tritium control in the blanket system; (8) the sulfur trioxide fluidized-bed composer; (9) preliminary cost estimates; and (10) fuels beyond hydrogen. (MOW)

Donohue, M.L.; Price, M.E. (eds.)

1984-07-01T23:59:59.000Z

153

Advanced Light Water Reactor Utility Requirements Document, Volume 2, Revision 8: ALWR Evolutionary Plant  

Science Conference Proceedings (OSTI)

EPRI's ALWR Program has been an industry-wide effort to establish the technical foundation for design of the advanced light water reactor (ALWR). This program included participation and sponsorship of several international utility companies and close cooperation with the U.S. Department of Energy. The cornerstone of the ALWR Program is a set of utility design requirements, which are contained in the ALWR Utility Requirements Document. The purpose of this document is to present a clear, complete statement...

1999-03-30T23:59:59.000Z

154

Advanced Light Water Reactor Utility Requirements Document, Volume 3, Revision 8: ALWR Passive Plant  

Science Conference Proceedings (OSTI)

EPRI's ALWR Program has been an industry-wide effort to establish the technical foundation for design of the advanced light water reactor (ALWR). This program included participation and sponsorship of several international utility companies and close cooperation with the U.S. Department of Energy. The cornerstone of the ALWR Program is a set of utility design requirements, which are contained in the ALWR Utility Requirements Document. The purpose of this document is to present a clear, complete statement...

1999-03-30T23:59:59.000Z

155

Evaluation of the applicability of existing nuclear power plant regulatory requirements in the U.S. to advanced small modular reactors.  

SciTech Connect

The current wave of small modular reactor (SMR) designs all have the goal of reducing the cost of management and operations. By optimizing the system, the goal is to make these power plants safer, cheaper to operate and maintain, and more secure. In particular, the reduction in plant staffing can result in significant cost savings. The introduction of advanced reactor designs and increased use of advanced automation technologies in existing nuclear power plants will likely change the roles, responsibilities, composition, and size of the crews required to control plant operations. Similarly, certain security staffing requirements for traditional operational nuclear power plants may not be appropriate or necessary for SMRs due to the simpler, safer and more automated design characteristics of SMRs. As a first step in a process to identify where regulatory requirements may be met with reduced staffing and therefore lower cost, this report identifies the regulatory requirements and associated guidance utilized in the licensing of existing reactors. The potential applicability of these regulations to advanced SMR designs is identified taking into account the unique features of these types of reactors.

LaChance, Jeffrey L.; Wheeler, Timothy A.; Farnum, Cathy Ottinger; Middleton, Bobby D.; Jordan, Sabina Erteza; Duran, Felicia Angelica; Baum, Gregory A.

2013-05-01T23:59:59.000Z

156

Digital computer code for simulating the dynamics of full-size dual-purpose desalting plants using a pressurized water reactor as a heat source  

SciTech Connect

A digital simulator was developed for use in calculating the dynamic response of full-size dual-purpose desalting plants. This simulator consists of a multistage flash (MSF) evaporator, a pressurized water reactor (PWR) as the heat source, a drumtype steam generator, and a turbine plant utilizing a back- pressure turbine. A bypass steam system was modeled to achieve flexible operation of the electric power and water portions of the combined plant. The proposed use of this simulator is to investigate various coupling and control schemes and to help determine possible problem areas in full back-pressure turbine dual-purpose desalting plant designs. (auth)

Delene, J.G.

1973-10-01T23:59:59.000Z

157

BOILING WATER REACTOR TRANSIENT INSTABILITY STUDIES OF RINGHALS 1 REACTOR USING TRACE COUPLED WITH PARCS.  

E-Print Network (OSTI)

??Reactor plant design often incorporates data and insight ascertained from computer code simulations of plant dynamics and reactor core behavior. Increasing utilization of data gathered (more)

Walls, Robert

2009-01-01T23:59:59.000Z

158

ASSESSMENT OF THE RADIONUCLIDE COMPOSITION OF "HOT PARTICLES" SAMPLED IN THE CHERNOBYL NUCLEAR POWER PLANT FOURTH REACTOR UNIT  

SciTech Connect

Fuel-containing materials sampled from within the Chernobyl Nuclear Power Plant (ChNPP) 4th Reactor Unit Confinement Shelter were spectroscopically studied for gamma and alpha content. Isotopic ratios for cesium, europium, plutonium, americium, and curium were identified and the fuel burnup in these samples was determined. A systematic deviation in the burnup values based on the cesium isotopes, in comparison with other radionuclides, was observed. The conducted studies were the first ever performed to demonstrate the presence of significant quantities of {sup 242}Cm and {sup 243}Cm. It was determined that there was a systematic underestimation of activities of transuranic radionuclides in fuel samples from inside of the ChNPP Confinement Shelter, starting from {sup 241}Am (and going higher), in comparison with the theoretical calculations.

Farfan, E.; Jannik, T.; Marra, J.

2011-10-01T23:59:59.000Z

159

An Estimate of the Cost of Electricity from Light Water Reactors and Fossil Plants with Carbon Capture and Sequestration  

SciTech Connect

As envisioned in this report, LIFE technology lends itself to large, centralized, baseload (or 'always on') electrical generation. Should LIFE plants be built, they will have to compete in the electricity market with other generation technologies. We consider the economics of technologies with similar operating characteristics: significant economies of scale, limited capacity for turndown, zero dependence on intermittent resources and ability to meet environmental constraints. The five generation technologies examined here are: (1) Light Water Reactors (LWR); (2) Coal; (3) Coal with Carbon Capture and Sequestration (CCS); (4) Natural Gas; and (5) Natural Gas with Carbon Capture and Sequestration. We use MIT's cost estimation methodology (Du and Parsons, 2009) to determine the cost of electricity at which each of these technologies is viable.

Simon, A J

2009-08-21T23:59:59.000Z

160

PATHFINDER ATOMIC POWER PLANT. BORON STAINLESS STEEL CONTROL RODS FOR PATHFINDER REACTOR. Final Summary Report  

SciTech Connect

An extensive development, fabrication, and design analysis program was completed on 2 wt% boron stainiess steel for Pathfinder boiler and superheater control rods. Results indicated that the control rods will operate satisfactorily in the reactor under all foreseeable conditions. It was concluded that the maximum local burnup on the boiler rods is limited to 1.0 core at.% and the superheater rods to 0.5 core at.% to remain well below the onset of any deleterious effects. All stresses imposed on the material were found to be below yield strength. The deflections were calculated and were not sufficient to restrict any movement of the cruciform during reactor operations and scrams. A welding development program produced a sound fabrication method for producing the control rods. A prototype rod was extensively tested in the hot control rod drive test, including a large number of scrams. The rod showed no deleterious effects. The rods were designed to accommodate any helium release and swelling that may occur. (M.C.G.)

Patterson, D.A.; Nehrig, D.A.; Klumb, R.H.; Peterson, T.E.

1963-02-15T23:59:59.000Z

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


161

BWR (boiling-water reactor) radiation control: In-plant demonstration at Vermont Yankee: Final report  

Science Conference Proceedings (OSTI)

Results of the RP1934 program, which was established by EPRI in 1981 to demonstrate the adequacy of BRAC program (RP819) principles for BWR radiation control at Vermont Yankee, are presented. Evaluations were performed of the effectiveness of optimization of purification system performance, control of feedwater dissolved oxygen concentrations, minimization of corrosion product and ionic transport, and improved startup, shutdown, and layup practices. The impact on shutdown radiation levels of these corrective actions was assessed based on extensive primary system radiation survey and component gamma scan data. Implementation of the BRAC recommendations was found to be insufficient to reduce the rate of activity buildup on out-of-core surfaces at Vermont Yankee, and additional corrective actions were found necessary. Specifically, replacement of cobalt-bearing materials in the control rod drive pins and rollers and feedwater regulating valves was pursued as was installation of electropolished 316 stainless steel during a recirculation piping replacement program. Aggressive programs to further reduce copper concentrations in the reactor water by improving condensate demineralizer efficiency and to minimize organic ingress to the power cycle by reducing organic concentrations in recycled radwaste also were undertaken. Evaluations of the impact on activity buildup of several pretreatment processes including prefilming in moist air, preexposure to high temperature water containing zinc, and electropolishing also were performed in a test loop installed in the reactor water cleanup system. A significant beneficial impact of electropolishing was shown to be present for periods up to 6000 hours.

Palino, G.F.; Hobart, R.L.; Sawochka, S.G.

1987-10-01T23:59:59.000Z

162

Advanced Nuclear Reactors | Department of Energy  

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

in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor...

163

Naval Spent Nuclear Fuel disposal Container System Description Document  

Science Conference Proceedings (OSTI)

The Naval Spent Nuclear Fuel Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers/waste packages are loaded and sealed in the surface waste handling facilities, transferred underground through the access drifts using a rail mounted transporter, and emplaced in emplacement drifts. The Naval Spent Nuclear Fuel Disposal Container System provides long term confinement of the naval spent nuclear fuel (SNF) placed within the disposal containers, and withstands the loading, transfer, emplacement, and retrieval operations. The Naval Spent Nuclear Fuel Disposal Container System provides containment of waste for a designated period of time and limits radionuclide release thereafter. The waste package maintains the waste in a designated configuration, withstands maximum credible handling and rockfall loads, limits the waste form temperature after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Each naval SNF disposal container will hold a single naval SNF canister. There will be approximately 300 naval SNF canisters, composed of long and short canisters. The disposal container will include outer and inner cylinder walls and lids. An exterior label will provide a means by which to identify a disposal container and its contents. Different materials will be selected for the waste package inner and outer cylinders. The two metal cylinders, in combination with the Emplacement Drift System, drip shield, and the 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 while the outer cylinder and outer cylinder lids will be made of high-nickel alloy.

N. E. Pettit

2001-07-13T23:59:59.000Z

164

Evaluation of Suitability of Selected Set of Coal Plant Sites for Repowering with Small Modular Reactors  

SciTech Connect

This report summarizes the approach that ORNL developed for screening a sample set of small coal stations for possible repowering with SMRs; the methodology employed, including spatial modeling; and initial results for these sample plants. The objective in conducting this type of siting evaluation is to demonstrate the capability to characterize specific sample coal plant sites to identify any particular issues associated with repowering existing coal stations with SMRs using OR-SAGE; it is not intended to be a definitive assessment per se as to the absolute suitability of any particular site.

Belles, Randy [ORNL; Copinger, Donald A [ORNL; Mays, Gary T [ORNL; Omitaomu, Olufemi A [ORNL; Poore III, Willis P [ORNL

2013-03-01T23:59:59.000Z

165

Cavitation and two-phase flow characteristics of SRPR (Savannah River Plant Reactor) pump. Final report  

SciTech Connect

The possible head degradation of the SRPR pumps may be attributable to two independent phenomena, one due to the inception of cavitation and the other due to the two-phase flow phenomena. The head degradation due to the appearance of cavitation on the pump blade is hardly likely in the conventional pressurized water reactor (PWR) since the coolant circulating line is highly pressurized so that the cavitation is difficult to occur even at LOCA (loss of coolant accident) conditions. On the other hand, the suction pressure of SRPR pump is order-of-magnitude smaller than that of PWR so that the cavitation phenomena, may prevail, should LOCA occur, depending on the extent of LOCA condition. In this study, therefore, both cavitation phenomena and two-phase flow phenomena were investigated for the SRPR pump by using various analytical tools and the numerical results are presented herein.

Not Available

1991-07-01T23:59:59.000Z

166

Thermal-Hydraulic Analyses of Heat Transfer Fluid Requirements and Characteristics for Coupling A Hydrogen Production Plant to a High-Temperature Nuclear Reactor  

DOE Green Energy (OSTI)

The Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the hightemperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant, may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood. Seven possible configurations for a system that transfers heat between the nuclear reactor and the hydrogen and/or electrical generation plants were identified. These configurations included both direct and indirect cycles for the production of electricity. Both helium and liquid salts were considered as the working fluid in the intermediate heat transport loop. Methods were developed to perform thermalhydraulic and cycle-efficiency evaluations of the different configurations and coolants. The thermalhydraulic evaluations estimated the sizes of various components in the intermediate heat transport loop for the different configurations. The relative sizes of components provide a relative indication of the capital cost associated with the various configurations. Estimates of the overall cycle efficiency of the various configurations were also determined. The evaluations determined which configurations and coolants are the most promising from thermal-hydraulic and efficiency points of view. These evaluations also determined which configurations and options do not appear to be feasible at the current time.

C. B. Davis; C. H. Oh; R. B. Barner; D. F. Wilson

2005-06-01T23:59:59.000Z

167

NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG 7 JUNE 2013 Quick Facts..............................................................................................................................................................4  

E-Print Network (OSTI)

NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG ­ 7 JUNE 2013 1 CONTENTS Quick Facts......................................................................................................................................................26 #12;NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG ­ 7 JUNE 2013 2 Curriculum Listing...............................................................................................................................................428 Energy Core Group

168

NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG 3 JAN 2013 Quick Facts..............................................................................................................................................................4  

E-Print Network (OSTI)

NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG ­ 3 JAN 2013 1 CONTENTS Quick Facts......................................................................................................................................................26 #12;NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG ­ 3 JAN 2013 2 Curriculum Listing ................................................................................................427 Energy Core Group

169

NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG 8 MARCH 2013 Quick Facts..............................................................................................................................................................4  

E-Print Network (OSTI)

NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG ­ 8 MARCH 2013 1 CONTENTS Quick Facts......................................................................................................................................................26 #12;NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG ­ 8 MARCH 2013 2 Curriculum Listing ................................................................................................428 Energy Core Group

170

Naval petroleum and oil shale reserves: Annual report of operations, FY 1987  

SciTech Connect

Production and reserves, development and exploration, revenues and expenditures, sales, environment and safety, and litigation are discussed for naval petroleum reserves numbers one through three and for naval oil shale reserves. 28 figs., 21 tabs. (ACT)

Not Available

1987-01-01T23:59:59.000Z

171

NPS-SCAT electrical power system ; Naval Postgraduate School Solar Cell Array Tester .  

E-Print Network (OSTI)

??The Naval Postgraduate School Solar Cell Array Tester (NPS-SCAT) seeks to expand the CubeSat knowledge base and provide learning possibilities at the Naval Postgraduate School. (more)

Dorn, Lawrence Tyrone.

2009-01-01T23:59:59.000Z

172

DOE - Office of Legacy Management -- Norfolk Naval Station - VA 05  

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

Norfolk Naval Station - VA 05 Norfolk Naval Station - VA 05 FUSRAP Considered Sites Site: NORFOLK NAVAL STATION (VA.05) Eliminated from consideration under FUSRAP Designated Name: Not Designated Alternate Name: None Location: Norfolk , Virginia VA.05-1 Evaluation Year: 1993 VA.05-1 Site Operations: Demonstration of extinguishing a uranium fire at the Fire Fighters School for AEC contractors. VA.05-3 VA.05-2 Site Disposition: Eliminated - Potential for contamination considered remote based on the limited quantity of materials handled VA.05-1 Radioactive Materials Handled: Yes Primary Radioactive Materials Handled: Uranium VA.05-2 Radiological Survey(s): Yes - Health and Safety Monitoring during operations only VA.05-2 Site Status: Eliminated from consideration under FUSRAP

173

Balance of Plant System Analysis and Component Design of Turbo-Machinery for High Temperature Gas Reactor Systems  

Science Conference Proceedings (OSTI)

The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the power conversion system have been verified with an industry-standard general thermal-fluid code Flownet. With respect to the dynamic model, bypass valve control and inventory control have been used as the primary control methods for the power conversion system. By performing simulation using the dynamic model with the designed control scheme, the combination of bypass and inventory control was optimized to assure system stability within design temperature and pressure limits. Bypass control allows for rapid control system response while inventory control allows for ultimate steady state operation at part power very near the optimum operating point for the system. Load transients simulations show that the indirect, three-shaft arrangement gas turbine power conversion system is stable and controllable. For the indirect cycle the intermediate heat exchanger (IHX) is the interface between the reactor and the turbomachinery systems. As a part of the design effort the IHX was identified as the key component in the system. Two technologies, printed circuit and compact plate-fin, were investigated that have the promise of meeting the design requirements for the system. The reference design incorporates the possibility of using either technology although the compact plate-fin design was chosen for subsequent analysis. The thermal design and parametric analysis with an IHX and recuperator using the plate-fin configuration have been performed. As a three-shaft arrangement, the turbo-shaft sets consist of a pair of turbine/compressor sets (high pressure and low pressure turbines with same-shaft compressor) and a power turbine coupled with a synchronous generator. The turbines and compressors are all axial type and the shaft configuration is horizontal. The core outlet/inlet temperatures are 900/520 C, and the optimum pressure ratio in the power conversion cycle is 2.9. The design achieves a plant net efficiency of approximately 48%.

Ronald G. Ballinger Chunyun Wang Andrew Kadak Neil Todreas

2004-08-30T23:59:59.000Z

174

NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapter 1, project number 669  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume 1, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

Not Available

1994-08-01T23:59:59.000Z

175

NRC review of Electric Power Research Institute`s advanced light water reactor utility requirements document. Passive plant designs, chapters 2-13, project number 669  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) is preparing a compendium of technical requirements, referred to as the {open_quotes}Advanced Light Water Reactor [ALWR] Utility Requirements Document{close_quotes}, that is acceptable to the design of an ALWR power plant. When completed, this document is intended to be a comprehensive statement of utility requirements for the design, construction, and performance of an ALWR power plant for the 1990s and beyond. The Requirements Document consists of three volumes. Volume I, {open_quotes}ALWR Policy and Summary of Top-Tier Requirements{close_quotes}, is a management-level synopsis of the Requirements Document, including the design objectives and philosophy, the overall physical configuration and features of a future nuclear plant design, and the steps necessary to take the proposed ALWR design criteria beyond the conceptual design state to a completed, functioning power plant. Volume II consists of 13 chapters and contains utility design requirements for an evolutionary nuclear power plant [approximately 1350 megawatts-electric (MWe)]. Volume III contains utility design requirements for nuclear plants for which passive features will be used in their designs (approximately 600 MWe). In April 1992, the staff of the Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, issued Volume 1 and Volume 2 (Parts 1 and 2) of its safety evaluation report (SER) to document the results of its review of Volumes 1 and 2 of the Requirements Document. Volume 1, {open_quotes}NRC Review of Electric Power Research Institute`s Advanced Light Water Reactor Utility Requirements Document - Program Summary{close_quotes}, provided a discussion of the overall purpose and scope of the Requirements Document, the background of the staff`s review, the review approach used by the staff, and a summary of the policy and technical issues raised by the staff during its review.

Not Available

1994-08-01T23:59:59.000Z

176

Approved for public release; distribution is unlimited. NAVAL POSTGRADUATE SCHOOL  

E-Print Network (OSTI)

1995 Thesis Advisor: Michael J. Zyda Thesis Co-Advisor: John S. Falby #12;Public reporting burden-18 REPORT DOCUMENTATION PAGE Form Approved OMB No. 0704-0188 1. AGENCY USE ONLY (Leave Blank) 2. REPORT DATEApproved for public release; distribution is unlimited. THESIS NAVAL POSTGRADUATE SCHOOL Monterey

Zyda, Michael

177

Distributed Energy Resources at Naval Base Ventura County Building 1512  

E-Print Network (OSTI)

LBNL-55340 Distributed Energy Resources at Naval Base Ventura County Building 1512 Prepared, of the U.S. Department of Energy under Contract No. DE-AC03-76SF00098. #12;Distributed Energy Resources and the Distributed Energy Program of DOE also provided prior funding to develop and validate the DER-CAM model

178

Estimating the Economic Benefits of Forward-Engaged Naval Forces  

Science Conference Proceedings (OSTI)

In preparing for the 1997 quadrennial defense review, US Navy leaders asked us if we could quantify the economic benefits of forward-engaged naval forces and communicate them to policy makers. Until this point, the only evidence of such benefits was ... Keywords: INDUSTRIES--PETROLEUM-NATURAL GAS, MILITARY--COST EFFECTIVENESS

Robert E. Looney; David A. Schrady; Ronald L. Brown

2001-07-01T23:59:59.000Z

179

Naval Postgraduate School Monterey, California 939435138  

E-Print Network (OSTI)

; Vlosky 2002 #12;s The first MDF plant started up four years ago and now there are 3 or 4, so furniture for exotic wood veneer layup. Pressing veneer on MDF will give the most defect-free, flat and smooth surface

180

THE ORNL GCR-3, A 750-Mw(e) GAS-COOLED CLAD-FUEL REACTOR POWER PLANT. A JOINT DESIGN STUDY  

SciTech Connect

ABS>An advanced, gas-cooled, clad-fuel reactor power plant to generate 750 Mw of electricity was designed as a study of the potential capability of that system. The graphitemoderated reactor generates 1908 Mw of heat in 1062 fuel channels 21 ft long for a power density of 5.5 kw/liter. Gas temperatures entering and leaving the reactor are 574 and 1150 deg F, respectively, operating at 420 psia. Steam at 2415 psia and 950 deg F with reheat to 1000 deg F drives a 763-Mw(e) turbogenerator and also four 31,000-hp blower drive turbines and the boiler feed pumps. Net thermal efficiency of the plant is 39.4%. Estimated direct cost of construction is 0,267,000, or 7 per kilowatt net electric output. Fuel-cycle costs at 20,000 Mwd per metric ton of uranium are 1.46 mills/ kwhr, operating and maintenance costs are 0.39 mill, and fixed charges range from 1.80 to 4.65 mills, depending on method of financing. Total power generation costs at an 80% load factor range from 3.65 to 6.50 mills/kwhr. (auth)

1963-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "naval reactor plants" 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

Measurement of theta13 with reactor neutrinos  

E-Print Network (OSTI)

? detectors ~1 km nuclear reactor Figure 2. Concept of a 2-Measurement of ? 13 with Reactor Neutrinos K.M. Heeger a ,power plant, a future reactor neutrino experiment has the

Heeger, Karsten M.; Freedman, Stuart J.; Kadel, Richard W.; Luk, Kam-Biu

2004-01-01T23:59:59.000Z

182

DOE - Office of Legacy Management -- Naval Research Laboratory - DC 02  

Office of Legacy Management (LM)

Research Laboratory - DC 02 Research Laboratory - DC 02 FUSRAP Considered Sites Site: NAVAL RESEARCH LABORATORY (DC.02 ) Eliminated from consideration under FUSRAP - Referred to DOD Designated Name: Not Designated Alternate Name: None Location: Washington , D.C. DC.02-4 Evaluation Year: 1987 DC.02-4 Site Operations: Research and development on thermal diffusion. DC.02-4 Site Disposition: Eliminated - No Authority - AEC licensed - Military facility DC.02-4 DC.02-1 Radioactive Materials Handled: Yes Primary Radioactive Materials Handled: Uranium, Thorium DC.02-2 DC.02-3 Radiological Survey(s): None Indicated Site Status: Eliminated from consideration under FUSRAP - Referred to DOD DC.02-4 Also see Documents Related to NAVAL RESEARCH LABORATORY DC.02-1 - AEC Memorandum and Source Material License No. C-3393;

183

Naval Construction Battalion Center Gulfport - Mississippi Power Partnership Success Story  

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

Construction Battalion Center Construction Battalion Center Gulfport - Mississippi Power Partnership Success Story Utilities Hardening Project Joe Bosco May 5, 2009 May 5, 2009 * Naval Construction Battalion Center * Established 1942 - Gulfport * Home of Atlantic Fleet Seabees Home of Atlantic Fleet Seabees * Mission: Prepare for & support all facets of the mobilization of construction forces * Naval Construction Battalion Center * 1,100 Acres * 9+ MVA; $3M/yr in Electricity 9+ MVA; $3M/yr in Electricity * One of two Battalion Centers in U.S. * Economic Impact - $500M Mississippi Power Company * Headquartered - Gulfport * Subsidiary y of Southern Comp pany y * Serves 23 counties Southeast Mississippi * 192,000 retail customers * * Generating capacity: 3 166 192 kW Generating capacity: 3,166,192 kW

184

Average effluent releases from U. S. nuclear power reactors, compared with those from fossil-fueled plants, in terms of currently applicable environmental standards  

SciTech Connect

From 3rd international congress of the International Radiation Protection Association meeting; Washington, District of Columbia, USA(9 Sep 1973). Between 1967 and 1972, eighteen second generation'' lightwater-cooled nuclear power plants, with capacities in the range of 500 to 800 MW(e) have been put into operation in the United States. These were in addition to ten smaller demonstration plants and one high-temperature gas-cooled nuclear power plant in operation at the start of this period. The reported yearly air effluent releases of radioactive gases, halogens and particulates, and liquid effluent fission and activation products and of tritium from these plants are evaluated on a Ci/10/sup 3/ MW(e) basis, and the overall yearly averages for the various types of reactors (boiling water (BWR), pressurized water (PWR) and high temperature gas-cooled (HTGR)! are compared. These and the amounts of effluents released from reference 1,000 MW(e) fossil-fueled plants are compared in terms of relative environmental concentrations and their relationship to the applicable U. S. environmental standards for the principal constituents in their respective plant air-effluent streams. 21 references. (auth)

Hull, A.P.

1973-09-19T23:59:59.000Z

185

ARMY GAS-COOLED REACTOR SYSTEMS PROGRAM. INITIAL FULL POWER AND LIMITED ENDURANCE TESTS OF THE ML-1 NUCLEAR POWER PLANT. Final Test Report  

SciTech Connect

The evaluation of the data generated during the full power and limited endurance tests of the ML-1 mobile nuclear power plant indicates that the reactor performs in accordance with the design specifications. During the 101 hr test period, the reactor attained a maximum power of 3.44 Mw( and 247 kw(e) was measured at the output shaft of the turbine-compressor set. No operating limits were exceeded during these tests and all systems performed satisfactorily Except for the known performance deficiency of the turbinecompressor set, which prevented the attainment of design output power, no operational, stability, or control problems were encountered. All test objectives were achieved and the tests were considered completely successful. (auth)

Kattchee, N.

1963-07-01T23:59:59.000Z

186

United States Naval Surface Warfare Center | Open Energy Information  

Open Energy Info (EERE)

Naval Surface Warfare Center Naval Surface Warfare Center Jump to: navigation, search Hydro | Hydrodynamic Testing Facilities Name United States Naval Surface Warfare Center Address Carderock, 9500 MacArthur Boulevard West Place Bethesda, Maryland Zip 20817 Sector Hydro Phone number (301) 227-1574 Website http://www.dt.navy.mil/hyd/fac Coordinates 38.9782231°, -77.1973878° Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"ROADMAP","zoom":14,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"600px","height":"350px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":38.9782231,"lon":-77.1973878,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

187

Naval Petroleum and Oil Shale Reserves annual report of operations for fiscal year 1996  

SciTech Connect

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.

1996-12-31T23:59:59.000Z

188

Underground collocation of nuclear power plant reactors and repository to facilitate the post-renaissance expansion of nuclear power  

SciTech Connect

Underground collocation of nuclear power reactors and the nuclear waste management facilities supporting those reactors, termed an underground nuclear park (UNP), appears to have several advantages compared to the conventional approach to siting reactors and waste management facilities. These advantages include the potential to lower reactor capital and operating cost, lower nuclear waste management cost, and increase margins of physical security and safety. Envirorunental impacts related to worker health, facility accidents, waste transportation, and sabotage and terrorism appear to be lower for UNPs compared to the current approach. In-place decommissioning ofUNP reactors appears to have cost, safety, envirorunental and waste disposal advantages. The UNP approach has the potential to lead to greater public acceptance for the deployment of new power reactors. Use of the UNP during the post-nuclear renaissance time frame has the potential to enable a greater expansion of U.S. nuclear power generation than might otherwise result. Technical and economic aspects of the UNP concept need more study to determine the viability of the concept.

Myers, Carl W [Los Alamos National Laboratory; Elkins, Ned Z [Los Alamos National Laboratory

2008-01-01T23:59:59.000Z

189

NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG 7 JUNE 2013 Quick Facts..............................................................................................................................................................4  

E-Print Network (OSTI)

NAVAL POSTGRADUATE SCHOOL ACADEMIC CATALOG ­ 7 JUNE 2013 1 CONTENTS Quick Facts...............................................................................................................................................435 Energy Core Group .............................................................................................................................................438 Certificate in Defense Energy - Curriculum 234

190

Distributed Energy Resources at Naval Base Ventura County Building 1512: A Sensitivity Analysis  

E-Print Network (OSTI)

Naval Base Ventura County Standby Generator Optimization20 Figure 7: Standby Charge Sensitivity Separate24 Figure 11: Standby Charge Sensitivity Analysis

Bailey, Owen C.; Marnay, Chris

2005-01-01T23:59:59.000Z

191

NUCLEAR REACTORS AND EARTHQUAKES  

SciTech Connect

A book is presented which supplies pertinent seismological information to engineers in the nuclear reactor field. Data are presented on the occurrence, intensity, and wave shapes. Techniques are described for evaluating the response of structures to such events. Certain reactor types and their modes of operation are described briefly. Various protection systems are considered. Earthquake experience in industrial and reactor plants is described. (D.L.C.)

1961-01-01T23:59:59.000Z

192

Review of Experience With the EPRI DFD Process: Decontamination for Decommissioning of Reactor Coolant Systems and Plant Components  

Science Conference Proceedings (OSTI)

This report describes recent experience using the EPRI DFD (decontamination for decommissioning) dilute chemical decontamination process on reactor coolant systems and components. This report updates previous EPRI reports with additional data on materials corrosion and waste management. A wide range of applications show that EPRI DFD is a versatile, cost-effective process for reducing radiation exposures and radioactive waste.

1998-12-31T23:59:59.000Z

193

Production reactor characteristics  

SciTech Connect

Reactors for the production of special nuclear materials share many similarities with commercial nuclear power plants. Each relies on nuclear fission, uses uranium fuel, and produces large quantities of thermal power. However, there are some important differences in production reactor characteristics that may best be discussed in terms of mission, role, and technology.

Thiessen, C.W.; Hootman, H.E.

1990-01-01T23:59:59.000Z

194

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine.

Isaksson, Juhani (Karhula, FI)

1996-01-01T23:59:59.000Z

195

Pressurized fluidized bed reactor  

DOE Patents (OSTI)

A pressurized fluid bed reactor power plant includes a fluidized bed reactor contained within a pressure vessel with a pressurized gas volume between the reactor and the vessel. A first conduit supplies primary gas from the gas volume to the reactor, passing outside the pressure vessel and then returning through the pressure vessel to the reactor, and pressurized gas is supplied from a compressor through a second conduit to the gas volume. A third conduit, comprising a hot gas discharge, carries gases from the reactor, through a filter, and ultimately to a turbine. During normal operation of the plant, pressurized gas is withdrawn from the gas volume through the first conduit and introduced into the reactor at a substantially continuously controlled rate as the primary gas to the reactor. In response to an operational disturbance of the plant, the flow of gas in the first, second, and third conduits is terminated, and thereafter the pressure in the gas volume and in the reactor is substantially simultaneously reduced by opening pressure relief valves in the first and third conduits, and optionally by passing air directly from the second conduit to the turbine. 1 fig.

Isaksson, J.

1996-03-19T23:59:59.000Z

196

Naval Petroleum and Oil Shale Reserves. Annual report of operations, Fiscal year 1992  

SciTech Connect

During fiscal year 1992, the reserves generated $473 million in revenues, a $181 million decrease from the fiscal year 1991 revenues, primarily due to significant decreases in oil and natural gas prices. Total costs were $200 million, resulting in net cash flow of $273 million, compared with $454 million in fiscal year 1991. From 1976 through fiscal year 1992, the Naval Petroleum and Oil Shale Reserves generated more than $15 billion in revenues and a net operating income after costs of $12.5 billion. In fiscal year 1992, production at the Naval Petroleum Reserves at maximum efficient rates yielded 26 million barrels of crude oil, 119 billion cubic feet of natural gas, and 164 million gallons of natural gas liquids. From April to November 1992, senior managers from the Naval Petroleum and Oil Shale Reserves held a series of three workshops in Boulder, Colorado, in order to build a comprehensive Strategic Plan as required by Secretary of Energy Notice 25A-91. Other highlights are presented for the following: Naval Petroleum Reserve No. 1--production achievements, crude oil shipments to the strategic petroleum reserve, horizontal drilling, shallow oil zone gas injection project, environment and safety, and vanpool program; Naval Petroleum Reserve No. 2--new management and operating contractor and exploration drilling; Naval Petroleum Reserve No. 3--steamflood; Naval Oil Shale Reserves--protection program; and Tiger Team environmental assessment of the Naval Petroleum and Oil Shale Reserves in Colorado, Utah, and Wyoming.

1992-12-31T23:59:59.000Z

197

EA-0531: Proposed Natural Gas Protection Program for Naval Oil Shale  

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

31: Proposed Natural Gas Protection Program for Naval Oil 31: Proposed Natural Gas Protection Program for Naval Oil Shale Reserves Nos. 1 and 3, Garfield County, Colorado EA-0531: Proposed Natural Gas Protection Program for Naval Oil Shale Reserves Nos. 1 and 3, Garfield County, Colorado SUMMARY This EA evaluates the environmental impacts of a proposal for a Natural Gas Protection Program for Naval Oil Shale Reserves Nos. 1 and 3 which would be implemented over a five-year period that would encompass a total of 200 wells in Garfield County, Colorado. PUBLIC COMMENT OPPORTUNITIES None available at this time. DOCUMENTS AVAILABLE FOR DOWNLOAD August 9, 1991 EA-0531: Final Environmental Assessment Proposed Natural Gas Protection Program for Naval Oil Shale Reserves Nos. 1 and 3 August 9, 1991 EA-0531: Finding of No Significant Impact

198

Fallon Naval Air Station Geothermal Area | Open Energy Information  

Open Energy Info (EERE)

Page Page Edit with form History Facebook icon Twitter icon » Fallon Naval Air Station Geothermal Area Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Geothermal Resource Area: Fallon Naval Air Station Geothermal Area Contents 1 Area Overview 2 History and Infrastructure 3 Regulatory and Environmental Issues 4 Exploration History 5 Well Field Description 6 Geology of the Area 7 Geofluid Geochemistry 8 NEPA-Related Analyses (0) 9 Exploration Activities (0) 10 References Loading map... {"minzoom":false,"mappingservice":"googlemaps3","type":"TERRAIN","zoom":6,"types":["ROADMAP","SATELLITE","HYBRID","TERRAIN"],"geoservice":"google","maxzoom":false,"width":"500px","height":"300px","centre":false,"title":"","label":"","icon":"","visitedicon":"","lines":[],"polygons":[],"circles":[],"rectangles":[],"copycoords":false,"static":false,"wmsoverlay":"","layers":[],"controls":["pan","zoom","type","scale","streetview"],"zoomstyle":"DEFAULT","typestyle":"DEFAULT","autoinfowindows":false,"kml":[],"gkml":[],"fusiontables":[],"resizable":false,"tilt":0,"kmlrezoom":false,"poi":true,"imageoverlays":[],"markercluster":false,"searchmarkers":"","locations":[{"text":"","title":"","link":null,"lat":39.38,"lon":-118.65,"alt":0,"address":"","icon":"","group":"","inlineLabel":"","visitedicon":""}]}

199

Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750800C Reactor Outlet Temperature  

Science Conference Proceedings (OSTI)

This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750800C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

John Collins

2009-08-01T23:59:59.000Z

200

Materials Reliability Program: Sensitivity Studies for Functionality Analysis of Reactor Internals of Westinghouse-Designed Power Plants (MRP-356)  

Science Conference Proceedings (OSTI)

In developing the Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Rev. 0), a semi-empirical material behavior model for irradiated austenitic stainless steels was developed for the degradation mechanisms applicable to the internals. Functionality analyses predicting aging up to 60 years were performed using the material model. MRP-230 results indicate that susceptibility to irradiation-assisted stress corrosion cracking (IASCC) is governed by two competing ...

2013-08-13T23:59:59.000Z

Note: This page contains sample records for the topic "naval reactor plants" 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

Plant Design and Cost Assessment of Forced Circulation Lead-Bismuth Cooled Reactor with Conventional Power Conversion Cycles  

E-Print Network (OSTI)

Cost of electricity is the key factor that determines competitiveness of a power plant. Thus the proper selection, design and optimization of the electric power generating cycle is of main importance. This report makes an ...

Dostal, Vaclav

202

Reactor Thermal-Hydraulics  

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

Thermal-Hydraulics Thermal-Hydraulics Dr. Tanju Sofu, Argonne National Laboratory In a power reactor, the energy produced in fission reaction manifests itself as heat to be removed by a coolant and utilized in a thermodynamic energy conversion cycle to produce electricity. A simplified schematic of a typical nuclear power plant is shown in the diagram below. Primary coolant loop Steam Reactor Heat exchanger Primary pump Secondary pump Condenser Turbine Water Although this process is essentially the same as in any other steam plant configuration, the power density in a nuclear reactor core is typically four orders of magnitude higher than a fossil fueled plant and therefore it poses significant heat transfer challenges. Maximum power that can be obtained from a nuclear reactor is often limited by the

203

Commander, Naval Base ATTN: Ms. Cheryl Barnett Building N-26  

Office of Legacy Management (LM)

.J>?j 1.2 1990 .J>?j 1.2 1990 Commander, Naval Base ATTN: Ms. Cheryl Barnett Building N-26 Code N 9 E Norfolk, Virginia 23511-6002 Dear Ms. Barnett: I enjoyed speaking with you on the phone. The Department of Energy (DOE) has established its Formerly Utilized Sites Remedial Action Program (FUSRAP) to identify sites formerly utilized by its predecessor agencies in the early days of the nation's atomic energy program and to determine the potential for these sites to contain radiological contamination, related to DOE's past activities, which may require remedial action. When necessary, radiological surveys of individual sites are performed to provide the data necessary to make this necessary determination. As we discussed, in July 1956, the Atomic Energy Commission (a DOE

204

Light Water Reactor Sustainability (LWRS) Initiative Science-Based R&D to Extend Nuclear Plant Operation  

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

Nuclear Energy Nuclear Energy Updates Dr. Pete Lyons Acting Assistant Secretary for Nuclear Energy U.S. Department of Energy December 9, 2010 NEAC Meeting Leadership Changes Pete Miller retired Pete Lyons - Acting NE-1 Shane Johnson - Acting NE-2 Dennis Miotla - Acting COO Monica Regalbuto - Acting DAS for Fuel Cycle Technologies John Herczeg- Acting ADAS for Fuel Cycle Technologies John Kelly - DAS for Nuclear Reactor Technologies Bob Boudreau- Acting ADAS International Nuclear Energy Coop Monica Regalbuto John Kelly NE University Programs (NEUP) - Overview and FY 2011 Schedule NEUP FY 2011 Solicitations Schedule RPA/FOA Pre- Applications Proposals Due Awards Announced R&D (PS and Blue Sky) Oct. '10 Dec. '10 Feb. '11 May '11 Integrated Research Projects (IRP) Dec. '10 Late Jan '11

205

Light Water Reactor Sustainability (LWRS) Initiative Science-Based R&D to Extend Nuclear Plant Operation  

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

9, 2010 9, 2010 New Program Proposal for Fiscal Year 2011 - Modified Open Cycle Carter "Buzz" Savage Nuclear Energy Advisory Committee Meeting April 29, 2010 Washington, DC April 29, 2010 Recycle of Used Fuel Option to recycle used fuel has been the subject of much debate and discussion. Nonproliferation issues and economics have limited recycle options. Recycle of used fuel enables increased utilization of uranium resource and potential waste management benefits. - Once through fuel cycle uses less than 1% of energy value of the uranium. Courtesy AREVA 2 April 29, 2010 Summary of Fuel Cycle Options 3 Once-Through Fuel Cycle - One pass through reactor, used fuel directly disposed in a geologic repository. Modified Open Cycle - No or limited separations steps and

206

Lead-Cooled Fast Reactor (LFR) Design: Safety, Neutronics, Thermal Hydraulics, Structural Mechanics, Fuel, Core, and Plant Design  

Science Conference Proceedings (OSTI)

The idea of developing fast spectrum reactors with molten lead (or lead alloy) as a coolant is not a new one. Although initially considered in the West in the 1950s, such technology was not pursued to completion because of anticipated difficulties associated with the corrosive nature of these coolant materials. However, in the Soviet Union, such technology was actively pursued during the same time frame (1950s through the 1980s) for the specialized role of submarine propulsion. More recently, there has been a renewal of interest in the West for such technology, both for critical systems as well as for Accelerator Driven Subcritical (ADS) systems. Meanwhile, interest in the former Soviet Union, primarily Russia, has remained strong and has expanded well beyond the original limited mission of submarine propulsion. This section reviews the past and current status of LFR development.

Smith, C

2010-02-22T23:59:59.000Z

207

Slurry reactor design studies  

SciTech Connect

The objective of these studies was to perform a realistic evaluation of the relative costs of tublar-fixed-bed and slurry reactors for methanol, mixed alcohols and Fischer-Tropsch syntheses under conditions where they would realistically be expected to operate. The slurry Fischer-Tropsch reactor was, therefore, operated at low H{sub 2}/CO ratio on gas directly from a Shell gasifier. The fixed-bed reactor was operated on 2.0 H{sub 2}/CO ratio gas after adjustment by shift and CO{sub 2} removal. Every attempt was made to give each reactor the benefit of its optimum design condition and correlations were developed to extend the models beyond the range of the experimental pilot plant data. For the methanol design, comparisons were made for a recycle plant with high methanol yield, this being the standard design condition. It is recognized that this is not necessarily the optimum application for the slurry reactor, which is being proposed for a once-through operation, coproducing methanol and power. Consideration is also given to the applicability of the slurry reactor to mixed alcohols, based on conditions provided by Lurgi for an Octamix{trademark} plant using their standard tubular-fixed reactor technology. 7 figs., 26 tabs.

Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

1990-06-01T23:59:59.000Z

208

Georgia Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

209

Arkansas Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

210

Iowa Nuclear Profile - Power Plants  

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

Iowa nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

211

Ohio Nuclear Profile - Power Plants  

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

Ohio nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

212

Vermont Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

213

Florida Nuclear Profile - Power Plants  

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

Florida nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

214

Virginia Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

215

Washington Nuclear Profile - Power Plants  

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

Washington nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

216

Missouri Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

217

Nebraska Nuclear Profile - Power Plants  

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

Nebraska nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

218

Tennessee Nuclear Profile - Power Plants  

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

Tennessee nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

219

Connecticut Nuclear Profile - Power Plants  

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

Connecticut nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

220

Minnesota Nuclear Profile - Power Plants  

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

Minnesota nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

Note: This page contains sample records for the topic "naval reactor plants" 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

California Nuclear Profile - Power Plants  

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

California nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

222

Arizona Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

223

Massachusetts Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

224

Kansas Nuclear Profile - Power Plants  

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

Kansas nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

225

Alabama Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

226

Wisconsin Nuclear Profile - Power Plants  

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

Wisconsin nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

227

Texas Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

228

Michigan Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

229

Mississippi Nuclear Profile - Power Plants  

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

Mississippi nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State...

230

Advanced Reactor Technologies | Department of Energy  

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

Advanced Reactor Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies Advanced Reactor Technologies The Office of Advanced Reactor Technologies (ART) sponsors research, development and deployment (RD&D) activities through its Next Generation Nuclear Plant (NGNP), Advanced Reactor Concepts (ARC), and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. The Office of Nuclear Energy (NE) will pursue these advancements through RD&D activities at the Department of Energy (DOE) national laboratories and U.S. universities, as well as through collaboration with industry and international partners. These activities will focus on advancing scientific

231

22.312 Engineering of Nuclear Reactors, Fall 2004  

E-Print Network (OSTI)

Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

Buongiorno, Jacopo, 1971-

232

22.312 Engineering of Nuclear Reactors, Fall 2002  

E-Print Network (OSTI)

Engineering principles of nuclear reactors, emphasizing power reactors. Power plant thermodynamics, reactor heat generation and removal (single-phase as well as two-phase coolant flow and heat transfer), and structural ...

Todreas, Neil E.

233

Extension of the supercritical carbon dioxide brayton cycle to low reactor power operation: investigations using the coupled anl plant dynamics code-SAS4A/SASSYS-1 liquid metal reactor code system.  

SciTech Connect

Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO{sub 2} cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO{sub 2}. It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO{sub 2} heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO{sub 2}-to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO{sub 2} turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the calculations reveal that the compressor conditions are calculated to approach surge such that the need for a surge control system for each compressor is identified. Thus, it is demonstrated that the S-CO{sub 2} cycle can operate in the initial decay heat removal mode even with autonomous reactor control. Because external power is not needed to drive the compressors, the results show that the S-CO{sub 2} cycle can be used for initial decay heat removal for a lengthy interval in time in the absence of any off-site electrical power. The turbine provides sufficient power to drive the compressors. Combined with autonomous reactor control, this represents a significant safety advantage of the S-CO{sub 2} cycle by maintaining removal of the reactor power until the core decay heat falls to levels well below those for which the passive decay heat removal system is designed. The new control strategy is an alternative to a split-shaft layout involving separate power and compressor turbines which had previously been identified as a promising approach enabling heat removal from a SFR at low power levels. The current results indicate that the split-shaft configuration does not provide any significant benefits for the S-CO{sub 2} cycle over the current single-shaft layout with shaft speed control. It has been demonstrated that when connected to the grid the single-shaft cycle can effectively follow the load over the entire range. No compressor speed variation is needed while power is delivered to the grid. When the system is disconnected from the grid, the shaft speed can be changed as effectively as it would be with the split-shaft arrangement. In the split-shaft configuration, zero generator power means disconnection of the power turbine, such that the resulting system will be almost identical to the single-shaft arrangement. Without this advantage of the split-shaft configuration, the economic benefits of the single-shaft arrangement, provided by just one turbine and lower losses at the design point, are more important to the overall cycle performance. Therefore, the single-shaft

Moisseytsev, A.; Sienicki, J. J. (Nuclear Engineering Division)

2012-05-10T23:59:59.000Z

234

Management of the aging of critical safety-related concrete structures in light-water reactor plants  

SciTech Connect

The Structural Aging Program has the overall objective of providing the USNRC with an improved basis for evaluating nuclear power plant safety-related structures for continued service. The program consists of a management task and three technical tasks: materials property data base, structural component assessment/repair technology, and quantitative methodology for continued-service determinations. Objectives, accomplishments, and planned activities under each of these tasks are presented. Major program accomplishments include development of a materials property data base for structural materials as well as an aging assessment methodology for concrete structures in nuclear power plants. Furthermore, a review and assessment of inservice inspection techniques for concrete materials and structures has been complete, and work on development of a methodology which can be used for performing current as well as reliability-based future condition assessment of concrete structures is well under way. 43 refs., 3 tabs.

Naus, D.J.; Oland, C.B. (Oak Ridge National Lab., TN (USA)); Arndt, E.G. (Nuclear Regulatory Commission, Washington, DC (USA))

1990-01-01T23:59:59.000Z

235

EA-1236: Preparation for Transfer of Ownership of Naval Petroleum Reserve  

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

236: Preparation for Transfer of Ownership of Naval Petroleum 236: Preparation for Transfer of Ownership of Naval Petroleum Reserve No. 3, Natrona County, WY EA-1236: Preparation for Transfer of Ownership of Naval Petroleum Reserve No. 3, Natrona County, WY SUMMARY This EA evaluates activities that DOE would conduct in anticipation of possible transfer of Naval Petroleum Reserve No. 3 (NPR-3) out of Federal operation. Proposed activities would include accelerated plugging and abandoning of uneconomic wells, complete reclamation and restoration of abandoned sites including dismantling surface facilities, batteries, roads, test satellites, electrical distribution systems and associated power poles, when they are no longer needed for production, and the development of the Rocky Mountain Oilfield Testing Center (RMOTC).

236

Development Wells At Fallon Naval Air Station Area (Sabin, Et Al., 2010) |  

Open Energy Info (EERE)

Naval Air Station Area (Sabin, Et Al., 2010) Naval Air Station Area (Sabin, Et Al., 2010) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Development Wells At Fallon Naval Air Station Area (Sabin, Et Al., 2010) Exploration Activity Details Location Fallon Naval Air Station Area Exploration Technique Development Wells Activity Date Usefulness not indicated DOE-funding Unknown Notes As was mentioned previously, the Navy signed a development contract with Ormat in 2005 to produce power from a potential resource on the SE corner of the main side portion of NAS Fallon. Additionally the GPO began additional exploration activities on the Bombing Range 16 in collaboration with the Great Basin Center for Geothermal Energy. The introduction of $9.1M of Recovery Act funds in early 2009 led to a broadening as well as an

237

Sale of the Elk Hills Naval Petroleum Reserve | Department of Energy  

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

Services » Petroleum Reserves » Naval Reserves » Sale of the Elk Services » Petroleum Reserves » Naval Reserves » Sale of the Elk Hills Naval Petroleum Reserve Sale of the Elk Hills Naval Petroleum Reserve Energy Secretary Federico Pena (left) and Occidental Petroleum's David Hentschel sign the historic transfer agreement with Patricia Godley, DOE's Assistant Secretary for Fossil Energy, who orchestrated the sale, looking on. Energy Secretary Federico Pena (left) and Occidental Petroleum's David Hentschel sign the historic transfer agreement with Patricia Godley, DOE's Assistant Secretary for Fossil Energy, who orchestrated the sale, looking on. On February 5, 1998, the Department of Energy and Occidental Petroleum Corporation concluded the largest divestiture of federal property in the history of the U.S. government.

238

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

239

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

240

EA-1008: Continued Development of Naval Petroleum Reserve No. 3 (Sitewide), Natrona County, Wyoming  

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

This EA evaluates the environmental impacts of the proposal to continue development of the U.S. Department of Energy's Naval Petroleum Reserve No. 3 located in Natrona County, Wyoming over the next...

Note: This page contains sample records for the topic "naval reactor plants" 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

Development of A Mesoscale Ensemble Data Assimilation System at The Naval Research Laboratory  

Science Conference Proceedings (OSTI)

An ensemble Kalman filter (EnKF) has been adopted and implemented at the Naval Research Laboratory (NRL) for mesoscale and storm-scale data assimilation to study the impact of ensemble assimilation of high-resolution observations, including those ...

Qingyun Zhao; Fuqing Zhang; Teddy Holt; Craig H. Bishop; Qin Xu

242

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

243

Analyzing the effects of component reliability on naval Integrated Power System quality of service  

E-Print Network (OSTI)

The Integrated Power System (IPS) is a key enabling technology for future naval vessels and their advanced weapon systems. While conventional warship designs utilize separate power systems for propulsion and shipboard ...

Hawbaker, Benjamin F. (Benjamin Forrest)

2008-01-01T23:59:59.000Z

244

Advanced Nuclear Reactors | Department of Energy  

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

Advanced Nuclear Advanced Nuclear Reactors Advanced Nuclear Reactors Turbulent Flow of Coolant in an Advanced Nuclear Reactor Visualizing Coolant Flow in Sodium Reactor Subassemblies Sodium-cooled Fast Reactor (SFR) Coolant Flow At the heart of a nuclear power plant is the reactor. The fuel assembly is placed inside a reactor vessel where all the nuclear reactions occur to produce the heat and steam used for power generation. Nonetheless, an entire power plant consists of many other support components and key structures like coolant pipes; pumps and tanks including their surrounding steel framing; and concrete containment and support structures. The Reactors Product Line within NEAMS is concerned with modeling the reactor vessel as well as those components of a complete power plant that

245

Southwest Division, Naval Facilities Engineering Command, Demand Side Management Program Implementation  

E-Print Network (OSTI)

This paper covers some of the major aspects of the development and execution of the Southwest Division, Naval Facilities Engineering Command (SOUTHWESTNAVFACENGCOM) Energy and Water Program. The program covers Naval and Marine facilities in 14 western states. It started from zero in 1992 and has grown to a program which has identified and is in the process of implementing energy and water savings projects totaling over $115,000,000.

Gates, G. G.

1997-04-01T23:59:59.000Z

246

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations, which govern the operation of commercial nuclear power plants, require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including

247

U.S. Department of Energy Naval Reactors Laboratory Field Office  

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

use soil cleanup objectives. The ICM will be conducted by a DOE-EM SPRU contractor (URS). URS work will entail the preparation of an excavation plan in accordance with the ICM...

248

Nuclear Reactors  

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

Reactors Nuclear reactors created not only large amounts of plutonium needed for the weapons programs, but a variety of other interesting and useful radioisotopes. They produced...

249

NALDA (Naval Aviation Logistics Data Analysis) CAI (computer aided instruction)  

SciTech Connect

Data Systems Engineering Organization (DSEO) personnel developed a prototype computer aided instruction CAI system for the Naval Aviation Logistics Data Analysis (NALDA) system. The objective of this project was to provide a CAI prototype that could be used as an enhancement to existing NALDA training. The CAI prototype project was performed in phases. The task undertaken in Phase I was to analyze the problem and the alternative solutions and to develop a set of recommendations on how best to proceed. The findings from Phase I are documented in Recommended CAI Approach for the NALDA System (Duncan et al., 1987). In Phase II, a structured design and specifications were developed, and a prototype CAI system was created. A report, NALDA CAI Prototype: Phase II Final Report, was written to record the findings and results of Phase II. NALDA CAI: Recommendations for an Advanced Instructional Model, is comprised of related papers encompassing research on computer aided instruction CAI, newly developing training technologies, instructional systems development, and an Advanced Instructional Model. These topics were selected because of their relevancy to the CAI needs of NALDA. These papers provide general background information on various aspects of CAI and give a broad overview of new technologies and their impact on the future design and development of training programs. The paper within have been index separately elsewhere.

Handler, B.H. (Oak Ridge K-25 Site, TN (USA)); France, P.A.; Frey, S.C.; Gaubas, N.F.; Hyland, K.J.; Lindsey, A.M.; Manley, D.O. (Oak Ridge Associated Universities, Inc., TN (USA)); Hunnum, W.H. (North Carolina Univ., Chapel Hill, NC (USA)); Smith, D.L. (Memphis State Univ., TN (USA))

1990-07-01T23:59:59.000Z

250

Renewable Energy Optimization Report for Naval Station Newport  

DOE Green Energy (OSTI)

In 2008, the U.S. Environmental Protection Agency (EPA) launched the RE-Powering America's Land initiative to encourage the development of renewable energy (RE) on potentially contaminated land and mine sites. As part of this effort, EPA is collaborating with the U.S. Department of Energy's (DOE's) National Renewable Energy Laboratory (NREL) to evaluate RE options at Naval Station (NAVSTA) Newport in Newport, Rhode Island. NREL's Renewable Energy Optimization (REO) tool was utilized to identify RE technologies that present the best opportunity for life-cycle cost-effective implementation while also serving to reduce energy-related carbon dioxide emissions and increase the percentage of RE used at NAVSTA Newport. The technologies included in REO are daylighting, wind, solar ventilation preheating (SVP), solar water heating, photovoltaics (PV), solar thermal (heating and electric), and biomass (gasification and cogeneration). The optimal mix of RE technologies depends on several factors including RE resources; technology cost and performance; state, utility, and federal incentives; and economic parameters (discount and inflation rates). Each of these factors was considered in this analysis. Technologies not included in REO that were investigated separately per NAVSTA Newport request include biofuels from algae, tidal power, and ground source heat pumps (GSHP).

Robichaud, R.; Mosey, G.; Olis, D.

2012-02-01T23:59:59.000Z

251

Naval Undersea Warfare Center Division Newport utilities metering, Phase 1  

SciTech Connect

Pacific Northwest Laboratory developed this report for the US Navy`s Naval Undersea Warfare Center Division Newport, Rhode Island (NUWC). The purpose of the report was to review options for metering electricity and steam used in the NUWC compound, and to make recommendations to NUWC for implementation under a follow-on project. An additional NUWC concern is a proposed rate change by the servicing utility, Newport Electric, which would make a significant shift from consumption to demand billing, and what effect that rate change would have on the NUWC utility budget. Automated, remote reading meters are available which would allow NUWC to monitor its actual utility consumption and demand for both the entire NUWC compound and by end-use in individual buildings. Technology is available to perform the meter reads and manipulate the data using a personal computer with minimal staff requirement. This is not meant to mislead the reader into assuming that there is no requirement for routine preventive maintenance. All equipment requires routine maintenance to maintain its accuracy. While PNL reviewed the data collected during the site visit, however, it became obvious that significant opportunities exist for reducing the utility costs other than accounting for actual consumption and demand. Unit costs for both steam and electricity are unnecessarily high, and options are presented in this report for reducing them. Additionally, NUWC has an opportunity to undertake a comprehensive energy resource management program to significantly reduce its energy demand, consumption, and costs.

Carroll, D.M.

1992-11-01T23:59:59.000Z

252

Naval Undersea Warfare Center Division Newport utilities metering, Phase 1  

SciTech Connect

Pacific Northwest Laboratory developed this report for the US Navy's Naval Undersea Warfare Center Division Newport, Rhode Island (NUWC). The purpose of the report was to review options for metering electricity and steam used in the NUWC compound, and to make recommendations to NUWC for implementation under a follow-on project. An additional NUWC concern is a proposed rate change by the servicing utility, Newport Electric, which would make a significant shift from consumption to demand billing, and what effect that rate change would have on the NUWC utility budget. Automated, remote reading meters are available which would allow NUWC to monitor its actual utility consumption and demand for both the entire NUWC compound and by end-use in individual buildings. Technology is available to perform the meter reads and manipulate the data using a personal computer with minimal staff requirement. This is not meant to mislead the reader into assuming that there is no requirement for routine preventive maintenance. All equipment requires routine maintenance to maintain its accuracy. While PNL reviewed the data collected during the site visit, however, it became obvious that significant opportunities exist for reducing the utility costs other than accounting for actual consumption and demand. Unit costs for both steam and electricity are unnecessarily high, and options are presented in this report for reducing them. Additionally, NUWC has an opportunity to undertake a comprehensive energy resource management program to significantly reduce its energy demand, consumption, and costs.

Carroll, D.M.

1992-11-01T23:59:59.000Z

253

Multilayer ALD Coating of Light Water Reactor Zirconium Alloy ...  

Science Conference Proceedings (OSTI)

Abstract Scope, The accident at Fukushima Daiichi nuclear power plant raised concerns about nuclear reactors safety. The plant experienced an accident in...

254

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A power plant is described comprising a turbine and employing round cylindrical fuel rods formed of BeO and UO/sub 2/ and stacks of hexagonal moderator blocks of BeO provided with passages that loosely receive the fuel rods so that coolant may flow through the passages over the fuels to remove heat. The coolant may be helium or steam and fiows through at least one more heat exchanger for producing vapor from a body of fluid separate from the coolant, which fluid is to drive the turbine for generating electricity. By this arrangement the turbine and directly associated parts are free of particles and radiations emanating from the reactor. (AEC)

Daniels, F.

1962-12-18T23:59:59.000Z

255

Fossil fuel furnace reactor  

DOE Patents (OSTI)

A fossil fuel furnace reactor is provided for simulating a continuous processing plant with a batch reactor. An internal reaction vessel contains a batch of shale oil, with the vessel having a relatively thin wall thickness for a heat transfer rate effective to simulate a process temperature history in the selected continuous processing plant. A heater jacket is disposed about the reactor vessel and defines a number of independent controllable temperature zones axially spaced along the reaction vessel. Each temperature zone can be energized to simulate a time-temperature history of process material through the continuous plant. A pressure vessel contains both the heater jacket and the reaction vessel at an operating pressure functionally selected to simulate the continuous processing plant. The process yield from the oil shale may be used as feedback information to software simulating operation of the continuous plant to provide operating parameters, i.e., temperature profiles, ambient atmosphere, operating pressure, material feed rates, etc., for simulation in the batch reactor.

Parkinson, William J. (Los Alamos, NM)

1987-01-01T23:59:59.000Z

256

Reactor physics design of supercritical CO?-cooled fast reactors  

E-Print Network (OSTI)

Gas-Cooled Fast Reactors (GFRs) are among the GEN-IV designs proposed for future deployment. Driven by anticipated plant cost reduction, the use of supercritical CO? (S-CO?) as a Brayton cycle working fluid in a direct ...

Pope, Michael A. (Michael Alexander)

2004-01-01T23:59:59.000Z

257

Carl Vinson and pre-war naval legislation 1932-1940  

E-Print Network (OSTI)

At the end of World War I, the United States Navy was in the midst of a building program designed to make it a "Navy Second to None." However, the post-war desire to avoid involvement in another international conflict led the United States to retreat from the aggressive naval policy of President Woodrow Wilson. America initiated and supported conferences which led to the Washington Naval Treaty of 1922 and the London Naval Treaty of 1930, both of which limited the maximum strength of the world's naval powers. In addition, throughout the 1920's and early 1930's the United States failed to build up to the levels allowed by the treaties, causing the Navy to slowly age toward obsolescence. This situation changed in the period from 1933 to 1940 as Congressman Carl Vinson of Georgia, chairman of the House Naval Affairs Committee, drafted and saw to the passage of legislation designed to once again provide the United States with a powerful navy as part of a strong national defense. Vinson obtained passage of naval construction legislation in 1934, 1937, 1938, and 1940, and supported the annual and supplemental appropriation acts which funded the newly authorized construction. In doing so, Vinson overcame opposition from pacifists and isolationists in Congress who feared America would be drawn into another European war. Vinson also had to persuade fiscal conservatives opposed to increasing the budget deficit that funding naval construction was in the nation's best interests. Additionally, Vinson obtained the often reluctant support of President Franklin Roosevelt to insure final approval of his legislation. This study examines Vinson's efforts to provide the Navy the ships with which it eventually fought World War II. it looks at committee hearings, 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 construction legislation, Carl Vinson was the driving force behind the expansion of the Navy. Without Vinson's pre-war legislation, the Navy could not have been prepared to effectively fight the Second World War.

Svonavec, Stephen Charles

1994-01-01T23:59:59.000Z

258

Reactor operation environmental information document  

Science Conference Proceedings (OSTI)

This volume is a reactor operation environmental information document for the Savannah River Plant. Topics include meteorology, surface hydrology, transport, environmental impacts, and radiation effects. 48 figs., 56 tabs. (KD)

Bauer, L.R.; Hayes, D.W.; Hunter, C.H.; Marter, W.L.; Moyer, R.A.

1989-12-01T23:59:59.000Z

259

This document was downloaded on May 22, 2013 at 14:23:58 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

Selection Procedures 20 Naval Officers 20 Other U.S. Military Officers 20 International Students 20 Civilian

260

ch_10  

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

S1W Naval Nuclear Reactor Prototype Project Officer 29 years; including experience in Nuclear Power Plant Operations and maintenance, radioactive and hazardous materials...

Note: This page contains sample records for the topic "naval reactor plants" 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

This document was downloaded on May 22, 2013 at 14:29:51 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

- graduate School a naval university, un- ified in policy, procedure and purpose. In addition to its Naval PROCEDURES U.S. NAVAL OFFICERS U.S. Navy officers interested in admis- sion to one of the curricula offered in the submission. Requests for admission or questions regarding admission procedures should be directed to the Dean

262

This document was downloaded on May 22, 2013 at 14:28:55 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

, procedures and objectives. In 1973, the Naval Postgraduate School, together with the Naval War College. ADMISSIONS PROCEDURES U.S. NAVAL OFFICERS U.S. Navy officers interested in ad- mission to one- ity of quotas assigned to each country. The procedures for application are con- tained in OPNAV

263

This document was downloaded on May 22, 2013 at 14:29:10 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

, procedures and objectives. In 1973, the Naval Postgraduate School, together with the Naval War College or local bookstores, or from other students. ADMISSIONS PROCEDURES U.S. NAVAL OFFICERS U.S. Navy officers. Their admission is subject to availabil- ity of quotas assigned to each country. The procedures for application

264

This document was downloaded on May 22, 2013 at 14:28:34 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

, procedures and objectives. In 1973, the Naval Postgraduate School, together with the Naval War College Navy West Coast Match racing cham- pionships. ADMISSIONS PROCEDURES U.S. NAVAL OFFICERS U.S. Navy, and procedural guidance for the Navy's graduate education program. Included is material concerning officer

265

This document was downloaded on May 22, 2013 at 14:29:21 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

, procedures and objectives. In 1973, the Naval Postgraduate School, together with the Naval War College MILITARY OFFICERS ADMISSIONS PROCEDURES U.S. NAVAL OFFICERS U.S. Navy officers interested in ad- Military- ity of quotas assigned to each country. The procedures for application are con- 8 #12;GENERAL

266

This document was downloaded on May 22, 2013 at 14:28:45 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

, procedures and objectives. In 1973, the Naval Postgraduate School, together with the Naval War College- change system and stocks all required supplies. ADMISSIONS PROCEDURES U.S. NAVAL OFFICERS ACADEMIC of quotas assigned to each country. The procedures for application are con- 10 #12;GENERAL INFORMATION

267

This document was downloaded on May 22, 2013 at 14:29:42 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

the Naval Postgraduate School, in effect, a naval university, unified in policies, procedures and objec bookstores, or from other students. ADMISSIONS PROCEDURES U.S. NAVAL OFFICERS U. S. Navy officers interested. The procedures for application are contained in OPNAV INSTRUCTION 4950. IE. Corre- spondence must be processed

268

This document was downloaded on May 22, 2013 at 14:28:23 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

, procedures and objectives. In 1973, the Naval Postgraduate School, together with the Naval War College Navy West Coast Match racing cham- pionships. ADMISSIONS PROCEDURES U.S. NAVAL OFFICERS U.S. Navy, policy, and procedural guidance for the Navy's graduate education program. Included is material

269

NUCLEAR REACTOR  

DOE Patents (OSTI)

A boiling-water nuclear reactor is described wherein control is effected by varying the moderator-to-fuel ratio in the reactor core. This is accomplished by providing control tubes containing a liquid control moderator in the reactor core and providing means for varying the amount of control moderatcr within the control tubes.

Treshow, M.

1961-09-01T23:59:59.000Z

270

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor in which at least a portion of the moderator is in the form of movable refractory balls is described. In addition to their moderating capacity, these balls may serve as carriers for fissionable material or fertile material, or may serve in a coolant capacity to remove heat from the reactor. A pneumatic system is used to circulate the balls through the reactor.

Daniels, F.

1959-10-27T23:59:59.000Z

271

North Carolina Nuclear Profile - Power Plants  

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

Carolina nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

272

New Hampshire Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (nw)","Net generation (thousand mwh)","Share of State nuclear net...

273

New Jersey Nuclear Profile - Power Plants  

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

nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net...

274

Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors  

SciTech Connect

The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors.

1978-09-01T23:59:59.000Z

275

Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.  

SciTech Connect

STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal resistance of a gas-filled gap.

Moisseytsev, A.; Sienicki, J. J.

2007-03-08T23:59:59.000Z

276

Rancho Seco Reactor Vessel Segmentation Experience Report  

Science Conference Proceedings (OSTI)

This report documents the approach taken by Sacramento Municipal Utility District (SMUD) in the segmentation and disposal of the Reactor Vessel from the Rancho Seco Nuclear Generating Station (RSNGS). The location of the Rancho Seco plant placed major constraints on the shipping options available for large plant components (Steam Generators and Reactor Vessel). This report details the engineering evaluations leading to the segmentation and disposal of the Reactor Vessel (RV). It describes the key element...

2008-03-18T23:59:59.000Z

277

Gas-cooled reactors  

SciTech Connect

Experience to date with operation of high-temperature gas-cooled reactors has been quite favorable. Despite problems in completion of construction and startup, three high-temperature gas-cooled reactor (HTGR) units have operated well. The Windscale Advanced Gas-Cooled Reactor (AGR) in the United Kingdom has had an excellent operating history, and initial operation of commercial AGRs shows them to be satisfactory. The latter reactors provide direct experience in scale-up from the Windscale experiment to fullscale commercial units. The Colorado Fort St. Vrain 330-MWe prototype helium-cooled HTGR is now in the approach-to-power phase while the 300-MWe Pebble Bed THTR prototype in the Federal Republic of Germany is scheduled for completion of construction by late 1978. THTR will be the first nuclear power plant which uses a dry cooling tower. Fuel reprocessing and refabrication have been developed in the laboratory and are now entering a pilot-plant scale development. Several commercial HTGR power station orders were placed in the U.S. prior to 1975 with similar plans for stations in the FRG. However, the combined effects of inflation, reduced electric power demand, regulatory uncertainties, and pricing problems led to cancellation of the 12 reactors which were in various stages of planning, design, and licensing.

Schulten, R.; Trauger, D.B.

1976-01-01T23:59:59.000Z

278

CONVECTION REACTOR  

DOE Patents (OSTI)

An homogeneous nuclear power reactor utilizing convection circulation of the liquid fuel is proposed. The reactor has an internal heat exchanger looated in the same pressure vessel as the critical assembly, thereby eliminating necessity for handling the hot liquid fuel outside the reactor pressure vessel during normal operation. The liquid fuel used in this reactor eliminates the necessity for extensive radiolytic gas rocombination apparatus, and the reactor is resiliently pressurized and, without any movable mechanical apparatus, automatically regulates itself to the condition of criticality during moderate variations in temperature snd pressure and shuts itself down as the pressure exceeds a predetermined safe operating value.

Hammond, R.P.; King, L.D.P.

1960-03-22T23:59:59.000Z

279

Nuclear Reactor Severe Accident Experiments  

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

Nuclear Reactor Severe Accident Experiments Nuclear Reactor Severe Accident Experiments Capabilities Engineering Experimentation Reactor Safety Testing and Analysis Overview Nuclear Reactor Severe Accident Experiments MAX NSTF SNAKE Aerosol Experiments System Components Laser Applications Robots Applications Other Facilities Other Capabilities Work with Argonne Contact us For Employees Site Map Help Join us on Facebook Follow us on Twitter NE on Flickr Nuclear Reactor Severe Accident Experiments 1 2 3 4 5 6 7 We perform experiments simulating reactor core melt phenomena in which molten core debris ("corium") erodes the concrete floor of a containment building. This occurred during the Fukushima nuclear power plant accident though the extent of concrete damage is yet unknown. This video shows the top view of a churning molten pool of uranium oxide at 2000°C (3600°F) seen during an experiment at Argonne. Corium behaves much like lava.

280

Modularity Approach Modular Pebble Bed Reactor (MPBR)  

E-Print Network (OSTI)

· On--line Refueling #12;4/23/03 MIT NED MPBR Reference Plant Modular Pebble Bed Reactor Thermal Power ­ Reduces Location Requirements #12;4/23/03 MIT NED MPBR · Plant "Farm": ~10 MPBR Systems per "Power Plant modularity principles to the design, construction and operation of advanced nuclear energy plants · To employ

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


281

Naval Research Laboratory Multiscale Targeting Guidance for T-PARC and TCS-08  

Science Conference Proceedings (OSTI)

As part of The Observing System Research and Predictability Experiment (THORPEX) Pacific Asian Regional Campaign (T-PARC) and the Office of Naval Researchs (ONRs) Tropical Cyclone Structure-08 (TCS-08) experiments, a variety of real-time ...

Carolyn A. Reynolds; James D. Doyle; Richard M. Hodur; Hao Jin

2010-04-01T23:59:59.000Z

282

Mitigation action plan sale of Naval Petroleum Reserve No. 1 (Elk Hills) Kern County, California  

Science Conference Proceedings (OSTI)

Naval Petroleum Reserve No. 1 (NPR-1, also called {open_quotes}Elk Hills{close_quotes}), a Federally-owned oil and gas production field in Kern County, California, was created by an Executive Order issued by President Taft on September 2, 1912. He signed another Executive Order on December 13, 1912, to establish Naval Petroleum Reserve No. 2 (NPR-2), located immediately south of NPR-1 and containing portions of the town of Taft, California. NPR-1 was not developed until the 1973-74 oil embargo demonstrated the nation`s vulnerability to oil supply interruptions. Following the embargo, Congress passed the Naval Petroleum Reserves Production Act of 1976 which directed that the reserve be explored and developed to its fall economic potential at the {open_quotes}maximum efficient rate{close_quotes} (MER) of production. Since Elk Hills began full production in 1976, it has functioned as a commercial operation, with total revenues to the Federal government through FY 1996 of $16.4 billion, compared to total exploration, development and production costs of $3.1 billion. In February 1996, Title 34 of the National Defense Authorization Act for Fiscal Year 1996 (P.L. 104-106), referred to as the Elk Hills Sales Statute, directed the Secretary of Energy to sell NPR-1 by February 10, 1998.The Secretary was also directed to study options for enhancing the value of the other Naval Petroleum and Oil Shale Reserve properties such as NPR-2, located adjacent to NPR-1 in Kern County- Naval Petroleum Reserve No. 3 (NPR-3) located in Natrona County, Wyoming; Naval Oil Shale Reserves No. 1 and No. 3 (NOSR-1 and NOSR-3) located in Garfield County, Colorado; and Naval Oil Shale Reserve No. 2 (NOSR-2) located in Uintah and Carbon Counties, Utah. The purpose of these actions was to remove the Federal government from the inherently non-Federal function of operating commercial oil fields while making sure that the public would obtain the maximum value from the reserves.

NONE

1998-01-01T23:59:59.000Z

283

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in electron anti-neutrino detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties and their relevance to reactor anti-neutrino experiments.

Z. Djurcic; J. A. Detwiler; A. Piepke; V. R. Foster Jr.; L. Miller; G. Gratta

2008-08-06T23:59:59.000Z

284

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

E-Print Network (OSTI)

Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in electron anti-neutrino detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties and their relevance to reactor anti-neutrino experiments.

Djurcic, Z; Piepke, A; Foster, V R; Miller, L; Gratta, G

2008-01-01T23:59:59.000Z

285

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A neutronic reactor in which neutron moderation is achieved primarily in its reflector is described. The reactor structure consists of a cylindrical central "island" of moderator and a spherical moderating reflector spaced therefrom, thereby providing an annular space. An essentially unmoderated liquid fuel is continuously passed through the annular space and undergoes fission while contained therein. The reactor, because of its small size, is particularly adapted for propulsion uses, including the propulsion of aircraft. (AEC)

Fraas, A.P.; Mills, C.B.

1961-11-21T23:59:59.000Z

286

A laboratory and pilot plant scaled continuous stirred reactor separator for the production of ethanol from sugars, corn grits/starch or biomass streams  

DOE Green Energy (OSTI)

An improved bio-reactor has been developed to allow the high speed, continues, low energy conversion of various substrates to ethanol. The Continuous Stirred Reactor Separator (CSRS) incorporates gas stripping of the ethanol using a recalculating gas stream between cascading stirred reactors in series. We have operated a 4 liter lab scale unit, and built and operated a 24,000 liter pilot scale version of the bioreactor. High rates of fermentation are maintained in the reactor stages using a highly flocculent yeast strain. Ethanol is recovered from the stripping gas using a hydrophobic solvent absorber (isothermal), after which the gas is returned to the bioreactor. Ethanol can then be removed from the solvent to recover a highly concentrated ethanol product. We have applied the lab scale CSRS to sugars (glucose/sucrose), molasses, and raw starch with simultaneous saccharification and fermentation of the starch granules (SSF). The pilot scale CSRS has been operated as a cascade reactor using dextrins as a feed. Operating data from both the lab and pilot scale CSRS are presented. Details of how the system might be applied to cellulosics, with some preliminary data are also given.

Dale, M.C.; Lei, Shuiwang; Zhou, Chongde

1995-10-01T23:59:59.000Z

287

A COMPARISON OF THE NUCLEAR DEFENSE CAPABILITIES ON NUCLEAR AND COAL-FIRED POWER PLANTS. FUEL COST STUDY VARIOUS REACTORS AT 100 AND 300 Mwe  

SciTech Connect

Appendices C and D may further be identified as SL1925 and CF-61-12- 20(Rev.), respectively. A comparative report is presented in which the economics and feasibility of plant protection from nuclear attack by plant hardening, remote siting, and utilization of optional fueling concepts for the coal-fired plant are evaluated. (J.R.D.)

Gift, E.H.

1962-05-29T23:59:59.000Z

288

Reactor Internals Segmentation Experience Report  

Science Conference Proceedings (OSTI)

Several U.S. nuclear power plants entered decommissioning in the 1990's. Based on current information, the next group of plants will not begin this process for nearly a decade. This report provides detailed information on projects involving the segmentation of the Reactor Vessel Internal Subassemblies conducted at a number of plants undergoing decommissioning. Utilities have found this task very challenging from both a technological and radiation exposure perspective. This report documents the experience...

2007-07-10T23:59:59.000Z

289

REACTOR COOLING  

DOE Patents (OSTI)

A nuclear reactor with provisions for selectively cooling the fuel elements is described. The reactor has a plurality of tubes extending throughout. Cylindrical fuel elements are disposed within the tubes and the coolant flows through the tubes and around the fuel elements. The fuel elements within the central portion of the reactor are provided with roughened surfaces of material. The fuel elements in the end portions of the tubes within the reactor are provlded with low conduction jackets and the fuel elements in the region between the central portion and the end portions are provided with smooth surfaces of high heat conduction material.

Quackenbush, C.F.

1959-09-29T23:59:59.000Z

290

Audit Report: OIG-0884 | Department of Energy  

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

OIG-0884 Audit Report: OIG-0884 April 12, 2013 Management of Naval Reactors' Cyber Security Program The Naval Reactors Program (Naval Reactors), an organization within the National...

291

A Pilot Plant Scale Reactor/Separator for Ethanol from Cellulosics. ERIP/DOE Quarterly Reports 5 and 6, October 1, 1998 through March 30, 1999  

DOE Green Energy (OSTI)

The objective of this project was to develop and demonstrate a continuous low energy process for the conversion of cellulosics to ethanol. BPI's process involves a proprietary low temperature pretreatment step which allows recycle of the pretreatment chemicals and recovery of a lignin stream. The pretreated biomass is then converted to glucans and xylans enzymatically and these sugars simultaneously fermented to ethanol (SSF) in BPI's Continuous Stirred Reactor Separator (CSRS). The CSRS is a multi stage bio-reactor where the glucans are first converted to ethanol using a high temperature tolerant yeast, followed by xylan SSF on the lower stages using a second xylose fermenting yeast strain. Ethanol is simultaneously removed from the bio-reactor stages, speeding the fermentation, and allowing the complete utilization of the biomass.

Dale, M. Clark; Moelhman, Mark

1999-09-30T23:59:59.000Z

292

A Pilot Plant Scale Reactor/Separator for Ethanol from Cellulosics. ERIP/DOE Quarterly Reports 7, 8 and Final report  

DOE Green Energy (OSTI)

The objective of this project was to develop and demonstrate a continuous low energy process for the conversion of cellulosics to ethanol. BPI's process involves a proprietary low temperature pretreatment step which allows recycle of the pretreatment chemicals and recovery of a lignin stream. The pretreated biomass is then converted to glucans and xylans enzymatically and these sugars simultaneously fermented to ethanol (SSF) in BPI's Continuous Stirred Reactor Separator (CSRS). The CSRS is a multi stage bio-reactor where the glucans are first converted to ethanol using a high temperature tolerant yeast stran, followed by xylan SSF on the lower stages using a second xylose fermenting yeast strain. Ethanol is simultaneously removed from the bio-reactor stages, speeding the fermentation, and allowing the complete utilization of the biomass.

Cale, M. Clark; Moelhman, Mark

1999-09-30T23:59:59.000Z

293

Performance and Economics of a Wind-Diesel Hybrid Energy System: Naval Air Landing Field, San Clemente Island, California  

DOE Green Energy (OSTI)

This report provides an overview of the wind resource, economics and operation of the recently installed wind turbines in conjunction with diesel power for the Naval Air Landing Field (NALF), San Clemente Island (SCI), California Project. The primary goal of the SCI wind power system is to operate with the existing diesel power plant and provide equivalent or better power quality and system reliability than the existing diesel system. The wind system is also intended to reduce, as far as possible, the use of diesel fuel and the inherent generation of nitrogen-oxide emissions and other pollutants. The first two NM 225/30 225kW wind turbines were installed and started shake-down operations on February 5, 1998. This report describes the initial operational data gathered from February 1998 through January 1999, as well as the SCI wind resource and initial cost of energy provided by the wind turbines on SCI. In support of this objective, several years of data on the wind resources of San Clemente Island were collected and compared to historical data. The wind resource data were used as input to economic and feasibility studies for a wind-diesel hybrid installation for SCI.

McKenna, E. (National Renewable Energy Laboratory); Olsen, T. (Timothy Olsen Consulting)

1999-07-06T23:59:59.000Z

294

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program:  

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

Assessment of High Value Surveillance Materials Assessment of High Value Surveillance Materials Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Assessment of High Value Surveillance Materials The reactor pressure vessel (RPV) in a light-water reactor (LWR) represents the first line of defense against a release of radiation in case of an accident. Thus, regulations that govern the operation of commercial nuclear power plants require conservative margins of fracture toughness, both during normal operation and under accident scenarios. In the unirradiated condition, the RPV has sufficient fracture toughness such that failure is implausible under any postulated condition, including pressurized thermal shock (PTS) in pressurized water reactors (PWR). In the irradiated condition, however, the fracture toughness of the RPV may be severely

295

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor for isotope production is described. This reactor is designed to provide a maximum thermal neutron flux in a region adjacent to the periphery of the reactor rather than in the center of the reactor. The core of the reactor is generally centrally located with respect tn a surrounding first reflector, constructed of beryllium. The beryllium reflector is surrounded by a second reflector, constructed of graphite, which, in tune, is surrounded by a conventional thermal shield. Water is circulated through the core and the reflector and functions both as a moderator and a coolant. In order to produce a greatsr maximum thermal neutron flux adjacent to the periphery of the reactor rather than in the core, the reactor is designed so tbat the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the materials in the reflector is approximately twice the ratio of neutron scattering cross section to neutron absorption cross section averaged over all of the material of the core of the reactor.

Wigner, E.P.

1958-04-22T23:59:59.000Z

296

Success Story: Naval Medical Center San Diego Co-Generation Project  

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

Success Story Success Story Success Story Naval Medical Center San Diego Naval Medical Center San Diego Co-Generation Project Co-Generation Project Karen Jackson, SDG&E Karen Jackson, SDG&E Project Manager Project Manager Edward Thibodo, NAVFAC SW Edward Thibodo, NAVFAC SW Energy Team Contract Energy Team Contract ' ' s Lead s Lead NAVFAC Contractor NAVFAC Contractor ' ' s Guide: s Guide:   Partnering Philosophy Partnering Philosophy - - " " We W are partners e are partners in every contract we award. Partnering is in every contract we award. Partnering is an attitude that we both work hard to an attitude that we both work hard to develop, an it requires both of us to take develop, an it requires both of us to take some extra risk and trust one another. some extra risk and trust one another.

297

DOE - Office of Legacy Management -- Naval Office at the University of New  

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

Office at the University of Office at the University of New Mexico - NM 0-03 FUSRAP Considered Sites Site: NAVAL OFFICE AT THE UNIVERSITY OF NEW MEXICO (NM.0-03) Eliminated from further consideration under FUSRAP - Referred to DOD Designated Name: Not Designated Alternate Name: None Location: Albuquerque , New Mexico NM.0-03-1 Evaluation Year: 1987 NM.0-03-1 Site Operations: Site was a transshipment station for equipment to the Los Alamos site. NM.0-03-1 Site Disposition: Eliminated - Referred to DOD NM.0-03-1 Radioactive Materials Handled: None Indicated Primary Radioactive Materials Handled: None NM.0-03-1 Radiological Survey(s): None Indicated Site Status: Eliminated from further consideration under FUSRAP - Referred to DOD NM.0-03-1 Also see Documents Related to NAVAL OFFICE AT THE UNIVERSITY OF NEW MEXICO

298

DOE - Office of Legacy Management -- Naval Proving Ground - VA 0-01  

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

Proving Ground - VA 0-01 Proving Ground - VA 0-01 FUSRAP Considered Sites Site: NAVAL PROVING GROUND (VA.0-01 ) Eliminated from further consideration under FUSRAP - Referred to DOD Designated Name: Not Designated Alternate Name: None Location: Dahlgren , Virginia VA.0-01-1 Evaluation Year: 1987 VA.0-01-1 Site Operations: Site operations were not specified; this site was identified on the 1954 Accountable Station Lists. VA.0-01-1 VA.0-01-2 Site Disposition: Eliminated - Referred to DOD VA.0-01-1 Radioactive Materials Handled: None Indicated Primary Radioactive Materials Handled: None Radiological Survey(s): None Indicated Site Status: Eliminated from further consideration under FUSRAP - Referred to DOD VA.0-01-1 Also see Documents Related to NAVAL PROVING GROUND VA.0-01-1 - DOE Letter; Fiore to Schafer; Referral of DOD or Former

299

DOE - Office of Legacy Management -- Naval Ordnance Laboratory - MD 0-03  

Office of Legacy Management (LM)

Ordnance Laboratory - MD 0-03 Ordnance Laboratory - MD 0-03 FUSRAP Considered Sites Site: NAVAL ORDNANCE LABORATORY (MD.0-03 ) Eliminated from further consideration under FUSRAP - Referred to DOD Designated Name: Not Designated Alternate Name: Naval Ordnance Laboratory - White Oak Location: White Oak Area , Silver Spring , Maryland MD.0-03-1 MD.0-03-2 Evaluation Year: 1987 MD.0-03-2 Site Operations: Research and development - may have involved radioactive materials because the site was identified on a 1955 Accountability Station List. MD.0-03-1 Site Disposition: Eliminated - NRC licensed MD.0-03-2 Radioactive Materials Handled: None Specifically Identified Primary Radioactive Materials Handled: None specifically indicated Radiological Survey(s): None Indicated Site Status: Eliminated from further consideration under FUSRAP - Referred to DOD MD.0-03-2

300

Investigation on the continued production of the Naval Petroleum Reserves beyond April 5, 1991  

SciTech Connect

The authority to produce the Naval Petroleum Reserves (NPRs) is due to expire in April 1991, unless extended by Presidential finding. As provided in the Naval Petroleum Reserves Production act of 1976 (Public Law 94-258), the President may continue production of the NPRs for a period of up to three years following the submission to Congress, at least 180 days prior to the expiration of the current production period, of a report that determines that continued production of the NPRs is necessary and a finding by the President that continued production is in the national interest. This report assesses the need to continue production of the NPRs, including analyzing the benefits and costs of extending production or returning to the shut-in status that existed prior to 1976. This continued production study considers strategic, economic, and energy issues at the local, regional, and national levels. 15 figs., 13 tabs.

Not Available

1990-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "naval reactor plants" 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

Naval Petroleum and Oil Shale Reserves annual report of operations, Fiscal year 1991  

Science Conference Proceedings (OSTI)

During FY 1991 the Reserves generating $654 million in revenues, a $52 million increase from the Fy 1990 revenues, reflecting the increase in FY 1991 oil prices during the Gulf War. Total costs were $200 million, resulting in net cash flow of $454 million, compared with $423 million in FY 1990. Revenues for FY 1992 are expected to decrease, reflecting a decrease in production and prices. In FY 1991, production at the NPRs at maximum efficient rates yielded 28 million barrels of crude oil, 125 billion cubic feet of natural gas, and 183 million gallons of natural gas liquids. Additional highlights on the following topics are included: legislative change, Naval Petroleum Reserve-1 (NPR-1) exploration, NPR-1 horizontal drilling, NPR-1 shallow oil zone gas injection project, NPR-1 FY 1991-1997 long range plan, NPR-1 environment and safety, NPR-2 exploration drilling, NPR-3 steamflood, Naval Oil Shale Reserves gas migration prevention.

Not Available

1991-01-01T23:59:59.000Z

302

Naval Petroleum and Oil Shale Reserves annual report of operations, Fiscal year 1991  

Science Conference Proceedings (OSTI)

During FY 1991 the Reserves generating $654 million in revenues, a $52 million increase from the Fy 1990 revenues, reflecting the increase in FY 1991 oil prices during the Gulf War. Total costs were $200 million, resulting in net cash flow of $454 million, compared with $423 million in FY 1990. Revenues for FY 1992 are expected to decrease, reflecting a decrease in production and prices. In FY 1991, production at the NPRs at maximum efficient rates yielded 28 million barrels of crude oil, 125 billion cubic feet of natural gas, and 183 million gallons of natural gas liquids. Additional highlights on the following topics are included: legislative change, Naval Petroleum Reserve-1 (NPR-1) exploration, NPR-1 horizontal drilling, NPR-1 shallow oil zone gas injection project, NPR-1 FY 1991-1997 long range plan, NPR-1 environment and safety, NPR-2 exploration drilling, NPR-3 steamflood, Naval Oil Shale Reserves gas migration prevention.

Not Available

1991-12-31T23:59:59.000Z

303

U.S. Department of Energy Naval Petroleum and Oil Shale Reserves combined financial statements, September 30, 1996 and 1995  

Science Conference Proceedings (OSTI)

The Naval Petroleum and Oil Shale Reserves (NPOSR) produces crude oil and associated hydrocarbons from the Naval Petroleum Reserves (NPR) numbered 1, 2, and 3, and the Naval Oil Shale Reserves (NOSR) numbered 1, 2, and 3 in a manner to achieve the greatest value and benefits to the US taxpayer. NPOSR consists of the Naval Petroleum Reserve in California (NPRC or Elk Hills), which is responsible for operations of NPR-1 and NPR-2; the Naval Petroleum Oil Shale Reserve in Colorado, Utah, and Wyoming (NPOSR-CUW), which is responsible for operations of NPR-3, NOSR-1, 2, and 3 and the Rocky Mountain Oilfield Testing Center (RMOTC); and NPOSR Headquarters in Washington, DC, which is responsible for overall program direction. Each participant shares in the unit costs and production of hydrocarbons in proportion to the weighted acre-feet of commercially productive oil and gas formations (zones) underlying the respective surface lands as of 1942. The participating shares of NPR-1 as of September 30, 1996 for the US Government and Chevron USA, Inc., are listed. This report presents the results of the independent certified public accountants` audit of the Department of Energy`s (Department) Naval Petroleum and Oil Shale Reserves (NPOSR) financial statements as of September 30, 1996.

NONE

1997-03-01T23:59:59.000Z

304

System Design and Analysis of a 900-MW(thermal) Lead-Cooled Fast Reactor  

Science Conference Proceedings (OSTI)

Technical Paper / Special Issue on the 2008 International Congress on Advances in Nuclear Power Plants / Fission Reactors

Sang Ji Kim; Yonghee Kim; Sergi Hong; Chung Ho Cho; Jae-Hyuk Eoh; Jong Bum Kim; Myung Hwan Wi; Kwi Seok Ha; Eui Kwang Kim

305

NUCLEAR PLANT OPERATIONS AND  

E-Print Network (OSTI)

reactor Unit 4 of the Ringhals Nuclear Power Plant (Sweden) during fuel cycle 16 is analyzed--has been benchmarked against measurements.30 At the Ringhals nuclear power plant, this measurement is car a measurement performed at the PWR Unit 4 of the Ring hals Nuclear Power Plant was available to us

Demazière, Christophe

306

NUCLEAR PLANT OPERATIONS AND  

E-Print Network (OSTI)

reactor Unit 4 of the Ringhals Nuclear Power Plant (Sweden) during fuel cycle 16 is analyzed reactivity effects--has been benchmarked against measurements.30 At the Ringhals nuclear power plant a measurement performed at the PWR Unit 4 of the Ring- hals Nuclear Power Plant was available to us

Demazière, Christophe

307

Louisiana Nuclear Profile - Power Plants  

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

Louisiana nuclear power plants, summer capacity and net generation, 2010" "Plant NameTotal Reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear...

308

Reactor closure design for a pool-type fast reactor  

SciTech Connect

The reactor closure is the topmost structural part of a reactor module. For a pool-type fast reactor it is an especially important structure because it provides the interface between the primary coolant system and the main access area above the closure. The reactor closure comprises a stationary deck, a rotatable plug, the boundary elements of primary system and containment penetrations for equipment and auxiliary systems. This paper evaluates two different reactor closure design concepts, referred to as ''warm'' deck and ''hot'' deck, for a pool-type fast reactor with respect to their design features, technical merits, and economic benefits. The evaluation also includes functional, structural, and thermal analyses of the two deck design concepts. Issues related to their fabrication and shipping to the plant site are also addressed. The warm deck is a thick solid steel plate with under-the-deck insulation consisting of many layers of steel plates. The hot deck is a box-type structure consisting of a bottom plate reinforced with vertical ribs and cylinders. For insulation and radiation shielding, the region of the hot deck above the bottom plate is filled with steel balls. Conventional insulation is added on the top to further reduce heat loss into area above the deck. The design choice of the closure deck is strongly dependent on design features of the reactor; especially on the reactor module support. While the warm deck is preferable with the top support, the hot deck is better suited for the bottom support design of the module.

Chung, H.; Seidensticker, R.W.; Kann, W.J.; Bump, T.R.; Schatmeier, C.

1986-01-01T23:59:59.000Z

309

Licensed reactor nuclear safety criteria applicable to DOE reactors  

SciTech Connect

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC (Nuclear Regulatory Commission) licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor.

Not Available

1991-04-01T23:59:59.000Z

310

STUDY OF REMOTE MILITARY POWER APPLICATIONS. REPORT NO. 12. EVALUATION AND SELECTION OF APPLICABLE REACTOR CONCEPTS  

SciTech Connect

An evaluation of the reactor concepts under consideration for remote military power plants is presented. The concepts include water-cooled and - moderated reactors, both direct and indirect cycle. organic-cooled and -moderated reactors, heavy-water-cooled and -moderated reactors. gas-cooled reactors, sodium- cooled graphite-moderated reactors, fast reactors, and fluid-fuel reactors. The limitations and advantages, technological status, economics, and future potential of each reactor are reviewed. From the reviews it is concluded that direct-cycle boiling-water and pressurized-water reactors are most suitable for applications requiring power before 1965. (C.J.G.)

1960-01-01T23:59:59.000Z

311

NUCLEAR REACTOR  

DOE Patents (OSTI)

A heterogeneous, natural uranium fueled, solid moderated, gas cooled reactor is described, in which the fuel elements are in the form of elongated rods and are dlsposed within vertical coolant channels ln the moderator symmetrically arranged as a regular lattice in groups. This reactor employs control rods which operate in vertical channels in the moderator so that each control rod is centered in one of the fuel element groups. The reactor is enclosed in a pressure vessel which ls provided with access holes at the top to facilitate loading and unloadlng of the fuel elements, control rods and control rod driving devices.

Moore, R.V.; Bowen, J.H.; Dent, K.H.

1958-12-01T23:59:59.000Z

312

High Temperature Gas Reactors Briefing to  

E-Print Network (OSTI)

· Nuclear Power 2010 · Next Generation Nuclear Plant (NGNP) · Generation IV Nuclear Plants · NRC Regulatory Specifications · Rated Power per Module 165-175 MW(e) depending on injection temperature · Eight-pack Plant 1320 - Indirect Cycle - Core Options Available - Waste Minimization #12;Modular Pebble Bed Reactor Thermal Power

313

STEAM-COOLED POWER REACTOR EVALUATION, STEAM-COOLED FAST BREEDER REACTOR  

SciTech Connect

Conceptual design and economic studies of a steamcooled fast breeder reactor that can also be used as a source of power are presented. Two reactor plant sizes were considered: a 300-Mw(e) central power station plant and a 40 Mw(e) plant. It was concluded that attractive economics and good breeding characteristics breeding ratios from 1.27 to 1.42) can be achieved in steam- cooled PuO/sub 2/UO/sub 2/ fueled fast reactors. Low capital costs can be obtained by a compact reactor core and the absence of large heat exchangers and complicated process systems. Reactor design data are discussed. Analysis showed that these reactors can be prevented from going prompt critical, when fully flooded, by incorporating a tolerable amount of high resonance absorption materials such as hafnium or indium. An increase in reactivity on loss of coolant was indicated by preliminary calculations. (M.C.G.)

Sofer, G.; Hankel, R.; Goldstein, L.; Birman, G.

1961-04-15T23:59:59.000Z

314

Materials Reliability Program: Sensitivity Studies for Functionality Analysis of Reactor Internals of Babcock and Wilcox (B&W-) Designed Power Plants (MRP-355)  

Science Conference Proceedings (OSTI)

In preparing the Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0), a semi-empirical material behavior model for irradiated austenitic stainless steels was developed for the degradation mechanisms applicable to the internals. Functionality analyses that predicted aging up to 60 years were performed using this material model. These functionality analyses are part of the technical bases for the guidelines. For the material model and ...

2013-07-25T23:59:59.000Z

315

Nuclear Reactor Technologies | Department of Energy  

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

Reactor Technologies Reactor Technologies Nuclear Reactor Technologies TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority TVA Watts Bar Nuclear Power Plant | Photo courtesy of Tennessee Valley Authority Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Small Modular Reactor Technologies Small modular reactors can also be made in factories and transported to sites where they would be ready to "plug and play" upon arrival, reducing both capital costs and construction times. The smaller size also makes these reactors ideal for small electric grids and for locations that

316

EVALUATION OF MERCURY COOLED BREEDER REACTORS  

SciTech Connect

A technical and economic evaluation of a mercury-cooled fast breeder reactor is presented. The objectives of the program were to establish the technical feasibility of a fast breeder reactor cooled with boiling mercury and to evaluate the long-range potential of such a reactor power plant for production of economic power. Details of the conceptual design of a 100-Mw(e) reactor and system are discussed. The power cost from a mercury cooled fast breeder reactor was estimated as 21.4 mills/kwh which is competitive with the power cost for the initial Enrico Fermi plant. It was concluded that this reactor concept is technically feasible and has promising long-range economic potential. (M.C.G.)

Battles, D.W.

1960-12-14T23:59:59.000Z

317

Light Water Reactor Sustainability Program - Non-Destructive...  

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

Remaining Useful Life of Aging Cables in Nuclear Power Plants Light Water Reactor Sustainability Program - Non-Destructive Evaluation R&D Roadmap for Determining Remaining Useful...

318

IMPROVEMENTS IN POWER PLANT  

SciTech Connect

A power plant for nuclear reactors is designed for improved cycle efficiency. In addition to the usual heat exchanger for heat transfer from gaseous reactor coolant to water for vaporization, a second heat exchanger is provided between the first heat exchanger and a point betwveen the intermediate- pressure and low-pressure turbine stages. In this way, interstage reheating of the steam is obtained without passage of the steam back to the first heat exchanger. (D.L.C.) Research Reactors

Peters, M.C.

1961-10-11T23:59:59.000Z

319

Update; Sodium advanced fast reactor (SAFR) concept  

SciTech Connect

This paper reports on the sodium advanced fast reactor (SAFR) concept developed by the team of Rockwell International, Combustion Engineering, and Bechtel during the 3-year period extending from January 1985 to December 1987 as one element in the U.S. Department of Energy's Advanced Liquid Metal Reactor Program. In January 1988, the team was expanded to include Duke Engineering and Services, Inc., and the concept development was extended under DOE's Program for Improvement in Advanced Modular LMR Design. The SAFR plant concept employs a 450-MWe pool-type liquid metal cooled reactor as its basic module. The reactor assembly module is a standardized shop-fabricated unit that can be shipped to the plant site by barge for installation. Shop fabrication minimizes nuclear-grade field fabrication and reduces the plant construction schedule. Reactor modules can be used individually or in multiples at a given site to supply the needed generating capacity.

Oldenkamp, R.D.; Brunings, J.E. (Rockwell International Corp., Canoga Park, CA (USA)); Guenther, E. (Combustion Engineering, Windsor, CT (US)); Hren, R. (Bechtel National Inc., San Francisco, CA (US))

1988-01-01T23:59:59.000Z

320

Reactor Materials  

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

The reactor materials crosscut effort will enable the development of innovative and revolutionary materials and provide broad-based, modern materials science that will benefit all four DOE-NE...

Note: This page contains sample records for the topic "naval reactor plants" 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

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor incorporating seed and blanket assemblies is designed. Means are provided for obtaining samples of the coolant from the blanket assemblies and for varying the flow of coolant through the blanket assemblies. (AEC)

Sherman, J.; Sharbaugh, J.E.; Fauth, W.L. Jr.; Palladino, N.J.; DeHuff, P.G.

1962-10-23T23:59:59.000Z

322

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A nuclear reactor is described wherein horizontal rods of thermal- neutron-fissionable material are disposed in a body of heavy water and extend through and are supported by spaced parallel walls of graphite.

Wigner, E.P.

1960-11-22T23:59:59.000Z

323

REACTOR SHIELD  

DOE Patents (OSTI)

Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

1959-02-17T23:59:59.000Z

324

SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS  

E-Print Network (OSTI)

SIGNAL GROUPING FOR CONDITION MONITORING OF NUCLEAR POWER PLANT COMPONENTS Piero Baraldi between those used to monitor the reactor coolant pump of a Pressurized Water Reactor (PWR) is considered Monitoring, Empirical Modeling, Power Plants, Safety Critical Nuclear Instrumentation, Autoassociative models

325

Physical protection of power reactors  

SciTech Connect

Sandia Laboratories has applied a systematic approach to designing physical protection systems for nuclear facilities to commercial light-water reactor power plants. A number of candidate physical protection systems were developed and evaluated. Focus is placed on the design of access control subsystems at each of three plant layers: the protected area perimeter, building surfaces, and vital areas. Access control refers to barriers, detectors, and entry control devices and procedures used to keep unauthorized personnel and contraband out of the plant, and to control authorized entry into vital areas within the plant.

Darby, J.L.

1979-01-01T23:59:59.000Z

326

NUCLEAR REACTOR  

DOE Patents (OSTI)

High temperature reactors which are uniquely adapted to serve as the heat source for nuclear pcwered rockets are described. The reactor is comprised essentially of an outer tubular heat resistant casing which provides the main coolant passageway to and away from the reactor core within the casing and in which the working fluid is preferably hydrogen or helium gas which is permitted to vaporize from a liquid storage tank. The reactor core has a generally spherical shape formed entirely of an active material comprised of fissile material and a moderator material which serves as a diluent. The active material is fabricated as a gas permeable porous material and is interlaced in a random manner with very small inter-connecting bores or capillary tubes through which the coolant gas may flow. The entire reactor is divided into successive sections along the direction of the temperature gradient or coolant flow, each section utilizing materials of construction which are most advantageous from a nuclear standpoint and which at the same time can withstand the operating temperature of that particular zone. This design results in a nuclear reactor characterized simultaneously by a minimum critiral size and mass and by the ability to heat a working fluid to an extremely high temperature.

Grebe, J.J.

1959-07-14T23:59:59.000Z

327

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

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

328

High Temperature Gas Reactors The Next Generation ?  

E-Print Network (OSTI)

HPT CCS Reactor CBCS #12;14 Integrated Plant Systems #12;15 Differences Between LWRS · Higher Thermal - Not shown Fresh Fuel Storage Used Fuel Storage Tanks #12;39 MPBR Specifications Thermal Power 250 MW Core temperatures about 1670 C. #12;MPBRBUSBARGENERATIONCOSTS(`92$) ReactorThermal Power (MWt) 10x250 Net Efficiency

329

Selective purge for hydrogenation reactor recycle loop  

SciTech Connect

Processes and apparatus for providing improved contaminant removal and hydrogen recovery in hydrogenation reactors, particularly in refineries and petrochemical plants. The improved contaminant removal is achieved by selective purging, by passing gases in the hydrogenation reactor recycle loop or purge stream across membranes selective in favor of the contaminant over hydrogen.

Baker, Richard W. (Palo Alto, CA); Lokhandwala, Kaaeid A. (Union City, CA)

2001-01-01T23:59:59.000Z

330

Research reactors - an overview  

SciTech Connect

A broad overview of different types of research and type reactors is provided in this paper. Reactor designs and operating conditions are briefly described for four reactors. The reactor types described include swimming pool reactors, the High Flux Isotope Reactor, the Mark I TRIGA reactor, and the Advanced Neutron Source reactor. Emphasis in the descriptions is placed on safety-related features of the reactors. 7 refs., 7 figs., 2 tabs.

West, C.D.

1997-03-01T23:59:59.000Z

331

On reactor type comparisons for the next generation of reactors  

SciTech Connect

In this paper, we present a broad comparison of studies for a selected set of parameters for different nuclear reactor types including the next generation. This serves as an overview of key parameters which provide a semi-quantitative decision basis for selecting nuclear strategies. Out of a number of advanced reactor designs of the LWR type, gas cooled type, and FBR type, currently on the drawing board, the Advanced Light Water Reactors (ALWR) seem to have some edge over other types of the next generation of reactors for the near-term application. This is based on a number of attributes related to the benefit of the vast operating experience with LWRs coupled with an estimated low risk profile, economics of scale, degree of utilization of passive systems, simplification in the plant design and layout, modular fabrication and manufacturing. 32 refs., 1 fig., 3 tabs.

Alesso, H.P.; Majumdar, K.C.

1991-08-22T23:59:59.000Z

332

Modular Pebble Bed Reactor High Temperature Gas Reactor  

E-Print Network (OSTI)

For 1150 MW Combined Heat and Power Station Oil Refinery Hydrogen Production Desalinization Plant VHTR;Equipment Layout #12;Modular Pebble Bed Reactor Thermal Power 250 MW Core Height 10.0 m Core Diameter 3.5 m · License by Test · Expert I&C System - Hands free operation #12;MIT MPBR Specifications Thermal Power 250

333

Brookhaven Medical Research Reactor  

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

Medical Research Reactor BMRR The last of the Lab's reactors, the Brookhaven Medical Research Reactor (BMRR), was shut down in December 2000. The BMRR was a three megawatt...

334

NEUTRONIC REACTOR  

DOE Patents (OSTI)

This patent relates to neutronic reactors of the heterogeneous water cooled type, and in particular to a fuel element charging and discharging means therefor. In the embodiment illustrated the reactor contains horizontal, parallel coolant tubes in which the fuel elements are disposed. A loading cart containing a magnzine for holding a plurality of fuel elements operates along the face of the reactor at the inlet ends of the coolant tubes. The loading cart is equipped with a ram device for feeding fuel elements from the magazine through the inlot ends of the coolant tubes. Operating along the face adjacent the discharge ends of the tubes there is provided another cart means adapted to receive irradiated fuel elements as they are forced out of the discharge ends of the coolant tubes by the incoming new fuel elements. This cart is equipped with a tank coataining a coolant, such as water, into which the fuel elements fall, and a hydraulically operated plunger to hold the end of the fuel element being discharged. This inveation provides an apparatus whereby the fuel elements may be loaded into the reactor, irradiated therein, and unloaded from the reactor without stopping the fiow of the coolant and without danger to the operating personnel.

Ohlinger, L.A.; Wigner, E.P.; Weinberg, A.M.; Young, G.J.

1958-09-01T23:59:59.000Z

335

ENGINEERING STUDY ON UNDERGROUND CONSTRUCTION OF NUCLEAR POWER REACTORS  

SciTech Connect

The advantages, disadvantages, and cost of constructing a auclear power reactor underground are outlinedData on underground construction of hydroelectric plants, other structures, and underground reactor projects in Norway and Sweden are reviewed. A hypothetical underground Experimental Boiling Water Reactor design and sketch are given with cost estimates(T.R.H.)

Beck, C.

1958-04-15T23:59:59.000Z

336

POWER REACTOR  

DOE Patents (OSTI)

A fast nuclear reactor system ls described for producing power and radioactive isotopes. The reactor core is of the heterogeneous, fluid sealed type comprised of vertically arranged elongated tubular fuel elements having vertical coolant passages. The active portion is surrounded by a neutron reflector and a shield. The system includes pumps and heat exchangers for the primary and secondary coolant circuits. The core, primary coolant pump and primary heat exchanger are disposed within an irapenforate tank which is filled with the primary coolant, in this case a liquid metal such as Na or NaK, to completely submerge these elements. The tank is completely surrounded by a thick walled concrete shield. This reactor system utilizes enriched uranium or plutonium as the fissionable material, uranium or thorium as a diluent and thorium or uranium containing less than 0 7% of the U/sup 235/ isotope as a fertile material.

Zinn, W.H.

1958-07-01T23:59:59.000Z

337

REACTOR CONTROL  

DOE Patents (OSTI)

A control system employed with a high pressure gas cooled reactor in which a control rod is positioned for upward and downward movement into the neutron field from a position beneath the reactor is described. The control rod is positioned by a coupled piston cylinder releasably coupled to a power drive means and the pressurized coolant is directed against the lower side of the piston. The coolant pressure is offset by a higher fiuid pressure applied to the upper surface of the piston and means are provided for releasing the higher pressure on the upper side of the piston so that the pressure of the coolant drives the piston upwardly, forcing the coupled control rod into the ncutron field of the reactor. (AEC)

Fortescue, P.; Nicoll, D.

1962-04-24T23:59:59.000Z

338

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor of the homogeneous liquid fuel type is described wherein the fissionable isotope is suspended or dissolved in a liquid moderator such as water. The reactor core is comprised essentially of a spherical vessel for containing the reactive composition surrounded by a reflector, preferably of beryllium oxide. The reactive composition may be an ordinary water solution of a soluble salt of uranium, the quantity of fissionable isotope in solution being sufficient to provide a critical mass in the vessel. The liquid fuel is stored in a tank of non-crtttcal geometry below the reactor vessel and outside of the reflector and is passed from the tank to the vessel through a pipe connecting the two by air pressure means. Neutron absorbing control and safety rods are operated within slots in the reflector adjacent to the vessel.

Christy, R.F.

1958-07-15T23:59:59.000Z

339

Catalytic reactor  

DOE Patents (OSTI)

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

Aaron, Timothy Mark (East Amherst, NY); Shah, Minish Mahendra (East Amherst, NY); Jibb, Richard John (Amherst, NY)

2009-03-10T23:59:59.000Z

340

NEUTRONIC REACTORS  

DOE Patents (OSTI)

A method is presented for loading and unloading rod type fuel elements of a neutronic reactor of the heterogeneous, solld moderator, liquid cooled type. In the embodiment illustrated, the fuel rods are disposed in vertical coolant channels in the reactor core. The fuel rods are loaded and unloaded through the upper openings of the channels which are immersed in the coolant liquid, such as water. Unloading is accomplished by means of a coffer dam assembly having an outer sleeve which is placed in sealing relation around the upper opening. A radiation shield sleeve is disposed in and reciprocable through the coffer dam sleeve. A fuel rod engaging member operates through the axial bore in the radiation shield sleeve to withdraw the fuel rod from its position in the reactor coolant channel into the shield, the shield snd rod then being removed. Loading is accomplished in the reverse procedure.

Wigner, E.P.; Young, G.J.

1958-10-14T23:59:59.000Z

Note: This page contains sample records for the topic "naval reactor plants" 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

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent covers a power-producing nuclear reactor in which fuel rods of slightly enriched U are moderated by heavy water and cooled by liquid metal. The fuel rods arranged parallel to one another in a circle are contained in a large outer closed-end conduit that extends into a tank containing the heavy water. Liquid metal is introduced into the large conduit by a small inner conduit that extends within the circle of fuel rods to a point near the lower closed end of the outer conduit. (AEC) Production Reactors

Young, G.

1963-01-01T23:59:59.000Z

342

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A nuclear reactor which uses uranium in the form of elongated tubes as fuel elements and liquid as a coolant is described. Elongated tubular uranium bodies are vertically disposed in an efficient neutron slowing agent, such as graphite, for example, to form a lattice structure which is disposed between upper and lower coolant tanks. Fluid coolant tubes extend through the uranium bodies and communicate with the upper and lower tanks and serve to convey the coolant through the uranium body. The reactor is also provided with means for circulating the cooling fluid through the coolant tanks and coolant tubes, suitable neutron and gnmma ray shields, and control means.

Wigner, E.P.; Weinberg, A.W.; Young, G.J.

1958-04-15T23:59:59.000Z

343

Naval Petroleum and Oil Shale Reserves. Annual report of operations, Fiscal year 1993  

Science Conference Proceedings (OSTI)

During fiscal year 1993, the reserves generated $440 million in revenues, a $33 million decrease from the fiscal year 1992 revenues, primarily due to significant decreases in oil and natural gas prices. Total costs were $207 million, resulting in net cash flow of $233 million, compared with $273 million in fiscal year 1992. From 1976 through fiscal year 1993, the Naval Petroleum and Oil Shale Reserves generated $15.7 billion in revenues for the US Treasury, with expenses of $2.9 billion. The net revenues of $12.8 billion represent a return on costs of 441 percent. See figures 2, 3, and 4. In fiscal year 1993, production at the Naval Petroleum and Oil Shale Reserves at maximum efficient rates yielded 25 million barrels of crude oil, 123 billion cubic feet of natural gas, and 158 million gallons of natural gas liquids. The Naval Petroleum and Oil Shale Reserves has embarked on an effort to identify additional hydrocarbon resources on the reserves for future production. In 1993, in cooperation with the US Geological Survey, the Department initiated a project to assess the oil and gas potential of the program`s oil shale reserves, which remain largely unexplored. These reserves, which total a land area of more than 145,000 acres and are located in Colorado and Utah, are favorably situated in oil and gas producing regions and are likely to contain significant hydrocarbon deposits. Alternatively the producing assets may be sold or leased if that will produce the most value. This task will continue through the first quarter of fiscal year 1994.

Not Available

1993-12-31T23:59:59.000Z

344

Naval petroleum and oil shale reserves: Annual report of operations, Fiscal Year 1986  

Science Conference Proceedings (OSTI)

Market prices for curde oil experienced their greatest decline in history during 1986, with substantial effect on the Naval Petroleum and Oil Shale Reserves. Sales procedures which had served the Department well in prior years during periods when oil prices were stable or rising were found inadequate to cope with these declines, and new sales procedures were developed and implemented. Congressional concern that the Government receive fair prices from Naval Petroleum Reserves (NPR) petroleum sales also led to Public Law No. 99-413, enacted on August 29, 1986, which amended title 10, United States Code, Chapter 641. This law sets a minimum price, using two criteria, under which petroleum from the NPR cannot be sold. Notwithstanding the decline in oil and gas prices, revenues from the sale of NPR petroleum totaled $763 million in 1986. Although this represented a 41% decline from 1985 receipts, the 1986 revenues well exceeded net program expenditures of $157 million. Because of the decline in prevailing oil prices during the second quarter of 1986, major cost reductions of $3.4 million were implemented at NPR-3, and even greater reductions are planned for 1987 to ensure the profitability of that field. The decline in energy prices also affected plans to protect natural gas underlying Naval Oil Shale Reserve No. 3 (NOSR-3). Two of 14 planned wells were drilled and are available for production. Unfortunately, efforts to sell gas from these wells have been unsuccessful, and further drilling has been delayed until the gas market is stronger. 16 figs., 20 tabs.

Not Available

1987-01-01T23:59:59.000Z

345

Power Burst Facility (PBF) Reactor Reactor Decommissioning  

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

Reactor Decommissioning Click here to view Click here to view Reactor Decommissioning Click on an image to enlarge A crane removes the reactor vessel from the Power Burst Facility...

346

Rocky Mountain Oilfield Testing Center RMOTC at the Naval Petroleum Reserve No. 3  

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

RMOTC RMOTC The Rocky Mountain Oilfield Testing Center (RMOTC), is an operating oil field focusing on environmentally-balanced energy technologies and alternatives, and is the premiere energy testing and demonstration field in the nation. 3 3 * the opportunity to explore environmentally- balanced solutions to the nation's energy issues * opportunities to develop, demonstrate, and evaluate a variety of energy related technologies * a chance to collaborate with top professionals in the energy, environmental technology, and engineering fields * shared industry knowledge through technology transfer via reports, journal articles, and presentations Located within the Naval Petroleum Reserve No. 3 (NPR-3) near Casper, Wyoming, RMOTC offers: RMOTC Offers Solutions 4 4 The Administration and Engineering

347

Quality assurance assessment of new efficient lighting systems for Naval ships. Final report  

SciTech Connect

Ballasts and lamps, which have been selected to replace existing lamp/ballast systems based on improved performance, were tested to determine if they meet standard Naval MIL specifications. Fifty ballasts manufactured by Advance Transformer Corporation and Universal Manufacturing Corp., and 100 lamps manufactured by GTE were tested to determine their quality assurance and durability. These components met all of the MIL specifications that lamp/ballast systems in use must meet. In addition, these new systems have an improved system efficacy, 62 lumens per watt, and lower third harmonics, which will reduce the need for generating capacity for lighting on ships.

Verderber, R.

1984-08-01T23:59:59.000Z

348

France gets nuclear fusion plant France will get to host the project to build a 10bn-euro (6.6bn) nuclear fusion reactor, in  

E-Print Network (OSTI)

the construction costs will be shouldered by the EU. "We believe that the Iter project should start as soon energy programme in 1959. ITER - NUCLEAR FUSION PROJECT Project estimated to cost 10bn euros and will runFrance gets nuclear fusion plant France will get to host the project to build a 10bn-euro (£6.6bn

349

NRC proposes new reactor embrittlement regulations  

SciTech Connect

The Nuclear Regulatory Commission has proposed amending its regulations regarding embrittlement of pressure vessels in commercial nuclear power reactors. The amendments, which include clarifying the pressurized thermal shock requirements, changing the fracture toughness requirements and establishing new requirements for thermal annealing of a reactor pressure vessel, were prompted in part by NRC`s 1990 inspection of the reactor pressure vessel at the Yankee Rowe nuclear plant in Massachusetts. NRC`s proposed changes would establish a screening criterion for both the rate of irradiation embrittlement or RPV materials and fracture toughness, above which a plant cannot continue to be operated safely.

Newman, P.

1994-10-13T23:59:59.000Z

350

A pilot plant scale reactor/separator for ethanol from cellulosics. Quarterly report No. 1 & 2, October 1, 1997--March 30, 1998  

SciTech Connect

The basic objective of this project is to develop and demonstrate a continuous, low energy process for the conversion of cellulosics to ethanol. This process involves a pretreatment step followed by enzymatic release of sugars and the consecutive saccharification/fermentation of cellulose (glucans) followed by hemi-cellulose (glucans) in a multi-stage continuous stirred reactor separator (CSRS). During year 1, pretreatment and bench scale fermentation trials will be performed to demonstrate and develop the process, and during year 2, a 130 L or larger process scale unit will be operated to demonstrate the process using straw or cornstalks. Co-sponsors of this project include the Indiana Biomass Grants Program, Bio-Process Innovation, Xylan Inc as a possible provider of pretreated biomass.

Dale, M.C.

1998-06-01T23:59:59.000Z

351

ALPR PRELIMINARY DESIGN STUDY (ARGONNE LOW POWER REACTOR). PHASE I  

SciTech Connect

A preliminary design study, Phase I of the ALPR . project, has been made in accordance with the Army Reactors Branch specifications for a nuclear ''package'' power plant with a 200 to 260 kw electric and 400 kw heating capacity..The plant is to be installed at the Idaho Reactor Testing Station as a prototype for remote arctic installations. The ''conventiornl'' power plant as well as the exterior reactor components are described, and cost estimates are given. ''Nuclear'' components of the reactor are described. (auth)

Treshow, M.; Hamer, E.; Pearlman, H.; Rossin, D.; Shaftman, D.

1956-04-20T23:59:59.000Z

352

NEUTRONIC REACTOR  

DOE Patents (OSTI)

A reactor is described comprising a plurality of horizontal trays containing a solution of a fissionable material, the trays being sleeved on a vertical tube which contains a vertically-reciprocable control rod, a gas-tight chamber enclosing the trays, and means for conducting vaporized moderator from the chamber and for replacing vaporized moderator in the trays. (AEC)

Wigner, E.P.

1962-12-25T23:59:59.000Z

353

Neutronic reactor  

DOE Patents (OSTI)

A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield.

Wende, Charles W. J. (West Chester, PA)

1976-08-17T23:59:59.000Z

354

NUCLEAR REACTOR  

DOE Patents (OSTI)

A nuclear reactor is described that includes spaced vertical fuel elements centrally disposed in a pressure vessel, a mass of graphite particles in the pressure vessel, means for fluidizing the graphite particles, and coolant tubes in the pressure vessel laterally spaced from the fuel elements. (AEC)

Post, R.G.

1963-05-01T23:59:59.000Z

355

NUCLEAR REACTOR  

DOE Patents (OSTI)

This patent relates to a combination useful in a nuclear reactor and is comprised of a casing, a mass of graphite irapregnated with U compounds in the casing, and at least one coolant tube extending through the casing. The coolant tube is spaced from the mass, and He is irtroduced irto the space between the mass and the coolant tube. (AEC)

Starr, C.

1963-01-01T23:59:59.000Z

356

NEUTRONIC REACTOR  

DOE Patents (OSTI)

BS>A reactor cooled by water, biphenyl, helium, or other fluid with provision made for replacing the fuel rods with the highest plutonium and fission product content without disassembling the entire core and for promptly cooling the rods after their replacement in order to prevent build-up of heat from fission product activity is described.

Creutz, E.C.; Ohlinger, L.A.; Weinberg, A.M.; Wigner, E.P.; Young, G.J.

1959-10-27T23:59:59.000Z

357

NEUTRONIC REACTORS  

DOE Patents (OSTI)

The design of control rods for nuclear reactors are described. In this design the control rod consists essentially of an elongated member constructed in part of a neutron absorbing material and having tube means extending therethrough for conducting a liquid to cool the rod when in use.

Anderson, H.L.

1958-10-01T23:59:59.000Z

358

Daya Bay Reactor Neutrino Project at NERSC  

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

Daya Bay Reactor Neutrino Daya Bay Reactor Neutrino Experiment Daya Bay Reactor Neutrino Experiment Daya Bay is an international neutrino-oscillation experiment designed to determine the last unknown neutrino mixing angle θ13 using anti-neutrinos produced by the Daya Bay and Ling Ao Nuclear Power Plant reactors. The experiment is being built by blasting three kilometers of tunnel through the granite rock under the mountains where the power plants are located. Data collection is now scheduled to start in in 2011. On the PDSF cluster at NERSC, Daya Bay performs simulations of the detectors, reactors, and surrounding mountains to help design and anticipate detector properties and behavior. Once real data are available, Daya Bay will be using NERSC to analyze data and NERSC HPSS will be the central U.S. repository for all raw

359

This document was downloaded on May 22, 2013 at 14:29:32 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

- graduate School in effect, a naval univer- sity, unified in policies, procedures and ob- jectives. In 1973;GENERAL INFORMATION ADMISSIONS PROCEDURES U.S. NAVAL OFFICERS U.S. Navy officers interested in admis- sion to availability of quotas assigned to each country. The procedures for application are contained in OPNAV

360

This document was downloaded on May 22, 2013 at 14:24:38 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

Organizations 18 Student Council 18 Selection Procedures 19 Naval Officers 19 Other U.S. Military Officers 19 in matters involving curricula, facilities, procedures and policies deemed worthy of attention. The OSAC. SELECTION PROCEDURES NAVAL OFFICERS Selection for the Navy fully funded graduate education program is based

Note: This page contains sample records for the topic "naval reactor plants" 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

This document was downloaded on May 22, 2013 at 14:25:15 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

20 Selection Procedures 20 Naval Officers 20 Other U.S. Military Officers 20 International Military administration. It functions in an advisory capacity in matters involving curricula, facilities, procedures graduate education program. SELECTION PROCEDURES NAVAL OFFICERS Selection for the Navy fully funded

362

This document was downloaded on May 22, 2013 at 14:23:11 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

Center for Executive Education 11 NPS Curricula Summary 13 ADMISSIONS 16 Selection Procedures 16 Naval SELECTION PROCEDURES NAVAL OFFICERS Selection for the Navy fully funded graduate education program is based may be admitted to most curricula. The procedures for application are available from the Security

363

This document was downloaded on May 22, 2013 at 14:27:00 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

15 Faculty Organizations 16 Student Council 17 Admissions 18 Selection Procedures 18 Naval Officers, procedures and policies deemed worthy of attention. The OSAC is comprised of thirty-five student for the Navy's fully funded graduate education program. SELECTION PROCEDURES NAVAL OFFICERS Selection

364

This document was downloaded on May 22, 2013 at 14:23:43 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

Organizations 17 Student Council 17 ADMISSIONS 18 Selection Procedures 18 Naval Officers 18 Other U.S. Military. It functions in an advisory capacity in matters involving curricula, facilities, procedures and policies deemed funded graduate education program. SELECTION PROCEDURES NAVAL OFFICERS Selection for the Navy fully

365

This document was downloaded on May 22, 2013 at 14:24:49 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

Council 18 Selection Procedures 19 Naval Officers 19 Other U.S. Military Officers 19 International involving curricula, facilities, procedures and policies deemed worthy of attention. The OSAC is comprised. SELECTION PROCEDURES NAVAL OFFICERS Selection for the Navy fully funded graduate education program is based

366

This document was downloaded on May 22, 2013 at 14:24:27 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

Selection Procedures 20 Naval Officers 20 Other U.S. Military Officers 20 International Military Officers 20 administration. It functions in an advisory capacity in matters involving curricula, facilities, procedures funded graduate education program. SELECTION PROCEDURES NAVAL OFFICERS Selection for the Navy fully

367

CIVILIAN POWER REACTOR PROGRAM. PART III. STATUS REPORT ON GAS-COOLED REACTORS AS OF 1959. Book 8  

SciTech Connect

The technology of natural-uranium-fueled graphitemoderated gas-cooled reactor power plants is summarized for its relevance to the technology of enriched-fuel graphite-moderated systems. The technology of D/sub 2/Omoderated gas-cooled reactors is also summarized. Estimated technical performance parameters are given for the enriched-fuel prototype and for a large natural- uraniumfueled plant. Current technical status is discussed in terms of reactor physics, heat transfer and fluid flow, core materials, components, plant design and conctruction, and hazards. Detailed tables of characteristics for various reactors are given. An extensive bibliography is included. (W.D.M.)

1960-01-01T23:59:59.000Z

368

Work Breakdown Structure and Plant/Equipment Designation System Numbering Scheme for the High Temperature Gas- Cooled Reactor (HTGR) Component Test Capability (CTC)  

SciTech Connect

This white paper investigates the potential integration of the CTC work breakdown structure numbering scheme with a plant/equipment numbering system (PNS), or alternatively referred to in industry as a reference designation system (RDS). Ideally, the goal of such integration would be a single, common referencing system for the life cycle of the CTC that supports all the various processes (e.g., information, execution, and control) that necessitate plant and equipment numbers be assigned. This white paper focuses on discovering the full scope of Idaho National Laboratory (INL) processes to which this goal might be applied as well as the factors likely to affect decisions about implementation. Later, a procedure for assigning these numbers will be developed using this white paper as a starting point and that reflects the resolved scope and outcome of associated decisions.

Jeffrey D Bryan

2009-09-01T23:59:59.000Z

369

Uncertainties in the Anti-neutrino Production at Nuclear Reactors  

Science Conference Proceedings (OSTI)

Anti-neutrino emission rates from nuclear reactors are determined from thermal power measurements and fission rate calculations. The uncertainties in these quantities for commercial power plants and their impact on the calculated interaction rates in {bar {nu}}{sub e} detectors is examined. We discuss reactor-to-reactor correlations between the leading uncertainties, and their relevance to reactor {bar {nu}}{sub e} experiments.

Djurcic, Zelimir; Detwiler, Jason A.; Piepke, Andreas; Foster Jr., Vince R.; Miller, Lester; Gratta, Giorgio

2008-08-06T23:59:59.000Z

370

INSIGHTS INTO THE ROLE OF THE OPERATOR IN ADVANCED REACTORS.  

SciTech Connect

NUCLEAR POWER PLANT PERSONNEL PLAY A VITAL ROLE IN THE PRODUCTIVE, EFFICIENT, AND SAFE GENERATION OF ELECTRIC POWER, WHETHER FOR CONVENTIONAL LIGHT WATER REACTORS OR NEW ADVANCED REACTORS. IT IS WIDELY RECOGNIZED THAT HUMAN ACTIONS THAT DEPART FROM OR FAIL TO ACHIEVE WHAT SHOULD BE DONE CAN BE IMPORTANT CONTRIBUTORS TO THE RISK ASSOCIATED WITH THE OPERATION OF NUCLEAR POWER PLANTS. ADVANCED REACTORS ARE EXPECTED TO PRESENT A CONCEPT OF OPERATI...

PERSENSKY, J.; LEWIS, P.; O' HARA, J.

2005-11-13T23:59:59.000Z

371

Rapid starting methanol reactor system  

DOE Patents (OSTI)

The invention relates to a methanol-to-hydrogen cracking reactor for use with a fuel cell vehicular power plant. The system is particularly designed for rapid start-up of the catalytic methanol cracking reactor after an extended shut-down period, i.e., after the vehicular fuel cell power plant has been inoperative overnight. Rapid system start-up is accomplished by a combination of direct and indirect heating of the cracking catalyst. Initially, liquid methanol is burned with a stoichiometric or slightly lean air mixture in the combustion chamber of the reactor assembly. The hot combustion gas travels down a flue gas chamber in heat exchange relationship with the catalytic cracking chamber transferring heat across the catalyst chamber wall to heat the catalyst indirectly. The combustion gas is then diverted back through the catalyst bed to heat the catalyst pellets directly. When the cracking reactor temperature reaches operating temperature, methanol combustion is stopped and a hot gas valve is switched to route the flue gas overboard, with methanol being fed directly to the catalytic cracking reactor. Thereafter, the burner operates on excess hydrogen from the fuel cells.

Chludzinski, Paul J. (38 Berkshire St., Swampscott, MA 01907); Dantowitz, Philip (39 Nancy Ave., Peabody, MA 01960); McElroy, James F. (12 Old Cart Rd., Hamilton, MA 01936)

1984-01-01T23:59:59.000Z

372

Economic analysis of nuclear reactors  

SciTech Connect

The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U/sub 3/O/sub 8/ is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented.

Owen, P.S.; Parker, M.B.; Omberg, R.P.

1979-05-01T23:59:59.000Z

373

IMPROVEMENTS IN OR RELATING TO STEAM GENERATING PLANT  

SciTech Connect

A nuclear power plant is designed using a heavy-watermoderated, steam- cooled reactor. In this plant, feed water is heated by the moderator and reactor steam to form feed steam, which is then superheated by superheated reactor steam and expanded through a nozzle. The feed steam issuing from the nozzie has added to it the superheated reactor steam, and the resulting steam is compressed, heated further in the reactor, and part of it passed to the turbine. (D.L.C.)

Bauer, S.G.; Jubb, D.H.

1962-10-10T23:59:59.000Z

374

X-10 Graphite Reactor | Department of Energy  

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

X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor X-10 Graphite Reactor When President Roosevelt in December 1942 authorized the Manhattan Project, the Oak Ridge site in eastern Tennessee had already been obtained and plans laid for an air-cooled experimental pile, a pilot chemical separation plant, and support facilities. The X-10 Graphite Reactor, designed and built in ten months, went into operation on November 4, 1943. The X-10 used neutrons emitted in the fission of uranium-235 to convert uranium-238 into a new element, plutonium-239. The reactor consists of a huge block of graphite, measuring 24 feet on each side, surrounded by several feet of high-density concrete as a radiation shield. The block is pierced by 1,248 horizontal diamond-shaped channels in

375

NUCLEAR POWER PLANT  

DOE Patents (OSTI)

A nuclear power plant for use in an airless environment or other environment in which cooling is difficult is described. The power plant includes a boiling mercury reactor, a mercury--vapor turbine in direct cycle therewith, and a radiator for condensing mercury vapor. (AEC)

Carter, J.C.; Armstrong, R.H.; Janicke, M.J.

1963-05-14T23:59:59.000Z

376

Economic Analysis of the Modular Pebble Bed Reactor  

E-Print Network (OSTI)

$) Reactor Thermal Power (MWt) 10 x 250 Net Efficiency (%) 45.3% Net Electrical Rating (Mwe) 1100 Capacity Turbomachinery Ten-Unit MPBR Plant Layout (Top View) (distances in meters) Equip Access Hatch Equip Access Hatch for 1100 MWe plant $2,296 million #12;Plant Construction · Construction Plan / Techniques · Plant Physical

377

Light Water Reactor Sustainability Technical Documents | Department of  

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

Reactor Technologies » Light Water Reactor Reactor Technologies » Light Water Reactor Sustainability Program » Light Water Reactor Sustainability Technical Documents Light Water Reactor Sustainability Technical Documents April 30, 2013 LWRS Program and EPRI Long-Term Operations Program - Joint R&D Plan To address the challenges associated with pursuing commercial nuclear power plant operations beyond 60 years, the U.S. Department of Energy's (DOE) Office of Nuclear Energy (NE) and the Electric Power Research Institute (EPRI) have established separate but complementary research and development programs: DOE-NE's Light Water Reactor Sustainability (LWRS) Program and EPRI's Long-Term Operations (LTO) Program. April 30, 2013 Light Water Reactor Sustainability Program - Integrated Program Plan The Light Water Reactor Sustainability (LWRS) Program is a research and

378

INITIAL TESTING AND OPERATION OF THE ARGONNE LOW POWER REACTOR (ALPR)  

SciTech Connect

The major events of a program designed to test and operate the completed reactor power plant and associated equipment are described. The design and construction phases of the project, component installation, preliminary systems testing, zero-power experiments, areas affected by the design parameters, reactor operation, plant safety, and reactor operator training are covered. (W.D.M.)

Hamer, E.E. ed.

1959-12-01T23:59:59.000Z

379

REACTOR UNLOADING  

DOE Patents (OSTI)

This patent is related to gas cooled reactors wherein the fuel elements are disposed in vertical channels extending through the reactor core, the cooling gas passing through the channels from the bottom to the top of the core. The invention is a means for unloading the fuel elements from the core and comprises dump values in the form of flat cars mounted on wheels at the bottom of the core structure which support vertical stacks of fuel elements. When the flat cars are moved, either manually or automatically, for normal unloading purposes, or due to a rapid rise in the reproduction ratio within the core, the fuel elements are permtted to fall by gravity out of the core structure thereby reducing the reproduction ratio or stopping the reaction as desired.

Leverett, M.C.

1958-02-18T23:59:59.000Z

380

NUCLEAR REACTOR  

DOE Patents (OSTI)

A neuclear reactor is described of the heterogeneous type and employing replaceable tubular fuel elements and heavy water as a coolant and moderator. A pluraltty of fuel tubesa having their axes parallel, extend through a tank type pressure vessel which contatns the liquid moderator. The fuel elements are disposed within the fuel tubes in the reaetive portion of the pressure vessel during normal operation and the fuel tubes have removable plug members at each end to permit charging and discharging of the fuel elements. The fuel elements are cylindrical strands of jacketed fissionable material having helical exterior ribs. A bundle of fuel elements are held within each fuel tube with their longitudinal axes parallel, the ribs serving to space them apart along their lengths. Coolant liquid is circulated through the fuel tubes between the spaced fuel elements. Suitable control rod and monitoring means are provided for controlling the reactor.

Treshow, M.

1958-08-19T23:59:59.000Z

Note: This page contains sample records for the topic "naval reactor plants" 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

Neutronic reactor  

DOE Patents (OSTI)

A graphite-moderated, water-cooled nuclear reactor including a plurality of rectangular graphite blocks stacked in abutting relationship in layers, alternate layers having axes which are normal to one another, alternate rows of blocks in alternate layers being provided with a channel extending through the blocks, said channeled blocks being provided with concave sides and having smaller vertical dimensions than adjacent blocks in the same layer, there being nuclear fuel in the channels.

Lewis, Warren R. (Richland, WA)

1978-05-30T23:59:59.000Z

382

NUCLEAR REACTORS  

DOE Patents (OSTI)

An active portion assembly for a fast neutron reactor is described wherein physical distortions resulting in adverse changes in the volume-to-mass ratio are minimized. A radially expandable locking device is disposed within a cylindrical tube within each fuel subassembly within the active portion assembly, and clamping devices expandable toward the center of the active portion assembly are disposed around the periphery thereof. (AEC)

Koch, L.J.; Rice, R.E. Jr.; Denst, A.A.; Rogers, A.J.; Novick, M.

1961-12-01T23:59:59.000Z

383

REACTOR CONTROL  

DOE Patents (OSTI)

This patent relates to nuclear reactors of the type which utilize elongited rod type fuel elements immersed in a liquid moderator and shows a design whereby control of the chain reaction is obtained by varying the amount of moderator or reflector material. A central tank for containing liquid moderator and fuel elements immersed therein is disposed within a surrounding outer tank providing an annular space between the two tanks. This annular space is filled with liquid moderator which functions as a reflector to reflect neutrons back into the central reactor tank to increase the reproduction ratio. Means are provided for circulating and cooling the moderator material in both tanks and additional means are provided for controlling separately the volume of moderator in each tank, which latter means may be operated automatically by a neutron density monitoring device. The patent also shows an arrangement for controlling the chain reaction by injecting and varying an amount of poisoning material in the moderator used in the reflector portion of the reactor.

Ruano, W.J.

1957-12-10T23:59:59.000Z

384

Light Water Reactors Technology Development - Nuclear Reactors  

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

Light Water Reactors Light Water Reactors About Director's Welcome Organization Achievements Highlights Fact Sheets, Brochures & Other Documents Multimedia Library Visit Argonne Work with Argonne Contact us Nuclear Energy Why Nuclear Energy? Why are some people afraid of Nuclear Energy? How do nuclear reactors work? Cheaper & Safer Nuclear Energy Helping to Solve the Nuclear Waste Problem Nuclear Reactors Nuclear Reactors Early Exploration Training Reactors Basic and Applied Science Research LWR Technology Development BORAX-III lighting Arco, Idaho (Press Release) Heavy Water and Graphite Reactors Fast Reactor Technology Integral Fast Reactor Argonne Reactor Tree CP-1 70th Anniversary CP-1 70th Anniversary Argonne's Nuclear Science and Technology Legacy Argonne's Nuclear Science and Technology Legacy

385

GAS COOLED NUCLEAR REACTOR STUDY. Final Report  

SciTech Connect

An investigntion was made of the performance of a gas-cooled reactor, designed to provide a source of high temperature heat to a stream of helium. This reactor, in turn, is used as a source of heat for the air stream in a gas- turbine power plant. The reactor design was predicted primarily on the requirement for transferring a large amount of heat to the helium stream with a pressure drop low enough that it will not represent a major loss of power in the power plant. The mass of uranium e uired far criticality under various circumstances was investigated by multigroup calculations, both on desk calculators and on an IBM-704 machine. The gasturbine power plant perfarmance was studied based on a Studebaker-Packard-designed gas-turbine power plant for the propulsion of destroyer-escort vessels. A small experimental program was carried out to study some effects of helium on graphite and on structural steels. (auth)

Thompson, A.S.

1956-07-31T23:59:59.000Z

386

ESBWR... An Evolutionary Reactor Design  

Science Conference Proceedings (OSTI)

GE's latest evolution of the Boiling Water Reactor, the ESBWR, combines improvements in safety with design simplification and component standardization to produce a safer, more reliable nuclear power plant, with lower projected construction costs than plants in operation today. The ESBWR program started in the early 1990's when GE was developing the Simplified Boiling Water Reactor (SBWR). GE stopped this program because the power output of the SBWR was too small to generate the right economics for a new build project. The program was a success however, because the design proved many of the passive safety technology developments that are being utilized in the ESBWR. By harnessing these design concepts and testing results from the original SBWR and construction and operating experience from the Advanced Boiling Water Reactor (ABWR), the ESBWR design team has produced a simplified reactor with a standardized design and first-rate economics. Significant simplification of plant systems is achieved in the ESBWR. As a result, operating and maintenance staff requirements are reduced; low-level waste generation is reduced; dose rates are reduced; operational reliability is improved; and plant safety and security are improved. Each of these improvements provide distinct and unique advantages to the ESBWR design. First, fewer active components (in particular, active safety systems) reduce the maintenance and online surveillance requirements, thereby reducing operational exposure and dose rates. Second, fewer demands on plant operators and safety systems reduce plant operating staff while still providing direct improvements in accident and transient response. Finally, reductions in building volumes and required manufactured components shorten the length of time needed for ESBWR construction, resulting in improved financial returns for plant owners. The ESBWR is designed to meet the needs of nuclear power plant owners today and into the future, with a 60-year design life. Through design simplification and standardization, ESBWR offers improved safety, increased reliability, and ease of operation. Yet compared to current nuclear power plants, the ESBWR requires only a fraction of traditional plant operating and maintenance staff, offers faster construction and lower costs of construction, while also reducing operational costs. (authors)

Gamble, Robert E. [GE Energy - Nuclear, 1989 Little Orchard St., San Jose, CA 95125 (United States); Hinds, David H.; Hucik, Steven A.; Maslak, Chris E. [GE Energy - Nuclear, 3901 Castle Hayne Road A-30, Wilmington, NC 28402 (United States)

2006-07-01T23:59:59.000Z

387

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

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.

1994-06-01T23:59:59.000Z

388

Operational safety enhancement of Soviet-designed nuclear reactors via development of nuclear power plant simulators and transfer of related technology  

SciTech Connect

The US Department of Energy (DOE), under the US government`s International Nuclear Safety Program (INSP), is implementing a program of developing and providing simulators for many of the Russian and Ukrainian Nuclear Power Plants (NPPs). Pacific Northwest National Laboratory (PNNL) and Brookhaven National Laboratory (BNL) manage and provide technical oversight of the various INSP simulator projects for DOE. The program also includes a simulator technology transfer process to simulator design organizations in Russia and Ukraine. Training programs, installation of new simulators, and enhancements in existing simulators are viewed as providing a relatively fast and cost-effective technology transfer that will result in measurable improvement in the safety culture and operation of NPPs. A review of this program, its present status, and its accomplishments are provided in this paper.

Kohut, P.; Epel, L.G.; Tutu, N.K. [and others

1998-08-01T23:59:59.000Z

389

Light Water Reactor Sustainability (LWRS) Program Non-Destructive Evaluation (NDE) R&D Roadmap for Determining Remaining Useful Life of Aging Cables in Nuclear Power Plants  

SciTech Connect

The purpose of the non-destructive evaluation (NDE) R&D Roadmap for Cables is to support the Materials Aging and Degradation (MAaD) R&D pathway. The focus of the workshop was to identify the technical gaps in detecting aging cables and predicting their remaining life expectancy. The workshop was held in Knoxville, Tennessee, on July 30, 2012, at Analysis and Measurement Services Corporation (AMS) headquarters. The workshop was attended by 30 experts in materials, electrical engineering, U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy (DOE) National Laboratories (Oak Ridge National Laboratory, Pacific Northwest National Laboratory, Argonne National Laboratory, and Idaho National Engineering Laboratory), NDE instrumentation development, universities, commercial NDE services and cable manufacturers, and Electric Power Research Institute (EPRI). The motivation for the R&D roadmap comes from the need to address the aging management of in-containment cables at nuclear power plants (NPPs).

Simmons, Kevin L.; Ramuhalli, Pradeep; Brenchley, David L.; Coble, Jamie B.; Hashemian, Hash; Konnik, Robert; Ray, Sheila

2012-09-14T23:59:59.000Z

390

Development of Technologies on Innovative Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors (10) Application to a Small District-Heating Reactor  

SciTech Connect

A steam injector (SI) is a simple, compact and passive pump and also acts as a high-performance direct-contact compact heater. This provides SI with capability to use as a passive ECCS pump and also as a direct-contact feedwater heater that heats up feedwater by using extracted steam from the turbine. In order to develop a high reliability passive ECCS pump and a compact feedwater heater, it is necessary to quantify the characteristics between physical properties of the flow field. We carried out experiments to observe the internal behavior of the water jet as well as measure the velocity of steam jet using a laser Doppler velocimetry. Its performance depends on the phenomena of steam condensation onto the water jet surface and heat transfer in the water jet due to turbulence on to the phase-interface. The analysis was also conducted by using a CFD code with the separate two-phase flow models. With regard to the simplified feed-water system, size of four-stage SI system is almost the same as the model SI that had done the steam and water test that pressures were same as that of current ABWR. The authors also conducted the hot water supply system test in the snow for a district heating. With regard to the SI core cooling system, the performance tests results showed that the low-pressure SI core cooling system will decrease the PCT to almost the same as the saturation temperature of the steam pressure in a pressure vessel. As it is compact equipment, SI is expected to bring about great simplification and materials-saving effects, while its simple structure ensures high reliability of its operation, thereby greatly contributing to the simplification of the power plant for not only an ABWR power plant but also a small PWR/ BWR for district heating system. (authors)

Tadashi Narabayashi; Yoichiro Shimadu; Toshiiro Murase; Masatoshi Nagai [Hokkaido University, Kita-ku, Sapporo (Japan); Michitsugu Mori; Shuichi Ohmori [Tokyo Electric Power Company (Japan)

2006-07-01T23:59:59.000Z

391

CIVILIAN POWER REACTOR PROGRAM. PART III. BOOK I. STATUS REPORT ON FAST REACTORS AS OF 1959  

SciTech Connect

A description of fast reactor types is given and the objectives of the AEC program on fast reactors are outlined. General research and development programs completed and in process on physics, fuel and materials, heat transfer, fluid flow, coolant chemistry, reactor safety, and components and systems are discussed. Reactors completed, under design and construction, and under study are described. (W.D.M.) A general description of ship and power plant, tabulation of principle characteristics, economics, potential advances, and necessary research and development are discussed in detail. (W.D.M.)

1960-01-01T23:59:59.000Z

392

Graphite Reactor | ornl.gov  

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

Graphite Reactor Graphite Reactor 'In the early, desperate days of World War II, the United States launched the top-secret, top-priority Manhattan Project...' In the early, desperate days of U.S. involvement in World War II, American scientists began to fear that the German discovery of uranium fission in 1939 might enable the Nazis to develop a super bomb. Afraid of losing this crucial race, the United States launched the top-secret, top-priority Manhattan Project. The plan was to create two atomic weapons-one fueled by plutonium, the other by enriched uranium. Hanford, Washington, was selected as the site for plutonium production, but before large reactors could be built there, a pilot plant was necessary to prove the feasibility of scaling up from laboratory experiments. A secluded, rural area near Clinton, Tennessee, was

393

Nuclear reactor  

DOE Patents (OSTI)

A nuclear reactor in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assmblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters and the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters and the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance.

Pennell, William E. (Greensburg, PA); Rowan, William J. (Monroeville, PA)

1977-01-01T23:59:59.000Z

394

ELECTRONUCLEAR REACTOR  

DOE Patents (OSTI)

An electronuclear reactor is described in which a very high-energy particle accelerator is employed with appropriate target structure to produce an artificially produced material in commercial quantities by nuclear transformations. The principal novelty resides in the combination of an accelerator with a target for converting the accelerator beam to copious quantities of low-energy neutrons for absorption in a lattice of fertile material and moderator. The fertile material of the lattice is converted by neutron absorption reactions to an artificially produced material, e.g., plutonium, where depleted uranium is utilized as the fertile material.

Lawrence, E.O.; McMillan, E.M.; Alvarez, L.W.

1960-04-19T23:59:59.000Z

395

Photocatalytic reactor  

DOE Patents (OSTI)

A photocatalytic reactor for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane.

Bischoff, Brian L. (Knoxville, TN); Fain, Douglas E. (Oak Ridge, TN); Stockdale, John A. D. (Knoxville, TN)

1999-01-01T23:59:59.000Z

396

Passive cooling system for top entry liquid metal cooled nuclear reactors  

SciTech Connect

This patent describes a passive cooling system for liquid metal cooled, top entry loop nuclear fission reactors. It comprises: a liquid metal cooled nuclear reactor plant; a passive cooling system; and a secondary passive cooling system.

Boardman, C.E.; Hunsbedt, A.; Hui, M.M.

1992-10-27T23:59:59.000Z

397

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

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

Request for Records Disposition Authority Leave Blank (NARA Use Only) (See Instructions on reverse) Job Number I / {£. 0- _~ To. National Archives and Records Administration (NIR) NI-'-r 3 7- 6 6 J Washington, DC 20408 Date Received 1 From (Agencyor establishment) Department of Energy Notification to Agency 2 MajorSubdivrsion In accordance with the provisions of 44 Assistant Secretary for Fossil Energy USC 3303a. the disposition request. In- cluding amendments. ISapproved except for 3 Minorsubcrvrsron 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 Telephone (Includearea code) [ Pamela Gentel 301-903-1856 6 Agency Certification

398

Major General Harold Holesinger The Adjutant General Illinois Kilitary and Naval Dept.  

Office of Legacy Management (LM)

General Harold Holesinger General Harold Holesinger The Adjutant General Illinois Kilitary and Naval Dept. 1301 North MacArthur Boulevard Springfield, Illinois 62702-2399 Dear General Holesinger: I am enclosing a copy of the radiological survey report for the National Guard Armory, Chicago, Illinois. Although the data ncted in the report indicate levels of radioactivity in excess of current guidelines, the radioactive residues presently there do not pose a health hazard provided they were not disturbed in the past and will not be disturbed in the future; i.e., no excavation, building, or construction that would disturb the areas in which contamination was found. Based on these data and on an authority record review, the National Guard Armory property is being authorized for remedial action and will be

399

Rocky Mountain Oilfield Testing Center RMOTC at the Naval Petroleum Reserve No. 3  

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

OILFIELD TESTING OILFIELD TESTING OILFIELD TESTING CENTER CENTER 2 2 HISTORY OF TEAPOT DOME Mark Milliken 3 3 TEAPOT DOME LOCATION 4 4 Salt Creek 670 MMBBLS 722 BCF Teapot Dome 27 MMBBLS 57 BCF N P R - 3 Cumulative Production 5 5 The Great White Fleet December 1907 - February 1909 6 6 THE END OF COAL-FIRED SHIPS * 2-week cruising time. * Labor and time intensive cleaning and reloading. * At the mercy of foreign countries for coal supply. * 1912: All battleships will be oil-powered. Great White Fleet 7 7 NAVAL PETROLEUM RESERVES ARE BORN * 1908: Dr. Otis Smith, USGS Director, recommends DOI retain oil lands for fuel reserve for Navy. * 1909: Taft withdraws 3,000,000 acres in Wyoming and California. * 1910: Concern over the President's authority to withdraw lands, so Congress passed the Pickett Act. * 1910: Taft issues a 2nd land withdrawal executive

400

FEMP ESPC Success Story - U.S. Naval Station, Guantanamo Bay, Cuba  

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

a m a m ESPC Success Stories Environmental Stewardship and Cost Savings These photographs chronicle the installation of the wind turbines at John Paul Jones Hill, Guantanamo Bay. The four wind turbine towers are about 185 feet high. The blade lengths are 90 feet. The top of the blades are about 275 feet off the g round. The blades rotate at a maximum of 22 RPM, or a rotation every three seconds. This translates to a blade tip speed of 140 mph. During construction there were as many as 20 workers on the project. However, operating the wind turbines will only take one part-time staff-person who will check on them daily. Photos courtesy of: Jeffrey M. Johnston, Public Works Officer, Guantanamo Bay; Paul DelSignore, NFESC; Daniel Ingold, NORESCO. U.S. NAVAL STATION

Note: This page contains sample records for the topic "naval reactor plants" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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401

DOE - Office of Legacy Management -- Naval Supply Depot AEC Warehouse - NY  

Office of Legacy Management (LM)

Supply Depot AEC Warehouse - Supply Depot AEC Warehouse - NY 36 FUSRAP Considered Sites Site: NAVAL SUPPLY DEPOT, AEC WAREHOUSE (NY.36) Eliminated from further consideration under FUSRAP - Referred to DOD Designated Name: Not Designated Alternate Name: None Location: Building 546 , Scotia , New York NY.36-1 Evaluation Year: 1987 NY.36-1 Site Operations: This facility served as a storage and transshipment point for feed materials between the Hanford and commercial metal fabricators in the northeastern states. NY.36-1 NY.36-2 NY.36-3 Site Disposition: Eliminated - Referred to DOD NY.36-1 Radioactive Materials Handled: Yes Primary Radioactive Materials Handled: Uranium, Thorium Metals NY.36-1 NY.36-2 NY.36-3 Radiological Survey(s): None Indicated Site Status: Eliminated from further consideration under FUSRAP - Referred to DOD NY.36-1

402

Rocky Mountain Oilfield Testing Center RMOTC at the Naval Petroleum Reserve No. 3  

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

& Renewable Energy Scoping Meeting & Renewable Energy Scoping Meeting March 26, 2004 2 2 RMOTC The Rocky Mountain Oilfield Testing Center (RMOTC), is an operating oil field focusing on environmentally-balanced energy technologies and alternatives, and is the premiere energy testing and demonstration field in the nation. 3 3 * the opportunity to explore environmentally- balanced solutions to the nation's energy issues * opportunities to develop, demonstrate, and evaluate a variety of energy related technologies * a chance to collaborate with top professionals in the energy, environmental technology, and engineering fields * shared industry knowledge through technology transfer via reports, journal articles, and presentations Located within the Naval Petroleum Reserve No. 3 (NPR-3) near Casper, Wyoming, RMOTC offers:

403

Quality assurance assessment of new efficient lighting systems for naval ships. Final report. Revision  

SciTech Connect

Ballasts and lamps, which have been selected to replace existing lamp/ballast systems based on improved performance, were tested to determine if they meet standard Naval MIL specifications. Fifty ballasts manufactured by Advance Transformer Corporation and Universal Manufacturing Corporation and 100 lamps manufactured by GTE were tested to determine their quality assurance and durability. These components met all of the MIL specifications that lamp/ballast systems in use must meet. These new systems have an improved system efficacy, 62 lumens per watt, and lower third harmonics, which will reduce the need for generating capacity for lighting on ships. An addendum to the original study describes the assessment of the performance of the advanced ballast system with a new F-17 fluorescent lamp. The results indicate the system performs reliably and shows increased efficacy. This new lamp/ballast system reduces the harmonic content to within Navy limits, and improves the power factor, resulting in a 50% reduction in line current.

Verderber, R.R.; Morse, O.; Dumm, C.

1984-12-01T23:59:59.000Z

404

AUTOMATED RADIOLOGICAL MONITORING AT A RUSSIAN MINISTRY OF DEFENCE NAVAL SITE.  

SciTech Connect

The Arctic Military Environmental Cooperation (AMEC) Program is a cooperative effort between the military establishments of the Kingdom of Norway, the Russian Federation, and the US. This paper discusses joint activities conducted over the past year among Norwegian, Russian, and US technical experts on a project to develop, demonstrate and implement automated radiological monitoring at Russian Navy facilities engaged in the dismantlement of nuclear-powered strategic ballistic missile launching submarines. Radiological monitoring is needed at these facilities to help protect workers engaged in the dismantlement program and the public living within the footprint of routine and accidental radiation exposure areas. By providing remote stand-alone monitoring, the Russian Navy will achieve added protection due to the defense-in-depth strategy afforded by local (at the site), regional (Kola) and national-level (Moscow) oversight. The system being implemented at the Polyaminsky Russian Naval Shipyard was developed from a working model tested at the Russian Institute for Nuclear Safety, Moscow, Russia. It includes Russian manufactured terrestrial and underwater gamma detectors, smart controllers for graded sampling, radio-modems for offsite transmission of the data, and a data fusion/display system: The data fusion/display system is derived from the Norwegian Picasso AMEC Environmental Monitoring software package. This computer package allows monitoring personnel to review the real-time and historical status of monitoring at specific sites and objects and to establish new monitoring protocols as required, for example, in an off-normal accident situation. Plans are being developed to implement the use of this system at most RF Naval sites handling spent nuclear fuel.

MOSKOWITZ,P.D.; POMERVILLE,J.; GAVRILOV,S.; KISSELEV,V.; DANIYLAN,V.; BELIKOV,A.; EGORKIN,A.; SOKOLOVSKI,Y.; ENDREGARD,M.; KROSSHAVN,M.; SUNDLING,C.V.; YOKSTAD,H.

2001-02-25T23:59:59.000Z

405

AUTOMATED RADIOLOGICAL MONITORING AT A RUSSIAN MINISTRY OF DEFENSE NAVAL SITE.  

SciTech Connect

The Arctic Military Environmental Cooperation (AMEC) Program is a cooperative effort between the military establishments of the Kingdom of Norway, the Russian Federation, and the US. This paper discusses joint activities conducted over the past year among Norwegian, Russian, and US technical experts on a project to develop, demonstrate and implement automated radiological monitoring at Russian Navy facilities engaged in the dismantlement of nuclear-powered strategic ballistic missile launching submarines. Radiological monitoring is needed at these facilities to help protect workers engaged in the dismantlement program and the public living within the footprint of routine and accidental radiation exposure areas. By providing remote stand-alone monitoring, the Russian Navy will achieve added protection due to the defense-in-depth strategy afforded by local (at the site), regional (Kola) and national-level (Moscow) oversight. The system being implemented at the Polyaminsky Russian Naval Shipyard was developed from a working model tested at the Russian Institute for Nuclear Safety, Moscow, Russia. It includes Russian manufactured terrestrial and underwater gamma detectors, smart controllers for graded sampling, radio-modems for offsite transmission of the data, and a data fusion/display system: The data fusion/display system is derived from the Norwegian Picasso AMEC Environmental Monitoring software package. This computer package allows monitoring personnel to review the real-time and historical status of monitoring at specific sites and objects and to establish new monitoring protocols as required, for example, in an off-normal accident situation. Plans are being developed to implement the use of this system at most RF Naval sites handling spent nuclear fuel.

MOSKOWITZ,P.D.; POMERVILLE,J.; GAVRILOV,S.; KISSELEV,V.; DANIYLAN,V.; BELIKOV,A.; EGORKIN,A.; SOKOLOVSKI,Y.; ENDREGARD,M.; KROSSHAVN,M.; SUNDLING,C.V.; YOKSTAD,H.

2001-02-25T23:59:59.000Z

406

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors  

Science Conference Proceedings (OSTI)

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are too costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized.

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.; Trapp, T.J.; Spinrad, B.I.

1981-09-01T23:59:59.000Z

407

SHARP: Reactor Performance and Safety Simulation Suite  

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

SHARP SHARP Argonne National Laboratory's Reactor Performance and Safety Simulation Suite SHARP could save millions in nuclear reactor design and development... The Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite of codes enables virtual design and engineering of nuclear plant behavior that would be impractical from a traditional experimental approach. ...by leveraging the computational power of one of the world's most powerful supercomputers. Exploiting the power of Argonne Leadership Computing Facility's near-petascale computers, researchers have developed a set of simulation tools that provide a highly detailed description of the reactor core and the nuclear plant behavior. This enables the efficient and precise design of tomorrow's safe and clean nuclear energy sources.

408

CIVILIAN POWER REACTOR PROGRAM. PART II. ECONOMIC POTENTIAL AND DEVELOPMENT PROGRAM. HEAVY WATER-MODERATED POWER REACTOR  

SciTech Connect

The reactor design which forms the base for the current economic status of D/sub 2/O-moderated reactors was estimated from developments in several reactor programs. However, since a heavy water-moderated reactor was not operated on natural U fuel at power reactor conditions, considerable improvement from this current status can be foreseen. A summary of improvements is presented concerning the concept which would result solely from operation of succeeding generation plants without a parallel development program, and improvements which would result from the successful completion of the development program as presented. One plant size was used in the evaluation of plant potential, with a 300 Mw/sub e/ nominal rating. The boiling D/sub 2/O-cooled, pressure tube direct cycle plant design was used. The current development program is outlined; this work includes several items leading to the long-range development of the concept. (auth)

Hutton, J.H.; Davis, S.A.; Graves, C.C.; Duffy, J.G. comps.

1960-08-19T23:59:59.000Z

409

CONTROL MEANS FOR REACTOR  

DOE Patents (OSTI)

An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

Manley, J.H.

1961-06-27T23:59:59.000Z

410

Decommissioning: Reactor Pressure Vessel Internals Segmentation  

Science Conference Proceedings (OSTI)

Decommissioning a nuclear plant covers a wide variety of challenging projects. One of the most challenging areas is the removal and disposal of the reactor pressure vessel (RPV) and the RPV internals. This report describes commercial reactor pressure vessel segmentation projects that have been completed and discusses several projects that are still in the planning stages. The report also covers lessons learned from each project.

2001-10-11T23:59:59.000Z

411

Mechanical Cutting of Irradiated Reactor Internal Components  

SciTech Connect

This paper discusses the use of mechanical cutting methods to volume reduce and package irradiated reactor internal components. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods used for similar projects, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. (authors)

Anderson, M.G.; Fennema, J.A. [MOTA Corporation, West Columbia, SC (United States)

2007-07-01T23:59:59.000Z

412

Environmental Information Document: L-reactor reactivation  

SciTech Connect

Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program.

Mackey, H.E. Jr. (comp.)

1982-04-01T23:59:59.000Z

413

Nuclear reactor characteristics and operational history  

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

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 2. Ownership Data, Table 3. Characteristics and Operational History Table 1. Nuclear Reactor, State, Type, Net Capacity, Generation, and Capacity Factor PDF XLS Plant/Reactor Name Generator ID State Type 2009 Summer Capacity Net MW(e)1 2010 Annual Generation Net MWh2 Capacity Factor Percent3 Arkansas Nuclear One 1 AR PWR 842 6,607,090 90 Arkansas Nuclear One 2 AR PWR 993 8,415,588 97 Beaver Valley 1 PA PWR 892 7,119,413 91 Beaver Valley 2 PA PWR 885 7,874,151 102 Braidwood Generation Station 1 IL PWR 1,178 9,196,689 89

414

Nuclear reactor characteristics and operational history  

Gasoline and Diesel Fuel Update (EIA)

Nuclear > U.S. reactor operation status tables Nuclear > U.S. reactor operation status tables Nuclear Reactor Operational Status Tables Release date: November 22, 2011 Next release date: November 2012 See also: Table 1. Capacity and Generation, Table 2. Ownership Data Table 3. Nuclear Reactor Characteristics and Operational History PDF XLS Plant Name Generator ID Type Reactor Supplier and Model Construction Start Grid Connection Original Expiration Date License Renewal Application License Renewal Issued Extended Expiration Arkansas Nuclear One 1 PWR Babcock&Wilcox, Lower Loop 10/1/1968 8/17/1974 5/20/2014 2/1/2000 6/20/2001 5/20/2034 Arkansas Nuclear One 2 PWR Combustion Eng. 7/1/1971 12/26/1978 7/17/2018 10/15/2003 6/30/2005 7/17/2038

415

Self-actuating reactor shutdown system  

DOE Patents (OSTI)

A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

Barrus, Donald M. (San Jose, CA); Brummond, Willian A (Livermore, CA); Peterson, Leslie F. (Danville, CA)

1988-01-01T23:59:59.000Z

416

Role of research reactors in training of NPP personnel with special focus on training reactor VR-1  

Science Conference Proceedings (OSTI)

Research reactors play an important role in providing key personnel of nuclear power plants a hands-on experience from operation and experiments at nuclear facilities. Training of NPP (Nuclear Power Plant) staff is usually deeply theoretical with an extensive utilisation of simulators and computer visualisation. But a direct sensing of the reactor response to various actions can only improve the personnel awareness of important aspects of reactor operation. Training Reactor VR-1 and its utilization for training of NPP operators and other professionals from Czech Republic and Slovakia is described. Typical experimental exercises and good practices in organization of a training program are demonstrated. (authors)

Sklenka, L.; Rataj, J.; Frybort, J.; Huml, O. [Dept. of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical Univ. in Prague, V Holesovickach 2, Prague 8, 180 00 (Czech Republic)

2012-07-01T23:59:59.000Z

417

Missouri Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA)

snpt2mo Callaway Unit 1 1,190 8,996 100.0 Union Electric Co 1 Plant 1 Reactor Owner Note: Totals may not equal sum of components due to independent rounding.

418

Connecticut Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA)

snpt2ct Millstone Unit 2, Unit 3 2,103 16,750 100.0 Dominion Nuclear Conn Inc 1 Plant 2 Reactors Owner Note: Totals may not equal sum of components due to independent ...

419

Arkansas Nuclear Profile - Power Plants  

U.S. Energy Information Administration (EIA)

snpt2ar Arkansas Nuclear One Unit 1, Unit 2 1,835 15,023 100.0 Entergy Arkansas Inc 1 Plant 2 Reactors Owner Note: Totals may not equal sum of ...

420

Nuclear Reactor Accidents  

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

Reactor Accidents The accidents at the Three Mile Island (TMI) and Chernobyl nuclear reactors have triggered particularly intense concern about radiation hazards. The TMI accident,...

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


421

Principles of Reactor Physics  

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

Nuclear Reactor Physics M A Smith Argonne National Laboratory Nuclear Engineering Division Phone: 630-252-9747, Email: masmith@anl.gov Abstract: Nuclear reactor physics deals with...

422

Light Water Reactor Sustainability (LWRS) Program | Department of Energy  

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

Light Water Reactor Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program The Light Water Reactor Sustainability (LWRS) Program is developing the scientific basis to extend existing nuclear power plant operating life beyond the current 60-year licensing period and ensure long-term reliability, productivity, safety, and security. The program is conducted in collaboration with national laboratories, universities, industry, and international partners. Idaho National Laboratory serves as the Technical Integration Office and coordinates the research and development (R&D) projects in the following pathways: Materials Aging and Degradation Assessment, Advanced Instrumentation, Information, and Control Systems

423

Prospects for Tokamak Fusion Reactors  

SciTech Connect

This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

Sheffield, J.; Galambos, J.

1995-04-01T23:59:59.000Z

424

AN ORGANIC MODERATED REACTOR FOR PROCESS HEAT  

SciTech Connect

A review is given of the potentialities of an organic moderated reactor for the supply of heat to factories having a large demand for low-pressure steam, together with a requirement for power produced by back-pressure turbine generation. By choosing a suitable steam cycle it is possible to cover the range of typical power demand/head load ratios found in the chemical industry. The economic developments of a reactor installed for such a duty are briefly considered to show the reductions in operating cost that might be brought about during the operating lifetlme of the reactor and its associated plant. (auth)

Baines, B.D.; Conway-Jones, J.M.

1962-02-01T23:59:59.000Z

425

Heat pipe reactors for space power applications  

SciTech Connect

A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kW(e) and operate in the temperature range 1200 to 1700/sup 0/K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO/sub 2/. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor.

Koenig, D.R.; Ranken, W.A.; Salmi, E.W.

1977-01-01T23:59:59.000Z

426

"The Fourth Dimension of Naval Tactics": The U.S. Navy and Public Relations, 1919-1939  

E-Print Network (OSTI)

Prior to 1917, the United States Navy only utilized public relations techniques during times of war or to attract recruits into naval service. Following World I, the Navy confronted several daunting problems, including the postwar demobilization of naval assets, the proposed creation of an independent air service, and a public desire for naval arms limitation which many officers believed would endanger the Navy's ability to fulfill its missions. These issues threatened the generous support that the Navy had received from Congress for a quarter of a century, and also hampered the service's attempts to incorporate new weapons systems into its arsenal and recruit high-quality manpower. In response to these challenges, the Navy developed a peacetime public relations capability during the interwar period, despite the fact that many senior naval officials placed a low priority on public relations. Their attitude led subordinates in different parts of the Navy Department to perform public relations tasks despite lacking official orders to carry out such work. Such efforts were haphazard, redundant, handicapped by tradition, and dependent largely upon individual initiative. To augment its meager capabilities, the Navy relied upon external groups, such as the Navy League, to lobby the public for naval expansion. The service also developed formal and informal ties to the mass media, particularly the rapidly expanding motion picture industry. These disparate elements attempted to convince the public that the Navy was a haven for morally upright masculine behavior, a service able to integrate aircraft and submarines into its force structure and keep their operators safe, and a vital national asset with value beyond basic national defense. During the interwar period, the Navy expanded and reorganized the ways in which it courted public opinion. By forging ties with motion picture studios, radio broadcasters, and the print media, it was able to improve the image of the service, attract high quality recruits, and gained the public support for its drive to gain the resources needed to modernize and expand the fleet. During the same era, naval officials became more adept at minimizing the negative impact of the accidents linked to the development of aviation and submarines. Developments of the era laid the foundation for the institutional development of public relations and enhanced media relations during World War II and in the decades that followed.

Wadle, Ryan David

2011-05-01T23:59:59.000Z

427

EA-1889-FEA-FONSI-2012.pdf  

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

UNITED STATES UNITED STATES DEPARTMENT OF THE NAVY FINAL ENVIRONMENTAL ASSESSMENT ON THE DISPOSAL OF DECOMMISSIONED, DEFUELED NAVAL REACTOR PLANTS FROM USS ENTERPRISE (CVN 65) AUGUST 2012 FINAL USS ENTERPRISE EA RESPONSIBLE AGENCIES: Lead Federal Agency: U.S. Department of the Navy Cooperating Agency: U.S. Department of Energy TITLE: Final Environmental Assessment on the Disposal of Decommissioned, Defueled, Naval Reactor Plants from USS ENTERPRISE (CVN 65) FINAL USS ENTERPRISE EA S-1 SUMMARY: The United States Navy proposes to remove the reactor compartments of USS ENTERPRISE (CVN 65), hereafter referred to as ENTERPRISE, at Puget Sound Naval Shipyard & Intermediate Maintenance Facility (PSNS & IMF), prepare the reactor compartments for disposal

428

This document was downloaded on May 22, 2013 at 14:34:19 Author(s) Naval Postgraduate School (U.S.)  

E-Print Network (OSTI)

to the Naval Auxiliary Air Station to house trans- sonic and supersonic wind tunnels and jet com- bustion pits tunnels and jet com- bustion pits. The third and next building increment is expected to comprise a General

429

Reactor and method of operation  

DOE Patents (OSTI)

A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor.

Wheeler, John A. (Princeton, NJ)

1976-08-10T23:59:59.000Z

430

Using Net-Zero Energy Projects to Enable Sustainable Economic Redevelopment at the Former Brunswick Air Naval Base  

SciTech Connect

A Study Prepared in Partnership with the Environmental Protection Agency for the RE-Powering America's Land Initiative: Siting Renewable Energy on Potentially Contaminated Land and Mine Sites. The Brunswick Naval Air Station is a naval air facility and Environmental Protection Agency (EPA) Super Fund site that is being cleaned up, and closed down. The objective of this report is not only to look at the economics of individual renewable energy technologies, but also to look at the systemic benefits that can be gained when cost-effective renewable energy technologies are integrated with other systems and businesses in a community; thus multiplying the total monetary, employment, and quality-of-life benefits they can provide to a community.

Huffman, S.

2011-10-01T23:59:59.000Z

431

Mechanical cutting of irradiated reactor internal components  

Science Conference Proceedings (OSTI)

Mechanical cutting methods to volume reduce and package reactor internal components are now a viable solution for stakeholders challenged with the retirement of first generation nuclear facilities. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods, inclusive of plasma arc and abrasive water-jet cutting, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. Reactor internal components were segmented, packaged, and removed from the reactor building for shipment or storage, allowing the reactor cavity to be drained and follow-on reactor segmentation activities to proceed in the dry state. Area exposure rates at the work positions during the segmentation process were generally 1 mR per hr. Radiological exposure documented for the underwater segmentation processes totaled 13 person rem. The reactor internals weighing 343,000 pounds were segmented into over 200 pieces for maximum shipping package efficiency and produced 5,600 lb of stainless steel chips and shavings which were packaged in void spaces of existing disposal containers, therefore creating no additional disposal volume. Because no secondary waste was driven into suspension in the reactor cavity water, the water was free released after one pass through a charcoal bed and ion exchange filter system. Mechanical cutting techniques are capable of underwater segmentation of highly radioactive components on a large scale. This method minimized radiological exposure and costly water cleanup while creating no secondary waste.

Anderson, Michael G. [MOTA Corporation: 3410 Sunset Boulevard, West Columbia, SC, 29169 (United States)

2008-01-15T23:59:59.000Z

432

Reactor safety method  

DOE Patents (OSTI)

This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

Vachon, Lawrence J. (Clairton, PA)

1980-03-11T23:59:59.000Z

433

NEUTRONIC REACTOR MANIPULATING DEVICE  

DOE Patents (OSTI)

A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

Ohlinger, L.A.

1962-08-01T23:59:59.000Z

434

EPRI NMAC Maintainability Review of the International Gas-Turbine Modular Helium Reactor Power Conversion Unit  

Science Conference Proceedings (OSTI)

This report provides information of interest to the designers of modular helium-reactor-driven gas turbines and persons considering the purchase of this type of plant.

2001-02-01T23:59:59.000Z

435

Environmental Survey preliminary report, Naval Petroleum and Oil Shale Reserves in Colorado, Utah, and Wyoming, Casper, Wyoming  

Science Conference Proceedings (OSTI)

This report presents the preliminary environmental findings from the first phase of the Environmental Survey of the United States Department of Energy (DOE) Naval Petroleum and Oil Shale Reserves in Colorado, Utah, and Wyoming (NPOSR-CUW) conducted June 6 through 17, 1988. NPOSR consists of the Naval Petroleum Reserve No. 3 (NPR-3) in Wyoming, the Naval Oil Shale Reserves No. 1 and 3 (NOSR-1 and NOSR-3) in Colorado and the Naval Oil Shale Reserve No. 2 (NOSR-2) in Utah. NOSR-2 was not included in the Survey because it had not been actively exploited at the time of the on-site Survey. The Survey is being conducted by an interdisciplinary team of environmental specialists, lead and managed by the Office of Environment, Safety and Health's Office of Environmental Audit. Individual team specialists are outside experts being supplied by a private contractor. The objective of the Survey is to identify environmental problems and areas of environmental risk associated with NPOSR. The Survey covers all environmental media and all areas of environmental regulation. It is being performed in accordance with the DOE Environmental Survey Manual. This phase of the Survey involves the review of existing site environmental data, observations of the operations carried on at NPOSR and interviews with site personnel. The Survey team has developed a Sampling and Analysis Plan to assist in further assessing specific environmental problems identified at NOSR-3 during the on-site Survey. There were no findings associated with either NPR-3 or NOSR-1 that required Survey-related sampling and Analysis. The Sampling and Analysis Plan will be executed by Idaho National Engineering Laboratory. When completed, the results will be incorporated into the Environmental Survey Summary report. The Summary Report will reflect the final determinations of the NPOSR-CUW Survey and the other DOE site-specific Surveys. 110 refs., 38 figs., 24 tabs.

Not Available

1989-02-01T23:59:59.000Z

436

Report to the President on agreements and programs relating to the Naval Petroleum and Oil Shale Reserves  

SciTech Connect

The Department of Energy monitors commercial natural gas production activities along the boundaries of Naval Oil Shale Reserve No. 1 and Naval Oil Shale Reserve No. 3, which are located in Garfield County, Colorado, and were created in the early part of this century to provide a future source of shale oil for the military. In response to the private sector`s drilling of natural gas wells along the south and southwest boundaries of the Reserves, which began in the early 1980`s, the Department developed a Natural Gas Protection Program to protect the Government`s resources from drainage due to the increasing number of commercial gas wells contiguous to Naval Oil Shale Reserve No. 3. This report provides an update of the Gas Protection Program being implemented and the agreements that have been placed in effect since December 19, 1991, and also includes the one communitized well containing Naval Petroleum Reserve No. 3 lands. The Protection Program employs two methods to protect the Government`s resources: (1) sharing with the private sector in the costs and production of wells by entering into ``communitization`` agreements; and (2) drilling wholly-owned Government wells to ``offset`` commercial wells that threaten to drain natural gas from the Reserves. The methods designed to protect the Government`s resources are achieving their objective of abating gas drainage and migration. As a result of the Protection Program, the Department of Energy is able to produce natural gas and either sell its share on the open market or transfer it for use at Government facilities. The Natural Gas Protection Program is a reactive, ongoing program that is continually revised as natural gas transportation constraints, market conditions, and nearby commercial production activities change.

1994-08-01T23:59:59.000Z

437

Systems analysis of the CANDU 3 Reactor  

SciTech Connect

This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

1993-07-01T23:59:59.000Z

438

Final sitewide environmental assessment for continued development of Naval Petroleum Reserve No. 3 (NPR-3), Natrona County, Wyoming  

SciTech Connect

The Secretary of Energy is required by law to explore, prospect, conserve, develop, use, and operate the Naval Petroleum and Oil Shale Reserves. The Naval Petroleum Reserves Production Act of 1976 (Public Law 94-258), requires that the Naval Petroleum Reserves be produced at their maximum efficient rate (MER), consistent with sound engineering practices, for a period of six years. To fulfill this mission, DOE is proposing continued development activities which would include the drilling of approximately 250 oil production and injection (gas, water, and steam) wells, the construction of between 25 and 30 miles of associated gas, water, and steam pipelines, the installation of several production and support facilities, and the construction of between 15 and 20 miles of access roads. These drilling and construction estimates include any necessary activities related to the operation of the Rocky Mountain Oilfield Testing Center (RMOTC). The purpose of RMOTC will be to provide facilities and necessary support to government and private industry for testing and evaluating new oilfield and environmental technologies, and to transfer these results to the petroleum industry through seminars and publications. Continued development activities either have no potential to result in adverse environmental impacts or would only result in adverse impacts that could be readily mitigated. The small amounts of disturbed surface area will be reclaimed to its original natural state when production operations terminate. The preparation of an environmental impact statement is not required, and the DOE is issuing this Finding of No Significant Impact (FONSI). 73 refs.

1995-07-01T23:59:59.000Z

439

Materials Reliability Program: Reactor Pressure Vessel Integrity Primer (MRP-278)  

Science Conference Proceedings (OSTI)

This primer is based on two earlier Electric Power Research Institute (EPRI) reports: Reactor Vessel Embrittlement Management Handbook: A Handbook for Managing Reactor Vessel Embrittlement and Vessel Integrity (TR-101975-T2) and Primer: Fracture Mechanics in the Nuclear Power Industry (NP-5792-SR, Rev. 1). The information in those earlier reports has been updated extensively and focuses on todays reactor pressure vessel (RPV) embrittlement, integrity, and plant license renewal issues. This RPV integrity ...

2010-06-09T23:59:59.000Z

440

An Engineering Test Reactor  

SciTech Connect

A relatively inexpensive reactor for the specific purpose of testing a sub-critical portion of another reactor under conditions that would exist during actual operation is discussed. It is concluded that an engineering tool for reactor development work that bridges the present gap between exponential and criticality experiments and the actual full scale operating reactor is feasible. An example of such a test reactor which would not entail development effort to ut into operation is depicted.

Fahrner, T.; Stoker, R.L.; Thomson, A.S.

1951-03-16T23:59:59.000Z

Note: This page contains sample records for the topic "naval reactor plants" 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.


441

Reactor Pressure Vessel Task of Light Water Reactor Sustainability...  

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

Reactor Pressure Vessel Task of Light Water Reactor Sustainability Program: Milestone Report on Materials and Machining of Specimens for the ATR-2 Experiment Reactor Pressure...

442

REVIEW OF THE STATUS OF SUPERCRITICAL WATER REACTOR TECHNOLOGY  

SciTech Connect

Supercritical water-reactor design studies are reviewed. The status of supercritical water technology relative to heat transfer and fluid flow, water chemistry, internal deposition on heated surfaces, plant power cycles, and reactor construction materials is reviewed. The direct cycle was found to offer the highest probability for achieving economic power. (C.J.G.)

Marchaterre, J.F.; Petrick, M.

1960-08-01T23:59:59.000Z

443

Economic characteristics of a smaller, simpler reactor  

SciTech Connect

Reduced load growth and heightened concern with economic risk has led to an expressed utility preference for smaller capacity additions. The Modular High Temperature Reactor (MHTGR) plant has been developed as a small, simple plant that has limited financial risk and is economically competitive with comparatively sized coal plants. Competitive economics is achieved by the simplifications made possible in a small MHTGR, reduction in the quantity of nuclear grade construction and design standardization and certification. Assessments show the MHTGR plant to have an economic advantage over coal plants for plant sizes from 270 MWe to 1080 MWe. Financial risk is limited by small unit sizes and short lead times that allow incremental deployment. Evaluations show the MHTGR incremental deployment capability to reduce negative cash flows by almost a factor of 2 relative to that required by a single large nuclear plant.

LaBar, M.; Bowers, H.

1988-01-01T23:59:59.000Z

444

Advanced Light Water Reactor utility requirements document  

SciTech Connect

The ALWR Requirements Document is a primary work product of the EPRI Program. This document is an extensive compilation of the utility requirements for design, construction and performance of advanced light water reactor power plants for the 1990s and beyond. The Requirements Document's primary emphasis is on resolution of significant problems experienced at existing nuclear power plants. It is intended to be used with companion documents, such as utility procurement specifications, which would cover the remaining detailed technical requirements applicable to new plant projects. The ALWR Requirements Document consists of several major parts. This volume is Part I, The Executive Summary. It is intended to serve as a concise, management level synopsis of advanced light water reactors including design objectives and philosophy, overall configuration and features and the steps necessary to proceed from the conceptual design stage to a completed, functioning power plant.

1986-06-01T23:59:59.000Z

445

Materials Degradation in Light Water Reactors: Life After 60 | Department  

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

Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the

446

Materials Degradation in Light Water Reactors: Life After 60 | Department  

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

Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Materials Degradation in Light Water Reactors: Life After 60 Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the

447

DOE - Office of Legacy Management -- Westinghouse Advanced Reactors...  

Office of Legacy Management (LM)

PA.10-4 Site Operations: 1960s and 1970s - Produced light water and fast breeder reactor fuels on a development and pilot plant scale. Closed in 1979. PA.10-2 PA.10-3 Site...

448

Existing reactor expansion study basis  

SciTech Connect

The latest HAPO Five Year Program review, HW-59633, forecasts substantial increases in Pu production from the eight existing Hanford reactors over the next several years. These production increases would be attained by a combination of several methods which include increased reactor power levels resulting from higher process water flow rates and coolant bulk outlet temperatures, improved time operated efficiency, higher conversion ratios, and reduced transient reactivity losses. In order to provide a realistic basis for budgeting to meet these or other increased production goals, it is necessary that a study program be undertaken to determine in general terms the plant changes required to support these forecasted levels, to evaluate the economic and technical feasibility of achieving the process conditions, and to present an integrated program for achieving these objectives. This study program will necessarily consider the interrelated effects of a number of various facets of reactor and water plant process conditions, operational requirements, and proposed development programs. The purpose of this document is to present a plan for the execution of the proposed study. Included in this outline are a review of the basic study considerations, problem assignments and schedules, and manpower and cost estimates for the performance of the study.

Heacock, H.W.

1959-06-24T23:59:59.000Z

449

Distributed Energy Resources at Naval Base Ventura County Building1512: A Sensitivity Analysis  

Science Conference Proceedings (OSTI)

This report is the second of a two-part study by BerkeleyLab of a DER (distributed energy resources) system at Navy Base VenturaCounty (NBVC). First, a preliminary assessment ofthe cost effectivenessof distributed energy resources at Naval Base Ventura County (NBVC)Building 1512 was conducted in response to the base s request for designassistance to the Federal Energy Management Program (Bailey and Marnay,2004). That report contains a detailed description of the site and theDER-CAM (Consumer Adoption Model) parameters used. This second reportcontains sensitivity analyses of key parameters in the DER system modelof Building 1512 at NBVC and additionally considers the potential forabsorption-powered refrigeration.The prior analysis found that under thecurrent tariffs, and given assumptions about the performance andstructure of building energy loads and available generating technologycharacteristics, installing a 600 kW DER system with absorption coolingand recovery heat capabilities could deliver cost savings of about 14percent, worth $55,000 per year. However, under current conditions, thisstudy also suggested that significant savings could be obtained ifBuilding 1512 changed from its current direct access contract to a SCETOU-8 (Southern California Edison time of use tariff number 8) ratewithout installing a DER system. Evaluated on this tariff, the potentialsavings from installation of a DER system would be about 4 percent of thetotal bill, or $16,000 per year.

Bailey, Owen C.; Marnay, Chris

2005-06-05T23:59:59.000Z

450

Oil shale resources of the Naval Oil Shale Reserve No. 1, Colorado  

SciTech Connect

The resource of potential oil represented by Green River Formation oil shale on Naval Oil Shale Reserve No. 1 (NOSR No. 1) in the southeast corner of Colorado's Piceance Creek Basin is evaluated in detail. NOSR No. 1 is the site of intensive long-term oil-shale development studies and is the source of innumerable oil-shale samples for all manner of testing. A brief history of these studies is presented. This oil-shale resource is defined from oil-yield assay data on 33 cores plotted as histograms and correlated into cross sections. Contour maps of thickness, richness and oil resource in place are presented for the Mahogany Zone, the rich zone in the Mahogany zone, and for 2 units beneath and 5 units above the Mahogany zone. Total oil shale resource on NOSR No. 1 is 20.4 billion barrels of which 17.4 billion barrels are particularly suitable for development by vertical modified in-place processes. A previously unknown Mahogany zone outcrop providing much additional development access is described. Now under sole control of the US Department of Energy (DOE), NOSR No. 1 offers DOE a unique site for oil shale testing and development.

Smith, J.W.; Beard, T.N.; Trudell, L.G.

1979-06-01T23:59:59.000Z

451

Development of a reactor engineering workstation at Seabrook station  

SciTech Connect

The reactor engineers at Seabrook station are responsible for supporting plant operation with respect to the current reactor core design. Advanced assembly designs, complex reactor core loading patterns, and emphasis on efficient and safe operation puts a greater demand on the reactor engineer. The traditional use of static data constants and coarse core modeling, in light of the more complex fuel and core designs of today, results in less than optimum monitoring and predicting tools for the reactor engineer. The incorporation of an advanced three-dimensional nodal code with thermal feedbacks and detailed spatial modeling along with the ability to follow current operational history on a state-of-the-art workstation provides the reactor engineer with a dynamic core monitoring and predictive tool. This approach allows for more accurate and efficient completion of the reactor engineer's tasks. Yankee Atomic Electric Company (YAEC) is currently in the process of providing advanced reactor physics nodal methods to the reactor engineers at Seabrook station. The scope of this project is to supply a reactor engineering workstation with a simplified user interface to an advanced nodal core model as part of an on-line core monitor/predictor for standard reactor engineering calculations. It uses the Studsvik Core Management System (CMS), which primarily consists of the CASMO-3 cross-section generating code and the SIMULATE-3 three-dimensional two-group nodal reactor analysis code.

Tremblay, M.A.; Gorski, J.P. (Yankee Atomic Electric Co., Bolton, MA (United States)); Gurney, P.V. (New Hampshire Yankee, Seabrook, NY (United States))

1992-01-01T23:59:59.000Z

452

An Account of Oak Ridge National Laboratory's Thirteen Research Reactors  

Science Conference Proceedings (OSTI)

The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

Rosenthal, Murray Wilford [ORNL

2009-08-01T23:59:59.000Z

453

Uncertainty quantification approaches for advanced reactor analyses.  

SciTech Connect

The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

Briggs, L. L.; Nuclear Engineering Division

2009-03-24T23:59:59.000Z

454

Memorandum on Chemical Reactors and Reactor Hazards  

SciTech Connect

Two important problems in the investigation of reactor hazards are the chemical reactivity of various materials employed in reactor construction and the chracteristics of heat transfer under transient conditions, specifically heat transfer when driven by an exponentially increasing heat source (exp t/T). Although these problems are independent of each other, when studied in relation to reactor hazards they may occur in a closely coupled sequence. For example the onset of a dangerous chemical reactor may be due to structural failure of various reactor components under an exponentially rising heat source originating with a runaway nuclear reactor. For this reason, these two problems should eventually be studied together after an exploratory experimental survey has been made in which they are considered separately.

Mills, M.M.; Pearlman, H.; Ruebsamen, W.; Steele, G., Chrisney, J.

1951-07-05T23:59:59.000Z

455

Small Modular Reactors (468th Brookhaven Lecture)  

SciTech Connect

With good reason, much more media attention has focused on nuclear power plants than solar farms, wind farms, or hydroelectric plants during the past month and a half. But as nations around the world demand more energy to power everything from cell phone batteries to drinking water pumps to foundries, nuclear plants are the only non-greenhouse-gas producing option that can be built to operate almost anywhere, and can continue to generate power during droughts, after the sun sets, and when winds die down. To supply this demand for power, designers around the world are competing to develop more affordable nuclear reactors of the future: small modular reactors. Brookhaven Lab is working with DOE to ensure that these reactors are designed to be safe for workers, members of surrounding communities, and the environment and to ensure that the radioactive materials and technology will only be used for peaceful purposes, not weapons. In his talk, Bari will discuss the advantages and challenges of small modular reactors and what drives both international and domestic interest in them. He will also explain how Brookhaven Lab and DOE are working to address the challenges and provide a framework for small modular reactors to be commercialized.

Bari, Robert

2011-04-20T23:59:59.000Z

456

Exploding the myths about the fast breeder reactor  

SciTech Connect

This paper discusses the facts and figures about the effects of conservation policies, the benefits of the Clinch River Breeder Reactor demonstration plant, the feasibility of nuclear weapons manufacture from reactor-grade plutonium, diversion of plutonium from nuclear plants, radioactive waste disposal, and the toxicity of plutonium. The paper concludes that the U.S. is not proceeding with a high confidence strategy for breeder development because of a variety of false assumptions.

Burns, S.

1979-01-01T23:59:59.000Z

457

Performance and safety design of the advanced liquid metal reactor  

SciTech Connect

The Advanced Liquid Metal Reactor (ALMR) program led by General Electric is developing, under U.S. Department of Energy sponsorship, a conceptual design for an advanced sodium-cooled liquid metal reactor plant. This design is intended to improve the already excellent level of plant safety achieved by the nuclear power industry while at the same time providing significant reductions in plant construction and operating costs. In this paper, the plant design and performance are reviewed, with emphasis on the ALMR's unique passive design safety features and its capability to utilize as fuel the actinides in LWR spent fuel.

Berglund, R.C.; Magee, P.M.; Boardman, C.E.; Gyorey, G.L. (General Electric Co., San Jose, CA (United States). Advanced Nuclear Technology)

1991-01-01T23:59:59.000Z

458

Updated comparison of economics of fusion reactors with advanced fission reactors  

Science Conference Proceedings (OSTI)

The projected cost of electricity (COE) for fusion is compared with that from current and advanced nuclear fission and coal-fired plants. Fusion cost models were adjusted for consistency with advanced fission plants and the calculational methodology and cost factors follow guidelines recommended for cost comparisons of advanced fission reactors. The results show COEs of about 59--74 mills/kWh for the fusion designs considered. In comparison, COEs for future fission reactors are estimated to be in the 43--54 mills/kWh range with coal-fired plant COEs of about 53--69 mills/kWh ($2--3/GJ coal). The principal cost driver for the fusion plants relative to fission plants is the fusion island cost. Although the estimated COEs for fusion are greater than those for fission or coal, the costs are not so high as to preclude fusion's competitiveness as a safe and environmentally sound alternative.

Delene, J.G.

1990-01-01T23:59:59.000Z

459

Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.  

Science Conference Proceedings (OSTI)

The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage, and cleaning stations-have accumulated satisfactory construction and operation experiences. In addition, two special issues for future development are described in this report: large capacity interim storage and transuranic-bearing fuel handling.

Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

2009-03-01T23:59:59.000Z

460

22.39 Integration of Reactor Design, Operations, and Safety, Fall 2005  

E-Print Network (OSTI)

This course integrates studies of reactor physics and engineering sciences into nuclear power plant design. Topics include materials issues in plant design and operations, aspects of thermal design, fuel depletion and ...

Todreas, Neil E.

Note: This page contains sample records for the topic "naval reactor plants" 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.


461

EBR-2 (Experimental Breeder Reactor-2), IFR (Integral Fast Reactor) prototype testing programs  

SciTech Connect

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs.

Lehto, W.K.; Sackett, J.I.; Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (USA). EBR-II Div. Argonne National Lab., IL (USA)); Planchon, H.P.; Lambert, J.D.B. (Argonne National Lab., IL (USA))

1990-01-01T23:59:59.000Z

462

Manhattan Project: F Reactor Plutonium Production Complex  

Office of Scientific and Technical Information (OSTI)

F REACTOR PLUTONIUM PRODUCTION COMPLEX F REACTOR PLUTONIUM PRODUCTION COMPLEX Hanford Engineer Works, 1945 Resources > Photo Gallery Plutonium production area, Hanford, ca. 1945 The F Reactor plutonium production complex at Hanford. The "boxy" building between the two water towers on the right is the plutonium production reactor; the long building in the center of the photograph is the water treatment plant. The photograph was reproduced from Henry DeWolf Smyth, Atomic Energy for Military Purposes: The Official Report on the Development of the Atomic Bomb under the Auspices of the United States Government, 1940-1945 (Princeton, NJ: Princeton University Press, 1945). The Smyth Report was commissioned by Leslie Groves and originally issued by the Manhattan Engineer District. Princeton University Press reprinted it in book form as a "public service" with "reproduction in whole or in part authorized and permitted."

463

THE EXPERIMENTAL BERYLLIUM OXIDE REACTOR. MARITIME GAS-COOLED REACTOR PROGRAM  

SciTech Connect

LIUM OXIDE REACTOR. MARITIME GAS-COOLED The Experimental Beryllium Oxide Reactor, EBOR, will be constructed at the National Reactor Testing Station as the AEC portion of the joint Maritime Administration--AEC Maritime Gas Cooled Reactor Program. The ultimate goal of the Program is the development of nuclear power plants employing a helium cooled and beryllium oxide moderated reactor directly coupled to a closed cycle gas turbine. The objective is to obtain compact nuclear engines suitable for use either in a merchant ship propulsion system or an intermediate size central station power plant in the 20 to 100 Mw(e) size range. The EBOR is a l0 Mw(t) test of the basic fuel element and moderator designs. It is capable of being up-graded in power at a later date to a test of the nuclear reactor turbine concept. The objective of the experiment is outlined. The principal reactor components to be tested and the test facility are described. (auth)

Moore, W.C.

1961-07-01T23:59:59.000Z

464

Gas Reactor Technology R&D  

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

U.S. Department of Energy to Invest U.S. Department of Energy to Invest up to $7.3 Million for "Deep-Burn" Gas-Reactor Technology R&D Artist's rendering of Nuclear Plant An artist's rendering of the Next Generation Nuclear Plant concept. The U.S. Department of Energy today announced a Funding Opportunity Announcement (FOA) valued at $7.3 million for universities, commercial entities, National Laboratories with expertise in the concept of nuclear fuel "Deep-Burn" in which plutonium and higher transuranics recycled from spent nuclear fuel are destroyed. The funding opportunity seeks to establish the technological foundations that will support the role of the very-high-temperature, gas-cooled reactor (VHTR) in the nuclear fuel cycle -- which is one of the prototype reactors being researched/developed under

465

A Small, Clean, Stable Fusion Power Plant ---- Inventor Samuel...  

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

Small, Clean, Stable Fusion Power Plant ---- Inventor Samuel A. Cohen This invention discloses improvements in magnetic fusion reactor design and operational modes that reduce...

466

PL FINAL DESIGN REPORT. VOLUME II. PLANT DRAWINGS  

SciTech Connect

Plant drawings for the final design for the Army Reactor (PL-2) are presented. Two hundred and twenty-eight figures are included. (M.C.G.)

Combustion Engineering, Inc. Nuclear Div., Windsor, Conn.

1961-06-30T23:59:59.000Z

467

How much electricity does a typical nuclear power plant generate ...  

U.S. Energy Information Administration (EIA)

... (kWh). There were 65 nuclear power plants with 104 operating nuclear reactors that generated a total of 790 billion kilowatt-hours (kWh), ...

468

Photo of the Week: The Sixth Zero Power Reactor | Department of Energy  

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

The Sixth Zero Power Reactor The Sixth Zero Power Reactor Photo of the Week: The Sixth Zero Power Reactor November 13, 2013 - 4:41pm Addthis In 1958, Argonne National Laboratory began the construction of several zero-power reactors (ZPRs), which are nuclear fission reactors that don't actually generate any power. Scientists developed ZPRs to assess the performance of various