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Sample records for naval reactor plants

  1. naval reactors | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    on Energy and Water Development, visited the Naval Reactors Facility (NRF) at the... ... propulsion plants use a pressurized-water reactor design that has two basic systems: ...

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

    National Nuclear Security Administration (NNSA)

    More About NNSA's Naval Reactors Office The Naval Nuclear Propulsion Program provides militarily effective nuclear propulsion plants and ensures their safe, reliable and long-lived ...

  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. EA-1889: Disposal of Decommissioned, Defueled Naval Reactor Plants from USS Enterprise (CVN 65) at the Hanford Site, Richland, Washington

    Broader source: Energy.gov [DOE]

    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.

  5. Naval Nuclear Propulsion Plants | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    | (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 radiation, shielding is placed around the reactor to protect the crew. Despite close proximity to a reactor core, a typical crewmember receives less exposure to radiation than one who remains ashore and works in an office building. In naval nuclear propulsion plants, fissioning of uranium atoms in the reactor

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

    National Nuclear Security Administration (NNSA)

    Administration | (NNSA) More About NNSA's Naval Reactors Office The Naval Nuclear Propulsion Program provides militarily effective nuclear propulsion plants and ensures their safe, reliable and long-lived operation. This mission requires the combination of fully trained U.S. Navy men and women with ships that excel in endurance, stealth, speed, and independence from supply chains. The Naval Nuclear Propulsion Program provides militarily effective nuclear propulsion plants and ensures their

  7. Congressional Delegation visits Naval Reactors Facility | National...

    National Nuclear Security Administration (NNSA)

    Chuck Fleischmann of the House Appropriations Subcommittee on Energy and Water Development, visited the Naval Reactors Facility (NRF) at the Idaho National Laboratory (INL). ...

  8. 2012 Annual Planning Summary for Naval Reactors | Department...

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

    Naval Reactors 2012 Annual Planning Summary for Naval Reactors The ongoing and projected Environmental Assessments and Environmental Impact Statements for 2012 and 2013 within the ...

  9. Special Analysis: Naval Reactor Waste Disposal Pad

    SciTech Connect (OSTI)

    Cook, J.R.

    2003-03-31

    This report presents the results of a special study of the Naval Reactor Waste Disposal Pad located within the boundary of the E-Area Low-Level Waste Facility at the Savannah River Site.

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

    National Nuclear Security Administration (NNSA)

    and works in an office building. U.S. naval nuclear propulsion plants use a pressurized-water reactor design that has two basic systems: the primary system and the secondary...

  11. Reactor Safety Planning for Prometheus Project, for Naval Reactors Information

    SciTech Connect (OSTI)

    P. Delmolino

    2005-05-06

    The purpose of this letter is to submit to Naval Reactors the initial plan for the Prometheus project Reactor Safety work. The Prometheus project is currently developing plans for cold physics experiments and reactor prototype tests. These tests and facilities may require safety analysis and siting support. In addition to the ground facilities, the flight reactor units will require unique analyses to evaluate the risk to the public from normal operations and credible accident conditions. This letter outlines major safety documents that will be submitted with estimated deliverable dates. Included in this planning is the reactor servicing documentation and shipping analysis that will be submitted to Naval Reactors.

  12. 2014 Annual Planning Summary for the NNSA Naval Reactors

    Broader source: Energy.gov [DOE]

    The ongoing and projected Environmental Assessments and Environmental Impact Statements for 2014 and 2015 within the NNSA Naval Reactors.

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

    National Nuclear Security Administration (NNSA)

    Administration | (NNSA) 30 - Deputy Administrator for Naval Reactors NA 30 - Naval Reactors FY15 Year End Report Semi Annual Report FY14 Year End Report Semi Annual Report NX 3 - Naval Reactors Laboratory Field Office FY15 Year End

  14. EIS-0275: Disposal of the S1C Prototype Reactor Plant, Hanford Site, Richland, WA (Navy Document)

    Broader source: Energy.gov [DOE]

    This EIS analyzes the Office of Naval Reactors (Naval Reactors) proposed action to dismantle the defueled S1C Prototype reactor plant.

  15. About Naval Reactors | National Nuclear Security Administration | (NNSA)

    National Nuclear Security Administration (NNSA)

    About Naval Reactors What Is the Naval Nuclear Propulsion Program? The Naval Nuclear Propulsion Program comprises the military and civilian personnel who design, build, operate, maintain, and manage the nuclear-powered ships and the many facilities that support the U.S. nuclear-powered naval fleet. The Program has cradle-to-grave responsibility for all naval nuclear propulsion matters. Program responsibilities are delineated in Presidential Executive Order 12344 of February 1, 1982, and

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

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

    ... Specifically, although the site transitioned to training employees using an online service, Naval Reactors Federal employees did not have the necessary application licenses needed ...

  17. NA 30 - Deputy Administrator for Naval Reactors | National Nuclear...

    National Nuclear Security Administration (NNSA)

    Us Our Operations Management and Budget Office of Civil Rights Workforce Statistics NA 30 - Deputy Administrator for Naval Reactors NA 30 - Deputy Administrator for...

  18. 1996 environmental monitoring report for the Naval Reactors Facility

    SciTech Connect (OSTI)

    1996-12-31

    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).

  19. Fuel Cell Power Plant Experience Naval Applications

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

    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

  20. 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 (OSTI)

    D. Kokkinos

    2005-04-28

    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.

  1. Fuel Cell Power Plant Experience Naval Applications | Department of Energy

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

    Plant Experience Naval Applications Fuel Cell Power Plant Experience Naval Applications Presented at the DOE-DOD Shipboard APU Workshop on March 29, 2011. apu2011_8_wolak.pdf (1.51 MB) More Documents & Publications Fuel Cell Power Plants Biofuel Case Study - Tulare, CA Fuel Cell Power Plants Renewable and Waste Fuels Co-production of Hydrogen and Electricity (A Developer's Perspective)

  2. Statement on Defense Nuclear Nonproliferation and Naval Reactors Activities

    National Nuclear Security Administration (NNSA)

    before the House Committee on Appropriations Subcommittee on Energy & Water Development | National Nuclear Security Administration | (NNSA) Defense Nuclear Nonproliferation and Naval Reactors Activities before the House Committee on Appropriations Subcommittee on Energy & Water Development February 26, 2013 INTRODUCTION Chairman Frelinghuysen, Ranking Member Kaptur, and distinguished members of the Subcommittee, thank you for having me here today to discuss the National Nuclear

  3. Naval Reactors Facility environmental monitoring report, calendar year 2001

    SciTech Connect (OSTI)

    2002-12-31

    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.

  4. Naval Reactors Facility Environmental Monitoring Report, Calendar Year 2003

    SciTech Connect (OSTI)

    2003-12-31

    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.

  5. Naval Reactors Facility environmental monitoring report, calendar year 1999

    SciTech Connect (OSTI)

    2000-12-01

    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).

  6. 1997 environmental monitoring report for the Naval Reactors Facility

    SciTech Connect (OSTI)

    1997-12-31

    The results of the radiological and nonradiological environmental monitoring programs for 1997 at the Naval Reactors Facility (NRF) are presented in this report. The results obtained from the environmental monitoring programs verify that releases to the environment from operations at NRF were in accordance with state and federal regulations. Evaluation of the environmental data confirms that the operation of NRF continues to have no adverse effect on the quality of the environment or the health and safety of the general public. Furthermore, a conservative assessment of radiation exposure to the general public as a result of NRF operations demonstrated that the dose received by any member of the public was well below the most restrictive dose limits prescribed by the Environmental Protection Agency (EPA) and the Department of Energy (DOE).

  7. Naval Reactors Facility environmental monitoring report, calendar year 2000

    SciTech Connect (OSTI)

    2001-12-01

    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).

  8. EIS-0259: Disposal of Decommissioned, Defueled Cruiser, Ohio Class and Los Angeles Class Naval Reactor Plants, Hanford Site, Richland (adopted from Navy)

    Broader source: Energy.gov [DOE]

    This EIS analyzes the alternate ways for disposing of decommissioned, defieled reactor compliments from U.S. Navy nuclear-powered cruisers, (Bainbridge, Truxtun, Long Beach, California Class and Virginia Class) and Los Angeles Class, and Ohio Class submarines.

  9. Audit Report - Naval Reactors Information Technology System Developmen...

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

    (EBS) project that included procurement, finance, human resources and logistics modules. ... The procurement module alone is expected to cost approximately 12.8 million, and Naval ...

  10. NEUTRONIC REACTOR POWER PLANT

    DOE Patents [OSTI]

    Metcalf, H.E.

    1962-12-25

    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)

  11. naval reactors

    National Nuclear Security Administration (NNSA)

    6%2A en Powering the Nuclear Navy http:www.nnsa.energy.govourmissionpoweringnavy

    Page...

  12. naval reactors

    National Nuclear Security Administration (NNSA)

    6%2A en Powering the Nuclear Navy http:nnsa.energy.govourmissionpoweringnavy

    Page...

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

    National Nuclear Security Administration (NNSA)

    use a pressurized-water reactor design that has two basic systems: the primary system and the secondary system. The primary system circulates ordinary water in an all-welded, ...

  14. Safeguards Issues at Nuclear Reactors and Enrichment Plants ...

    Office of Scientific and Technical Information (OSTI)

    Safeguards Issues at Nuclear Reactors and Enrichment Plants Citation Details In-Document Search Title: Safeguards Issues at Nuclear Reactors and Enrichment Plants Authors: Boyer, ...

  15. Safeguards Issues at Nuclear Reactors and Enrichment Plants ...

    Office of Scientific and Technical Information (OSTI)

    Safeguards Issues at Nuclear Reactors and Enrichment Plants Citation Details In-Document Search Title: Safeguards Issues at Nuclear Reactors and Enrichment Plants You are ...

  16. Accident Investigation of the June 17, 2012, Construction Accident- Structural Steel Collapse at The Over pack Storage Expansion #2 at the Naval Reactors Facility at the Idaho National Laboratory, Idaho Falls, Idaho

    Broader source: Energy.gov [DOE]

    This report documents the Naval Reactors investigation into the collapse ofa partially-erected spent fuel storage building, Overpack Storage Expansion #2 (OSE2), at the Naval Reactors Facility. The Accident Investigation Board inspected the scene, collected physical and photographic evidence, interviewed involved personnel, and reviewed relevant documents to determine the key causes of the accident. Based on the information gathered during the investigation, the Board identified several engineering and safety deficiencies that need to be addressed to prevent recurrence.

  17. Plant maintenance and advanced reactors, 2007

    SciTech Connect (OSTI)

    Agnihotri, Newal

    2007-09-15

    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.

  18. NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 05 NUCLEAR FUELS...

    Office of Scientific and Technical Information (OSTI)

    Title list of documents made publicly available, January 1-31, 1998 NONE 21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 05 NUCLEAR FUELS; BIBLIOGRAPHIES; NUCLEAR POWER PLANTS;...

  19. Generic small modular reactor plant design.

    SciTech Connect (OSTI)

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

    2012-12-01

    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.

  20. Superfund Record of Decision (EPA Region 5): Naval Industrial Reserve Ordnance Plant, Fridley, MN. (First remedial action), September 1990

    SciTech Connect (OSTI)

    Not Available

    1990-09-28

    The 82.6-acre Naval Industrial Reserve Ordnance Plant (NIROP) site is a weapons system manufacturing facility in Fridley, Minnesota, which began operations in 1940. The site is a government-owned, contractor-operated, plant located just north of the FMC Corp. During the 1970s, paint sludge and chlorinated solvents were disposed of onsite in pits and trenches. In 1981, State investigations identified TCE in onsite water supply wells drawing from the Prairie DuChien/Jordan aquifer, and the wells were shut down. In 1983, EPA found drummed waste in the trenches or pits at the northern portion of the site, and as a result, during 1983 and 1984, the Navy authorized an installation restoration program, during which approximately 1,200 cubic yards of contaminated soil and 42 drums were excavated and landfilled offsite. The Record of Decision (ROD) addresses the remediation of a shallow ground water operable unit. The primary contaminants of concern affecting the ground water are VOCs including PCE, TCE, toluene, and xylene.

  1. Plant maintenance and advanced reactors issue, 2008

    SciTech Connect (OSTI)

    Agnihotri, Newal

    2009-09-15

    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.

  2. Gas Reactor Plant Analyzer and Simulator for Hydrogen Production

    Energy Science and Technology Software Center (OSTI)

    2004-01-01

    This software is used to study and analyze various configurations of plant equipment for gas cooled nuclear reactor applications. The user of this software would likely be interested in optimizing the economic, safety, and operating performance of this type of reactor. The code provides the capability for the user through his input to configure networks of nuclear reactor components. The components available include turbine, compressor, heat exchanger, reactor core, coolers, bypass valves, and control systems.

  3. Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

    SciTech Connect (OSTI)

    MH Lane

    2006-02-15

    This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

  4. Liquid metal cooled nuclear reactor plant system

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    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.

  5. P.C. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR...

    Office of Scientific and Technical Information (OSTI)

    Erosioncorrosion-induced pipe wall thinning in US Nuclear Power Plants Wu, P.C. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; BWR TYPE REACTORS; PIPES; CORROSION; EROSION;...

  6. Nuclear Naval Propulsion: A Feasible Proliferation Pathway?

    SciTech Connect (OSTI)

    Swift, Alicia L.

    2014-01-31

    There is no better time than now to close the loophole in Article IV of the Nuclear Non-proliferation Treaty (NPT) that excludes military uses of fissile material from nuclear safeguards. Several countries have declared their intention to pursue and develop naval reactor technology, including Argentina, Brazil, Iran, and Pakistan, while other countries such as China, India, Russia, and the United States are expanding their capabilities. With only a minority of countries using low enriched uranium (LEU) fuel in their naval reactors, it is possible that a state could produce highly enriched uranium (HEU) under the guise of a nuclear navy while actually stockpiling the material for a nuclear weapon program. This paper examines the likelihood that non-nuclear weapon states exploit the loophole to break out from the NPT and also the regional ramifications of deterrence and regional stability of expanding naval forces. Possible solutions to close the loophole are discussed, including expanding the scope of the Fissile Material Cut-off Treaty, employing LEU fuel instead of HEU fuel in naval reactors, amending the NPT, creating an export control regime for naval nuclear reactors, and forming individual naval reactor safeguards agreements.

  7. Interactive nuclear plant analyzer for VVER-440 reactor

    SciTech Connect (OSTI)

    Shier, W.; Horak, W.; Kennett, R.

    1992-05-01

    This document discusses an interactive nuclear plant analyzer (NPA) which has been developed for a VVER-440, Model 213 reactor for use in the training of plant personnel, the development and verification of plant operating procedures, and in the analysis of various anticipated operational occurrences and accident scenarios. This NPA is operational on an IBM RISC-6000 workstation and utilizes the RELAP5/MOD2 computer code for the calculation of the VVER-440 reactor response to the interactive commands initiated by the NPA operator.

  8. N.R. 20 FOSSIL-FUELED POWER PLANTS; 21 SPECIFIC NUCLEAR REACTORS...

    Office of Scientific and Technical Information (OSTI)

    20 FOSSIL-FUELED POWER PLANTS; 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 14 SOLAR ENERGY; 15 GEOTHERMAL ENERGY; GEOTHERMAL POWER PLANTS; COMPUTERIZED SIMULATION; HEAT...

  9. 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 (OSTI)

    M. Chen; CM Regan; D. Noe

    2006-01-09

    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.

  10. Analysis of reactor trips originating in balance of plant systems

    SciTech Connect (OSTI)

    Stetson, F.T.; Gallagher, D.W.; Le, P.T.; Ebert, M.W. )

    1990-09-01

    This report documents the results of an analysis of balance-of-plant (BOP) related reactor trips at commercial US nuclear power plants of a 5-year period, from January 1, 1984, through December 31, 1988. The study was performed for the Plant Systems Branch, Office of Nuclear Reactor Regulation, US Nuclear Regulatory Commission. The objectives of the study were: to improve the level of understanding of BOP-related challenges to safety systems by identifying and categorizing such events; to prepare a computerized data base of BOP-related reactor trip events and use the data base to identify trends and patterns in the population of these events; to investigate the risk implications of BOP events that challenge safety systems; and to provide recommendations on how to address BOP-related concerns in regulatory context. 18 refs., 2 figs., 27 tabs.

  11. Medium Power Lead Alloy Fast Reactor Balance of Plant Options

    SciTech Connect (OSTI)

    Vaclav Dosta; Pavel Hejzlar; Neil E. Todreas; Jacopo Buongiorno

    2004-09-01

    Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature (~550°C) compared to that of light water reactors (~300°C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO2 (S-CO2) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545°C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312°C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO2 cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be the S-CO2 recompression cycle, because it

  12. Nuclear plant-aging research on reactor protection systems

    SciTech Connect (OSTI)

    Meyer, L.C.

    1988-01-01

    This report presents the rsults of a review of the Reactor Trip System (RTS) and the Engineered Safety Feature Actuating System (ESFAS) operating experiences reported in Licensee Event Reports (LER)s, the Nuclear Power Experience data base, Nuclear Plant Reliability Data System, and plant maintenance records. Our purpose is to evaluate the potential significance of aging, including cycling, trips, and testing as contributors to degradation of the RTS and ESFAS. Tables are presented that show the percentage of events for RTS and ESFAS classified by cause, components, and subcomponents for each of the Nuclear Steam Supply System vendors. A representative Babcock and Wilcox plant was selected for detailed study. The US Nuclear Regulatory Commission's Nuclear Plant Aging Research guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure modes for the RTS and ESFAS as generic instrumentation and control systems. Functional indicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed.

  13. Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

    SciTech Connect (OSTI)

    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.

    2006-12-11

    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.

  14. Naval Waste Package Design Report

    SciTech Connect (OSTI)

    M.M. Lewis

    2004-03-15

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

  15. Dual-phase reactor plant with partitioned isolation condenser

    DOE Patents [OSTI]

    Hui, Marvin M.

    1992-01-01

    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.

  16. Fast reactor power plant design having heat pipe heat exchanger

    DOE Patents [OSTI]

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

    1984-08-30

    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.

  17. Fast reactor power plant design having heat pipe heat exchanger

    DOE Patents [OSTI]

    Huebotter, Paul R.; McLennan, George A.

    1985-01-01

    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.

  18. Initiating Events for Multi-Reactor Plant Sites

    SciTech Connect (OSTI)

    Muhlheim, Michael David; Flanagan, George F.; Poore, III, Willis P.

    2014-09-01

    Inherent in the design of modular reactors is the increased likelihood of events that initiate at a single reactor affecting another reactor. Because of the increased level of interactions between reactors, it is apparent that the Probabilistic Risk Assessments (PRAs) for modular reactor designs need to specifically address the increased interactions and dependencies.

  19. EIS-0274: Disposal of S3G and D1G Prototype Reactor Plants

    Broader source: Energy.gov [DOE]

    This EIS analyzes the options and alternatives for the handling of the S3G and D1G Prototype reactor plants. Alternatives include their of prompt dismantlement, a deferred dismantlement alternative, and a no action alternative of keeping the defueled S3G and D1G Prototype reactor plants in protective storage indefinitely.

  20. REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

    DOE Patents [OSTI]

    Loeb, E.

    1961-01-17

    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.

  1. Knowledge and abilities catalog for nuclear power plant operators: Pressurized water reactors. Revision 1

    SciTech Connect (OSTI)

    1995-08-01

    This document provides the basis for the development of content-valid licensing examinations for reactor operators and senior reactor operators. The examinations developed using the PWR catalog will cover those topics listed under Title 10, (ode of Federal Regulations Part 55. The PWR catalog contains approximately 5100 knowledge and ability (K/A) statements for reactor operators and senior reactor operators. The catalog is organized into six major sections: Catalog Organization; Generic Knowledge and Abilities; Plant Systems; Emergency and Abnormal Plant Evolutions; Components and Theory.

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

    Office of Legacy Management (LM)

    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

  3. EIS-0108: L-Reactor Operation, Savannah River Plant, Aiken, South Carolina

    Broader source: Energy.gov [DOE]

    This Environmental Impact Statement (EIS) was prepared to provide environmental input into the proposed decision to restart L-Reactor operation at the Savannah River Plant (SRP). The Savannah River Plant is a major U.S. Department of Energy (DOE) installation for the production of defense nuclear materials. The proposed restart of L–Reactor would provide defense nuclear materials (i.e. , plutonium) to wet current and near-term needs for national defense purposes.

  4. FY 2012 Budget Hearing Testimony on Nuclear Nonproliferation and Naval

    National Nuclear Security Administration (NNSA)

    Reactor Programs before the House Appropriations Committee, Energy and Water Development Subcommittee | National Nuclear Security Administration | (NNSA) on Nuclear Nonproliferation and Naval Reactor Programs before the House Appropriations Committee, Energy and Water Development Subcommittee March 02, 2011 Chairman Frelinghuysen and Ranking Member Pastor, thank you for the opportunity to join you today to discuss the investments the President has requested for our nuclear nonproliferation

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

    SciTech Connect (OSTI)

    M. J. Russell

    2006-06-01

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1993-05-13

    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.

  7. Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

    SciTech Connect (OSTI)

    Not Available

    1986-09-01

    The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) 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 and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised.

  8. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    SciTech Connect (OSTI)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers.

  9. Naval Reactors | Y-12 National Security Complex

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

    Current Issues & Trends See more › Hurricane activity drives declines in Gulf of Mexico natural gas production productionweatherGulf of Mexico Gulf Coast's first ethane shipment soon to leave for Europe exportsEuropeethane As Japan and South Korea import less LNG, other Asian countries begin to import more LNGChinaIndiaJapanSouth Korea Future U.S. tight oil and shale gas production depends on resources, technology, markets productionshaledrillingtight oilAEO2016 Asian LNG imports increase

  10. Naval reactors in need of redesign

    SciTech Connect (OSTI)

    Kramer, David

    2015-05-15

    Nonproliferation concerns should propel US Navy to switch to safer nuclear fuel, says FAS task force.

  11. Optimal Coupling of a Nuclear Reactor and a Thermal Desalination Plant

    SciTech Connect (OSTI)

    Caruso, G.; Naviglio, A.; Nisan, S.; Bielak, B.; Cinotti, L.; Humphries, J.R.; Martins, N.; Volpi, L.

    2002-07-01

    The present study, performed in the framework of the EURODESAL Project (5. EU FWP), deals with the analysis of the 'optimum' coupling of a PWR and of a HTGR plant with a thermal desalination plant, based on the Multiple Effects process. The reference reactors are the AP600 and the PWR900 as Pressurized reactors and the GT-MHR as Gas reactor. The calculations performed show that there are several technical solutions allowing to couple PWRs and GRs to a ME desalination plant. The optimization criteria concern the technical feasibility of the coupling, producing the maximum quantity of fresh water at the lower cost, without unacceptable reduction of the electrical power produced and without undue health hazard for population. (authors)

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

    Office of Environmental Management (EM)

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

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

    SciTech Connect (OSTI)

    Wayne Moe

    2013-05-01

    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).

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

    SciTech Connect (OSTI)

    B. Levine

    2006-01-27

    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.

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

    SciTech Connect (OSTI)

    1995-08-01

    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.

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

    SciTech Connect (OSTI)

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

    1981-09-15

    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.

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

    SciTech Connect (OSTI)

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

    2010-11-01

    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 322°C and 750°C, 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.

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

    SciTech Connect (OSTI)

    L.E. Demick

    2010-09-01

    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.

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

    SciTech Connect (OSTI)

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

    2007-03-21

    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

  20. Categorical Exclusion Determinations: Naval Nuclear Propulsion Program |

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

    Department of Energy Naval Nuclear Propulsion Program Categorical Exclusion Determinations: Naval Nuclear Propulsion Program Categorical Exclusion Determinations issued by Naval Nuclear Propulsion Program. DOCUMENTS AVAILABLE FOR DOWNLOAD September 25, 2015 CX-014279: Categorical Exclusion Determination Lower Level Guard Post Replacement Project CX(s) Applied: B1.11, B1.15, B2.2 Date: 09/25/2015 Location(s): New York Offices(s): Naval Nuclear Propulsion Program July 6, 2015 CX-013878:

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

    SciTech Connect (OSTI)

    Wood, RT

    2004-09-27

    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.

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

    SciTech Connect (OSTI)

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

    2008-08-01

    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 540°C and 900°C, 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%.

  3. Medium-Power Lead-Alloy Fast Reactor Balance-of-Plant Options

    SciTech Connect (OSTI)

    Dostal, Vaclav [Massachusetts Institute of Technology (United States); Hejzlar, Pavel [Massachusetts Institute of Technology (United States); Todreas, Neil E. [Massachusetts Institute of Technology (United States); Buongiorno, Jacopo [Idaho National Engineering and Environmental Laboratory (United States)

    2004-09-15

    Proper selection of the power conversion cycle is a very important step in the design of a nuclear reactor. Due to the higher core outlet temperature ({approx}550 deg. C) compared to that of light water reactors ({approx}300 deg. C), a wide portfolio of power cycles is available for the lead alloy fast reactor (LFR). Comparison of the following cycles for the LFR was performed: superheated steam (direct and indirect), supercritical steam, helium Brayton, and supercritical CO{sub 2} (S-CO{sub 2}) recompression. Heat transfer from primary to secondary coolant was first analyzed and then the steam generators or heat exchangers were designed. The direct generation of steam in the lead alloy coolant was also evaluated. The resulting temperatures of the secondary fluids are in the range of 530-545 deg. C, dictated by the fixed space available for the heat exchangers in the reactor vessel. For the direct steam generation situation, the temperature is 312 deg. C. Optimization of each power cycle was carried out, yielding net plant efficiency of around 40% for the superheated steam cycle while the supercritical steam and S-CO{sub 2} cycles achieved net plant efficiency of 41%. The cycles were then compared based on their net plant efficiency and potential for low capital cost. The superheated steam cycle is a very good candidate cycle given its reasonably high net plant efficiency and ease of implementation based on the extensive knowledge and operating experience with this cycle. Although the supercritical steam cycle net plant efficiency is slightly better than that of the superheated steam cycle, its high complexity and high pressure result in higher capital cost, negatively affecting plant economics. The helium Brayton cycle achieves low net plant efficiency due to the low lead alloy core outlet temperature, and therefore, even though it is a simpler cycle than the steam cycles, its performance is mediocre in this application. The prime candidate, however, appears to be

  4. The Next Generation Nuclear Plant Graphite Creep Experiment Irradiation in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Blaine Grover

    2010-10-01

    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 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, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack 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 have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. 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

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

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2008-04-01

    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 900°C 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, 2¼Cr-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

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

    SciTech Connect (OSTI)

    J. K. Wright; R. N. Wright

    2010-07-01

    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 750°C, 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.

  7. REACTOR

    DOE Patents [OSTI]

    Szilard, L.

    1963-09-10

    A breeder reactor is described, including a mass of fissionable material that is less than critical with respect to unmoderated neutrons and greater than critical with respect to neutrons of average energies substantially greater than thermal, a coolant selected from sodium or sodium--potassium alloys, a control liquid selected from lead or lead--bismuth alloys, and means for varying the quantity of control liquid in the reactor. (AEC)

  8. REACTOR

    DOE Patents [OSTI]

    Christy, R.F.

    1961-07-25

    A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

  9. United States Department of Energy`s reactor core protection evaluation methodology for fires at RBMK and VVER nuclear power plants. Revision 1

    SciTech Connect (OSTI)

    1997-06-01

    This document provides operators of Soviet-designed RBMK (graphite moderated light water boiling water reactor) and VVER (pressurized light water reactor) nuclear power plants with a systematic Methodology to qualitatively evaluate plant response to fires and to identify remedies to protect the reactor core from fire-initiated damage.

  10. Reactor Chamber and Balance-of-Plant Characteristics for a Fast-Ignition Heavy-Ion Fusion Power Plant

    SciTech Connect (OSTI)

    Medin, Stanislav; Churazov, Mikhail; Koshkarev, Dmitri; Sharkov, Boris; Orlov, Yurii; Suslin, Viktor; Zemskov, Eugeni

    2003-05-15

    The concept of a fast-ignition heavy-ion fusion (FIHIF) power plant involves a cylindrical target and superhigh energy ion beams. The driver produces one plus/minus charge state multimass platinum ions with energy of 100 GeV. The driver efficiency and the target gain are taken as 0.25 and 100, respectively. The preliminary data on the energy fluxes delivered to the reactor chamber wall by the 500-MJ fusion yield are presented. The reactor chamber designed has two sections. In the first section, the microexplosions occur, and in the second section of bigger volume the expansion and condensation of vapors take place. The response of the blanket and the thin liquid film at the first-wall surface is evaluated. Lithium-lead eutectic is taken as a coolant. The evaporated mass and the condensation time are estimated, taking into account major thermophysical effects. The estimated neutron spectrum from the FIHIF target gives an average neutron energy of 11.9 MeV. The mechanical stresses in the construction material due to neutron energy release are evaluated. The outlet coolant chamber temperature is taken as 550 deg. C. The heat conversion system consisting of three coolant loops provides a net efficiency of the FIHIF power plant of 0.37.

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

    SciTech Connect (OSTI)

    Phillip Mills

    2012-02-01

    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.

  12. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    SciTech Connect (OSTI)

    Kambe, Mitsuru [Central Research Institute of Electric Power Industry (CRIEPI), 2-11-1, Iwado Kita, Komae-shi, Tokyo, 201-8511 (Japan); Tsunoda, Hirokazu [Mitsubishi Research Institute, Inc. 3-6, Otemachi 2-chome, Chiyoda-ku, Tokyo, 100-8141 (Japan); Mishima, Kaichiro [Research Reactor Institute, Kyoto University, Kumatori-cho, Sennan-gun, Osaka, 590-20494 (Japan); Iwamura, Takamichi [Japan Atomic Energy Research Institute, 2-4, Shirakata-shirane, Tokai-mura, Naka-gun, Ibaraki-ken, 319-1195 (Japan)

    2002-07-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B{sub 4}C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

  13. REACTOR

    DOE Patents [OSTI]

    Roman, W.G.

    1961-06-27

    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.

  14. 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 (OSTI)

    1993-09-15

    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.

  15. REACTORS

    DOE Patents [OSTI]

    Spitzer, L. Jr.

    1961-10-01

    Thermonuclear reactors, methods, and apparatus are described for controlling and confining high temperature plasma. Main axial confining coils in combination with helical windings provide a rotational transform that avoids the necessity of a figure-eight shaped reactor tube. The helical windings provide a multipolar helical magnetic field transverse to the axis of the main axial confining coils so as to improve the effectiveness of the confining field by counteracting the tendency of the more central lines of force in the stellarator tube to exchange positions with the magnetic lines of force nearer the walls of the tube. (AEC)

  16. Westinghouse Small Modular Reactor balance of plant and supporting systems design

    SciTech Connect (OSTI)

    Memmott, M. J.; Stansbury, C.; Taylor, C.

    2012-07-01

    The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

  17. Naval Research Laboratory Technology Marketing Summaries - Energy...

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

    Naval Research Laboratory 8 Technology Marketing Summaries Category Title and Abstract Laboratories Date Solar Photovoltaic Find More Like This Sputtered Thin Film Photovoltaics ...

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

    SciTech Connect (OSTI)

    1997-12-01

    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.

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

    SciTech Connect (OSTI)

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

    1984-01-01

    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.

  20. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect (OSTI)

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

    1999-01-01

    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.

  1. The application of Plant Reliability Data Information System (PRINS) to CANDU reactor

    SciTech Connect (OSTI)

    Hwang, S. W.; Lim, Y. H.; Park, H. C.

    2012-07-01

    As risk-informed applications (RIAs) are actively implanted in the nuclear industry, an issue associated with technical adequacy of Probabilistic Safety Assessment (PSA) arises in its modeling and data sourcing. In Korea, PSA for all Korean NPPs has been completed and KHNP(Korea Hydro and Nuclear Power Plant Company) developed the database called the Plant Reliability Data Information System (PRinS). It has several characteristics that distinguish it from other database system such as NPRDs (INPO,1994), PRIS (IAEA), and SRDF (EdF). This database has the function of systematic data management such as automatic data-gathering, periodic data deposition and updating, statistical analysis including Bayesian method, and trend analysis of failure rate or unavailability. In recent PSA for CANDU reactor, the component failure data of EPRI ALWR URD and Component Reliability Database were preferentially used as generic data set. The error factor for most component failure data was estimated by using the information NUREG/CR-4550 and NUREG/CR-4639. Also, annual trend analysis was performed for the functional losses of components using the statistical analysis and chart module of PRinS. Furthermore, the database has been updated regularly and maintained as a living program to reflect the current status. This paper presents the failure data analysis using PRinS which provides Bayesian analysis on main components in the CANDU reactor. (authors)

  2. Development Wells At Fallon Naval Air Station Area (Sabin, Et...

    Open Energy Info (EERE)

    Fallon Naval Air Station Area (Sabin, Et Al., 2010) Jump to: navigation, search GEOTHERMAL ENERGYGeothermal Home Exploration Activity: Development Wells At Fallon Naval Air Station...

  3. Naval Spent Fuel Rail Shipment Accident Exercise Objectives

    Office of Environmental Management (EM)

    NAVAL SPENT FUEL RAIL SHIPMENT ACCIDENT EXERCISE OBJECTIVES * Familiarize stakeholders with the Naval spent fuel ACCIDENT EXERCISE OBJECTIVES Familiarize stakeholders with the ...

  4. 2013 Federal Energy and Water Management Award Winner Naval Sea...

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

    Naval Sea Systems Command 2013 Federal Energy and Water Management Award Winner Naval Sea Systems Command PDF icon fewm13nswcphiladelphiahighres.pdf PDF icon ...

  5. Reactor

    DOE Patents [OSTI]

    Evans, Robert M.

    1976-10-05

    1. A neutronic reactor having a moderator, coolant tubes traversing the moderator from an inlet end to an outlet end, bodies of material fissionable by neutrons of thermal energy disposed within the coolant tubes, and means for circulating water through said coolant tubes characterized by the improved construction wherein the coolant tubes are constructed of aluminum having an outer diameter of 1.729 inches and a wall thickness of 0.059 inch, and the means for circulating a liquid coolant through the tubes includes a source of water at a pressure of approximately 350 pounds per square inch connected to the inlet end of the tubes, and said construction including a pressure reducing orifice disposed at the inlet ends of the tubes reducing the pressure of the water by approximately 150 pounds per square inch.

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

    SciTech Connect (OSTI)

    J. M. Beck; L. F. Pincock

    2011-04-01

    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.

  7. Naval Petroleum Reserves | Department of Energy

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

    For much of the 20th century, the Naval Petroleum and Oil Shale Reserves served as a ... 1900s, the government-owned petroleum and oil shale properties were originally envisioned ...

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

    SciTech Connect (OSTI)

    Curtis Smith; Scott Beck; Bill Galyean

    2005-09-01

    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.

  9. Turning points in reactor design

    SciTech Connect (OSTI)

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  10. Boiler Upgrades and Decentralizing Steam Systems Save Water and Energy at Naval Air Station Oceana

    Office of Energy Efficiency and Renewable Energy (EERE)

    Case study details Naval Air Station Oceana findings that its heating needs could be met more efficiently by replacing its central plant with a combination of distributed boilers and ground source heat pumps. The results saved more than 1 million MBtu in energy and 19,574 Kgal of water annually.

  11. The use of LBB concept in French fast reactors: Application to SPX plant

    SciTech Connect (OSTI)

    Turbat, A.; Deschanels, H.; Sperandio, M.

    1997-04-01

    The leak before break (LBB) concept was not used at the design level for SUPERPHENIX (SPX), but different studies have been performed or are in progress concerning different components : Main Vessel (MV), pipings. These studies were undertaken to improve the defense in depth, an approach used in all French reactors. In a first study, the LBB approach has been applied to the MV of SPX plant to verify the absence of risk as regards the core supporting function and to help in the definition of in-service inspection (ISI) program. Defining a reference semi-elliptic defect located in the welds of the structure, it is verified that the crack growth is limited and that the end-of-life defect is smaller than the critical one. Then it is shown that the hoop welds (those which are the most important for safety) located between the roof and the triple point verify the leak-before-break criteria. However, generally speaking, the low level of membrane primary stresses which is favorable for the integrity of the vessel makes the application of the leak-before-break concept more difficult due to small crack opening areas. Finally, the extension of the methodology to the secondary pipings of SPX incorporating recent European works of DCRC is briefly presented.

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

    SciTech Connect (OSTI)

    Not Available

    1994-01-15

    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.

  13. System Evaluation and Economic Analysis of a Nuclear Reactor Powered High-Temperature Electrolysis Hydrogen-Production Plant

    SciTech Connect (OSTI)

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

    2010-06-01

    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 540°C and 900°C, 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 47.1% 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%.

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

    SciTech Connect (OSTI)

    Lata

    1996-09-26

    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.

  15. Investigation of plant control strategies for the supercritical C0{sub 2}Brayton cycle for a sodium-cooled fast reactor using the plant dynamics code.

    SciTech Connect (OSTI)

    Moisseytsev, A.; Sienicki, J.

    2011-04-12

    The development of a control strategy for the supercritical CO{sub 2} (S-CO{sub 2}) Brayton cycle has been extended to the investigation of alternate control strategies for a Sodium-Cooled Fast Reactor (SFR) nuclear power plant incorporating a S-CO{sub 2} Brayton cycle power converter. The SFR assumed is the 400 MWe (1000 MWt) ABR-1000 preconceptual design incorporating metallic fuel. Three alternative idealized schemes for controlling the reactor side of the plant in combination with the existing automatic control strategy for the S-CO{sub 2} Brayton cycle are explored using the ANL Plant Dynamics Code together with the SAS4A/SASSYS-1 Liquid Metal Reactor (LMR) Analysis Code System coupled together using the iterative coupling formulation previously developed and implemented into the Plant Dynamics Code. The first option assumes that the reactor side can be ideally controlled through movement of control rods and changing the speeds of both the primary and intermediate coolant system sodium pumps such that the intermediate sodium flow rate and inlet temperature to the sodium-to-CO{sub 2} heat exchanger (RHX) remain unvarying while the intermediate sodium outlet temperature changes as the load demand from the electric grid changes and the S-CO{sub 2} cycle conditions adjust according to the S-CO{sub 2} cycle control strategy. For this option, the reactor plant follows an assumed change in load demand from 100 to 0 % nominal at 5 % reduction per minute in a suitable fashion. The second option allows the reactor core power and primary and intermediate coolant system sodium pump flow rates to change autonomously in response to the strong reactivity feedbacks of the metallic fueled core and assumed constant pump torques representing unchanging output from the pump electric motors. The plant behavior to the assumed load demand reduction is surprising close to that calculated for the first option. The only negative result observed is a slight increase in the intermediate

  16. Application of the LBB concept to nuclear power plants with WWER 440 and WWER 1000 reactors

    SciTech Connect (OSTI)

    Zdarek, J.; Pecinka, L.

    1997-04-01

    Leak-before-break (LBB) analysis of WWER type reactors in the Czech and Sloval Republics is summarized in this paper. Legislative bases, required procedures, and validation and verification of procedures are discussed. A list of significant issues identified during the application of LBB analysis is presented. The results of statistical evaluation of crack length characteristics are presented and compared for the WWER 440 Type 230 and 213 reactors and for the WWER 1000 Type 302, 320 and 338 reactors.

  17. The Naval Petroleum and Oil Shale Reserves | Department of Energy

    Energy Savers [EERE]

    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 ...

  18. 2013 Annual Planning Summary for the Naval Nuclear Propulsion Program

    Broader source: Energy.gov [DOE]

    The ongoing and projected Environmental Assessments and Environmental Impact Statements for 2013 and 2014 within the Naval Nuclear Propulsion Program.

  19. H Reactor - Hanford Site

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

    About Us Projects & Facilities H Reactor About Us About Hanford Cleanup Hanford History ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  20. C Reactor - Hanford Site

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

    C Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  1. F Reactor - Hanford Site

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

    About Us Projects & Facilities F Reactor About Us About Hanford Cleanup Hanford History ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  2. N Reactor - Hanford Site

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

    Projects & Facilities N Reactor About Us About Hanford Cleanup Hanford History Hanford ... 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and ...

  3. Naval Nuclear Propulsion | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    Naval Nuclear Propulsion Klotz visits Bettis Atomic Power Laboratory Lt. Gen. Frank G. Klotz, DOE Undersecretary for Nuclear Security and NNSA Administrator, visited the Bettis Atomic Power Laboratory in West Mifflin, PA on July 2, 2015. Gen. Klotz toured through several test facilities where Bettis personnel reviewed ongoing development efforts to qualify

  4. Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1995--September 30, 1996

    SciTech Connect (OSTI)

    1996-12-31

    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 1996 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.

  5. Advanced light water reactor plants System 80+{trademark} design certification program. Annual progress report, October 1, 1994--September 30, 1995

    SciTech Connect (OSTI)

    1998-09-01

    The purpose of this report is to provide the status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1995 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.

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

    SciTech Connect (OSTI)

    Donohue, M.L.; Price, M.E.

    1984-07-01

    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.

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

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    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.

  8. EIS-0259: Final Environmental Impact Statement

    Broader source: Energy.gov [DOE]

    Disposal of Decommissioned, Defueled Cruiser, Ohio Class, Los Angeles and Class Naval Reactor Plants

  9. EIS-0259: Record of Decision

    Broader source: Energy.gov [DOE]

    Disposal of Decommissioned, Defueled Cruiser, Ohio Class, Los Angeles and Class Naval Reactor Plants

  10. A probabilistic evaluation of the safety of Babcock and Wilcox nuclear reactor power plants with emphasis on historically observed operational events

    SciTech Connect (OSTI)

    Hsu, C.J.; Youngblood, R.W.; Fitzpatrick, R.G.; Amico, P.J.

    1989-03-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Reactor Regulation, Division of Engineering and System Technology (A/D for Systems), US Nuclear Regulatory Commission. This study was requested by the NRC to assist their staff in assessing the risk significance of features of the Babcock and Wilcox (B and W) reactor plant design in the light of recent operational events. This study focuses on a critical review of submissions from the B and W Owners Group (BWOG) and as an independent assessment of the risk significance of ''Category C'' events at each operating B and W reactor. Category C events are those in which system conditions reach limits which require significant safety system and timely operator response to mitigate. A precursor study for each of the major B and W historical Category C events also was carried out. In addition, selected PRAs for B and W reactor plants and plants with other pressurized water reactor (PWR) designs were reviewed to appraise their handling of Category C events, thereby establishing a comparison between the risk profiles of B and W reactor plants and those of other PWR designs. The effectiveness of BWOG recommendations set forth in Appendix J of the BWOG SPIP (Safety and Performance Improvement Program) report (BAW-1919) also was evaluated. 49 refs., 21 figs., 52 tabs.

  11. DESIGN ANALYSIS FOR THE NAVAL SNF WASTE PACKAGE

    SciTech Connect (OSTI)

    T.L. Mitchell

    2000-05-31

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

  12. United States Naval Surface Warfare Center | Open Energy Information

    Open Energy Info (EERE)

    Warfare Center Jump to: navigation, search Hydro | Hydrodynamic Testing Facilities Name United States Naval Surface Warfare Center Address Carderock, 9500 MacArthur Boulevard...

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

    SciTech Connect (OSTI)

    Donohue, M.L.; Price, M.E.

    1984-07-01

    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)

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

    SciTech Connect (OSTI)

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

    2013-05-01

    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.

  15. CX-008819: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Naval Reactors Facility Parking Lot Expansion General Plant Project CX(s) Applied: B1.15 Date: 06/20/2012 Location(s): Idaho Offices(s): Naval Nuclear Propulsion Program, Naval Reactors

  16. Sec. Moniz to Georgia, Energy Department Scheduled to Close on Loan Guarantees to Construct New Nuclear Power Plant Reactors

    Broader source: Energy.gov [DOE]

    Project represents first new nuclear reactors to begin construction in the United States in three decades

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

    SciTech Connect (OSTI)

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

    2013-03-01

    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.

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

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

    Department of Energy 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:

  19. Light water reactor program

    SciTech Connect (OSTI)

    Franks, S.M.

    1994-12-31

    The US Department of Energy`s Light Water Reactor Program is outlined. The scope of the program consists of: design certification of evolutionary plants; design, development, and design certification of simplified passive plants; first-of-a-kind engineering to achieve commercial standardization; plant lifetime improvement; and advanced reactor severe accident program. These program activities of the Office of Nuclear Energy are discussed.

  20. NNSA Administrator Addresses Stockpile Management Plans at New...

    National Nuclear Security Administration (NNSA)

    Naval Reactors began reactor and propulsion plant design in fiscal year 2010 for the OHIO Replacement submarine to support the Navy's schedule. Reactor plant components will be ...

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

    SciTech Connect (OSTI)

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

    2005-06-01

    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

  2. Daya Bay Reactor Neutrino Experiment

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

    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...

  3. Light Water Reactor Sustainability Program: Computer-based procedure for field activities: results from three evaluations at nuclear power plants

    SciTech Connect (OSTI)

    Oxstrand, Johanna; Bly, Aaron; LeBlanc, Katya

    2014-09-01

    Nearly all activities that involve human interaction with the systems of a nuclear power plant are guided by procedures. The paper-based procedures (PBPs) currently used by industry have a demonstrated history of ensuring safety; however, improving procedure use could yield tremendous savings in increased efficiency and safety. One potential way to improve procedure-based activities is through the use of computer-based procedures (CBPs). Computer-based procedures provide the opportunity to incorporate context driven job aids, such as drawings, photos, just-in-time training, etc into CBP system. One obvious advantage of this capability is reducing the time spent tracking down the applicable documentation. Additionally, human performance tools can be integrated in the CBP system in such way that helps the worker focus on the task rather than the tools. Some tools can be completely incorporated into the CBP system, such as pre-job briefs, placekeeping, correct component verification, and peer checks. Other tools can be partly integrated in a fashion that reduces the time and labor required, such as concurrent and independent verification. Another benefit of CBPs compared to PBPs is dynamic procedure presentation. PBPs are static documents which limits the degree to which the information presented can be tailored to the task and conditions when the procedure is executed. The CBP system could be configured to display only the relevant steps based on operating mode, plant status, and the task at hand. A dynamic presentation of the procedure (also known as context-sensitive procedures) will guide the user down the path of relevant steps based on the current conditions. This feature will reduce the user’s workload and inherently reduce the risk of incorrectly marking a step as not applicable and the risk of incorrectly performing a step that should be marked as not applicable. As part of the Department of Energy’s (DOE) Light Water Reactors Sustainability Program

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

    SciTech Connect (OSTI)

    Ballinger, Ronald G.; Wang, Chun Yun; Kadak, Andrew; Todreas, Neil; Mirick, Bradley; Demetri, Eli; Koronowski, Martin

    2004-08-30

    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

  5. DOE - Office of Legacy Management -- Naval Ordnance Plant - MI...

    Office of Legacy Management (LM)

    Eliminated from further consideration under FUSRAP - Referred to DoD for action Designated ... MI.0-03-1 Site Disposition: Eliminated - No Authority - Referred to DoD MI.0-03-1 ...

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

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

    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. RMOTC...

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

    Office of Legacy Management (LM)

    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

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

    SciTech Connect (OSTI)

    Not Available

    1994-08-01

    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.

  9. Reactor safety assessment system

    SciTech Connect (OSTI)

    Sebo, D.E.; Bray, M.A.; King, M.A.

    1987-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system under development for the United States Nuclear Regulatory Commission (USNRC). RSA is designed for use at the USNRC Operations Center in the event of a serious incident at a licensed nuclear power plant. RSAS is a situation assessment expert system which uses plant parametric data to generate conclusions for use by the NRC Reactor Safety Team. RSAS uses multiple rule bases and plant specific setpoint files to be applicable to all licensed nuclear power plants in the United States. RSAS currently covers several generic reactor categories and multiple plants within each category.

  10. Advanced Nuclear Reactors | Department of Energy

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

    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

  11. High Energy Utilization, Co-Generation Nuclear power Plants With Static Energy Conversion

    SciTech Connect (OSTI)

    El-Genk, Mohamed S.; Tournier, Jean-Michel P.

    2002-07-01

    In addition to being cost effective, very small nuclear power plants with static energy conversion could meet the needs and the energy mix in underdeveloped countries and remote communities, which may include electricity, residential and industrial space heating, seawater desalination, and/or high temperature process heat or steam for industrial uses. These plants are also an attractive option in naval, marine, and undersea applications, when the absence of a sound signature is highly desirable. An Analysis is performed of Gas Cooled Reactor (CGR) and Liquid Metal Cooled Reactor (LMR), very small nuclear power plants with static energy conversion, using a combination of options. These include Alkali Metal Thermal-to-Electric Converters (AMTECs) and both single segment and segmented thermoelectric converters. The total energy utilization of these plants exceeds 88%. It includes the fraction of the reactor's thermal power converted into electricity and delivered to the Grid at 6.6 kVA and those used for residential and industrial space heating at {approx}370 K, seawater desalination at 400 K, and/or high temperature process heat or steam at {approx}850 K. In addition to its inherently high reliability, modularity, low maintenance and redundancy, static energy conversion used in the present study could deliver electricity to the Grid at a net efficiency of 29.5%. A LMR plant delivers 2-3 times the fraction of the reactor thermal power converted into electricity in a GCR plant, but could not provide for both seawater desalination and high temperature process heat/steam concurrently, which is possible in GCR plants. The fraction of the reactor's thermal power used for non-electrical power generation in a GCR plant is {approx} 10 - 15% higher than in a LMR plant. (authors)

  12. Naval Spent Nuclear Fuel disposal Container System Description Document

    SciTech Connect (OSTI)

    N. E. Pettit

    2001-07-13

    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.

  13. Report

    National Nuclear Security Administration (NNSA)

    Naval Nuclear Laboratory - Knolls Site/Kesselring Site Naval Nuclear Laboratory - Bettis Site Los Alamos National Laboratory (LANL) Nevada National Security Site Naval Nuclear Laboratory - Naval Reactors Facility Sandia National Laboratories (SNL) Lawrence Livermore National Laboratory (LLNL) Savannah River Site (SRS) DOE/NNSA Headquarters Pantex Plant (PX) Albuquerque Complex Headquarters National Security Laboratories Plants and Sites Naval Nuclear Laboratories The Nuclear Security Enterprise

  14. Naval Petroleum and Oil Shale Reserves. Annual report of operations

    SciTech Connect (OSTI)

    Not Available

    1982-10-01

    The Naval Petroleum and Oil Shale Reserves (NPOSR), created to provide a source of liquid fuels for the armed forces during national emergencies, were established by a series of Executive Orders between 1912 and 1924. Following the 1973 to 1974 Arab Oil Embargo, which demonstrated the Nation's vulnerability to oil supply interruptions, the Congress authorized and directed in 1974 that the Reserves be explored and developed to their full economic and productive potential. In October 1981, the President notified the Congress of his decision to extend production of the Naval Petroleum Reserves to April 6, 1985. That decision became final when the Congress did not exercise its authority to disapprove the action. With regard to the Naval Oil Shale Reserves (NOSRs), a program was initiated in 1977 to examine the resource for development and subsequent production should national defense requirements so dictate.

  15. Thermal Evaluation for the Naval SNF Waste Package

    SciTech Connect (OSTI)

    T.L. Mitchell

    2000-04-25

    The purpose of this calculation is to evaluate the thermal performance of the naval long spent nuclear fuel (SNF) waste package (WP) under multiple disposal conditions in a monitored geologic repository (MGR). The scope of this calculation is limited to determination of thermal temperature profiles upon the surface of, and within, the naval long SNF WP. The objective is to develop a temperature profile history within the WP, at time increments up to 10,000 years of emplacement. The results of this calculation are intended to support the Naval SNF WP Analysis and Model Report (AMR) for Site Recommendation (SR). This calculation was performed to the specifications within its Technical Development Plan (TDP) (Ref. 8.16). This calculation is developed and documented in accordance with the AP-3.12Q/REV. 0IICN. 0 procedure, Calculations.

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

    SciTech Connect (OSTI)

    Myers, Carl W; Elkins, Ned Z

    2008-01-01

    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.

  17. Plutonium Uranium Extraction Plant (PUREX) - Hanford Site

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

    About Us Projects & Facilities Plutonium Uranium Extraction Plant (PUREX) About Us About ... and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage ...

  18. Reduction-Oxidation Plant (REDOX) - Hanford Site

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

    Facilities Reduction-Oxidation Plant (REDOX) About Us About Hanford Cleanup Hanford ... and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage ...

  19. DOE - Office of Legacy Management -- U S Naval Radiological Defense

    Office of Legacy Management (LM)

    Laboratory - CA 0-06 Naval Radiological Defense Laboratory - CA 0-06 FUSRAP Considered Sites Site: U. S. NAVAL RADIOLOGICAL DEFENSE LABORATORY (CA.0-06) Eliminated from consideration under FUSRAP - Referred to the DoD Designated Name: Not Designated Alternate Name: None Location: San Francisco , California CA.0-06-1 Evaluation Year: 1987 CA.0-06-1 Site Operations: NRC licensed DoD facility which used small quantities of nuclear materials for R&D purposes and decontaminated ships.

  20. T Plant - Hanford Site

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

    T Plant About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact ... and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage ...

  1. Technical evaluation report on the monitoring of electric power to the reactor-protection system for the Brunswick Steam Electric Plant, Units 1 and 2

    SciTech Connect (OSTI)

    Selan, J.C.

    1982-04-26

    This report documents the technical evaluation of the monitoring of electric power to the reactor protection system (RPS) at the Brunswick Steam Electric Plant, Units 1 and 2. The evaluation is to determine if the proposed design modification will protect the RPS from abnormal voltage and frequency conditions which could be supplied from the power supplies and will meet certain requirements set forth by the Nuclear Regulatory Commission. The proposed design modifications with time delays verified by GE, will protect the RPS from sustained abnormal voltage and frequency conditions from the supplying sources.

  2. U.S. Naval Station, Guantanamo Bay, Cuba | Department of Energy

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

    Naval Station, Guantanamo Bay, Cuba U.S. Naval Station, Guantanamo Bay, Cuba Fact sheet describes the Energy Savings Performance Contract (ESPC) success story on environmental stewardship and cost savings at the U.S. Naval Station at Guantanamo Bay, Cuba. Download the U.S. Naval Station at Guantanamo Bay, Cuba fact sheet. (316.37 KB) More Documents & Publications Idaho Operations AMWTP Fact Sheet Heating Ventilation and Air Conditioning Efficiency Greenpower Trap Mufflerl System

  3. Virginia Nuclear Profile - Power Plants

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

    nuclear power plants, summer capacity and net generation, 2010" "Plant nametotal ... Electric & Power Co" "2 Plants 4 Reactors","3,501","26,572",100.0 "Note: ...

  4. B Reactor - Hanford Site

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

    Reactor About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area 324 Building 325 Building 400 Area/Fast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim Storage Area Canyon Facilities Cold Test Facility D and DR Reactors Effluent Treatment Facility Environmental Restoration Disposal Facility F Reactor H

  5. Final MTI Data Report: Dahlgren Naval Surface Warfare Center

    SciTech Connect (OSTI)

    Parker, M.J.

    2003-03-17

    During the period from February 2001 to August 2002, paved-surface (tarmac) temperatures were collected at the Dahlgren Naval Surface Warfare Center. This effort was led by the Savannah River Technology Center (SRTC), with the assistance of base personnel, as part of SRTC's ground truth mission for the U.S. Department of Energy's Multispectral Thermal Imager (MTI) satellite.

  6. Endangered species and cultural resources program, Naval Petroleum Reserves in California, annual report FY97

    SciTech Connect (OSTI)

    1998-05-01

    The Naval Petroleum Reserves in California (NPRC) are oil fields administered by the DOE in the southern San Joaquin Valley of California. Four federally endangered animal species and one federally threatened plant species are known to occur on NPRC: San Joaquin kit fox (Vulpes macrotis mutica), blunt-nosed leopard lizard (Gambelia silus), giant kangaroo rat (Dipodomys ingens), Tipton kangaroo rat (Dipodomys nitratoides), and Hoover`s wooly-star (Eriastrum hooveri). All five are protected under the Endangered Species Act (ESA) of 1973. The DOE/NPRC is obliged to determine whether actions taken by their lessees on Naval Petroleum Reserve No. 2 (NPR-2) will have any effects on endangered species or their habitats. The primary objective of the Endangered Species and Cultural Resources Program is to provide NPRC with the scientific expertise necessary for compliance with the ESA, the National Environmental Policy Act (NEPA), and the National Historic Preservation Act (NHPA). The specific objective of this report is to summarize progress, results, and accomplishments of the program during fiscal year 1997 (FY97).

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

    SciTech Connect (OSTI)

    H. Marr

    2006-10-25

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

  8. Preliminary results of calculations for heavy-water nuclear-power-plant reactors employing {sup 235}U, {sup 233}U, and {sup 232}Th as a fuel and meeting requirements of a nonproliferation of nuclear weapons

    SciTech Connect (OSTI)

    Ioffe, B. L.; Kochurov, B. P.

    2012-02-15

    A physical design is developed for a gas-cooled heavy-water nuclear reactor intended for a project of a nuclear power plant. As a fuel, the reactor would employ thorium with a small admixture of enriched uranium that contains not more than 20% of {sup 235}U. It operates in the open-cycle mode involving {sup 233}U production from thorium and its subsequent burnup. The reactor meets the conditions of a nonproliferation of nuclear weapons: the content of fissionable isotopes in uranium at all stages of the process, including the final one, is below the threshold for constructing an atomic bomb, the amount of product plutonium being extremely small.

  9. Audit Report: OIG-0884 | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) 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 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 design, operations and maintenance

  10. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani

    1996-01-01

    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.

  11. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19

    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.

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

    SciTech Connect (OSTI)

    John Collins

    2009-08-01

    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.

  13. The impact of fuel cladding failure events on occupational radiation exposures at nuclear power plants: Case study, PWR (pressurized-water reactor) during an outage

    SciTech Connect (OSTI)

    Moeller, M.P.; Martin, G.F.; Kenoyer, J.L.

    1987-08-01

    This report is the second in a series of case studies designed to evaluate the magnitude of increase in occupational radiation exposures at commercial US nuclear power plants resulting from small incidents or abnormal events. The event evaluated is fuel cladding failure, which can result in elevated primary coolant activity and increased radiation exposure rates within a plant. For this case study, radiation measurements were made at a pressurized-water reactor (PWR) during a maintenance and refueling outage. The PWR had been operating for 22 months with fuel cladding failure characterized as 105 pin-hole leakers, the equivalent of 0.21% failed fuel. Gamma spectroscopy measurements, radiation exposure rate determinations, thermoluminescent dosimeter (TLD) assessments, and air sample analyses were made in the plant's radwaste, pipe penetration, and containment buildings. Based on the data collected, evaluations indicate that the relative contributions of activation products and fission products to the total exposure rates were constant over the duration of the outage. This constancy is due to the significant contribution from the longer-lived isotopes of cesium (a fission product) and cobalt (an activation product). For this reason, fuel cladding failure events remain as significant to occupational radiation exposure during an outage as during routine operations. As documented in the previous case study (NUREG/CR-4485 Vol. 1), fuel cladding failure events increased radiation exposure rates an estimated 540% at some locations of the plant during routine operations. Consequently, such events can result in significantly greater radiation exposure rates in many areas of the plant during the maintenance and refueling outages than would have been present under normal fuel conditions.

  14. Energy use baselining study for the National Naval Medical Center

    SciTech Connect (OSTI)

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

    1992-04-01

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

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

    Office of Legacy Management (LM)

    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

  16. Results of detailed analyses performed on boring cores extracted from the concrete floors of the Fukushima Daiichi nuclear power plant reactor buildings

    SciTech Connect (OSTI)

    Maeda, Koji; Sasaki, S.; Kumai, M.; Sato, Isamu; Osaka, Masahiko; Fukushima, Mineo; Kawatsuma, Shinji; Goto, Tetsuo; Sakai, Hitoshi; Chigira, Takayuki; Murata, Hirotoshi

    2013-07-01

    Due to the massive earthquake and tsunami on March 11, 2011, and the following severe accident at the Fukushima Daiichi Nuclear Power Plant, concrete surfaces within the reactor buildings were exposed to radioactive liquid and vapor phase contaminants. In order to clarify the situation of this contamination in the reactor buildings of Units 1, 2 and 3, selected samples were transported to the Fuels Monitoring Facility in the Oarai Engineering Center of JAEA where they were subjected to analyses to determine the surface radionuclide concentrations and to characterize the radionuclide distributions in the samples. In particular, penetration of radiocesium in the surface coatings layer and sub-surface concrete was evaluated. The analysis results indicate that the situation of contamination in the building of Unit 2 was different from others, and the protective surface coatings on the concrete floors provided significant protection against radionuclide penetration. The localized penetration of contamination in the concrete floors was found to be confined within a millimeter of the surface of the coating layer of some millimeters. (authors)

  17. Reactor power for large displacement autonomous underwater vehicles...

    Office of Scientific and Technical Information (OSTI)

    Country of Publication: United States Language: English Subject: 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS autonomous underwater vehicle; reactor power Word Cloud More ...

  18. Self-Sustaining Thorium Boiling Water Reactors (Technical Report...

    Office of Scientific and Technical Information (OSTI)

    of the Hitachi RBWR core designs and sodium cooled fast reactor counterparts - the ARR and ABR; ... Language: English Subject: 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; ...

  19. 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 (OSTI)

    Moisseytsev, A.; Sienicki, J. J.

    2012-05-10

    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

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

    SciTech Connect (OSTI)

    1996-12-31

    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.

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

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

    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

  2. EIS-0453: Recapitalization of Infrastructure Supporting Naval Spent Nuclear Fuel Handling at the Idaho National Laboratory

    Broader source: Energy.gov [DOE]

    The Draft EIS evaluates the potential environmental impacts associated with recapitalizing the infrastructure needed to ensure the long-term capability of the Naval Nuclear Propulsion Program (NNPP) to support naval spent nuclear fuel handling capabilities provided by the Expended Core Facility (ECF). Significant upgrades are necessary to ECF infrastructure and water pools to continue safe and environmentally responsible naval spent nuclear fuel handling until at least 2060.

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

    SciTech Connect (OSTI)

    Smith, C

    2010-02-22

    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.

  4. Arkansas Nuclear Profile - Power Plants

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

    net generation, 2010" "Plant nametotal reactors","Summer capacity (mw)","Net generation ...,"1,835","15,023",100.0,"Entergy Arkansas Inc" "1 Plant 2 Reactors","1,835","15,023",100.0

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

    SciTech Connect (OSTI)

    Naus, D.J.; Oland, C.B. ); Arndt, E.G. )

    1990-01-01

    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.

  6. American National Standard: design requirements for light water reactor spent fuel storage facilities at nuclear power plants

    SciTech Connect (OSTI)

    Not Available

    1983-10-07

    This standard presents necessary design requirements for facilities at nuclear power plants for the storage and preparation for shipment of spent fuel from light-water moderated and cooled nuclear power stations. It contains requirements for the design of fuel storage pool; fuel storage racks; pool makeup, instrumentation and cleanup systems; pool structure and integrity; radiation shielding; residual heat removal; ventilation, filtration and radiation monitoring systems; shipping cask handling and decontamination; building structure and integrity; and fire protection and communication.

  7. Slurry reactor design studies

    SciTech Connect (OSTI)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. ); Akgerman, A. ); Smith, J.M. )

    1990-06-01

    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.

  8. 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 ...

  9. 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 ...

  10. Connecticut 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 ...

  11. 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 ...

  12. 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 ...

  13. Pennsylvania Nuclear Profile - Power Plants

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

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

  14. 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 ...

  15. 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 ...

  16. 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 ...

  17. 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 ...

  18. 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 ...

  19. 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 ...

  20. 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 ...

  1. Illinois Nuclear Profile - Power Plants

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

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

  2. 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 ...

  3. 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 ...

  4. 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 ...

  5. Maryland Nuclear Profile - Power Plants

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

    "Plant nametotal reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Calvert Cliffs Nuclear Power Plant ...

  6. A pilot plant scale reactor/separator for ethanol from cellulosics. ERIP/DOE quarterly report no. 3 and 4

    SciTech Connect (OSTI)

    Dale, M.C.; Moelhman, M.; Butters, R.

    1998-12-01

    The 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 simultaneous saccharification/fermentation (SSF) of cellulose (glucans) followed by hemi-cellulose (pentosans) in a multi-stage continuous stirred reactor separator (CSRS). During quarters 3 and 4, we have completed a literature survey on cellulase production, activated one strain of Trichoderma reesei. We continued developing our proprietary Steep Delignification (SD) process for biomass pretreatment. Some problems with fermentations were traces to bad cellulase enzyme. Using commercial cellulase enzymes from Solvay & Genecor, SSF experiments with wheat straw showed 41 g/L ethanol and free xylose of 20 g/L after completion of the fermentation. From corn stover, we noted 36 g/L ethanol production from the cellulose fraction of the biomass, and 4 g/L free xylose at the completion of the SSF. We also began some work with paper mill sludge as a cellulose source, and in some preliminary experiments obtained 23 g/L ethanol during SSF of the sludge. During year 2, a 130 L 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.

  7. SPECKLE INTERFEROMETRY AT THE U.S. NAVAL OBSERVATORY. XVII

    SciTech Connect (OSTI)

    Mason, Brian D.; Hartkopf, William I.; Wycoff, Gary L. E-mail: wih@usno.navy.mil

    2011-08-15

    The results of 3362 intensified CCD observations of double stars, made with the 26 inch refractor of the U.S. Naval Observatory, are presented. Each observation of a system represents a combination of over 2000 short-exposure images. These observations are averaged into 1970 mean relative positions and range in separation from 0.''78 to 72.''17, with a mean separation of 14.''76. This is the 17th in this series of papers and covers the period 2010 January 6 through December 20. Also presented are 10 pairs that are resolved for the first time.

  8. D and DR Reactors - Hanford Site

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

    300 Area 324 Building 325 Building 400 AreaFast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim ...

  9. K-Reactor readiness

    SciTech Connect (OSTI)

    Rice, P.D.

    1991-12-04

    This document describes some of the more significant accomplishments in the reactor restart program and details the magnitude and extent of the work completed to bring K-Reactor to a state of restart readiness. The discussion of restart achievements is organized into the three major categories of personnel, programs, and plant. Also presented is information on the scope and extent of internal and external oversight of the efforts, as well as some details on the startup plan.

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

    SciTech Connect (OSTI)

    Not Available

    1992-12-31

    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.

  11. Naval Petroleum Reserve No. 1 (Elk Hills): Supplemental environmental impact statement. Record of decision

    SciTech Connect (OSTI)

    Not Available

    1994-02-01

    Pursuant to the Council on Environmental Quality regulations, which implement the procedural provisions of the National Environmental Policy Act, and the US Department of Energy National Environmental Policy Act regulations, the Department of Energy, Office of Fossil Energy, is issuing a Record of Decision on the continued operation of Naval Petroleum Reserve No. 1, Kern County, California. The Department of Energy has decided to continue current operations at Naval Petroleum Reserve No. 1 and implement additional well drilling, facility development projects and other activities necessary for continued production of Naval Petroleum Reserve No. 1 in accordance with the requirements of the Naval Petroleum Reserves Production Act of 1976. The final Supplemental Environmental Impact Statement, entitled ``Petroleum Production at Maximum Efficient Rate, Naval Petroleum Reserve No. 1 (Elk Hills), Kern County, California (DOE/SEIS-0158),`` was released on September 3, 1993.

  12. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1962-12-18

    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)

  13. High Temperature Gas Reactors: Assessment of Applicable Codes...

    Office of Scientific and Technical Information (OSTI)

    Technical Report: High Temperature Gas Reactors: Assessment of Applicable Codes and ... applicable to HTGR plants, the operating history of past and present HTGR plants, and with ...

  14. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Zinn, W.H.; Anderson, H.L.

    1958-09-16

    Means are presenied for increasing the reproduction ratio of a gaphite- moderated neutronic reactor by diminishing the neutron loss due to absorption or capture by gaseous impurities within the reactor. This means comprised of a fluid-tight casing or envelope completely enclosing the reactor and provided with a valve through which the casing, and thereby the reactor, may be evacuated of atmospheric air.

  15. Fossil fuel furnace reactor

    DOE Patents [OSTI]

    Parkinson, William J.

    1987-01-01

    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.

  16. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    SciTech Connect (OSTI)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  17. SPECKLE INTERFEROMETRY AT THE U.S. NAVAL OBSERVATORY. XVIII

    SciTech Connect (OSTI)

    Mason, Brian D.; Hartkopf, William I.; Friedman, Elizabeth A. E-mail: wih@usno.navy.mil

    2012-05-15

    The results of 2490 intensified CCD observations of double stars, made with the 26 inch refractor of the U.S. Naval Observatory, are presented. Each observation of a system represents a combination of over 2000 short-exposure images. These observations are averaged into 1462 mean relative positions and range in separation from 0.''56 to 71.''80, with a mean separation of 14.''81. This is the 18th in this series of papers and covers the period 2011 January 3 through 2011 December 18. Also presented are four pairs which are resolved for the first time, thirteen other pairs which appear to be lost, and linear elements for four additional pairs.

  18. SPECKLE INTERFEROMETRY AT THE U.S. NAVAL OBSERVATORY. XIX

    SciTech Connect (OSTI)

    Mason, Brian D.; Hartkopf, William I.; Hurowitz, Haley M. E-mail: wih@usno.navy.mil

    2013-09-15

    The results of 2916 intensified CCD observations of double stars, made with the 26 inch refractor of the U.S. Naval Observatory, are presented. Each observation of a system represents a combination of over two thousand short-exposure images. These observations are averaged into 1584 mean relative positions and range in separation from 0.''54 to 98.''09, with a median separation of 11.''73. This is the 19th in this series of papers and covers the period 2012 January 5 through 2012 December 18. Also presented are 10 pairs that are reported for the first time, 17 pairs that appear to be lost, linear elements for 18 pairs, and orbital elements for 2 additional pairs.

  19. Wisconsin Nuclear Profile - Power Plants

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

    Wisconsin nuclear power plants, summer capacity and net generation, 2010" "Plant name..."8,291",62.4,"NextEra Energy Point Beach LLC" "2 Plants 3 Reactors","1,584","13,281",100.0

  20. B Plant - Hanford Site

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

    Plant About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area 324 Building 325 Building 400 Area/Fast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim Storage Area Canyon Facilities Cold Test Facility D and DR Reactors Effluent Treatment Facility Environmental Restoration Disposal Facility F Reactor H Reactor

  1. BOILING REACTORS

    DOE Patents [OSTI]

    Untermyer, S.

    1962-04-10

    A boiling reactor having a reactivity which is reduced by an increase in the volume of vaporized coolant therein is described. In this system unvaporized liquid coolant is extracted from the reactor, heat is extracted therefrom, and it is returned to the reactor as sub-cooled liquid coolant. This reduces a portion of the coolant which includes vaporized coolant within the core assembly thereby enhancing the power output of the assembly and rendering the reactor substantially self-regulating. (AEC)

  2. US Department of Energy Naval petroleum reserve number 1. Financial statement audit

    SciTech Connect (OSTI)

    1997-03-01

    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 numbered 1, 2, and 3 in a manner to achieve the greatest value and benefits to the United States taxpayer. NPOSR was established by a series of Executive Orders in the early 1900s as a future source of liquid fuels for the military. NPOSR remained largely inactive until Congress, responding to the Arab oil embargo of 1973-74, passed the Naval Petroleum Reserves Production Act of 1976. The law authorized production for six years. Thereafter, NPOSR production could be reauthorized by the President in three-year increments. Since enactment of the law, every President has determined that continuing NPOSR production is in the nation`s best interest. NPOSR currently is authorized to continue production through April 5, 2000.

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

    Broader source: Energy.gov [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...

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

    Broader source: Energy.gov [DOE]

    Final Sitewide Environmental Assessment (EA) This Sitewide EA evaluates activities that DOE would conduct in anticipation of possible transfer of Naval Petroleum Reserve No. 3 (NPR-3) out of Federal operation.

  5. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Daniels, F.

    1959-10-27

    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.

  6. Microsoft Word - Written Statement FY2011 NNSA OMB approved final03012...

    National Nuclear Security Administration (NNSA)

    ... The request will enable Naval Reactors to continue reactor plant design and development efforts begun in 2010 for procurement of long-lead reactor plant components in 2017, in ...

  7. EIS-0068: Development Policy Options for the Naval Oil Shale Reserves in Colorado

    Office of Energy Efficiency and Renewable Energy (EERE)

    The U.S. Department of Energy Office of Naval Petroleum and Oil Shale Reserves prepared this programmatic statement to examine the environmental and socioeconomic impacts of development projects on the Naval Oil Shale Reserve 1, and examine select alternatives, such as encouraging production from other liquid fuel resources (coal liquefaction, biomass, offshore oil and enhanced oil recovery) or conserving petroleum in lieu of shale oil production.

  8. 2013 Federal Energy and Water Management Award Winner Naval Sea Systems

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

    Command | Department of Energy Naval Sea Systems Command 2013 Federal Energy and Water Management Award Winner Naval Sea Systems Command fewm13_nswcphiladelphia_highres.pdf (5.43 MB) fewm13_nswcphiladelphia.pdf (1.75 MB) More Documents & Publications CX-005670: Categorical Exclusion Determination U.S. Navy Marine Diesel Engines and the Environment - Part 1 EIS-0259: Record of Decision

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

    Office of Legacy Management (LM)

    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

  10. DOE - Office of Legacy Management -- Naval Petroleum Reserve No 3 - 046

    Office of Legacy Management (LM)

    Petroleum Reserve No 3 - 046 FUSRAP Considered Sites Site: Naval Petroleum Reserve No. 3 (046) More information at http://www.fossil.energy.gov/ Designated Name: Not Designated under FUSRAP Alternate Name: Naval Petroleum Reserve No 3 Landfill/Landfarm Location: Natrona County, Wyoming Evaluation Year: Not considered for FUSRAP - in another program Site Operations: Energy research Site Disposition: Site managed by DOE Office of Fossil Energy Radioactive Materials Handled: Unknown Primary

  11. 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 (OSTI)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    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

  12. BDDR, a new CEA technological and operating reactor database

    SciTech Connect (OSTI)

    Soldevilla, M.; Salmons, S.; Espinosa, B.

    2013-07-01

    The new application BDDR (Reactor database) has been developed at CEA in order to manage nuclear reactors technological and operating data. This application is a knowledge management tool which meets several internal needs: -) to facilitate scenario studies for any set of reactors, e.g. non-proliferation assessments; -) to make core physics studies easier, whatever the reactor design (PWR-Pressurized Water Reactor-, BWR-Boiling Water Reactor-, MAGNOX- Magnesium Oxide reactor-, CANDU - CANada Deuterium Uranium-, FBR - Fast Breeder Reactor -, etc.); -) to preserve the technological data of all reactors (past and present, power generating or experimental, naval propulsion,...) in a unique repository. Within the application database are enclosed location data and operating history data as well as a tree-like structure containing numerous technological data. These data address all kinds of reactors features and components. A few neutronics data are also included (neutrons fluxes). The BDDR application is based on open-source technologies and thin client/server architecture. The software architecture has been made flexible enough to allow for any change. (authors)

  13. CONVECTION REACTOR

    DOE Patents [OSTI]

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

    1960-03-22

    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.

  14. 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 ...

  15. New York 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 ...

  16. South Carolina Nuclear Profile - Power Plants

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

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

  17. 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 ...

  18. 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 ...

  19. X-10 Graphite Reactor | Department of Energy

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

    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

  20. Alternative energy conversion demonstration laboratory at U. S. Naval Academy

    SciTech Connect (OSTI)

    Wu, C.

    1983-12-01

    This paper describes an alternative energy conversion demonstration laboratory which supplements classroom theory in a senior engineering elective course in energy conversion in the Department of Mechanical Engineering at the U.S. Naval Academy. Oil, nuclear energy, and other conventional sources of power have been the dominant sources for industrial society and the U.S. Navy, and will continue to be so for the foreseeable future. There are other possibilities, however, including wind power, solar power, ocean thermal power and tidal power. A need for alternative sources of energy for the Navy was recognized at the time of the Arab oil embargo in 1973, and an academic program in alternative energy has been developed to help satisfy that need. Specific demonstrations included in this paper are as follows: Mechanical modeling of the depletion of energy reserve, Computer graphic simulation of energy consumption and energy resource exhaust, Wind model, Thermax helius rotor wind machine, Solar breeze - an electric sailboat project, Vertical axis wind turbine, Helicopter, airplane propeller and windmill models test in wind tunnel, Ocean Thermal Energy Conversion Device Demonstration, Pneumatic Wave Energy Conversion Device Demonstration, Chemical Energy Storage Device Demonstration, Solar Energy Demonstration.

  1. Renewable Energy Optimization Report for Naval Station Newport

    SciTech Connect (OSTI)

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

    2012-02-01

    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).

  2. SPECKLE INTERFEROMETRY AT THE U.S. NAVAL OBSERVATORY. XVI

    SciTech Connect (OSTI)

    Mason, Brian D.; Hartkopf, William I.; Wycoff, Gary L. E-mail: wih@usno.navy.mil

    2011-05-15

    The results of 1031 speckle-interferometric observations of double stars, made with the 26 inch refractor of the U.S. Naval Observatory, are presented. Each speckle-interferometric observation of a system represents a combination of over two thousand short-exposure images. These observations are averaged into 457 mean relative positions and range in separation from 0.''15 to 16.''94, with a median separation of 3.''03. The range in V-band magnitudes for the primary (secondary) of observed targets is 3.1-12.9 (3.2-13.3). This is the sixteenth in a series of papers presenting measurements obtained with this system and covers the period 2009 January 12 through 2009 December 17. Included in these data are 12 older measurements whose positions were previously deemed possibly aberrant, but are no longer classified this way following a confirming observation. Also, 10 pairs with a single observation are herein confirmed. This paper also includes the first data obtained using a new ICCD with fiber optic cables.

  3. CX-009246: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Naval Reactors Facility Parking Lot Expansion General Plant Project CX(s) Applied: B1.15 Date: 06/20/2012 Location(s): Pennsylvania Offices(s): Naval Nuclear Propulsion Program, NRF

  4. CX-013759: Categorical Exclusion Determination

    Broader source: Energy.gov [DOE]

    Naval Reactors Facility (NRF) Production Support Complex 3rd Floor General Plant Project CX(s) Applied: B1.15Date: 04/27/2015 Location(s): OtherOffices(s): Naval Nuclear Propulsion Program

  5. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1961-11-21

    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)

  6. REACTOR COOLING

    DOE Patents [OSTI]

    Quackenbush, C.F.

    1959-09-29

    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.

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

    SciTech Connect (OSTI)

    Dale, M. Clark; Moelhman, Mark

    1999-09-30

    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.

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

    SciTech Connect (OSTI)

    Cale, M. Clark; Moelhman, Mark

    1999-09-30

    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.

  9. Conservation plan for protected species on Naval Petroleum Reserve No. 1, Kern County, California

    SciTech Connect (OSTI)

    Otten, M.R.M.; Cypher, B.L.

    1997-07-01

    Habitats in and around Naval Petroleum Reserve No. 1 (NPR-1) support populations of various vertebrates and plants, including a number of threatened and endangered species. Adequate conservation of habitats and species, particularly protected species, can be facilitated through development and implementation of management plans. This document provides a comprehensive plan for the conservation of protected species on NPR-1, through compliance with terms and conditions expressed in Biological Opinions rendered by the U.S. Fish and Wildlife Service for NPR-1 activities. Six conservation strategies by which threatened and endangered species have been, and will be, protected are described: population monitoring, mitigation strategies, special studies, operating guidelines and policies, information transfer and outreach, and the endangered species conservation area. Population monitoring programs are essential for determining population densities and for assessing the effects of oil field developments and environmental factors on protected species. Mitigation strategies (preactivity surveys and habitat reclamation) are employed to minimize the loss of important habitats components and to restore previously disturbed lands to conditions more suitable for species` use. A number of special studies were undertaken between 1985 and 1995 to investigate the effectiveness of a variety of population and habitat management techniques with the goal of increasing the density of protected species. Operating guidelines and policies governing routine oil field activities continue to be implemented to minimize the potential for the incidental take of protected species and minimize damage to wildlife habitats. Information transfer and outreach activities are important means by which technical and nontechnical information concerning protected species conservation on NPR-1 is shared with both the scientific and non-scientific public.

  10. Endangered species and cultural resources program, Naval Petroleum Reserves in California: Annual report FY95

    SciTech Connect (OSTI)

    1996-04-01

    In FY95, EG and G Energy Measurements, Inc. (EG and G/EM) continued to support efforts to protect endangered species and cultural resources at the Naval Petroleum Reserves in California (NPRC). These efforts are conducted to ensure NPRC compliance with regulations regarding the protection of listed species and cultural resources on Federal properties. Population monitoring activities are conducted annually for San Joaquin kit foxes, giant kangaroo rats, blunt-nosed leopard lizards, and Hoover`s wooly-star. To mitigate impacts of oil field activities on listed species, 674 preactivity surveys covering approximately 211 hectares (521 acres) were conducted in FY95. EG and G/EM also assisted with mitigating effects from third-party projects, primarily by conducting biological and cultural resource consultations with regulatory agencies. EG and G/EM has conducted an applied habitat reclamation program at NPRC since 1985. In FY95, an evaluation of revegetation rates on reclaimed and non-reclaimed disturbed lands was completed, and the results will be used to direct future habitat reclamation efforts at NPRC. In FY95, reclamation success was monitored on 50 sites reclaimed in 1985. An investigation of factors influencing the distribution and abundance of kit foxes at NPRC was initiated in FY94. Factors being examined include habitat disturbance, topography, grazing, coyote abundance, lagomorph abundance, and shrub density. This investigation continued in FY95 and a manuscript on this topic will be completed in FY96. Also, Eg and G/EM completed collection of field data to evaluate the effects of a well blow-out on plant and animal populations. A final report will be prepared in FY96. Finally, EG and G/EM completed a life table analysis on San Joaquin kit foxes at NPRC.

  11. K-East and K-West Reactors - Hanford Site

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

    300 Area 324 Building 325 Building 400 AreaFast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim ...

  12. THE FOURTH US NAVAL OBSERVATORY CCD ASTROGRAPH CATALOG (UCAC4)

    SciTech Connect (OSTI)

    Zacharias, N.; Finch, C. T.; Bartlett, J. L.; Girard, T. M.; Henden, A.; Monet, D. G.; Zacharias, M. I.

    2013-02-01

    The fourth United States Naval Observatory (USNO) CCD Astrograph Catalog, UCAC4, was released in 2012 August (double-sided DVD and CDS data center Vizier catalog I/322). It is the final release in this series and contains over 113 million objects; over 105 million of them with proper motions (PMs). UCAC4 is an updated version of UCAC3 with about the same number of stars also covering all-sky. Bugs were fixed, Schmidt plate survey data were avoided, and precise five-band photometry was added for about half the stars. Astrograph observations have been supplemented for bright stars by FK6, Hipparcos, and Tycho-2 data to compile a UCAC4 star catalog complete from the brightest stars to about magnitude R = 16. Epoch 1998-2004 positions are obtained from observations with the 20 cm aperture USNO Astrograph's 'red lens', equipped with a 4k by 4k CCD. Mean positions and PMs are derived by combining these observations with over 140 ground- and space-based catalogs, including Hipparcos/Tycho and the AC2000.2, as well as unpublished measures of over 5000 plates from other astrographs. For most of the faint stars in the southern hemisphere, the first epoch plates from the Southern Proper Motion program form the basis for PMs, while the Northern Proper Motion first epoch plates serve the same purpose for the rest of the sky. These data are supplemented by 2MASS near-IR photometry for about 110 million stars and five-band (B, V, g, r, i) APASS data for over 51 million stars. Thus the published UCAC4, as were UCAC3 and UCAC2, is a compiled catalog with the UCAC observational program being a major component. The positional accuracy of stars in UCAC4 at mean epoch is about 15-100 mas per coordinate, depending on magnitude, while the formal errors in PMs range from about 1 to 10 mas yr{sup -1} depending on magnitude and observing history. Systematic errors in PMs are estimated to be about 1-4 mas yr{sup -1}.

  13. EA-1889: Draft Environmental Assessment | Department of Energy

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

    Draft Environmental Assessment EA-1889: Draft Environmental Assessment Disposal of Decommissioned, Defueled Naval Reactor Plants from USS Enterprise (CVN 65) The Department of the...

  14. nr | National Nuclear Security Administration

    National Nuclear Security Administration (NNSA)

    nr On Womens Equality Day, we celebrate NNSA's talented Women in STEM About Naval Reactors What Is the Naval Nuclear Propulsion Program? The Naval Nuclear Propulsion Program comprises the military and civilian personnel who design, build, operate, maintain, and manage the nuclear-powered ships and the many facilities that support the U.S. nuclear-powered naval fleet. The Program... Powering the Nuclear Navy Concern for the Environment Protection of People Naval Nuclear Propulsion Plants Annual

  15. NUCLEAR REACTOR

    DOE Patents [OSTI]

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

    1958-12-01

    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.

  16. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect (OSTI)

    Not Available

    1991-04-01

    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.

  17. N Reactor RELAP5 model benchmark comparisons

    SciTech Connect (OSTI)

    Fletcher, C.D.; Bolander, M.A.

    1988-02-01

    This report documents work performed at the Idaho National Engineering Laboratory (INEL) in support of Westinghouse Hanford Company safety analyses for the N Reactor. The portion of the work reported here includes comparisons of RELAP5/MOD2-calculated data with measured plant data for: (1) a plant trip reactor transient from full power operation; and (2) a hot dump test performed prior to plant startup. These qualitative comparisons are valuable because they provide an indication of the capabilities of the RELAP5/MOD2 code to simulate operational and blowdonw transients in the N Reactor. 9 refs., 12 figs., 4 tabs.

  18. 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 ...

  19. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Hurwitz, H. Jr.; Brooks, H.; Mannal, C.; Payne, J.H.; Luebke, E.A.

    1959-03-24

    A reactor of the heterogeneous, liquid cooled type is described. This reactor is comprised of a central region of a plurality of vertically disposed elongated tubes surrounded by a region of moderator material. The central region is comprised of a central core surrounded by a reflector region which is surrounded by a fast neutron absorber region, which in turn is surrounded by a slow neutron absorber region. Liquid sodium is used as the primary coolant and circulates through the core which contains the fuel elements. Control of the reactor is accomplished by varying the ability of the reflector region to reflect neutrons back into the core of the reactor. For this purpose the reflector is comprised of moderator and control elements having varying effects on reactivity, the control elements being arranged and actuated by groups to give regulation, shim, and safety control.

  20. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Wigner, E.P.

    1960-11-22

    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.

  1. REACTOR SHIELD

    DOE Patents [OSTI]

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

    1959-02-17

    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.

  2. Reactor apparatus

    DOE Patents [OSTI]

    Echtler, J. Paul

    1981-01-01

    A reactor apparatus for hydrocracking a polynuclear aromatic hydrocarbonaceous feedstock to produce lighter hydrocarbon fuels by contacting the hydrocarbonaceous feedstock with hydrogen in the presence of a molten metal halide catalyst.

  3. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  4. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.

    1960-04-01

    A nuclear reactor is described consisting of blocks of graphite arranged in layers, natural uranium bodies disposed in holes in alternate layers of graphite blocks, and coolant tubes disposed in the layers of graphite blocks which do not contain uranium.

  5. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Miller, H.I.; Smith, R.C.

    1958-01-21

    This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

  6. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Vernon, H.C.

    1959-01-13

    A neutronic reactor of the heterogeneous, fluid cooled tvpe is described. The reactor is comprised of a pressure vessel containing the moderator and a plurality of vertically disposed channels extending in spaced relationship through the moderator. Fissionable fuel material is placed within the channels in spaced relationship thereto to permit circulation of the coolant fluid. Separate means are provided for cooling the moderator and for circulating a fluid coolant thru the channel elements to cool the fuel material.

  7. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  8. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Breden, C.R.; Dietrich, J.R.

    1961-06-20

    A water-soluble non-volatile poison may be introduced into a reactor to nullify excess reactivity. The poison is removed by passing a side stream of the water containing the soluble poison to an evaporation chamber. The vapor phase is returned to the reactor to decrease the concentration of soluble poison and the liquid phase is returned to increase the concentration of soluble poison.

  9. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-07-14

    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.

  10. State space modeling of reactor core in a pressurized water reactor

    SciTech Connect (OSTI)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W.; Shamsuddin, Mustaffa; Abdullah, M. Adib

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  11. Update; Sodium advanced fast reactor (SAFR) concept

    SciTech Connect (OSTI)

    Oldenkamp, R.D.; Brunings, J.E. ); Guenther, E. ); Hren, R. )

    1988-01-01

    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.

  12. Advanced Burner Reactor with Breed-and-Burn Thorium Blankets...

    Office of Scientific and Technical Information (OSTI)

    This study assesses the feasibility of designing Seed and Blanket (S&B) Sodium-cooled Fast ... Language: English Subject: 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS Word ...

  13. Foreign Research Reactor Spent Nuclear Fuel Acceptance Program

    National Nuclear Security Administration (NNSA)

    * Complete reactor control rod system. * Note: Does not include the steam turbine generator portion of the power plant. - Sensitive nuclear technology: Any information...

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

    SciTech Connect (OSTI)

    1998-01-01

    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.

  15. Significant issues and changes for ANSI/ASME OM-1 1981, part 1, ASME OMc code-1994, and ASME OM Code-1995, Appendix I, inservice testing of pressure relief devices in light water reactor power plants

    SciTech Connect (OSTI)

    Seniuk, P.J.

    1996-12-01

    This paper identifies significant changes to the ANSI/ASME OM-1 1981, Part 1, and ASME Omc Code-1994 and ASME OM Code-1995, Appendix I, {open_quotes}Inservice Testing of Pressure Relief Devices in Light-Water Reactor Power Plants{close_quotes}. The paper describes changes to different Code editions and presents insights into the direction of the code committee and selected topics to be considered by the ASME O&M Working Group on pressure relief devices. These topics include scope issues, thermal relief valve issues, as-found and as-left set-pressure determinations, exclusions from testing, and cold setpoint bench testing. The purpose of this paper is to describe some significant issues being addressed by the O&M Working Group on Pressure Relief Devices (OM-1). The writer is currently the chair of OM-1 and the statements expressed herein represents his personal opinion.

  16. Selective purge for hydrogenation reactor recycle loop

    DOE Patents [OSTI]

    Baker, Richard W.; Lokhandwala, Kaaeid A.

    2001-01-01

    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.

  17. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1958-09-01

    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.

  18. Catalytic reactor

    DOE Patents [OSTI]

    Aaron, Timothy Mark; Shah, Minish Mahendra; Jibb, Richard John

    2009-03-10

    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.

  19. POWER REACTOR

    DOE Patents [OSTI]

    Zinn, W.H.

    1958-07-01

    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.

  20. REACTOR CONTROL

    DOE Patents [OSTI]

    Fortescue, P.; Nicoll, D.

    1962-04-24

    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)

  1. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

    A bioconversion reactor for the anaerobic fermentation of organic material. The bioconversion reactor comprises a shell enclosing a predetermined volume, an inlet port through which a liquid stream containing organic materials enters the shell, and an outlet port through which the stream exits the shell. A series of vertical and spaced-apart baffles are positioned within the shell to force the stream to flow under and over them as it passes from the inlet to the outlet port. The baffles present a barrier to the microorganisms within the shell causing them to rise and fall within the reactor but to move horizontally at a very slow rate. Treatment detention times of one day or less are possible.

  2. NEUTRONIC REACTORS

    DOE Patents [OSTI]

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

    1958-10-14

    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.

  3. (Reactor dosimetry)

    SciTech Connect (OSTI)

    West, C.D.

    1990-09-13

    The lead in most aspects of research reactor design and use passed from the USA about 15 years ago, soon after the construction of the HFIR and HFBR. The Europeans have consistently upgraded and improved their existing facilities and have built new ones including the HFR at Grenoble and ORPHEE at Saclay. They studied ultra-high flux concepts ({approximately}10{sup 20}/m{sup {minus}2}{center dot}s{sup {minus}1}) about 10 years ago, and are in the design phase of a new, highly efficient medium flux reactor to be built at Garching, near Munich in Germany. A visit was made to Interatom, the firm -- the equivalent of the Architect/Engineer for the ANS project -- responsible, under contract to the Technical University of Munich, for the new Munich reactor design. There are many similarities to the ANS design, and we reviewed and discussed technical and safety aspects of the two reactors. A request was made for some new, hitherto proprietary, experimental data on reactor thermal hydraulics and cooling that will be very valuable to the ANS project. I presented a seminar on the ANS project. A visit was made to Kernforschungszentrum Karlsruhe and knowledge was gained from Dr. Kuchle, a true pioneer of ultra-high flux reactor concepts, of their work. Dr. Kuchle kindly reviewed the ANS reference core and cooling system design (with favorable conclusions). I then talked with researchers working on materials irradiation damage and activation of structural materials by neutron irradiation, both key issues for the ANS. I was shown some new techniques they have developed for testing materials irradiation effects at high fluences, in a short time, using accelerated particle beams.

  4. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Fermi, E.; Szilard, L.

    1957-09-24

    Reactors of the type employing plates of natural uranium in a moderator are discussed wherein the plates are um-formly disposed in parallel relationship to each other thereby separating the moderator material into distinct and individual layers. Each plate has an uninterrupted sunface area substantially equal to the cross-sectional area of the active portion of the reactor, the particular size of the plates and the volume ratio of moderator to uranium required to sustain a chain reaction being determinable from the known purity of these materials and other characteristics such as the predictable neutron losses due to the formation of radioactive elements of extremely high neutron capture cross section.

  5. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1958-04-15

    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.

  6. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Young, G.

    1963-01-01

    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

  7. U Plant - Hanford Site

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

    About Us Projects & Facilities U Plant About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area 324 Building 325 Building 400 Area/Fast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim Storage Area Canyon Facilities Cold Test Facility D and DR Reactors Effluent Treatment Facility Environmental Restoration

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

    SciTech Connect (OSTI)

    Dale, M.C.

    1998-06-01

    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.

  9. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.; Babcock, Dale F.; Menegus, Robert L.

    1983-01-01

    A nuclear reactor includes an active portion with fissionable fuel and neutron moderating material surrounded by neutron reflecting material. A control element in the active portion includes a group of movable rods constructed of neutron-absorbing material. Each rod is movable with respect to the other rods to vary the absorption of neutrons and effect control over neutron flux.

  10. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Starr, C.

    1963-01-01

    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)

  11. NEUTRONIC REACTOR

    DOE Patents [OSTI]

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

    1959-10-27

    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.

  12. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J.

    1976-08-17

    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.

  13. Neutronic reactor

    DOE Patents [OSTI]

    Carleton, John T.

    1977-01-25

    A graphite-moderated nuclear reactor includes channels between blocks of graphite and also includes spacer blocks between adjacent channeled blocks with an axis of extension normal to that of the axis of elongation of the channeled blocks to minimize changes in the physical properties of the graphite as a result of prolonged neutron bombardment.

  14. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, H.L.

    1958-10-01

    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.

  15. Assessment of impacts and evaluation of restoration methods on areas affected by a well blowout, Naval Petroleum Reserve No. 1, California

    SciTech Connect (OSTI)

    Warrick, G.D.; Kato, T.T.; Phillips, M.V.

    1996-12-01

    In June 1994, an oil well on Naval Petroleum Reserve No. 1 blew-out and crude oil was deposited downwind. After the well was capped, information was collected to characterize the release and to assess effects to wildlife and plants. Oil residue was found up to 13.7 km from the well site, but deposition was relatively light and the oil quickly dried to form a thin crust on the soil surface. Elevated levels of hydrocarbons were found in livers collected from Heermann`s kangaroo rats (Dipodomys heermanni) from the oiled area but polycyclic aromatic hydrocarbons (known carcinogens or mutagens) were not detected in the livers. Restoration techniques (surface modification and bioremediation) and natural recovery were evaluated within three portions of the oiled area. Herbaceous cover and production, and survival and vigor of desert saltbush (Atriplex polycarpa) were also monitored within each trapping grid.

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

    Broader source: Energy.gov [DOE]

    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...

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

    SciTech Connect (OSTI)

    Not Available

    1990-09-01

    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.

  18. Rapid starting methanol reactor system

    DOE Patents [OSTI]

    Chludzinski, Paul J.; Dantowitz, Philip; McElroy, James F.

    1984-01-01

    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.

  19. Light Water Reactor Sustainability Technical Documents | Department of

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

    Energy Nuclear Reactor Technologies » Light Water Reactor Sustainability Program » Light Water Reactor Sustainability Technical Documents Light Water Reactor Sustainability Technical Documents April 30, 2015 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

  20. Uncertainties in the Anti-neutrino Production at Nuclear Reactors

    SciTech Connect (OSTI)

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

    2008-08-06

    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.

  1. Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor

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

    Demonstration Case Study | Department of Energy (RISMC) Advanced Test Reactor Demonstration Case Study Risk Informed Safety Margin Characterization (RISMC) Advanced Test Reactor Demonstration Case Study Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A

  2. Investigation of waste rag generation at Naval Station Mayport. Project report, May 1990-July 1993

    SciTech Connect (OSTI)

    1995-08-01

    The report presents the results of an investigation examining pollution prevention alternatives for reducing the volume of waste rags generated at Naval Station Mayport, located near Jacksonville Beach, Florida. The report recommends five specific pollution prevention alternatives: better operating practices, installation of equipment cleaning stations to remove contaminants normally removed with rags; replacement of SERVE MART rags with disposable wipers; use of recyclable rats for oil and great removal; and confirmation that used rags are fully contaminated prior to disposal.

  3. DOE - Office of Legacy Management -- Naval Oil Shale Reserves Site - 013

    Office of Legacy Management (LM)

    Oil Shale Reserves Site - 013 FUSRAP Considered Sites Site: Naval Oil Shale Reserves Site (013 ) More information at http://www.fossil.energy.gov/ Designated Name: Not Designated under FUSRAP Alternate Name: None Location: Anvil Points, Colorado Evaluation Year: Not considered for FUSRAP - in another program Site Operations: Energy research Site Disposition: Site previously managed by DOE Office of Fossil Energy; transferred to Bureau of Land Management Radioactive Materials Handled: Unknown

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

    SciTech Connect (OSTI)

    1997-03-01

    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.

  5. NUCLEAR POWER PLANT

    DOE Patents [OSTI]

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

    1963-05-14

    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)

  6. REACTOR MONITORING

    DOE Patents [OSTI]

    Bugbee, S.J.; Hanson, V.F.; Babcock, D.F.

    1959-02-01

    A neutron density inonitoring means for reactors is described. According to this invention a tunnel is provided beneath and spaced from the active portion of the reactor and extends beyond the opposite faces of the activc portion. Neutron beam holes are provided between the active portion and the tunnel and open into the tunnel near the middle thereof. A carriage operates back and forth in the tunnel and is adapted to convey a neutron detector, such as an ion chamber, and position it beneath one of the neutron beam holes. This arrangement affords convenient access of neutron density measuring instruments to a location wherein direct measurement of neutron density within the piles can be made and at the same time affords ample protection to operating personnel.

  7. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Treshow, M.

    1958-08-19

    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.

  8. REACTOR UNLOADING

    DOE Patents [OSTI]

    Leverett, M.C.

    1958-02-18

    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.

  9. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wade, E.J.

    1958-09-16

    This patent relates to a reflector means for a neutronic reactor. A reflector comprised of a plurality of vertically movable beryllium control members is provided surrounding the sides of the reactor core. An absorber of fast neutrons comprised of natural uramum surrounds the reflector. An absorber of slow neutrons surrounds the absorber of fast neutrons and is formed of a plurality of beryllium blocks having natural uranium members distributcd therethrough. in addition, a movable body is positioned directly below the core and is comprised of a beryllium reflector and an absorbing member attached to the botiom thereof, the absorbing member containing a substance selected from the goup consisting of natural urantum and Th/sup 232/.

  10. Nuclear reactor

    DOE Patents [OSTI]

    Wade, Elman E.

    1979-01-01

    A nuclear reactor including two rotatable plugs and a positive top core holddown structure. The top core holddown structure is divided into two parts: a small core cover, and a large core cover. The small core cover, and the upper internals associated therewith, are attached to the small rotating plug, and the large core cover, with its associated upper internals, is attached to the large rotating plug. By so splitting the core holddown structures, under-the-plug refueling is accomplished without the necessity of enlarging the reactor pressure vessel to provide a storage space for the core holddown structure during refueling. Additionally, the small and large rotating plugs, and their associated core covers, are arranged such that the separation of the two core covers to permit rotation is accomplished without the installation of complex lifting mechanisms.

  11. NUCLEAR REACTOR

    DOE Patents [OSTI]

    Grebe, J.J.

    1959-12-15

    A reactor which is particularly adapted tu serve as a heat source for a nuclear powered alrcraft or rocket is described. The core of this reactor consists of a porous refractory modera;or body which is impregnated with fissionable nuclei. The core is designed so that its surface forms tapered inlet and outlet ducts which are separated by the porous moderator body. In operation a gaseous working fluid is circulated through the inlet ducts to the surface of the moderator, enters and passes through the porous body, and is heated therein. The hot gas emerges into the outlet ducts and is available to provide thrust. The principle advantage is that tremendous quantities of gas can be quickly heated without suffering an excessive pressure drop.

  12. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashby, J.W.

    1958-09-16

    ABS>A graphite moderator structure is presented for a nuclear reactor compriscd of an assembly of similarly orientated prismatic graphite blocks arranged on spaced longitudinal axes lying in common planes wherein the planes of the walls of the blocks are positioned so as to be twisted reintive to the planes of said axes so thatthe unlmpeded dtrect paths in direction wholly across the walls of the blocks are limited to the width of the blocks plus spacing between the blocks.

  13. Neutronic reactor

    DOE Patents [OSTI]

    Lewis, Warren R.

    1978-05-30

    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.

  14. NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  15. REACTOR CONTROL

    DOE Patents [OSTI]

    Ruano, W.J.

    1957-12-10

    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.

  16. Toward reactor monitoring with antineutrinos

    SciTech Connect (OSTI)

    Guillon, Benoit; Cormon, S.; Fallot, M.; Giot, L.; Martino, J.; Cribier, M.; Lasserre, T.

    2007-07-01

    The fundamental knowledge on neutrino properties acquired in recent years as well as the great experimental progress made on neutrino detection open nowadays the possibility of applied neutrino physics. Among it, the International Atomic Energy Agency (IAEA) asked to its member states to study the possibility of nuclear reactor monitoring applications, such as the thermal power measurement or the fuel composition bookkeeping. In this context, we report studies aiming at a better determination of the antineutrino energy spectrum emitted by nuclear power plants, necessary for reactor monitoring applications, but also for experiments studying the ground properties of these particles. (authors)

  17. 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 (OSTI)

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

    2012-09-14

    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).

  18. B Reactor | Department of Energy

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

    Operational Management History Manhattan Project Signature Facilities B Reactor B Reactor B Reactor Completed in September 1944, the B Reactor was the world's first ...

  19. Nuclear reactor

    DOE Patents [OSTI]

    Pennell, William E.; Rowan, William J.

    1977-01-01

    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.

  20. NUCLEAR REACTORS

    DOE Patents [OSTI]

    Long, E.; Ashley, J.W.

    1958-12-16

    A graphite moderator structure is described for a gas-cooled nuclear reactor having a vertical orlentation wherein the structure is physically stable with regard to dlmensional changes due to Wigner growth properties of the graphite, and leakage of coolant gas along spaces in the structure is reduced. The structure is comprised of stacks of unlform right prismatic graphite blocks positioned in layers extending in the direction of the lengths of the blocks, the adjacent end faces of the blocks being separated by pairs of tiles. The blocks and tiles have central bores which are in alignment when assembled and are provided with cooperatlng keys and keyways for physical stability.

  1. Photocatalytic reactor

    DOE Patents [OSTI]

    Bischoff, B.L.; Fain, D.E.; Stockdale, J.A.D.

    1999-01-19

    A photocatalytic reactor is described 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. 4 figs.

  2. Business Opportunities for Small Reactors

    SciTech Connect (OSTI)

    Minato, Akio; Nishimura, Satoshi; Brown, Neil W.

    2007-07-01

    This report assesses the market potential and identifies a number of potential paths for developing the small nuclear reactor business. There are several potential opportunities identified and evaluated. Selecting a specific approach for the business development requires additional information related to a specific market and sources of capital to support the investment. If and how a market for small nuclear plants may develop is difficult to predict because of the complexity of the economic and institutional factors that will influence such development. Key factors are; economics, safety, proliferation resistance and investment risk. The economic and political interest of any of the identified markets is also dependent on successful demonstration of the safety and reliability of small nuclear reactor. Obtaining a US-NRC Standard Design approval would be an important development step toward establishing a market for small reactors. (authors)

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

    Broader source: Energy.gov [DOE]

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

  4. Woo, H.H.; Chou, C.K. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED...

    Office of Scientific and Technical Information (OSTI)

    Piping-reliability analysis for pressurized-water-reactor feedwater lines Woo, H.H.; Chou, C.K. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; PIPES; CRACKS; RELIABILITY; PWR...

  5. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T.; Sahimi, Muhammad; Fayyaz-Najafi, Babak; Harale, Aadesh; Park, Byoung-Gi; Liu, Paul K. T.

    2011-03-01

    A hybrid adsorbent-membrane reactor in which the chemical reaction, membrane separation, and product adsorption are coupled. Also disclosed are a dual-reactor apparatus and a process using the reactor or the apparatus.

  6. 850/sup 0/C VHTR plant technical description

    SciTech Connect (OSTI)

    Not Available

    1980-06-01

    This report describes the conceptual design of an 842-MW(t) process heat very high temperature reactor (VHTR) plant having a core outlet temperature of 850/sup 0/C (1562/sup 0/F). The reactor is a variation of the high-temperature gas-cooled reactor (HTGR) power plant concept. The report includes a description of the nuclear heat source (NHS) and of the balance of reactor plant (BORP) requirements. The design of the associated chemical process plant is not covered in this report. The reactor design is similar to a previously reported VHTR design having a 950/sup 0/C (1742/sup 0/F) core outlet temperature.

  7. Control Means for Reactor

    DOE Patents [OSTI]

    Manley, J. H.

    1961-06-27

    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.

  8. Power Burst Facility (PBF) Reactor Reactor Decommissioning

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

    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 (top), then places it ...

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

    SciTech Connect (OSTI)

    Not Available

    1993-12-31

    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.

  10. Environmental Information Document: L-reactor reactivation

    SciTech Connect (OSTI)

    Mackey, H.E. Jr.

    1982-04-01

    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.

  11. Self-actuating reactor shutdown system

    DOE Patents [OSTI]

    Barrus, Donald M.; Brummond, Willian A; Peterson, Leslie F.

    1988-01-01

    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.

  12. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Stewart, H.B.

    1958-12-23

    A nuclear reactor of the type speclfically designed for the irradiation of materials is discussed. In this design a central cyllndrical core of moderating material ls surrounded by an active portlon comprlsed of an annular tank contalning fissionable material immersed ln a liquid moderator. The active portion ls ln turn surrounded by a reflector, and a well ls provided in the center of the core to accommodate the materlals to be irradiated. The over-all dimensions of the core ln at least one plane are equal to or greater than twice the effective slowing down length and equal to or less than twlce the effective diffuslon length for neutrons in the core materials.

  13. NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Wigner, E.P.

    1957-09-17

    A reactor of the type having coolant liquid circulated through clad fuel elements geometrically arranged in a solid moderator, such as graphite, is described. The core is enclosed in a pressure vessel and suitable shielding, wherein means is provided for circulating vapor through the core to superheat the same. This is accomplished by drawing off the liquid which has been heated in the core due to the fission of the fuel, passing it to a nozzle within a chamber where it flashes into a vapor, and then passing the vapor through separate tubes extending through the moderator to pick up more heat developed in the core due to the fission of the fuel, thereby producing superheated vapor.

  14. DOE - Office of Legacy Management -- Naval Gun Factory and Bureau of

    Office of Legacy Management (LM)

    Ordnance - DC 0-01 Gun Factory and Bureau of Ordnance - DC 0-01 FUSRAP Considered Sites Site: NAVAL GUN FACTORY AND BUREAU OF ORDNANCE (DC.0-01) Eliminated from consideration under FUSRAP - Referred to DOD Designated Name: Not Designated Alternate Name: None Location: Washington , D.C. DC.0-01-1 Evaluation Year: 1987 DC.0-01-1 Site Operations: Designed guns and nuclear projectiles. DC.0-01-1 Site Disposition: Eliminated - No Authority DC.0-01-1 Radioactive Materials Handled: None Indicated

  15. Report (Vertical)

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

    M236, M333, M473, M0512, M0538, M0541, M552

    M473, M0512, M0538, M0541

    Naval Nuclear Laboratory - Knolls Site/Kesselring Site Naval Nuclear Laboratory - Bettis Site Los Alamos National Laboratory (LANL) Nevada National Security Site Naval Nuclear Laboratory - Naval Reactors Facility Sandia National Laboratories (SNL) Lawrence Livermore National Laboratory (LLNL) Savannah River Site (SRS) DOE/NNSA Headquarters Pantex Plant (PX) Albuquerque Complex Headquarters National Security

  16. Nuclear reactor

    DOE Patents [OSTI]

    Yant, Howard W.; Stinebiser, Karl W.; Anzur, Gregory C.

    1977-01-01

    A nuclear reactor, particularly a liquid-metal breeder reactor, whose upper internals include outlet modules for channeling the liquid-metal coolant from selected areas of the outlet of the core vertically to the outlet plenum. The modules are composed of a highly-refractory, high corrosion-resistant alloy, for example, INCONEL-718. Each module is disposed to confine and channel generally vertically the coolant emitted from a subplurality of core-component assemblies. Each module has a grid with openings, each opening disposed to receive the coolant from an assembly of the subplurality. The grid in addition serves as a holdown for the assemblies of the corresponding subplurality preventing their excessive ejection upwardly from the core. In the region directly over the core the outlet modules are of such peripheral form that they nest forming a continuum over the core-component assemblies whose outlet coolant they confine. Each subassembly includes a chimney which confines the coolant emitted by its corresponding subassemblies to generally vertical flow between the outlet of the core and the outlet plenum. Each subplurality of assemblies whose emitted coolant is confined by an outlet module includes assemblies which emit lower-temperature coolant, for example, a control-rod assembly, or fertile assemblies, and assemblies which emit coolant of substantially higher temperature, for example, fuel-rod assemblies. The coolants of different temperatures are mixed in the chimneys reducing the effect of stripping (hot-cold temperature fluctuations) on the remainder of the upper internals which are composed typically of AISI-304 or AISI-316 stainless steel.

  17. Washington Nuclear Profile - Power Plants

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

    of State nuclear net generation (percent)","Owner" "Columbia Generating Station Unit 2","1,097","9,241",100.0,"Energy Northwest" "1 Plant 1 Reactor","1,097","9,241",100.0

  18. Mississippi Nuclear Profile - Power Plants

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

    mwh)","Share of State nuclear net generation (percent)","Owner" "Grand Gulf Unit 1","1,251","9,643",100.0,"System Energy Resources, Inc" "1 Plant 1 Reactor","1,251","9,643",100.0

  19. Massachusetts Nuclear Profile - Power Plants

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

    Power Station Unit 1",685,"5,918",100.0,"Entergy Nuclear Generation Co" "1 Plant 1 Reactor",685,"5,918",100.0 "Note: Totals may not equal sum of components due to independent ...

  20. Missouri Nuclear Profile - Power Plants

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

    "Callaway Unit 1","1,190","8,996",100.0,"Union Electric Co" "1 Plant 1 Reactor","1,190","8,996",100.0 "Note: Totals may not equal sum of components due to ...

  1. Kansas Nuclear Profile - Power Plants

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

    nuclear net generation (percent)","Owner" "Wolf Creek Generating Station Unit 1","1,160","9,556",100.0,"Wolf Creek Nuclear Optg Corp" "1 Plant 1 Reactor","1,160","9,556",100.0

  2. Vermont Nuclear Profile - Power Plants

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

    mwh)","Share of State nuclear net generation (percent)","Owner" "Vermont Yankee Unit 1",620,"4,782",100.0,"Entergy Nuclear Vermont Yankee" "1 Plant 1 Reactor",620,"4,782",100.0

  3. Iowa Nuclear Profile - Power Plants

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

    of State nuclear net generation (percent)","Owner" "Duane Arnold Energy Center Unit 1",601,"4,451",100.0,"NextEra Energy Duane Arnold LLC" "1 Plant 1 Reactor",601,"4,451",100.0

  4. Prospects for Tokamak Fusion Reactors

    SciTech Connect (OSTI)

    Sheffield, J.; Galambos, J.

    1995-04-01

    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.

  5. Multimegawatt Space Reactor Safety

    SciTech Connect (OSTI)

    Stanley, M.L. )

    1989-01-01

    The Multimegawatt (MMW) Space Reactor Project supports the Strategic Defense Initiative Office requirement to provide reliable, safe, cost-effective, electrical power in the MMW range. Specifically, power may be used for neutral particle beams, free electron lasers, electromagnetic launchers, and orbital transfer vehicles. This power plant technology may also apply to the electrical power required for other uses such as deep-space probes and planetary exploration. The Multimegawatt Space Reactor Project, the Thermionic Fuel Element Verification Program, and Centaurus Program all support the Multimegawatt Space Nuclear Power Program and form an important part of the US Department of Energy's (DOE's) space and defense power systems activities. A major objective of the MMW project is the development of a reference flight system design that provides the desired levels of public safety, health protection, and special nuclear material (SNM) protection when used during its designated missions. The safety requirements for the MMW project are a hierarchy of requirements that consist of safety requirements/regulations, a safety policy, general safety criteria, safety technical specifications, safety design specifications, and the system design. This paper describes the strategy and philosophy behind the development of the safety requirements imposed upon the MMW concept developers. The safety organization, safety policy, generic safety issues, general safety criteria, and the safety technical specifications are discussed.

  6. Reactor and method of operation

    DOE Patents [OSTI]

    Wheeler, John A.

    1976-08-10

    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.

  7. Maryland Nuclear Profile - Power Plants

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

    nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Calvert Cliffs Nuclear Power Plant Unit 1, Unit 2","1,705","13,994",100.0,"Calvert Cliffs Nuclear PP Inc" "1 Plant 2 Reactors","1,705","13,994",100.0 "Note: Totals

  8. In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

    SciTech Connect (OSTI)

    Upadhyaya, Belle; Hines, J. Wesley; Damiano, Brian; Mehta, Chaitanya; Collins, Price; Lish, Matthew; Cady, Brian; Lollar, Victor; de Wet, Dane; Bayram, Duygu

    2015-12-15

    plant parameters and the pump electrical signatures. Additionally, the reactor simulation is being used to generate normal operation data and data with instrumentation faults and process anomalies. A frequency controller was interfaced with the motor power supply in order to vary the electrical supply frequency. The experimental flow control loop was used to generate operational data under varying motor performance characteristics. Coolant leakage events were simulated by varying the bypass loop flow rate. The accuracy of motor power calculation was improved by incorporating the power factor, computed from motor current and voltage in each phase of the induction motor.- A variety of experimental runs were made for steady-state and transient pump operating conditions. Process, vibration, and electrical signatures were measured using a submersible pump with variable supply frequency. High correlation was seen between motor current and pump discharge pressure signal; similar high correlation was exhibited between pump motor power and flow rate. Wide-band analysis indicated high coherence (in the frequency domain) between motor current and vibration signals. - Wide-band operational data from a PWR were acquired from AMS Corporation and used to develop time-series models, and to estimate signal spectrum and sensor time constant. All the data were from different pressure transmitters in the system, including primary and secondary loops. These signals were pre-processed using the wavelet transform for filtering both low-frequency and high-frequency bands. This technique of signal pre-processing provides minimum distortion of the data, and results in a more optimal estimation of time constants of plant sensors using time-series modeling techniques.

  9. Light Water Reactor Sustainability (LWRS) Program | Department...

    Energy Savers [EERE]

    Nuclear Reactor Technologies Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) Program Light Water Reactor Sustainability (LWRS) ...

  10. EIS-0020: Crude Oil Transport Alternate From Naval Petroleum Reserve No. 1 Elk Hills/SOHIO Pipeline Connection Conveyance System, Terminal Tank Farm Relocation to Rialto, California

    Office of Energy Efficiency and Renewable Energy (EERE)

    The Office of Naval Petroleum and Oil Shale Reserves developed this supplement to a Department of Navy statement to evaluate the environmental impacts associated with a modified design of a proposed 250,000 barrels per day crude oil conveyance system from Naval Petroleum Reserve No. 1 to connect to the proposed SOHIO West Coast to Midcontinent Pipeline at Rialto, California.

  11. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21

    The small modular reactor program at the Savannah River Site and the Savannah River National Laboratory.

  12. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J.

    1980-03-11

    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.

  13. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B.

    2004-03-16

    A nuclear reactor comprising a cylindrical pressure vessel, an elongated annular core centrally disposed within and spaced from the pressure vessel, and a plurality of ducts disposed longitudinally of the pressure vessel about the periphery thereof, said core comprising an annular active portion, an annular reflector just inside the active portion, and an annular reflector just outside the active a portion, said annular active portion comprising rectangular slab, porous fuel elements radially disposed around the inner reflector and extending the length of the active portion, wedge-shaped, porous moderator elements disposed adjacent one face of each fuel element and extending the length of the fuel element, the fuel and moderator elements being oriented so that the fuel elements face each other and the moderator elements do likewise, adjacent moderator elements being spaced to provide air inlet channels, and adjacent fuel elements being spaced to provide air outlet channels which communicate with the interior of the peripheral ducts, and means for introducing air into the air inlet channels which passes through the porous moderator elements and porous fuel elements to the outlet channel.

  14. Systems analysis of the CANDU 3 Reactor

    SciTech Connect (OSTI)

    Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H.

    1993-07-01

    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.

  15. Jim Hoffman- Biography

    Broader source: Energy.gov [DOE]

    Mr. Hoffman has over 30 years experience in the nuclear industry. He served in the U.S. Navy in submarine reactor operations and concluded his naval career as a ship repair officer managing reactor plant repair, including de-fueling and decommissioning operations at Puget Sound Naval Shipyard.

  16. Economic characteristics of a smaller, simpler reactor

    SciTech Connect (OSTI)

    LaBar, M.; Bowers, H.

    1988-01-01

    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.

  17. An Account of Oak Ridge National Laboratory's Thirteen Research Reactors

    SciTech Connect (OSTI)

    Rosenthal, Murray Wilford

    2009-08-01

    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.

  18. DOE - Office of Legacy Management -- Westinghouse Advanced Reactors...

    Office of Legacy Management (LM)

    PA.10-1 Evaluation Year: Circa 1987 PA.10-1 PA.10-4 Site Operations: 1960s and 1970s - Produced light water and fast breeder reactor fuels on a development and pilot plant scale. ...

  19. EIS-0158-S2: Supplemental Environmental Impact Statement Naval Petroleum Reserve No. 1 (Elk Hills), Kern County, California

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy developed this statement, the supplement to DOE/EIS-0158, to analyze the environmental and socioeconomic impacts of the sale of Naval Petroleum Reserve No. 1 in Kern County, California to Occidental Petroleum Corporation.

  20. Small Modular Reactors (468th Brookhaven Lecture)

    SciTech Connect (OSTI)

    Bari, Robert

    2011-04-20

    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.

  1. Environmental Survey preliminary report, Naval Petroleum Reserves in California (NPRC), Tupman, California

    SciTech Connect (OSTI)

    Not Available

    1989-02-01

    This report presents the preliminary environmental findings from the first phase of the Environmental Survey of the US Department of Energy (DOE) Naval Petroleum Reserves 1 (NPR-1) and 2 (NPR-2) in California (NPRC), conducted May 9--20, 1988. The Survey is being conducted by an interdisciplinary team of environmental specialists, led 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 NPRC. The Survey covers all environmental media and all areas of environmental regulation. It is being performed in accordance with the DOE Environmental Survey Manual. The on-site phase of the Survey involved the review of existing site environmental data, observations of the operations carried on at NPRC, and interviews with site personnel. 120 refs., 28 figs., 40 tabs.

  2. Industrial hygiene survey report of worker exposures to organotins at Norfolk Naval Shipyard, Portsmouth, Virginia

    SciTech Connect (OSTI)

    Eissler, A.W.; Ferrel, T.W.; Bloom, T.F.; Fajen, J.M.

    1985-06-24

    Breathing-zone samples were analyzed for organotin compounds, copper, and xylene during spray application of organotin containing marine antifouling paint at Norfolk Naval Shipyard, Portsmouth, Virginia, March, 1984. The survey was part of a NIOSH study of occupational exposures to organotin compounds, conducted as a component of an assessment to determine the feasibility of conducting a study of reproductive effects. Company personnel records were reviewed. Work practices were observed. The authors conclude that a potential exists for exposures to organotins and copper. As all employees were wearing respiratory protective equipment, actual exposures may be less than that indicated by the analytical data. The facility could contribute 16 potentially exposed workers to the reproductive effects study.

  3. Assessment of Fleet Inventory for Naval Air Station Whidbey Island. Task 1

    SciTech Connect (OSTI)

    Schey, Stephen; Francfort, Jim

    2015-06-01

    Task 1includes a survey of the inventory of non-tactical fleet vehicles at Naval Air Station Whidbey Island (NASWI) to characterize the fleet. This information and characterization are used to select vehicles for monitoring that takes place during Task 2. This monitoring involves data logging of vehicle operation in order to identify the vehicle’s mission and travel requirements. Individual observations of these selected vehicles provide the basis for recommendations related to PEV adoption. It also identifies whether a battery electric vehicle or plug-in hybrid electric vehicle (collectively referred to as PEVs) can fulfill the mission requirements and provide observations related to placement of PEV charging infrastructure. This report provides the results of the assessments and observations of the current non-tactical fleet, fulfilling the Task 1 requirements.

  4. Technology gap analysis on sodium-cooled reactor fuel handling system supporting advanced burner reactor development.

    SciTech Connect (OSTI)

    Chikazawa, Y.; Farmer, M.; Grandy, C.; Nuclear Engineering Division

    2009-03-01

    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

  5. Probabilistic Safety Assessment of Tehran Research Reactor

    SciTech Connect (OSTI)

    Hosseini, Seyed Mohammad Hadi; Nematollahi, Mohammad Reza; Sepanloo, Kamran

    2004-07-01

    Probabilistic Safety Assessment (PSA) application is found to be a practical tool for research reactor safety due to intense involvement of human interactions in an experimental facility. In this paper the application of the Probabilistic Safety Assessment to the Tehran Research Reactor (TRR) is presented. The level 1 PSA application involved: Familiarization with the plant, selection of accident initiators, mitigating functions and system definitions, event tree constructions and quantification, fault tree constructions and quantification, human reliability, component failure data base development and dependent failure analysis. Each of the steps of the analysis given above is discussed with highlights from the selected results. Quantification of the constructed models is done using SAPHIRE software. This Study shows that the obtained core damage frequency for Tehran Research Reactor (8.368 E-6 per year) well meets the IAEA criterion for existing nuclear power plants (1E-4). But safety improvement suggestions are offered to decrease the most probable accidents. (authors)

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

    SciTech Connect (OSTI)

    Not Available

    1994-06-01

    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.

  7. NGNP Reactor Coolant Chemistry Control Study

    SciTech Connect (OSTI)

    Brian Castle

    2010-11-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor.

  8. US graphite reactor D&D experience

    SciTech Connect (OSTI)

    Garrett, S.M.K.; Williams, N.C.

    1997-02-01

    This report describes the results of the U.S. Graphite Reactor Experience Task for the Decommissioning Strategy Plan for the Leningrad Nuclear Power Plant (NPP) Unit 1 Study. The work described in this report was performed by the Pacific Northwest National Laboratory (PNNL) for the Department of Energy (DOE).

  9. Iowa Nuclear Profile - Power Plants

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

    Iowa nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Duane Arnold Energy Center Unit 1",601,"4,451",100.0,"NextEra Energy Duane Arnold LLC" "1 Plant 1 Reactor",601,"4,451",100.0

  10. Kansas Nuclear Profile - Power Plants

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

    Kansas nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Wolf Creek Generating Station Unit 1","1,160","9,556",100.0,"Wolf Creek Nuclear Optg Corp" "1 Plant 1 Reactor","1,160","9,556",100.0

  11. Massachusetts Nuclear Profile - Power Plants

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

    nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Pilgrim Nuclear Power Station Unit 1",685,"5,918",100.0,"Entergy Nuclear Generation Co" "1 Plant 1 Reactor",685,"5,918",100.0 "Note: Totals may not equal sum of components due to

  12. Mississippi Nuclear Profile - Power Plants

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

    Mississippi nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Grand Gulf Unit 1","1,251","9,643",100.0,"System Energy Resources, Inc" "1 Plant 1 Reactor","1,251","9,643",100.0

  13. Missouri Nuclear Profile - Power Plants

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

    nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Callaway Unit 1","1,190","8,996",100.0,"Union Electric Co" "1 Plant 1 Reactor","1,190","8,996",100.0 "Note: Totals may not equal sum of components due to

  14. Arizona Nuclear Profile - Power Plants

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

    nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Palo Verde Unit 1, Unit 2, Unit 3","3,937","31,200",100.0,"Arizona Public Service Co" "1 Plant 3 Reactors","3,937","31,200",100.0 "Note: Totals may not equal sum of

  15. Arkansas Nuclear Profile - Power Plants

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

    nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Arkansas Nuclear One Unit 1, Unit 2","1,835","15,023",100.0,"Entergy Arkansas Inc" "1 Plant 2 Reactors","1,835","15,023",100.0

  16. Connecticut Nuclear Profile - Power Plants

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

    Connecticut nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Millstone Unit 2, Unit 3","2,103","16,750",100.0,"Dominion Nuclear Conn Inc" "1 Plant 2 Reactors","2,103","16,750",100.0

  17. Attrition reactor system

    DOE Patents [OSTI]

    Scott, C.D.; Davison, B.H.

    1993-09-28

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur. 2 figures.

  18. Attrition reactor system

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Davison, Brian H. (Knoxvile, TN)

    1993-01-01

    A reactor vessel for reacting a solid particulate with a liquid reactant has a centrifugal pump in circulatory flow communication with the reactor vessel for providing particulate attrition, resulting in additional fresh surface where the reaction can occur.

  19. Hybrid plasmachemical reactor

    SciTech Connect (OSTI)

    Lelevkin, V. M. Smirnova, Yu. G.; Tokarev, A. V.

    2015-04-15

    A hybrid plasmachemical reactor on the basis of a dielectric barrier discharge in a transformer is developed. The characteristics of the reactor as functions of the dielectric barrier discharge parameters are determined.

  20. Reactor System Transient Code.

    Energy Science and Technology Software Center (OSTI)

    1999-07-14

    RELAP3B describes the behavior of water-cooled nuclear reactors during postulated accidents or power transients, such as large reactivity excursions, coolant losses or pump failures. The program calculates flows, mass and energy inventories, pressures, temperatures, and steam qualities along with variables associated with reactor power, reactor heat transfer, or control systems. Its versatility allows one to describe simple hydraulic systems as well as complex reactor systems.

  1. Period meter for reactors

    DOE Patents [OSTI]

    Rusch, Gordon K.

    1976-01-06

    An improved log N amplifier type nuclear reactor period meter with reduced probability for noise-induced scrams is provided. With the reactor at low power levels a sampling circuit is provided to determine the reactor period by measuring the finite change in the amplitude of the log N amplifier output signal for a predetermined time period, while at high power levels, differentiation of the log N amplifier output signal provides an additional measure of the reactor period.

  2. NEUTRONIC REACTOR SHIELDING

    DOE Patents [OSTI]

    Borst, L.B.

    1961-07-11

    A special hydrogenous concrete shielding for reactors is described. In addition to Portland cement and water, the concrete essentially comprises 30 to 60% by weight barytes aggregate for enhanced attenuation of fast neutrons. The biological shields of AEC's Oak Ridge Graphite Reactor and Materials Testing Reactor are particular embodiments.

  3. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  4. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-01-01

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  5. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E.; Grohmann, Karel; Himmel, Michael E.; Richard, Christopher J.

    1993-03-02

    A fermentation reactor and method for fermentation of materials having greater than about 10% solids. The reactor includes a rotatable shaft along the central axis, the shaft including rods extending outwardly to mix the materials. The reactor and method are useful for anaerobic digestion of municipal solid wastes to produce methane, for production of commodity chemicals from organic materials, and for microbial fermentation processes.

  6. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P.; Scahill, John W.

    1995-01-01

    An improved vortex reactor system for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor.

  7. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01

    The Advanced Test Reactor at Idaho National Laboratory is the foremost nuclear materials test reactor in the world. This virtual tour describes the reactor, how experiments are conducted, and how spent nuclear fuel is handled and stored. For more information about INL research, visit http://www.facebook.com/idahonationallaboratory.

  8. NUCLEAR REACTOR CONTROL SYSTEM

    DOE Patents [OSTI]

    Epler, E.P.; Hanauer, S.H.; Oakes, L.C.

    1959-11-01

    A control system is described for a nuclear reactor using enriched uranium fuel of the type of the swimming pool and other heterogeneous nuclear reactors. Circuits are included for automatically removing and inserting the control rods during the course of normal operation. Appropriate safety circuits close down the nuclear reactor in the event of emergency.

  9. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P.; Holley, John C.

    1982-01-01

    A reactor vessel support system includes a support ring at the reactor top supported through a box ring on a ledge of the reactor containment. The box ring includes an annular space in the center of its cross-section to reduce heat flow and is keyed to the support ledge to transmit seismic forces from the reactor vessel to the containment structure. A coolant channel is provided at the outside circumference of the support ring to supply coolant gas through the keyways to channels between the reactor vessel and support ledge into the containment space.

  10. Nuclear reactor overflow line

    DOE Patents [OSTI]

    Severson, Wayne J.

    1976-01-01

    The overflow line for the reactor vessel of a liquid-metal-cooled nuclear reactor includes means for establishing and maintaining a continuous bleed flow of coolant amounting to 5 to 10% of the total coolant flow through the overflow line to prevent thermal shock to the overflow line when the reactor is restarted following a trip. Preferably a tube is disposed concentrically just inside the overflow line extending from a point just inside the reactor vessel to an overflow tank and a suction line is provided opening into the body of liquid metal in the reactor vessel and into the annulus between the overflow line and the inner tube.

  11. Alternatives to proposed replacement production reactors

    SciTech Connect (OSTI)

    Cullingford, H.S.

    1981-06-01

    To insure adequate supplies of plutonium and tritium for defense purposes, an independent evaluation was made by Los Alamos National Laboratory of the numerous alternatives to the proposed replacement production reactors (RPR). This effort concentrated on the defense fuel cycle operation and its technical implications in identifying the principal alternatives for the 1990s. The primary options were identified as (1) existing commercial reactors, (2) existing and planned government-owned facilities (not now used for defense materials production), and (3) other RPRs (not yet proposed) such as CANDU or CANDU-type heavy-water reactors (HWR) for both plutonium and tritium production. The evaluation considered features and differences of various options that could influence choice of RPR alternatives. Barring a change in the US approach to civilian and defense fuel cycles and precluding existing commercial reactors at government-owned sites, the most significant alternatives were identified as a CANDU-type HWR at Savannah River Plant (SRP) site or the Three Mile Island commercial reactor with reprocessing capability at Barnwell Nuclear Fuel Plant and at SRP.

  12. Spinning fluids reactor

    DOE Patents [OSTI]

    Miller, Jan D; Hupka, Jan; Aranowski, Robert

    2012-11-20

    A spinning fluids reactor, includes a reactor body (24) having a circular cross-section and a fluid contactor screen (26) within the reactor body (24). The fluid contactor screen (26) having a plurality of apertures and a circular cross-section concentric with the reactor body (24) for a length thus forming an inner volume (28) bound by the fluid contactor screen (26) and an outer volume (30) bound by the reactor body (24) and the fluid contactor screen (26). A primary inlet (20) can be operatively connected to the reactor body (24) and can be configured to produce flow-through first spinning flow of a first fluid within the inner volume (28). A secondary inlet (22) can similarly be operatively connected to the reactor body (24) and can be configured to produce a second flow of a second fluid within the outer volume (30) which is optionally spinning.

  13. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling.

  14. Reactor water cleanup system

    DOE Patents [OSTI]

    Gluntz, D.M.; Taft, W.E.

    1994-12-20

    A reactor water cleanup system includes a reactor pressure vessel containing a reactor core submerged in reactor water. First and second parallel cleanup trains are provided for extracting portions of the reactor water from the pressure vessel, cleaning the extracted water, and returning the cleaned water to the pressure vessel. Each of the cleanup trains includes a heat exchanger for cooling the reactor water, and a cleaner for cleaning the cooled reactor water. A return line is disposed between the cleaner and the pressure vessel for channeling the cleaned water thereto in a first mode of operation. A portion of the cooled water is bypassed around the cleaner during a second mode of operation and returned through the pressure vessel for shutdown cooling. 1 figure.

  15. Fission reactors and materials

    SciTech Connect (OSTI)

    Frost, B.R.T.

    1981-12-01

    The American-designed boiling water reactor and pressurized water reactor dominate the designs currently in use and under construction worldwide. As in all energy systems, materials problems have appeared during service; these include stress-corrosion of stainless steel pipes and heat exchangers and questions regarding crack behavior in pressure vessels. To obtain the maximum potential energy from our limited uranium supplies is is essential to develop the fast breeder reactor. The materials in these reactors are subjected to higher temperatures and neutron fluxes but lower pressures than in the water reactors. The performance required of the fuel elements is more arduous in the breeder than in water reactors. Extensive materials programs are in progress in test reactors and in large test rigs to ensure that materials will be available to meet these conditions.

  16. Proceedings of the 1992 topical meeting on advances in reactor physics. Volume 2

    SciTech Connect (OSTI)

    Not Available

    1992-04-01

    This document, Volume 2, presents proceedings of the 1992 Topical Meeting on Advances in Reactor Physics on March 8--11, 1992 at Charleston, SC. Session topics were as follows: Transport Theory; Fast Reactors; Plant Analyzers; Integral Experiments/Measurements & Analysis; Core Computational Systems; Reactor Physics; Monte Carlo; Safety Aspects of Heavy Water Reactors; and Space-Time Core Kinetics. The individual reports have been cataloged separately. (FI)

  17. Passive cooling system for liquid metal cooled nuclear reactors with backup coolant flow path

    DOE Patents [OSTI]

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

    A liquid metal cooled nuclear fission reactor plant having a passive auxiliary safety cooling system for removing residual heat resulting from fuel decay during reactor shutdown, or heat produced during a mishap. This reactor plant is enhanced by a backup or secondary passive safety cooling system which augments the primary passive auxiliary cooling system when in operation, and replaces the primary system when rendered inoperable.

  18. Principles of providing inherent self-protection and passive safety characteristics of the SVBR-75/100 type modular reactor installation for nuclear power plants of different capacity and purpose

    SciTech Connect (OSTI)

    Toshinsky, G.I.; Komlev, O.G.; Novikova, N.N.; Tormyshev, I.V.; Stepanov, V.S.; Klimov, N.N.; Dedoul, A.V.

    2007-07-01

    The report presents a brief description of the reactor installation SVBR-75/100, states a concept of providing the RI safety and presents the basic results of the analysis of the most dangerous pre-accidental situations and beyond the design basis accidents, which have been obtained in the process of validating the RI safety. It has been shown that the safety functions concerning the accidental shutdown of the reactor, total blacking out of the NPP and localization of the accidental situation relating to the postulated simultaneous rupture of several steam-generator tubes are not subject to influence of the human factor and are entirely realized in a passive way. (authors)

  19. EA-1956: Site-Wide Environmental Assessment for the Divestiture of Rocky Mountain Oilfield Testing Center and Naval Petroleum Reserve No. 3, Natrona County, Wyoming

    Broader source: Energy.gov [DOE]

    DOE prepared an EA that assesses the potential environmental impacts of the proposed discontinuation of DOE operations at the Rocky Mountain Oilfield Testing Center (RMOTC) and the proposed divestiture of Naval Petroleum Reserve Number 3 (NPR-3)

  20. Using Net-Zero Energy Projects to Enable Sustainable Economic Redevelopment at the Former Brunswick Air Naval Base

    SciTech Connect (OSTI)

    Huffman, S.

    2011-10-01

    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.

  1. Nuclear reactor characteristics and operational history

    Gasoline and Diesel Fuel Update (EIA)

    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

  2. Report to the President on agreements and programs relating to the Naval Petroleum and Oil Shale Reserves

    SciTech Connect (OSTI)

    Not Available

    1994-08-01

    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.

  3. Environmental Survey preliminary report, Naval Petroleum and Oil Shale Reserves in Colorado, Utah, and Wyoming, Casper, Wyoming

    SciTech Connect (OSTI)

    Not Available

    1989-02-01

    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.

  4. Final sitewide environmental assessment for continued development of Naval Petroleum Reserve No. 3 (NPR-3), Natrona County, Wyoming

    SciTech Connect (OSTI)

    1995-07-01

    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.

  5. Naval Petroleum Reserve Number 1 financial statements September 30, 1997 and 1996 (with independent auditors` report thereon)

    SciTech Connect (OSTI)

    1997-12-31

    The Naval Petroleum and Oil Shale Reserves (NPOSR) produces crude oil and associated hydrocarbons from the Naval Petroleum Reserve No. 1 (NPR-1) in a manner to achieve the greatest value and benefits to the US taxpayer. As required by the 1996 National Defense Authorization Act, the Department of Energy offered NPR-1 for sale during FY 1997. DOE structured the sale so as to offer two types of ownership segments: one operatorship segment, consisting of 74% of the US interest in NPR-1, and 13 nonoperating segments, each consisting of 2% of the US interest. Potential purchasers could bid on one, some, or all of the segments. If a single purchaser wanted to buy all of the Government`s interest, then its bid would have to exceed the total of the highest bids for all of the individual segments. Bids were due October 1, 1997, at which time DOE received 22 bids from 15 parties acting alone or in concert. The report and management letter present the results of the independent certified public accountants` audits of the Department of Energy`s Naval Petroleum Reserve Number 1 (NPR-1) financial statements as of, and for the years ended, September 30, 1997 and 1996.

  6. United States-Russian laboratory-to-laboratory cooperation on protection, control, and accounting for naval nuclear materials

    SciTech Connect (OSTI)

    Sukhoruchkin, V.; Yurasov, N.; Goncharenko, Y.; Mullen, M.; McConnell, D.

    1996-12-31

    In March 1995, the Russian Navy contacted safeguards experts at the Kurchatov Institute (KI) and proposed the initiation of work to enhance nuclear materials protection, control, and accounting (MPC and A) at Russian Navy facilities. Because of KI`s successful experience in laboratory-to-laboratory MPC and A cooperation with US Department of Energy Laboratories, the possibility of US participation in the work with the Russian Navy was explored. Several months later, approval was received from the US Government and the Russian Navy to proceed with this work on a laboratory-to-laboratory basis through Kurchatov Institute. As a first step in the cooperation, a planning meeting occurred at KI in September, 1995. Representatives from the US Department of Energy (DOE), the US Department of Defense (DOD), the Russian Navy, and KI discussed several areas for near-term cooperative work, including a vulnerability assessment workshop and a planning study to identify and prioritize near-term MPC and A enhancements that might be implemented at Russian facilities which store or handle unirradiated highly enriched uranium fuel for naval propulsion applications. In subsequent meetings, these early proposals have been further refined and extended. This MPC and A cooperation will now include enhanced protection and control features for storage facilities and refueling service ships, computerized accounting systems for naval fuel, methods and equipment for rapid inventories, improved security of fresh fuel during truck transportation, and training. This paper describes the current status and future plans for MPC and A cooperation for naval nuclear materials.

  7. Endangered species and cultural resources program Naval petroleum Reserves in California. Annual report FY96

    SciTech Connect (OSTI)

    1997-07-01

    In FY96, Enterprise Advisory Services, Inc. (EASI) continued to support efforts to protect endangered species and cultural resources at the Naval Petroleum Reserves in California (NPRC). These efforts are conducted to ensure NPRC compliance with regulations regarding the protection of listed species and cultural resources on federal properties. Population monitoring activities were conducted for San Joaquin kit foxes, giant kangaroo rats, blunt-nosed leopard lizards, and Hoover`s wooly-star. Kit fox abundance and distribution was assessed by live-trapping over a 329-km{sup 2} area. Kit fox reproduction and mortality were assessed by radiocollaring and monitoring 22 adults and two pups. Reproductive success and litter size were determined through live-trapping and den observations. Rates and sources of kit fox mortality were assessed by recovering dead radiocollared kit foxes and conducting necropsies to determine cause of death. Abundance of coyotes and bobcats, which compete with kit foxes, was determined by conducting scent station surveys. Kit fox diet was assessed through analysis of fecal samples collected from live-trapped foxes. Abundance of potential prey for kit foxes was determined by conducting transect surveys for lagornorphs and live-trapping small mammals.

  8. Naval Petroleum Reserves in California site environmental report for calendar year 1989

    SciTech Connect (OSTI)

    Not Available

    1989-01-01

    This summary for Naval Petroleum Reserves in California (NPRC) is divided into NPR-1 and NPR-2. Monitoring efforts at NPR-1 include handling and disposal of oilfield wastes; environmental preactivity surveys for the protection of endangered species and archaeological resources; inspections of topsoil stockpiling; monitoring of revegetated sites; surveillance of production facilities for hydrocarbons and oxides of nitrogen (NO{sub x}) emissions; monitoring of oil spill prevention and cleanup; and monitoring of wastewater injection. No major compliance issues existed for NPR-1 during 1989. Oil spills are recorded, reviewed for corrective action, and reported. Environmental preactivity surveys for proposed projects which may disturb or contaminate the land are conducted to prevent damage to the federally protected San Joaquin kit fox, blunt-nosed leopard lizard, Tipton kangaroo rat and the giant kangaroo rat. Projects are adjusted or relocated as necessary to avoid impact to dens, burrows, or flat-bottomed drainages. A major revegetation program was accomplished in 1989 for erosion control enhancement of endangered species habitat. The main compliance issue on NPR-2 was oil and produced water discharges into drainages by lessees. An additional compliance issue on NPR-2 is surface refuse from past oilfield operations. 17 refs.

  9. Evaluation of EHD enhancement and thermoacoustic refrigeration for naval applications. Technical report, Jul-Sep 91

    SciTech Connect (OSTI)

    Memory, S.B.

    1991-12-01

    An evaluation has been made of two different techniques which could prove valuable for Naval refrigeration needs in the future. The first is electrohydrodynamic (EHD) enhancement of pool boiling and condensation heat transfer; this has been shown to provide significant enhancements for both modes of heat transfer under certain conditions and could provide increases in efficiency of present vapor-compression systems. EHD techniques are quite advanced and prototype condenser and evaporator bundles are currently being tested. The second technique is an alternative refrigeration technology called thermoacoustic refrigeration; alternative technologies have become increasingly attractive over recent years due to environmental concerns over CFCs. Thermoacoustic refrigeration uses acoustic power to pump heat from a low temperature source to a high temperature sink. It is still in the early stages of development and can presently accommodate only small thermal loads. However, its general principles of operation have been proven and its resent capacity and efficiency limitations are not seen as a problem in the long term. Electrohydrodynamic Enhancement, Boiling and Condensation, Thermoacoustic Refrigeration.

  10. Electric Vehicle Preparedness - Implementation Approach for Electric Vehicles at Naval Air Station Whidbey Island. Task 4

    SciTech Connect (OSTI)

    Schey, Stephen; Francfort, Jim

    2015-06-01

    Several U.S. Department of Defense base studies have been conducted to identify potential U.S. Department of Defense transportation systems that are strong candidates for introduction or expansion of plug-in electric vehicles (PEVs). This study is focused on the Naval Air Station Whidbey Island (NASWI) located in Washington State. Task 1 consisted of a survey of the non-tactical fleet of vehicles at NASWI to begin the review of vehicle mission assignments and types of vehicles in service. In Task 2, daily operational characteristics of vehicles were identified to select vehicles for further monitoring and attachment of data loggers. Task 3 recorded vehicle movements in order to characterize the vehicles’ missions. The results of the data analysis and observations were provided. Individual observations of the selected vehicles provided the basis for recommendations related to PEV adoption, i.e., whether a battery electric vehicle (BEV) or plug-in hybrid electric vehicle (PHEV) (collectively PEVs) can fulfill the mission requirements. It also provided the basis for recommendations related to placement of PEV charging infrastructure. This report focuses on an implementation plan for the near-term adoption of PEVs into the NASWI fleet.

  11. Limiting factors to advancing thermal-battery technology for naval applications

    SciTech Connect (OSTI)

    Davis, P.B.; Winchester, C.S.

    1991-10-01

    Thermal batteries are primary reserve electrochemical power sources using molten salt electrolyte which experience little effective aging while in storage or dormant deployment. Thermal batteries are primarily used in military applications, and are currently used in a wide variety of Navy devices such as missiles, torpedoes, decays, and training targets, usually as power supplies in guidance, propulsion, and Safe/Arm applications. Technology developments have increased the available energy and power density ratings by an order of magnitude in the last ten years. Present thermal batteries, using lithium anodes and metal sulfide cathodes, are capable of performing applications where only less rugged and more expensive silver oxide/zinc or silver/magnesium chloride seawater batteries could serve previously. Additionally, these batteries are capable of supplanting lithium/thionyl chloride reserve batteries in a variety of specifically optimized designs. Increases in thermal battery energy and power density capabilities are not projected to continue with the current available technology. Several battery designs are now at the edge of feasibility and safety. Since future naval systems are likely to require continued growth of battery energy and Power densities, there must be significant advances in battery technology. Specifically, anode alloy composition and new cathode materials must be investigated to allow for safe development and deployment of these high power, higher energy density batteries.

  12. New Hampshire Nuclear Profile - Power Plants

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

    "Seabrook Unit 1","1,247","10,910",100.0,"NextEra Energy Seabrook LLC" "1 Plant 1 Reactor","1,247","10,910",100.0 "Note: Totals may not equal sum of components due to ...

  13. ATOMIC POWER PLANT

    DOE Patents [OSTI]

    Daniels, F.

    1957-11-01

    This patent relates to neutronic reactor power plants and discloses a design of a reactor utilizing a mixture of discrete units of a fissionable material, such as uranium carbide, a neutron moderator material, such as graphite, to carry out the chain reaction. A liquid metal, such as bismuth, is used as the coolant and is placed in the reactor chamber with the fissionable and moderator material so that it is boiled by the heat of the reaction, the boiling liquid and vapors passing up through the interstices between the discrete units. The vapor and flue gases coming off the top of the chamber are passed through heat exchangers, to produce steam, for example, and thence through condensers, the condensed coolant being returned to the chamber by gravity and the non- condensible gases being carried off through a stack at the top of the structure.

  14. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Goett, J.J.

    1961-01-24

    A system is described which includes a neutronic reactor containing a dispersion of fissionable material in a liquid moderator as fuel and a conveyor to which a portion of the dispersion may be passed and wherein the self heat of the slurry evaporates the moderator. Means are provided for condensing the liquid moderator and returning it to the reactor and for conveying the dried fissionable material away from the reactor.

  15. THERMAL NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Spinrad, B.I.

    1960-01-12

    A novel thermal reactor was designed in which a first reflector formed from a high atomic weight, nonmoderating material is disposed immediately adjacent to the reactor core. A second reflector composed of a moderating material is disposed outwardly of the first reflector. The advantage of this novel reflector arrangement is that the first reflector provides a high slow neutron flux in the second reflector, where irradiation experiments may be conducted with a small effect on reactor reactivity.

  16. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, J.P.; Scahill, J.W.

    1995-05-09

    An improved vortex reactor system is described for affecting fast pyrolysis of biomass and Refuse Derived Fuel (RDF) feed materials comprising: a vortex reactor having its axis vertically disposed in relation to a jet of a horizontally disposed steam ejector that impels feed materials from a feeder and solids from a recycle loop along with a motive gas into a top part of said reactor. 12 figs.

  17. Small Modular Reactors (SMRs) | Department of Energy

    Energy Savers [EERE]

    Reactor Technologies Small Modular Reactors (SMRs) Small Modular Reactors (SMRs) ... to the NRC by late-2016 Complete reactor module final design by mid-2019 For more ...

  18. Nuclear Safeguards Considerations For The Pebble Bed Modular Reactor (PBMR)

    SciTech Connect (OSTI)

    Phillip Casey Durst; David Beddingfield; Brian Boyer; Robert Bean; Michael Collins; Michael Ehinger; David Hanks; David L. Moses; Lee Refalo

    2009-10-01

    High temperature reactors (HTRs) have been considered since the 1940s, and have been constructed and demonstrated in the United Kingdom (Dragon), United States (Peach Bottom and Fort Saint Vrain), Japan (HTTR), Germany (AVR and THTR-300), and have been the subject of conceptual studies in Russia (VGM). The attraction to these reactors is that they can use a variety of reactor fuels, including abundant thorium, which upon reprocessing of the spent fuel can produce fissile U-233. Hence, they could extend the stocks of available uranium, provided the fuel is reprocessed. Another attractive attribute is that HTRs typically operate at a much higher temperature than conventional light water reactors (LWRs), because of the use of pyrolytic carbon and silicon carbide coated (TRISO) fuel particles embedded in ceramic graphite. Rather than simply discharge most of the unused heat from the working fluid in the power plant to the environment, engineers have been designing reactors for 40 years to recover this heat and make it available for district heating or chemical conversion plants. Demonstrating high-temperature nuclear energy conversion was the purpose behind Fort Saint Vrain in the United States, THTR-300 in Germany, HTTR in Japan, and HTR-10 and HTR-PM, being built in China. This resulted in nuclear reactors at least 30% or more thermodynamically efficient than conventional LWRs, especially if the waste heat can be effectively utilized in chemical processing plants. A modern variant of high temperature reactors is the Pebble Bed Modular Reactor (PBMR). Originally developed in the United States and Germany, it is now being redesigned and marketed by the Republic of South Africa and China. The team examined historical high temperature and high temperature gas reactors (HTR and HTGR) and reviewed safeguards considerations for this reactor. The following is a preliminary report on this topic prepared under the ASA-100 Advanced Safeguards Project in support of the NNSA Next

  19. Tokamak reactor first wall

    DOE Patents [OSTI]

    Creedon, R.L.; Levine, H.E.; Wong, C.; Battaglia, J.

    1984-11-20

    This invention relates to an improved first wall construction for a tokamak fusion reactor vessel, or other vessels subjected to similar pressure and thermal stresses.

  20. Small Modular Reactors - SRSCRO

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

    smr Small Modular Reactors The Savannah River National Laboratory (SRNL) has announced several partnerships to bring refrigerator-sized modular nuclear reactors, known as Small Modular Reactors or SMRs, to the Savannah River Site facility and jump start development of the U.S. Energy Freedom CenterTM. Currently, all large commercial power reactors in the United States and most in the rest of the world are based on "light water" designs - that is, they use uranium fuel and ordinary

  1. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOE Patents [OSTI]

    King, L.D.P.

    1959-09-01

    A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

  2. Owners of nuclear power plants

    SciTech Connect (OSTI)

    Hudson, C.R.; White, V.S.

    1996-11-01

    Commercial nuclear power plants in this country can be owned by a number of separate entities, each with varying ownership proportions. Each of these owners may, in turn, have a parent/subsidiary relationship to other companies. In addition, the operator of the plant may be a different entity as well. This report provides a compilation on the owners/operators for all commercial power reactors in the United States. While the utility industry is currently experiencing changes in organizational structure which may affect nuclear plant ownership, the data in this report is current as of July 1996. The report is divided into sections representing different aspects of nuclear plant ownership.

  3. Owners of Nuclear Power Plants

    SciTech Connect (OSTI)

    Reid, R.L.

    2000-01-12

    Commercial nuclear power plants in this country can be owned by a number of separate entities, each with varying ownership proportions. Each of these owners may, in turn, have a parent/subsidiary relationship to other companies. In addition, the operator of the plant may be a different entity as well. This report provides a compilation on the owners/operators for all commercial power reactors in the United States. While the utility industry is currently experiencing changes in organizational structure which may affect nuclear plant ownership, the data in this report is current as of November 1999. The report is divided into sections representing different aspects of nuclear plant ownership.

  4. Brookhaven Graphite Research Reactor Workshop

    Broader source: Energy.gov [DOE]

    The Brookhaven Graphite Research Reactor (BGRR) was the first reactor built in the U.S. for peacetime atomic research following World War II.  Construction began in 1947 and the reactor started...

  5. Human Factors Aspects of Operating Small Reactors

    SciTech Connect (OSTI)

    OHara, J.M.; Higgins, J.; Deem, R.; Xing, J.; DAgostino, A.

    2010-11-07

    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. They are considering small modular reactors (SMRs) as one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can be designed to generate hydrogen, or to process heat. Many characteristics of SMRs are quite different from those of current plants, and so may require a concept of operations (ConOps) that also is different. The U.S. Nuclear Regulatory Commission (NRC) has begun examining the human factors engineering- (HFE) and ConOps- aspects of SMRs; if needed, they will formulate guidance to support SMR licensing reviews. We developed a ConOps model, consisting of the following dimensions: Plant mission; roles and responsibilities of all agents; staffing, qualifications, and training; management of normal operations; management of off-normal conditions and emergencies; and, management of maintenance and modifications. We are reviewing information on SMR design to obtain data about each of these dimensions, and have identified several preliminary issues. In addition, we are obtaining operations-related information from other types of multi-module systems, such as refineries, to identify lessons learned from their experience. Here, we describe the project's methodology and our preliminary findings.

  6. NEUTRONIC REACTOR BURIAL ASSEMBLY

    DOE Patents [OSTI]

    Treshow, M.

    1961-05-01

    A burial assembly is shown whereby an entire reactor core may be encased with lead shielding, withdrawn from the reactor site and buried. This is made possible by a five-piece interlocking arrangement that may be easily put together by remote control with no aligning of bolt holes or other such close adjustments being necessary.

  7. NEUTRONIC REACTOR SYSTEM

    DOE Patents [OSTI]

    Treshow, M.

    1959-02-10

    A reactor system incorporating a reactor of the heterogeneous boiling water type is described. The reactor is comprised essentially of a core submerged adwater in the lower half of a pressure vessel and two distribution rings connected to a source of water are disposed within the pressure vessel above the reactor core, the lower distribution ring being submerged adjacent to the uppcr end of the reactor core and the other distribution ring being located adjacent to the top of the pressure vessel. A feed-water control valve, responsive to the steam demand of the load, is provided in the feedwater line to the distribution rings and regulates the amount of feed water flowing to each distribution ring, the proportion of water flowing to the submerged distribution ring being proportional to the steam demand of the load. This invention provides an automatic means exterior to the reactor to control the reactivity of the reactor over relatively long periods of time without relying upon movement of control rods or of other moving parts within the reactor structure.

  8. REFLECTOR FOR NEUTRONIC REACTORS

    DOE Patents [OSTI]

    Fraas, A.P.

    1963-08-01

    A reflector for nuclear reactors that comprises an assembly of closely packed graphite rods disposed with their major axes substantially perpendicular to the interface between the reactor core and the reflector is described. Each graphite rod is round in transverse cross section at (at least) its interface end and is provided, at that end, with a coaxial, inwardly tapering hole. (AEC)

  9. Thermionic switched self-actuating reactor shutdown system

    DOE Patents [OSTI]

    Barrus, Donald M. (San Jose, CA); Shires, Charles D. (San Jose, CA); Brummond, William A. (Livermore, CA)

    1989-01-01

    A self-actuating reactor shutdown system incorporating a thermionic switched electromagnetic latch arrangement which is responsive to reactor neutron flux changes and to reactor coolant temperature changes. The system is self-actuating in that the sensing thermionic device acts directly to release (scram) the control rod (absorber) without reference or signal from the main reactor plant protective and control systems. To be responsive to both temperature and neutron flux effects, two detectors are used, one responsive to reactor coolant temperatures, and the other responsive to reactor neutron flux increase. The detectors are incorporated into a thermionic diode connected electrically with an electromagnetic mechanism which under normal reactor operating conditions holds the the control rod in its ready position (exterior of the reactor core). Upon reaching either a specified temperature or neutron flux, the thermionic diode functions to short-circuit the electromagnetic mechanism causing same to lose its holding power and release the control rod, which drops into the reactor core region under gravitational force.

  10. Use of microbes for paraffin cleanup at Naval Petroleum Reserve No. 3

    SciTech Connect (OSTI)

    Giangiacomo, L.; Khatib, A.

    1995-12-31

    Naval Petroleum Reserve No. 3 (NPR-3), also known as Teapot Dome, is a government-owned oil field in Natrona County, Wyoming. It is an asymmetrical anticline located on the western edge of the Powder River Basin, just south of the Salt Creek Anticline. Production started in 1922, and today the field is a marginally economic stripper field with average production of less than 3 BOPD (0.5 m{sup 3}/D) per well. Total field production is about 1,800 BOPD (286 m{sup 3}/D). The Second Wall Creek Formation was waterflooded from 1979 until June 1992 with poor results due to the extensive natural fracture system in this sandstone unit. Since water injection ceased, reservoir pressure has declined to very low levels. Liquids extraction and reinjection of the gas produced from high-GOR wells along the gas-oil contact continues, but the average gas cap pressure has fallen to approximately 150 psi (1.03 MPa) from an original pressure of 1,120 psi (7.72 MPa). Since the oil is highly paraffinic, wax deposition in the hydraulic fractures and the perforations has become a serious production problem. Microbial treatment was considered as a possible low-cost solution. Four wells were selected in the Second Wall Creek Reservoir with severe paraffin problems and production rates high enough to economically justify the treatment. Problems were experienced with the production of thick oil after approximately three months. This was interpreted to be a result of previously immobile paraffin being cleaned up. A slight decrease in the decline rate was seen in the wells, although some external factors cloud the interpretation. Microbial treatments were discontinued because of marginal economics. Three of the four wells produced additional oil and had a positive incremental cash flow. Oil viscosity tests did indicate that some positive microbial thinning was occurring, and changes to the treatment procedure may potentially yield more economic results in the future.

  11. Report of endangered species studies on Naval Petroleum Reserve No. 2, Kern County, California

    SciTech Connect (OSTI)

    O'Farrell, T.P.; Warrick, G.D.; Mathews, N.E.; Kato, T.T.

    1987-09-01

    Between 1983 and 1986 the size of the population of San Joaquin kit foxes (Vulpes macrotis mutica) on Naval Petroleum Reserve No. 2 (NPR-2), Kern County, California, was estimated semiannually using capture-recapture techniques. Although summer population estimates varied between 222 in 1983 and 121 in 1986, and winter estimates varies between 258 in 1984 and 91 in 1983, the population appeared to remain relatively stable at an apparent norm of 165. Kit foxes were abundant even in the intensely developed areas, and numbers and densities (1.12 to 2.49/sq mile) were consistently higher on NPR-2 than on neighboring NPR-1. The percentage of adult vixens that successfully raised pups was 55%, average litter size was 4.0 +- 0.0, and the sex ratio (M:F) of 25 pups was 1:1.5. Most (45.2%) foxes were killed by coyotes (Canis latrans), vehicles killed 6.4%, and 6.5% died of other causes. A cause could not be determined for 41.9% of the deaths. There was a general increase in coyote visitation rates at scent stations, but kit fox visitation rates generally decreased. Kit fox indices were consistently higher on NPR-2 than on NPR-1. Approximately 15% of the kit foxes on NPR-2 dispersed an average of 2.2 +- 0.2 miles. Average dispersal distance did not differ between the sexes. The longest dispersal was 6.9 miles. Proportionately more male than female pups dispersed. Remains of lagomorphs (jackrabbits and cottontails) and kangaroo rats had the highest frequency of occurrence in scats. Frequency of occurrence of lagomorph remains was greater in developed than in undeveloped habitats. Proportions of lagomorph remains increased and kangaroo rat remains decreased between 1983 and 1984. 62 refs., 9 figs., 24 tabs.

  12. Personnel radiation exposure in HTGR plants

    SciTech Connect (OSTI)

    Su, S.; Engholm, B.A.

    1980-01-01

    Occupational radiation exposures in high-temperature gas-cooled reactor (HTGR) plants were assessed. The expected rate of dose accumulations for a large HTGR steam cycle (HTGR-SC) unit is 0.07 man-rem/MW(e)y, while the design basis is 0.17 man-rem/MW(e)y. The comparable figure for actual light water reactor (LWR) experience is 1.3 man-rem/MW(e)y. The favorable HTGR occupational exposure is supported by results from the Peach Bottom Unit No. 1 HTGR and Fort St. Vrain HTGR plants and by operating experience at British gas-cooled reactor (GCR) stations.

  13. Risk Management for Sodium Fast Reactors.

    SciTech Connect (OSTI)

    Denman, Matthew R; Groth, Katrina; Cardoni, Jeffrey N; Wheeler, Timothy A.

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

  14. Some Aspects of Reactor Theory

    DOE R&D Accomplishments [OSTI]

    Weinberg, Alvin M.

    1952-10-10

    Some general remarks are made on reactor theory, particularly the asymptotic theory and multigroup methods. Unsolved reactor problems are also briefly discussed. (B.J.H.)

  15. Reactor Materials | Department of Energy

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

    reactor materials crosscut effort will enable the development of innovative and ... Research into specific degradation modes or material needs unique to a particular reactor ...

  16. Gas Reactor Technology R&D

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

    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

  17. Supercritical CO2 direct cycle Gas Fast Reactor (SC-GFR) concept.

    SciTech Connect (OSTI)

    Wright, Steven Alan; Parma, Edward J., Jr.; Suo-Anttila, Ahti Jorma; Al Rashdan, Ahmad; Tsvetkov, Pavel Valeryevich; Vernon, Milton E.; Fleming, Darryn D.; Rochau, Gary Eugene

    2011-05-01

    This report describes the supercritical carbon dioxide (S-CO{sub 2}) direct cycle gas fast reactor (SC-GFR) concept. The SC-GFR reactor concept was developed to determine the feasibility of a right size reactor (RSR) type concept using S-CO{sub 2} as the working fluid in a direct cycle fast reactor. Scoping analyses were performed for a 200 to 400 MWth reactor and an S-CO{sub 2} Brayton cycle. Although a significant amount of work is still required, this type of reactor concept maintains some potentially significant advantages over ideal gas-cooled systems and liquid metal-cooled systems. The analyses presented in this report show that a relatively small long-life reactor core could be developed that maintains decay heat removal by natural circulation. The concept is based largely on the Advanced Gas Reactor (AGR) commercial power plants operated in the United Kingdom and other GFR concepts.

  18. The IAEA international conference on fast reactors and related fuel cycles: highlights and main outcomes

    SciTech Connect (OSTI)

    Monti, S.; Toti, A.

    2013-07-01

    The 'International Conference on Fast Reactors and Related Fuel Cycles', which is regularly held every four years, represents the main international event dealing with fast reactors technology and related fuel cycles options. Main topics of the conference were new fast reactor concepts, design and simulation capabilities, safety of fast reactors, fast reactor fuels and innovative fuel cycles, analysis of past experience, fast reactor knowledge management. Particular emphasis was put on safety aspects, considering the current need of developing and harmonizing safety standards for fast reactors at the international level, taking also into account the lessons learned from the accident occurred at the Fukushima- Daiichi nuclear power plant in March 2011. Main advances in the several key areas of technological development were presented through 208 oral presentations during 41 technical sessions which shows the importance taken by fast reactors in the future of nuclear energy.

  19. Advanced Nuclear Technology: Advanced Light Water Reactors Utility Requirements Document Small Modular Reactors Inclusion Summary

    SciTech Connect (OSTI)

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

    This report summarizes a project by EPRI to include requirements for small modular light water reactors (smLWR) into the EPRI Utility Requirements Document (URD) for Advanced Light Water Reactors. The project was jointly funded by EPRI and the U.S. Department of Energy (DOE). The report covers the scope and content of the URD, the process used to revise the URD to include smLWR requirements, a summary of the major changes to the URD to include smLWR, and how to use the URD as revised to achieve value on new plant projects.

  20. Advanced Burner Reactor Preliminary NEPA Data Study.

    SciTech Connect (OSTI)

    Briggs, L. L.; Cahalan, J. E.; Deitrich, L. W.; Fanning, T. H.; Grandy, C.; Kellogg, R.; Kim, T. K.; Yang, W. S.; Nuclear Engineering Division

    2007-10-15

    The Global Nuclear Energy Partnership (GNEP) is a new nuclear fuel cycle paradigm with the goals of expanding the use of nuclear power both domestically and internationally, addressing nuclear waste management concerns, and promoting nonproliferation. A key aspect of this program is fast reactor transmutation, in which transuranics recovered from light water reactor spent fuel are to be recycled to create fast reactor transmutation fuels. The benefits of these fuels are to be demonstrated in an Advanced Burner Reactor (ABR), which will provide a representative environment for recycle fuel testing, safety testing, and modern fast reactor design and safeguard features. Because the GNEP programs will require facilities which may have an impact upon the environment within the meaning of the National Environmental Policy Act of 1969 (NEPA), preparation of a Programmatic Environmental Impact Statement (PEIS) for GNEP is being undertaken by Tetra Tech, Inc. The PEIS will include a section on the ABR. In support of the PEIS, the Nuclear Engineering Division of Argonne National Laboratory has been asked to provide a description of the ABR alternative, including graphics, plus estimates of construction and operations data for an ABR plant. The compilation of this information is presented in the remainder of this report. Currently, DOE has started the process of engaging industry on the design of an Advanced Burner Reactor. Therefore, there is no specific, current, vendor-produced ABR design that could be used for this PEIS datacall package. In addition, candidate sites for the ABR vary widely as to available water, geography, etc. Therefore, ANL has based its estimates for construction and operations data largely on generalization of available information from existing plants and from the environmental report assembled for the Clinch River Breeder Reactor Plant (CRBRP) design [CRBRP, 1977]. The CRBRP environmental report was chosen as a resource because it thoroughly

  1. Component failures that lead to reactor scrams. [PWR; BWR

    SciTech Connect (OSTI)

    Burns, E. T.; Wilson, R. J.; Lim, E. Y.

    1980-04-01

    This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation.

  2. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest cross-sectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor.

  3. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01

    This document summarizes the work performed by Pacific Northwest Laboratory (PNL) from January 1 through March 31, 1981, for the Division of Reactor Safety Research within the U.S. Nuclear Regulatory Commission (NRC). Evaluations of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibility of determining the strength of structural graphite, evaluating the feasibility of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the integrity of pressurized water reactor (PWR) steam generator tubes where service-induced degradation has been indicated. Experimental data and analytical models are being provided to aid in decision-making regarding pipeto- pipe impacts following postulated breaks in high-energy fluid system piping. Core thermal models are being developed to provide better digital codes to compute the behavior of full-scale reactor systems under postulated accident conditions. Fuel assemblies and analytical support are being provided for experimental programs at other facilities. These programs include loss-ofcoolant accident (LOCA) simulation tests at the NRU reactor, Chalk River, Canada; fuel rod deformation, severe fuel damage, and postaccident coolability tests for the ESSOR reactor Super Sara Test Program, Ispra, Italy; the instrumented fuel assembly irradiation program at Halden, Norway; and experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory (INEL). These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  4. Annual Reports | National Nuclear Security Administration | (NNSA)

    National Nuclear Security Administration (NNSA)

    Annual Reports Environmental Monitoring Report NT-16-1 - May 2016 - ENVIRONMENTAL MONITORING AND DISPOSAL OF RADIOACTIVE WASTES FROM U.S. NAVAL NUCLEAR-POWERED SHIPS AND THEIR SUPPORT FACILITIES Radiation Exposure Monitoring Report NT-16-2 - May 2016 - OCCUPATIONAL RADIATION EXPOSURE FROM U.S. NAVAL NUCLEAR PLANTS AND THEIR SUPPORT FACILITIES Report NT-16-3 - May 2016 - OCCUPATIONAL RADIATION EXPOSURE FROM NAVAL REACTORS' DEPARTMENT OF ENERGY FACILITIES Occupational Safety and Health Report

  5. Technical Feasibility Study for Deployment of Ground-Source Heat Pump Systems: Portsmouth Naval Shipyard -- Kittery, Maine

    SciTech Connect (OSTI)

    Hillesheim, M.; Mosey, G.

    2014-11-01

    The U.S. Environmental Protection Agency (EPA) Office of Solid Waste and Emergency Response, in accordance with the RE-Powering America's Lands initiative, engaged the U.S. Department of Energy's (DOE) National Renewable Energy Laboratory (NREL) to conduct feasibility studies to assess the viability of developing renewable energy generating facilities on contaminated sites. Portsmouth Naval Shipyard (PNSY) is a United States Navy facility located on a series of conjoined islands in the Piscataqua River between Kittery, ME and Portsmouth, NH. EPA engaged NREL to conduct a study to determine technical feasibility of deploying ground-source heat pump systems to help PNSY achieve energy reduction goals.

  6. Laser-driven fusion reactor

    DOE Patents [OSTI]

    Hedstrom, J.C.

    1973-10-01

    A laser-driven fusion reactor consisting of concentric spherical vessels in which the thermonuclear energy is derived from a deuterium-tritium (D + T) burn within a pellet'', located at the center of the vessels and initiated by a laser pulse. The resulting alpha -particle energy and a small fraction of the neutron energy are deposited within the pellet; this pellet energy is eventually transformed into sensible heat of lithium in a condenser outside the vessels. The remaining neutron energy is dissipated in a lithium blanket, located within the concentric vessels, where the fuel ingredient, tritium, is also produced. The heat content of the blanket and of the condenser lithium is eventually transferred to a conventional thermodynamic plant where the thermal energy is converted to electrical energy in a steam Rankine cycle. (Official Gazette)

  7. Reactor operation environmental information document

    SciTech Connect (OSTI)

    Wike, L.D.; Specht, W.L.; Mackey, H.E.; Paller, M.H.; Wilde, E.W.; Dicks, A.S.

    1989-12-01

    The Savannah River Site (SRS) is a large United States Department of Energy installation on the upper Atlantic Coastal Plain of South Carolina. The SRS contains diverse habitats, flora, and fauna. Habitats include upland terrestrial areas, varied wetlands including Carolina Bays, the Savannah River swamp system, and impoundment related and riparian wetlands, and the aquatic habitats of several stream systems, two large cooling reservoirs, and the Savannah River. These diverse habitats support a large variety of plants and animals including many commercially or recreational valuable species and several rare, threatened or endangered species. This volume describes the major habitats and their biota found on the SRS, and discuss the impacts of continued operation of the K, L, and P production reactors.

  8. Material unaccounted for at the Southwest Experimental Fast Oxide Reactor: The SEFOR MUF

    SciTech Connect (OSTI)

    Higinbotham, W.A.

    1994-11-07

    The U.S. Atomic Energy Commission contracted with the General Electric Company to design, construct, and operate the Southwest Experimental Fast Oxide Reactor (SEFOR) to measure the Doppler effect for fast neutron breeder reactors. It contracted with Nuclear Fuel Services to fabricate the fuel rods for the reactor. When the reactor went critical in May, 1969, it appeared that some of the mixed uranium-plutonium oxide (MOX) fuel rods did not contain the specified quantity of plutonium. The SEFOR operators soon found several fuel rods which appeared to be low in plutonium. The safeguards group at Brookhaven was asked to look into the problem and, if possible, determine how much plutonium was missing from the unirradiated rods and from the larger number which had been slightly irradiated in the reactor. It was decided that the plutonium content of the unirradiated and irradiated rods could be measured relative to a reference rod using a high resolution gamma-ray detector and also by neutron measurements using an auto-correlation circuit recently developed at the Naval Research Laboratory (NRL). During the next two years, Brookhaven personnel and C.V. Strain of NRL made several trips to the SEFOR reactor. About 250 of the 775 rods were measured by two or more methods, using a sodium-iodide detector, a high-resolution germanium detector, a neutron detector, or the reactor (to measure reactivity). The research team concluded that 4.6 {+-} 0.46 kg of plutonium was missing out of the 433 kg that the rods should have contained. This report describes the SEFOR experiment and the procedures used to determine the material unaccounted for, or MUF.

  9. Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

    SciTech Connect (OSTI)

    Pope, Michael A.; Gougar, Hans D.; Ryskamp, J. M.

    2015-03-01

    The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

  10. Fast Breeder Reactor studies

    SciTech Connect (OSTI)

    Till, C.E.; Chang, Y.I.; Kittel, J.H.; Fauske, H.K.; Lineberry, M.J.; Stevenson, M.G.; Amundson, P.I.; Dance, K.D.

    1980-07-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear safeguards. An Annex compares the cost of decommissioning light-water and fast breeder reactors. Separate abstracts are included for each of the parts.

  11. CONTROL FOR NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Lichtenberger, H.V.; Cameron, R.A.

    1959-03-31

    S>A control rod operating device in a nuclear reactor of the type in which the control rod is gradually withdrawn from the reactor to a position desired during stable operation is described. The apparatus is comprised essentially of a stop member movable in the direction of withdrawal of the control rod, a follower on the control rod engageable with the stop and means urging the follower against the stop in the direction of withdrawal. A means responsive to disengagement of the follower from the stop is provided for actuating the control rod to return to the reactor shut-down position.

  12. Microfluidic electrochemical reactors

    DOE Patents [OSTI]

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

    A microfluidic electrochemical reactor includes an electrode and one or more microfluidic channels on the electrode, where the microfluidic channels are covered with a membrane containing a gas permeable polymer. The distance between the electrode and the membrane is less than 500 micrometers. The microfluidic electrochemical reactor can provide for increased reaction rates in electrochemical reactions using a gaseous reactant, as compared to conventional electrochemical cells. Microfluidic electrochemical reactors can be incorporated into devices for applications such as fuel cells, electrochemical analysis, microfluidic actuation, pH gradient formation.

  13. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J.; Land, John T.; Misvel, Michael C.

    1994-01-01

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled.

  14. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, R.J.; Land, J.T.; Misvel, M.C.

    1994-06-07

    A nuclear reactor reflector is disclosed that comprises a stack of reflector blocks with vertical water flow passages to cool the reflector. The interface between blocks is opposite support points for reactor fuel rods. Water flows between the reflector and the reactor barrel from passages in a bottom block. The top block contains a flange to limit this flow and the flange has a slot to receive an alignment pin that is welded to the barrel. The pin is held in the slot by two removable shims. Alignment bars extend the length of the stack in slots machined in each block when the stack is assembled. 12 figs.

  15. COOLED NEUTRONIC REACTOR

    DOE Patents [OSTI]

    Binner, C.R.; Wilkie, C.B.

    1958-03-18

    This patent relates to a design for a reactor of the type in which a fluid coolant is flowed through the active portion of the reactor. This design provides for the cooling of the shielding material as well as the reactor core by the same fluid coolant. The core structure is a solid moderator having coolant channels in which are disposed the fuel elements in rod or slug form. The coolant fluid enters the chamber in the shield, in which the core is located, passes over the inner surface of said chamber, enters the core structure at the center, passes through the coolant channels over the fuel elements and out through exhaust ducts.

  16. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03

    The object of this invention is to provide a compact torus fusion reactor with dramatic simplification of plasma confinement design. Another object of this invention is to provide a compact torus fusion reactor with low magnetic field and small aspect ratio stable plasma confinement. In accordance with the principles of this invention there is provided a compact toroidal-type plasma confinement fusion reactor in which only the indispensable components inboard of a tokamak type of plasma confinement region, mainly a current conducting medium which carries electrical current for producing a toroidal magnet confinement field about the toroidal plasma region, are retained.

  17. NUCLEAR REACTOR FUEL SYSTEMS

    DOE Patents [OSTI]

    Thamer, B.J.; Bidwell, R.M.; Hammond, R.P.

    1959-09-15

    Homogeneous reactor fuel solutions are reported which provide automatic recombination of radiolytic gases and exhibit large thermal expansion characteristics, thereby providing stability at high temperatures and enabling reactor operation without the necessity of apparatus to recombine gases formed by the radiolytic dissociation of water in the fuel and without the necessity of liquid fuel handling outside the reactor vessel except for recovery processes. The fuels consist of phosphoric acid and water solutions of enriched uranium, wherein the uranium is in either the hexavalent or tetravalent state.

  18. Site Map | U.S. DOE Office of Science (SC)

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

    Home Site Map Front page Front page of National Nuclear Security Administration Main menu People Mission Powering the Nuclear Navy Concern for the Environment Protection of People Naval Nuclear Propulsion Plants Management and Administration Public Affairs More About NNSA's Naval Reactors Office Emergency Response Counterterrorism Recapitalizing Our Infrastructure Preventing Proliferation Managing the Stockpile Dismantlement and Disposition Stockpile Stewardship Program Quarterly Experiments

  19. Site Map | National Nuclear Security Administration | (NNSA)

    National Nuclear Security Administration (NNSA)

    Home Site Map Front page Front page of National Nuclear Security Administration Main menu People Mission Powering the Nuclear Navy Concern for the Environment Protection of People Naval Nuclear Propulsion Plants Management and Administration Public Affairs More About NNSA's Naval Reactors Office Emergency Response Counterterrorism Recapitalizing Our Infrastructure Preventing Proliferation Managing the Stockpile Dismantlement and Disposition Stockpile Stewardship Program Quarterly Experiments

  20. Report

    National Nuclear Security Administration (NNSA)

    Knolls Atomic Power Laboratory/ Kenneth A. Kesselring Site Bettis Atomic Power Laboratory Los Alamos National Laboratory (LANL) Nevada National Security Site Naval Reactors Facility Sandia National Laboratories (SNL) Lawrence Livermore National Laboratory (LLNL) Savannah River Site (SRS) DOE/NNSA Headquarters Pantex Plant (PX) Albuquerque Complex Headquarters National Security Laboratories Plants and Sites Naval Reactors Laboratories The Nuclear Security Enterprise DOE/NNSA is responsible for

  1. Microsoft Word - Enterprise EA (for public)2.docx

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

    enclosure to letter UNITED STATES DEPARTMENT OF THE NAVY DRAFT ENVIRONMENTAL ASSESSMENT ON THE DISPOSAL OF DECOMMISSIONED, DEFUELED NAVAL REACTOR PLANTS FROM USS ENTERPRISE (CVN 65) SEPTEMBER 2011 DRAFT USS ENTERPRISE EA RESPONSIBLE AGENCIES: Lead Federal Agency: U.S. Department of the Navy Cooperating Agency: U.S. Department of Energy TITLE: Draft Environmental Assessment on the Disposal of Decommissioned, Defueled, Naval Reactor Plants from USS ENTERPRISE (CVN 65) DRAFT USS ENTERPRISE EA i

  2. Small Modular Nuclear Reactors: Parametric Modeling of Integrated...

    Office of Environmental Management (EM)

    Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ... Small Modular Nuclear Reactors: Parametric Modeling of Integrated Reactor Vessel ...

  3. Measurements of the reactor neutron power in absolute units

    SciTech Connect (OSTI)

    Lebedev, G. V.

    2015-12-15

    The neutron power of the reactor of the Yenisei space nuclear power plant is measured in absolute units using the modernized method of correlation analysis during the ground-based tests of the Yenisei prototypes. Results of the experiments are given. The desired result is obtained in a series of experiments carried out at the stage of the plant preparation for tests. The acceptability of experimental data is confirmed by the results of measuring the reactor neutron power in absolute units at the nominal level by the thermal balance during the life cycle tests of the ground prototypes.

  4. P Reactor Grouting

    SciTech Connect (OSTI)

    2010-01-01

    Filling the P Reactor with grout. This seals the radioactive material and reduces the environmental footprint left from the Cold War. Project sponsored by the Recovery Act at the Savannah River Site.

  5. Reactor hot spot analysis

    SciTech Connect (OSTI)

    Vilim, R.B.

    1985-08-01

    The principle methods for performing reactor hot spot analysis are reviewed and examined for potential use in the Applied Physics Division. The semistatistical horizontal method is recommended for future work and is now available as an option in the SE2-ANL core thermal hydraulic code. The semistatistical horizontal method is applied to a small LMR to illustrate the calculation of cladding midwall and fuel centerline hot spot temperatures. The example includes a listing of uncertainties, estimates for their magnitudes, computation of hot spot subfactor values and calculation of two sigma temperatures. A review of the uncertainties that affect liquid metal fast reactors is also presented. It was found that hot spot subfactor magnitudes are strongly dependent on the reactor design and therefore reactor specific details must be carefully studied. 13 refs., 1 fig., 5 tabs.

  6. NEUTRONIC REACTOR STRUCTURE

    DOE Patents [OSTI]

    Daniels, F.

    1961-10-24

    A reactor core, comprised of vertical stacks of hexagonal blocks of beryllium oxide having axial cylindrical apertures extending therethrough and cylindrical rods of a sintered mixture of uranium dioxide and beryllium oxide, is described. (AEC)

  7. Next Generation Nuclear Plant Project Evaluation of Siting a HTGR Co-generation Plant on an Operating Commercial Nuclear Power Plant Site

    SciTech Connect (OSTI)

    L.E. Demick

    2011-10-01

    This paper summarizes an evaluation by the Idaho National Laboratory (INL) Next Generation Nuclear Plant (NGNP) Project of siting a High Temperature Gas-cooled Reactor (HTGR) plant on an existing nuclear plant site that is located in an area of significant industrial activity. This is a co-generation application in which the HTGR Plant will be supplying steam and electricity to one or more of the nearby industrial plants.

  8. Decommissioning experience from the Experimental Breeder Reactor-II.

    SciTech Connect (OSTI)

    Henslee, S.P.; Rosenberg, K.E.

    2002-03-28

    Consistent with the intent of this International Atomic Energy Agency technical meeting, decommissioning operating experience and contributions to the preparation for the Coordinated Research Project from Experimental Breeder Reactor-II activities will be discussed. This paper will review aspects of the decommissioning activities of the Experimental Breeder Reactor-II, make recommendations for future decommissioning activities and reactor system designs and discuss relevant areas of potential research and development. The Experimental Breeder Reactor-II (EBR-II) was designed as a 62.5 MWt, metal fueled, pool reactor with a conventional 19 MWe power plant. The productive life of the EBR-II began with first operations in 1964. Demonstration of the fast reactor fuel cycle, serving as an irradiation facility, demonstration of fast reactor passive safety and lastly, was well on its way to close the fast breeder fuel cycle for the second time when the Integral Fast Reactor program was prematurely ended in October 1994 with the shutdown of the EBR-II. The shutdown of the EBR-II was dictated without an associated planning phase that would have provided a smooth transition to shutdown. Argonne National Laboratory and the U.S. Department of Energy arrived at a logical plan and sequence for closure activities. The decommissioning activities as described herein fall into in three distinct phases.

  9. Molten metal reactors

    DOE Patents [OSTI]

    Bingham, Dennis N; Klingler, Kerry M; Turner, Terry D; Wilding, Bruce M

    2013-11-05

    A molten metal reactor for converting a carbon material and steam into a gas comprising hydrogen, carbon monoxide, and carbon dioxide is disclosed. The reactor includes an interior crucible having a portion contained within an exterior crucible. The interior crucible includes an inlet and an outlet; the outlet leads to the exterior crucible and may comprise a diffuser. The exterior crucible may contain a molten alkaline metal compound. Contained between the exterior crucible and the interior crucible is at least one baffle.

  10. Future reactor experiments

    SciTech Connect (OSTI)

    Wen, Liangjian

    2015-07-15

    The non-zero neutrino mixing angle θ{sub 13} has been discovered and precisely measured by the current generation short-baseline reactor neutrino experiments. It opens the gate of measuring the leptonic CP-violating phase and enables the neutrino mass ordering. The JUNO and RENO-50 proposals aim at resolving the neutrino mass ordering using reactors. The experiment design, physics sensitivity, technical challenges as well as the progresses of those two proposed experiments are reviewed in this paper.

  11. Compact power reactor

    DOE Patents [OSTI]

    Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

    1978-01-01

    There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

  12. Waste Treatment & Immobilization Plant Project - Hanford Site

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

    Treatment Plant About Us About Hanford Cleanup Hanford History Hanford Site Wide Programs Contact Us 100 Area 118-K-1 Burial Ground 200 Area 222-S Laboratory 242-A Evaporator 300 Area 324 Building 325 Building 400 Area/Fast Flux Test Facility 618-10 and 618-11 Burial Grounds 700 Area B Plant B Reactor C Reactor Canister Storage Building and Interim Storage Area Canyon Facilities Cold Test Facility D and DR Reactors Effluent Treatment Facility Environmental Restoration Disposal Facility F Reactor

  13. F Reactor Inspection

    SciTech Connect (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  14. F Reactor Inspection

    ScienceCinema (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-11-24

    Workers from Mission Support Alliance, LLC., removed the welds around the steel door of the F Reactor before stepping inside the reactor to complete its periodic inspection. This is the first time the Department of Energy (DOE) has had the reactor open since 2008. The F Reactor is one of nine reactors along the Columbia River at the Department's Hanford Site in southeastern Washington State, where environmental cleanup has been ongoing since 1989. As part of the Tri-Party Agreement, the Department completes surveillance and maintenance activities of cocooned reactors periodically to evaluate the structural integrity of the safe storage enclosure and to ensure confinement of any remaining hazardous materials. "This entry marks a transition of sorts because the Hanford Long-Term Stewardship Program, for the first time, was responsible for conducting the entry and surveillance and maintenance activities," said Keith Grindstaff, Energy Department Long-Term Stewardship Program Manager. "As the River Corridor cleanup work is completed and transitioned to long-term stewardship, our program will manage any on-going requirements."

  15. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    SciTech Connect (OSTI)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  16. NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

    DOE Patents [OSTI]

    West, J.M.; Weills, J.T.

    1960-03-15

    A method is given for operating a nuclear reactor having a negative coefficient of reactivity to compensate for the change in reactor reactivity due to the burn-up of the xenon peak following start-up of the reactor. When it is desired to start up the reactor within less than 72 hours after shutdown, the temperature of the reactor is lowered prior to start-up, and then gradually raised after start-up.

  17. A plan for implementation of innovative hazardous waste minimization techniques at an eastern US Naval Plating Shop

    SciTech Connect (OSTI)

    Walker, J.F. Jr.; Villiers-Fisher, J.F.; Brown, C.H. Jr.

    1987-01-01

    Oak Ridge National Laboratory (ORNL) was contracted by the Naval Energy and Environmental Support Activity (NEESA) to analyze the wastewater problems at a Naval Ordnance Station (NOS) plating shop in the eastern United States to recommend innovative wastewater treatment technologies for handling those problems and to implement the recommended treatment technology. Hexavalent chromium was identified as the major problem area at NOS. Water conservation measures were recommended which would reduce the volume of chromium-contaminated wastewater from approximately 300 L/min to approximately 20 L/min. A treatment scheme consisting of RO followed by evaporation of the RO concentrate steam was recommended. Paint-stripping operations at NOS potentially contaminate the wastewater with phenol, trichloroethane, and possibly other organics. However, the need for a treatment unit for removal of organics could not be established due to a lack of organic analytical data. A characterization study was therefore recommended for the NOS plating shop. If treatment for organics is necessary, the treatment unit might include two-stage filtration for removal of paint flakes or other solids, air stripping for removal of volatile organics, and carbon adsorption for removal of residual organics. 7 refs., 6 figs., 3 tabs.

  18. Naval Air Warfare Center, Aircraft Division at Warminster Environmental Materials Program. Phase 1. Interim report, October 1989-May 1992

    SciTech Connect (OSTI)

    Spadafora, S.J.; Hegedus, C.R.; Clark, K.J.; Eng, A.T.; Pulley, D.F.

    1992-06-24

    With the recent increase in awareness about the environment, there is an expanding concern of the deleterious effects of current materials and processes. Federal, state and local environmental agencies such as the EPA, State Air Resource Boards and local Air Quality Management Districts (AQMD) have issued legislation that restrict or prohibit the use and disposal of hazardous materials. National and local laws like the Clean Air and Clean Water Acts, Resource Conservation and Recovery Act, and AQMD regulations are examples of rules that govern the handling and disposal of hazardous materials and waste. The Department of Defense (DoD), in support of this effort, has identified the major generators of hazardous materials and hazardous waste to be maintenance depots and operations, particularly cleaning, pretreating, plating, painting and paint removal processes. Reductions of waste in these areas has been targeted as a primary goal in the DOD. The Navy is committed to significantly reducing its current hazardous waste generation and is working to attain a near zero discharge of hazardous waste by the year 2000. In order to attain these goals, the Naval Air Warfare Center Aircraft Division at Warminster has organized and is carrying out a comprehensive program in cooperation with the Naval Air Systems Command, the Air Force and the Department of Energy that deal with the elimination or reduction of hazardous materials. .... Environmental materials, Organic coatings, Inorganic pretreatments, Paint removal techniques, Cleaners, CFC'S.

  19. Public health assessment for Treasure Island Naval Station, Hunters Point Annex, San Francisco, San Francisco County, California, Region 9. Cerclis No. CA1170090087. Final report

    SciTech Connect (OSTI)

    Not Available

    1994-09-30

    Naval Station Treasure Island, Hunters Point Annex (HPA), an inactive Naval shipyard located on a peninsula in the San Francisco Bay, San Francisco, California, was listed for base closure in 1990. Metals, pesticides, radium-226, polychlorinated biphenyls (PCBs), polycyclic aromatic hydrocarbons (PAHs), volatile organic compounds, semivolatile organic compounds, petroleum products, and asbestos have been found in various media such as soil, groundwater, surface water, air, and sediments. Navy contractors have identified 58 HPA areas where there may be contamination; investigations at these areas are ongoing.

  20. Thermoacoustic Thermometry for Nuclear Reactor Monitoring

    SciTech Connect (OSTI)

    James A. Smith; Dale K. Kotter; Steven L. Garrett; Randall A. Ali

    2013-06-01

    On Friday, March 11, 2011, at 2:46pm (Japan Standard Trme), the Tohoku region on the east coast of northern Japan experi­enced what would become known as the largest earthquake in the country's history at magnitude 9.0 on the Richter scale. The Fukushima Daiichi nuclear power plant suffered exten­sive and irreversible damage. Six operating units were at the site, each with a boiling water reactor. When the earthquake struck, three of the six reactors were operating and the others were in a periodic inspection outage phase. In one reactor, all of the fuel had been relocated to a spent fuel pool in the reactor building. The seismic acceleration caused by the earthquake brought the three operating units to an automatic shutdown. Since there was damage to the power transmission lines, the emergency diesel generators (EDG) were automat­ically started to ensure continued cooling of the reactors and spent fuel pools. The situation was under control until the tsunami hit about forty-five minutes later with a maximum wave height of approximately 15 meters, which was three times taller than the sea wall of 5m. The influx of water submerged the EDGs, the electrical switchgear, and dc batteries, resulting in the total loss of power to five of the six reactors. The flooding also resulted in the loss of instrumentation that would have other­ wise been used to monitor and control the emergency. The ugly aftermath included high radiation exposure to operators at the nuclear power plants and early contamina­tion of food supplies and water within several restricted areas in Japan, where high radiation levels have rendered them un­safe for human habitation. While the rest of the story will remain a tragic history, it is this part of the series of unfortunate events that has inspired our research. It has indubitably highlighted the need for a novel sensor and instrumentation system that can withstand similar or worse conditions to avoid future catastrophe and assume damage

  1. Nebraska Nuclear Profile - Power Plants

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

    Nebraska nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Cooper Unit 1",767,"6,793",61.4,"Nebraska Public Power District" "Fort Calhoun Unit 1",478,"4,261",38.6,"Omaha Public Power District" "2 Plants 2

  2. Louisiana Nuclear Profile - Power Plants

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

    Louisiana nuclear power plants, summer capacity and net generation, 2010" "Plant Name/Total Reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (Pprcent)","Owner" "River Bend Unit 1",974,"8,363",44.9,"Entergy Gulf States - LA LLC" "Waterford 3 Unit 3","1,168","10,276",55.1,"Entergy Louisiana Inc" "2 Plants 2

  3. Georgia Nuclear Profile - Power Plants

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

    nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Edwin I Hatch Unit 1, Unit 2","1,759","13,902",41.5,"Georgia Power Co" "Vogtle Unit 1, Unit 2","2,302","19,610",58.5,"Georgia Power Co" "2 Plants 4

  4. Tennessee Nuclear Profile - Power Plants

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

    Tennessee nuclear power plants, summer capacity and net generation, 2010" "Plant name/total reactors","Summer capacity (mw)","Net generation (thousand mwh)","Share of State nuclear net generation (percent)","Owner" "Sequoyah Unit 1, Unit 2","2,278","18,001",64.9,"Tennessee Valley Authority" "Watts Bar Nuclear Plant Unit 1","1,123","9,738",35.1,"Tennessee Valley

  5. REACTOR GROUT THERMAL PROPERTIES

    SciTech Connect (OSTI)

    Steimke, J.; Qureshi, Z.; Restivo, M.; Guerrero, H.

    2011-01-28

    Savannah River Site has five dormant nuclear production reactors. Long term disposition will require filling some reactor buildings with grout up to ground level. Portland cement based grout will be used to fill the buildings with the exception of some reactor tanks. Some reactor tanks contain significant quantities of aluminum which could react with Portland cement based grout to form hydrogen. Hydrogen production is a safety concern and gas generation could also compromise the structural integrity of the grout pour. Therefore, it was necessary to develop a non-Portland cement grout to fill reactors that contain significant quantities of aluminum. Grouts generate heat when they set, so the potential exists for large temperature increases in a large pour, which could compromise the integrity of the pour. The primary purpose of the testing reported here was to measure heat of hydration, specific heat, thermal conductivity and density of various reactor grouts under consideration so that these properties could be used to model transient heat transfer for different pouring strategies. A secondary purpose was to make qualitative judgments of grout pourability and hardened strength. Some reactor grout formulations were unacceptable because they generated too much heat, or started setting too fast, or required too long to harden or were too weak. The formulation called 102H had the best combination of characteristics. It is a Calcium Alumino-Sulfate grout that contains Ciment Fondu (calcium aluminate cement), Plaster of Paris (calcium sulfate hemihydrate), sand, Class F fly ash, boric acid and small quantities of additives. This composition afforded about ten hours of working time. Heat release began at 12 hours and was complete by 24 hours. The adiabatic temperature rise was 54 C which was within specification. The final product was hard and displayed no visible segregation. The density and maximum particle size were within specification.

  6. Electrochemistry of Water-Cooled Nuclear Reactors

    SciTech Connect (OSTI)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

    This project developed a comprehensive mathematical and simulation model for calculating thermal hydraulic, electrochemical, and corrosion parameters, viz. temperature, fluid flow velocity, pH, corrosion potential, hydrogen injection, oxygen contamination, stress corrosion cracking, crack growth rate, and other important quantities in the coolant circuits of water-cooled nuclear power plants, including both Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). The model is being used to assess the three major operational problems in Pressurized Water Reactors (PWR), which include mass transport, activity transport, and the axial offset anomaly, and provide a powerful tool for predicting the accumulation of SCC damage in BWR primary coolant circuits as a function of operating history. Another achievement of the project is the development of a simulation tool to serve both as a training tool for plant operators and as an engineering test-bed to evaluate new equipment and operating strategies (normal operation, cold shut down and others). The development and implementation of the model allows us to estimate the activity transport or "radiation fields" around the primary loop and the vessel, as a function of the operating parameters and the water chemistry.

  7. Microchannel Reactor System for Catalytic Hydrogenation

    SciTech Connect (OSTI)

    Adeniyi Lawal; Woo Lee; Ron Besser; Donald Kientzler; Luke Achenie

    2010-12-22

    We successfully demonstrated a novel process intensification concept enabled by the development of microchannel reactors, for energy efficient catalytic hydrogenation reactions at moderate temperature, and pressure, and low solvent levels. We designed, fabricated, evaluated, and optimized a laboratory-scale microchannel reactor system for hydrogenation of onitroanisole and a proprietary BMS molecule. In the second phase of the program, as a prelude to full-scale commercialization, we designed and developed a fully-automated skid-mounted multichannel microreactor pilot plant system for multiphase reactions. The system is capable of processing 1 – 10 kg/h of liquid substrate, and an industrially relevant immiscible liquid-liquid was successfully demonstrated on the system. Our microreactor-based pilot plant is one-of-akind. We anticipate that this process intensification concept, if successfully demonstrated, will provide a paradigm-changing basis for replacing existing energy inefficient, cost ineffective, environmentally detrimental slurry semi-batch reactor-based manufacturing practiced in the pharmaceutical and fine chemicals industries.

  8. Integrated systems analysis of the PIUS reactor

    SciTech Connect (OSTI)

    Fullwood, F.; Kroeger, P.; Higgins, J.

    1993-11-01

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  9. Analysis of scrams and forced outages at boiling water reactors

    SciTech Connect (OSTI)

    Earle, R. T.; Sullivan, W. P.; Miller, K. R.; Schwegman, W. J.

    1980-07-01

    This report documents the results of a study of scrams and forced outages at General Electric Boiling Water Reactors (BWRs) operating in the United States. This study was conducted for Sandia Laboratories under a Light Water Reactor Safety Program which it manages for the United States Department of Energy. Operating plant data were used to identify the causes of scrams and forced outages. Causes of scrams and forced outages have been summarized as a function of operating plant and plant age and also ranked according to the number of events per year, outage time per year, and outage time per event. From this ranking, identified potential improvement opportunities were evaluated to determine the associated benefits and impact on plant availability.

  10. Experimental development of a multi-solid fluidized bed reactor concept

    SciTech Connect (OSTI)

    Litt, R.D.; Paisley, M.A.; Tewksbury, T.L.

    1990-02-01

    Battelle's Columbus Division is developing a coal mild gasification process based upon the Multi-Solid Fluidized bed reactor system to produce high quality liquid and gaseous products. This process uses 2-stages to gasify coal at high throughputs to produce a range of products in compact reactors without requiring an oxygen plant. 8 refs., 14 figs., 12 tabs.

  11. Pressurized fluidized bed reactor and a method of operating the same

    DOE Patents [OSTI]

    Isaksson, Juhani

    1996-01-01

    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.

  12. Pressurized fluidized bed reactor and a method of operating the same

    DOE Patents [OSTI]

    Isaksson, J.

    1996-02-20

    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.

  13. Advanced Gas Reactor Fuel Program's TRISO Particle Fuel Sets A New World Record For Irradiation Performance

    Broader source: Energy.gov [DOE]

    As part of the Office of Nuclear Energy's Next Generation Nuclear Plant (NGNP) Program, the Advanced Gas Reactor (AGR) Fuel Development Program has achieved a new international record for...

  14. W.B.; Allison, G.S. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED...

    Office of Scientific and Technical Information (OSTI)

    nuclear fuel bundle data for use in fuel bundle handling Weihermiller, W.B.; Allison, G.S. 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; FUEL ELEMENT CLUSTERS; REMOTE...

  15. Solar Thermochemical Advanced Reactor System, Wins R&D 100 Award

    Broader source: Energy.gov [DOE]

    Solar Thermochemical Advanced Reactor System, or STARS, converts natural gas and sunlight into a more energy-rich fuel called syngas, which power plants can burn to make electricity.

  16. CORAL: a stepping stone for establishing the Indian fast reactor fuel reprocessing technology

    SciTech Connect (OSTI)

    Venkataraman, M.; Natarajan, R.; Raj, Baldev

    2007-07-01

    The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR) spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)

  17. Hadlock, R.K.; Abbey, O.B. 22 GENERAL STUDIES OF NUCLEAR REACTORS...

    Office of Scientific and Technical Information (OSTI)

    on ultimate heat sinks--cooling ponds Hadlock, R.K.; Abbey, O.B. 22 GENERAL STUDIES OF NUCLEAR REACTORS; 20 FOSSIL-FUELED POWER PLANTS; COOLING PONDS; PERFORMANCE TESTING; NUCLEAR...

  18. Advanced Reactor Concepts Technical Review Panel Report | Department...

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

    The concepts included five fast reactors and three thermal reactors. As to reactor coolants, there were three sodium-cooled reactors, two gas-cooled reactors, one light ...

  19. REACTOR AND NOVEL METHOD

    DOE Patents [OSTI]

    Young, G.J.; Ohlinger, L.A.

    1958-06-24

    A nuclear reactor of the type which uses a liquid fuel and a method of controlling such a reactor are described. The reactor is comprised essentially of a tank for containing the liquid fuel such as a slurry of discrete particles of fissionnble material suspended in a heavy water moderator, and a control means in the form of a disc of neutron absorbirg material disposed below the top surface of the slurry and parallel thereto. The diameter of the disc is slightly smaller than the diameter of the tank and the disc is perforated to permit a flow of the slurry therethrough. The function of the disc is to divide the body of slurry into two separate portions, the lower portion being of a critical size to sustain a nuclear chain reaction and the upper portion between the top surface of the slurry and the top surface of the disc being of a non-critical size. The method of operation is to raise the disc in the reactor until the lower portion of the slurry has reached a critical size when it is desired to initiate the reaction, and to lower the disc in the reactor to reduce the size of the lower active portion the slurry to below criticality when it is desired to stop the reaction.

  20. Plant Operational Status - Pantex Plant

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

    Plant Operational Status Plant Operational Status Page Content Shift 1 - Day The Pantex Plant is open for normal Day Shift operations. Plant personnel are to report as assigned. Personnel may call 477-3000, Option 1 for additional details. Shift 2 - Swing The Pantex Plant is open for normal Swing Shift operations. Plant personnel are to report as assigned. Personnel may call 477-3000, Option 1 for additional details. Shift 3 - Grave The Pantex Plant is open for normal Graveyard Shift operations.