Sample records for mental breeder reactor

  1. 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.

    1980-07-01T23:59:59.000Z

    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.

  2. Light-Water Breeder Reactor

    DOE Patents [OSTI]

    Beaudoin, B. R.; Cohen, J. D.; Jones, D. H.; Marier, Jr, L. J.; Raab, H. F.

    1972-06-20T23:59:59.000Z

    Described is a light-water-moderated and -cooled nuclear breeder reactor of the seed-blanket type characterized by core modules comprising loosely packed blanket zones enriched with fissile fuel and axial zoning in the seed and blanket regions within each core module. Reactivity control over lifetime is achieved by axial displacement of movable seed zones without the use of poison rods in the embodiment illustrated. The seed is further characterized by a hydrogen-to-uranium-233 atom ratio in the range 10 to 200 and a uranium-233-to-thorium-232 atom ratio ranging from 0.012 to 0.200. The seed occupies from 10 to 35 percent of the core volume in the form of one or more individual islands or annuli. (NSA 26: 55130)

  3. Heterogeneous effects in fast breeder reactors

    E-Print Network [OSTI]

    Gregory, Michael Vladimir

    1973-01-01T23:59:59.000Z

    Heterogeneous effects in fast breeder reactors are examined through development of simple but accurate models for the calculation of a posteriori corrections to a volume-averaged homogeneous representation. Three distinct ...

  4. Fission reactor experiments for solid breeder blankets

    SciTech Connect (OSTI)

    Gierszewski, P.J.; Abdou, M.A.; Puigh, R.

    1986-11-01T23:59:59.000Z

    The testing needs for solid breeder blanket development are different from those for liquid breeder blankets. In particular, a reasonable number of moderate volume test sites in a neutron environment are needed. Existing fission reactors are shown to be able to provide this environment with reasonable simulation of many important blanket conditions. Three major additional fission reactor tests are identified beyond those presently underway. These are thermal behavior, advanced in-situ tritium recovery and nuclear submodule experiments.

  5. Technical note Coolant void worth in fast breeder reactors

    E-Print Network [OSTI]

    Technical note Coolant void worth in fast breeder reactors and accelerator-driven transuranium as a function of fuel composition and core geometry for several model fast breeder reactors and accelerator to be inferior to those employing sodium for pitch-to-diameter ratios exceeding 1.4. It is shown that in reactor

  6. Fusion reactor breeder material safety compatibility studies

    SciTech Connect (OSTI)

    Jeppson, D.W.; Cohen, S.; Muhlestein, L.D.

    1983-09-01T23:59:59.000Z

    Tritium breeder material selection for fusion reactors is strongly influenced by the desire to minimize safety and environmental concerns. Breeder material safety compatibility studies are being conducted to identify and characterize breeder-coolant-material interactions under postulated reactor accident conditions. Recently completed scoping compatibility tests indicate the following. 1. Ternary oxides (LiAlO/sub 2/, Li/sub 2/ZrO/sub 3/, Li/sub 2/SiO/sub 3/, Li/sub 4/SiO/sub 4/, and LiTiO/sub 3/) at postulated blanket operating temperatures are chemically compatible with water coolant, while liquid lithium and Li/sub 7/Pb/sub 2/ reactions with water generate heat, aerosol, and hydrogen. 2. Lithium oxide and 17Li-83Pb alloy react mildly with water requiring special precautions to control hydrogen release. 3. Liquid lithium reacts substantially, while 17Li83Pb alloy reacts mildly with concrete to produce hydrogen. 4. Liquid lithium-air reactions may present some major safety concerns. Additional scoping tests are needed, but the ternary oxides, lithium oxide, and 17Li-83Pb have definite safety advantages over liquid lithium and Li/sub 7/Pb/sub 2/. The ternary oxides present minimal safetyrelated problems when used with water as coolant, air or concrete; but they do require neutron multipliers, which may have safety compatibility concerns with surrounding materials. The combined favorable neutronics and minor safety compatibility concerns of lithium oxide and 17Li-83Pb make them prime candidates as breeder materials. Current safety efforts are directed toward assessing the compatibility of lithium oxide and the lithium-lead alloy with coolants and other materials.

  7. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    breeder reactors typically operate with an inner core of high fissile content surrounded by breeding blankets

  8. Evaluation of the parfait blanket concept for fast breeder reactors

    E-Print Network [OSTI]

    Ducat, Glenn Alexander

    1974-01-01T23:59:59.000Z

    An evaluation of the neutronic, thermal-hydraulic, mechanical and economic characteristics of fast breeder reactor configurations containing an internal blanket has been performed. This design, called the parfait blanket ...

  9. The economics of fuel depletion in fast breeder reactor blankets

    E-Print Network [OSTI]

    Brewer, Shelby Templeton

    1972-01-01T23:59:59.000Z

    A fast breeder reactor fuel depletion-economics model was developed and applied to a number of 1000 MWe UMBR case studies, involving radial blanket-radial reflector design, radial blanket fuel management, and sensitivity ...

  10. Molten Salt Breeder Reactors Academia Sinica, ITRI, NTHU

    E-Print Network [OSTI]

    Wang, Ming-Jye

    Molten Salt Breeder Reactors HX Team* Academia Sinica, ITRI, NTHU 6 April 2012 *F. H. Shu, M. J scenario for the future of world energy · Nuclear Reactors without Nuclear Reactions: ­ Thermal, Chemical ! 6 yr at 100% 600 yr at 100% tot 2,000 yr at 100% Moderator Water, slow n w absorption None, fast n

  11. Deployment Scenario of Heavy Water Cooled Thorium Breeder Reactor

    SciTech Connect (OSTI)

    Mardiansah, Deby; Takaki, Naoyuki [Course of Applied Science, School of Engineering, Tokai University (Japan)

    2010-06-22T23:59:59.000Z

    Deployment scenario of heavy water cooled thorium breeder reactor has been studied. We have assumed to use plutonium and thorium oxide fuel in water cooled reactor to produce {sup 233}U which will be used in thorium breeder reactor. The objective is to analysis the potential of water cooled Th-Pu reactor for replacing all of current LWRs especially in Japan. In this paper, the standard Pressurize Water Reactor (PWR) has been designed to produce 3423 MWt; (i) Th-Pu PWR, (ii) Th-Pu HWR (MFR = 1.0) and (iii) Th-Pu HWR (MFR 1.2). The properties and performance of the core were investigated by using cell and core calculation code. Th-Pu PWR or HWR produces {sup 233}U to introduce thorium breeder reactor. The result showed that to replace all (60 GWe) LWR by thorium breeder reactor within a period of one century, Th-Pu oxide fueled PWR has insufficient capability to produce necessary amount of {sup 233}U and Th-Pu oxide fueled HWR has almost enough potential to produce {sup 233}U but shows positive void reactivity coefficient.

  12. THE MATERIALS OF FAST BREEDER REACTORS

    E-Print Network [OSTI]

    Olander, Donald R.

    2013-01-01T23:59:59.000Z

    times larger in a fast reactor than in a thermal reactor,structural metals in a fast reactor will be subject to farof fuel ele- ments in fast reactors which are roughly one

  13. Nuclear breeder reactor fuel element with silicon carbide getter

    DOE Patents [OSTI]

    Christiansen, David W. (Kennewick, WA); Karnesky, Richard A. (Richland, WA)

    1987-01-01T23:59:59.000Z

    An improved cesium getter 28 is provided in a breeder reactor fuel element or pin in the form of an extended surface area, low density element formed in one embodiment as a helically wound foil 30 located with silicon carbide, and located at the upper end of the fertile material upper blanket 20.

  14. Light-water breeder reactors: preliminary safety and environmental information document. Volume III

    SciTech Connect (OSTI)

    Not Available

    1980-01-01T23:59:59.000Z

    Information is presented concerning prebreeder and breeder reactors based on light-water-breeder (LWBR) Type 1 modules; light-water backfit prebreeder supplying advanced breeder; light-water backfit prebreeder/seed-blanket breeder system; and light-water backfit low-gain converter using medium-enrichment uranium, supplying a light-water backfit high-gain converter.

  15. Feasibility study on the thorium fueled boiling water breeder reactor

    SciTech Connect (OSTI)

    PetrusTakaki, N. [Dept. of Applied Science, Tokai Univ., Kitakaname, Hiratsuka, Kanagawa 259-1292 (Japan)

    2012-07-01T23:59:59.000Z

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  16. Fuel Summary Report: Shippingport Light Water Breeder Reactor - Rev. 2

    SciTech Connect (OSTI)

    Olson, Gail Lynn; Mc Cardell, Richard Keith; Illum, Douglas Brent

    2002-09-01T23:59:59.000Z

    The Shippingport Light Water Breeder Reactor (LWBR) was developed by Bettis Atomic Power Laboratory to demonstrate the potential of a water-cooled, thorium oxide fuel cycle breeder reactor. The LWBR core operated from 1977-82 without major incident. The fuel and fuel components suffered minimal damage during operation, and the reactor testing was deemed successful. Extensive destructive and nondestructive postirradiation examinations confirmed that the fuel was in good condition with minimal amounts of cladding deformities and fuel pellet cracks. Fuel was placed in wet storage upon arrival at the Expended Core Facility, then dried and sent to the Idaho Nuclear Technology and Engineering Center for underground dry storage. It is likely that the fuel remains in good condition at its current underground dry storage location at the Idaho Nuclear Technology and Engineering Center. Reports show no indication of damage to the core associated with shipping, loading, or storage.

  17. Blanket management method for liquid metal fast breeder reactors

    SciTech Connect (OSTI)

    Carelli, M.D.

    1986-04-22T23:59:59.000Z

    The method is described of moving blanket assemblies during refueling in a heterogeneous-type core for a liquid-metal-cooled fast-breeder nuclear reactor to improve the performance thereof. The core consists of fissile-material-containing fuel assemblies and fertile-material-containing blanket assemblies, the blanket assemblies including a plurality of inner blanket assemblies positioned in predetermined different locations within the interior of the core and radial blanket assemblies positioned proximate the periphery of the core.

  18. Steam generator for liquid metal fast breeder reactor

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Garner, Daniel C. (Murrysville, PA); Wineman, Arthur L. (Greensburg, PA); Robey, Robert M. (North Huntingdon, PA)

    1985-01-01T23:59:59.000Z

    Improvements in the design of internal components of J-shaped steam generators for liquid metal fast breeder reactors. Complex design improvements have been made to the internals of J-shaped steam generators which improvements are intended to reduce tube vibration, tube jamming, flow problems in the upper portion of the steam generator, manufacturing complexities in tube spacer attachments, thermal stripping potentials and difficulties in the weld fabrication of certain components.

  19. Breeding nuclear fuels with accelerators: replacement for breeder reactors

    SciTech Connect (OSTI)

    Grand, P.; Takahashi, H.

    1984-01-01T23:59:59.000Z

    One application of high energy particle accelerators has been, and still is, the production of nuclear fuel for the nuclear energy industry; tantalizing because it would create a whole new industry. This approach to producing fissile from fertile material was first considered in the early 1950's in the context of the nuclear weapons program. A considerable development effort was expended before discovery of uranium ore in New Mexico put an end to the project. Later, US commitment to the Liquid Metal Fast Breeder Reactors (LMFBR) killed any further interest in pursuing accelerator breeder technology. Interest in the application of accelerators to breed nuclear fuels, and possibly burn nuclear wastes, revived in the late 1970's, when the LMFBR came under attack during the Carter administration. This period gave the opportunity to revisit the concept in view of the present state of the technology. This evaluation and the extensive calculational modeling of target designs that have been carried out are promising. In fact, a nuclear fuel cycle of Light Water Reactors and Accelerator Breeders is competitive to that of the LMFBR. At this time, however, the relative abundance of uranium reserves vs electricity demand and projected growth rate render this study purely academic. It will be for the next generation of accelerator builders to demonstate the competitiveness of this technology versus that of other nuclear fuel cycles, such as LMFBR's or Fusion Hybrid systems. 22 references, 1 figure, 5 tables.

  20. Experimental Breeder Reactor-II Primary Tank System Wash Water Workshop

    Broader source: Energy.gov [DOE]

    In 1994 Congress ordered the shutdown of the Experimental Breeder Reactor-II (EBR-II) and a closure project was initiated.

  1. Fuel Summary Report: Shippingport Light Water Breeder Reactor

    SciTech Connect (OSTI)

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

    1999-01-01T23:59:59.000Z

    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.

  2. Water chemistry of breeder reactor steam generators. [LMFBR

    SciTech Connect (OSTI)

    Simpson, J.L.; Robles, M.N.; Spalaris, C.N.; Moss, S.A.

    1980-08-01T23:59:59.000Z

    The water quality requirements will be described for breeder reactor steam generators, as well as specifications for balance of plant protection. Water chemistry details will be discussed for the following power plant conditions: feedwater and recirculation water at above and below 5% plant power, refueling or standby, makeup water, and wet layup. Experimental data will be presented from tests which included a departure from nucleate boiling experiment, the Few Tube Test, with a seven tube evaporator and three tube superheater, and a verification of control and on-line measurement of sodium ion in the ppB range. Sampling and instrumentation requirements to insure adherence to the specified water quality will be described. Evaporator cleaning criteria and data from laboratory testing of chemical cleaning solutions with emphasis on flow, chemical composition, and temperature will be discussed.

  3. EIS-0085-S: Liquid-Metal Fast Breeder Reactor Program, Supplemental

    Broader source: Energy.gov [DOE]

    The U.S. Department of Energy developed this supplemental statement to examine the reduced scope of the Liquid Metal Fast Breeder Reactor (LMFBR) program and the environmental impacts associated therewith, including a re-examination of the purpose, need and timing of the program, the present program structure, including reasonable program alternatives, and alternative electricity production technologies anticipated to be available within the same timeframe as the LMFBR technology option. This statement supplements ERDA-1535, Liquid Metal Fast Breeder Reactor Program.

  4. Decommissioning of Experimental Breeder Reactor - II Complex, Post Sodium Draining

    SciTech Connect (OSTI)

    J. A. (Bart) Michelbacher; S. Paul Henslee; Collin J. Knight; Steven R. sherman

    2005-09-01T23:59:59.000Z

    The Experimental Breeder Reactor - II (EBR-II) was shutdown in September 1994 as mandated by the United States Department of Energy. This sodium-cooled reactor had been in service since 1964. The bulk sodium was drained from the primary and secondary systems and processed. Residual sodium remaining in the systems after draining was converted into sodium bicarbonate using humid carbon dioxide. This technique was tested at Argonne National Laboratory in Illinois under controlled conditions, then demonstrated on a larger scale by treating residual sodium within the EBR-II secondary cooling system, followed by the primary tank. This process, terminated in 2002, was used to place a layer of sodium bicarbonate over all exposed surfaces of sodium. Treatment of the remaining EBR-II sodium is governed by the Resource Conservation and Recovery Act (RCRA). The Idaho Department of Environmental Quality issued a RCRA Operating Permit in 2002, mandating that all hazardous materials be removed from EBR-II within a 10 year period, with the ability to extend the permit and treatment period for another 10 years. A preliminary plan has been formulated to remove the remaining sodium and NaK from the primary and secondary systems using moist carbon dioxide, steam and nitrogen, and a water flush. The moist carbon dioxide treatment was resumed in May 2004. As of August 2005, approximately 60% of the residual sodium within the EBR-II primary tank had been treated. This process will continue through the end of 2005, when it is forecast that the process will become increasingly ineffective. At that time, subsequent treatment processes will be planned and initiated. It should be noted that the processes and anticipated costs associated with these processes are preliminary. Detailed engineering has not been performed, and approval for these methods has not been obtained from the regulator or the sponsors.

  5. Nuclear breeder reactor fuel element with axial tandem stacking and getter

    DOE Patents [OSTI]

    Gibby, Ronald L. (Richland, WA); Lawrence, Leo A. (Kennewick, WA); Woodley, Robert E. (Richland, WA); Wilson, Charles N. (Richland, WA); Weber, Edward T. (Kennewick, WA); Johnson, Carl E. (Elk Grove, IL)

    1981-01-01T23:59:59.000Z

    A breeder reactor fuel element having a tandem arrangement of fissile and fertile fuel with a getter for fission product cesium disposed between the fissile and fertile sections. The getter is effective at reactor operating temperatures to isolate the cesium generated by the fissile material from reacting with the fertile fuel section.

  6. Corrosion-resistant fuel cladding allow for liquid metal fast breeder reactors

    DOE Patents [OSTI]

    Brehm, Jr., William F. (Richland, WA); Colburn, Richard P. (Pasco, WA)

    1982-01-01T23:59:59.000Z

    An aluminide coating for a fuel cladding tube for LMFBRs (liquid metal fast breeder reactors) such as those using liquid sodium as a heat transfer agent. The coating comprises a mixture of nickel-aluminum intermetallic phases and presents good corrosion resistance to liquid sodium at temperatures up to 700.degree. C. while additionally presenting a barrier to outward diffusion of .sup.54 Mn.

  7. Atoms in Appalachia. Historical report on the Clinch River Breeder Reactor site

    SciTech Connect (OSTI)

    Schaffer, D

    1982-01-01T23:59:59.000Z

    The background information concerning the acquisition of the land for siting the Clinch River Breeder Reactor is presented. Historical information is also presented concerning the land acquisition for the Oak Ridge facilities known as the Manhattan Project during World War II.

  8. Feasibility of Water Cooled Thorium Breeder Reactor Based on LWR Technology

    SciTech Connect (OSTI)

    Takaki, Naoyuki; Permana, Sidik; Sekimoto, Hiroshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology 2-12-1, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

    2007-07-01T23:59:59.000Z

    The feasibility of Th-{sup 233}U fueled, homogenous breeder reactor based on matured conventional LWR technology was studied. The famous demonstration at Shipping-port showed that the Th-{sup 233}U fueled, heterogeneous PWR with four different lattice fuels was possible to breed fissile but its low averaged burn-up including blanket fuel and the complicated core configuration were not suitable for economically competitive reactor. The authors investigated the wide design range in terms of fuel cell design, power density, averaged discharge burn-up, etc. to determine the potential of water-cooled Th reactor as a competitive breeder. It is found that a low moderated (MFR=0.3) H{sub 2}O-cooled reactor with comparable burn-up with current LWR is feasible to breed fissile fuel but the core size is too large to be economical because of the low pellet power density. On the other hand, D{sub 2}O-cooled reactor shows relatively wider feasible design window, therefore it is possible to design a core having better neutronic and economic performance than H{sub 2}O-cooled. Both coolant-type cores show negative void reactivity coefficient while achieving breeding capability which is a distinguished characteristics of thorium based fuel breeder reactor. (authors)

  9. Ssessment methodology for proliferation resistant fast breeder reactor

    E-Print Network [OSTI]

    Singh, Mohit, S.M. Massachusetts Institute of Technology

    2014-01-01T23:59:59.000Z

    Due to perceived proliferation risks, current US fast reactor designs have avoided the use of uranium blankets. While reducing the amount of plutonium produced, this omission also restrains the reactor design space and has ...

  10. Status of EC solid breeder blanket designs and R&D for DEMO fusion reactors

    SciTech Connect (OSTI)

    Dalle Donne, M. [INR, Karlsruhe (Russian Federation); Anziedi, L.A. [C.R.E., Franscati (Italy); Kwast, H. [ECN, Petten (Netherlands)] [and others

    1994-12-31T23:59:59.000Z

    In the framework of the European Community Fusion Technology Program four blanket concepts for a DEMO reactor are being investigated. DEMO is the next step after ITER. It should ensure tritium self-sufficiency and operate at coolant temperatures high enough to have a reasonable plant efficiency. Further requirements have been specified for the four concepts, namely an average neutron wall load of 2.2 MW/m{sup 2}, a blanket lifetime of 20000 hours and the capability of the blanket segment to withstand the forces caused by a rapid distribution of the plasma current (20 MA to zero in 20 ms), so that after the disruption the segment can still allow a comparison of the various options, in view of reducing this number to two in 1995 and to design and develop modules and articles representative of the chosen blankets to be tested in ITER. The present paper deals with two solid breeder concepts. They have many features in common: both use high pressure helium as coolant and helium to purge the tritium from the breeder material, martensitic steel as structural material and beryllium as neutron multiplier. The configuration of the two blankets are however different: in the B.I.T. (Breeder Inside Tube) concept the breeder material is LiAlO{sub 2} or LiZrO{sub 3} in the form of annular pellets contained in tubes surrounded by beryllium blocks, the coolant helium being outside the tubes, whereas in the B.O.T. (Breeder out of Tube) the breeder and multiplier material are Li{sub 4}SiO{sub 4} and beryllium pebbles forming a mixed bed placed outside the tubes containing the coolant helium.

  11. Trace Fission Product Ratios for Nuclear Forensics Attribution of Weapons-Grade Plutonium from Fast Breeder Reactor Blankets

    E-Print Network [OSTI]

    Osborn, Jeremy

    2014-08-13T23:59:59.000Z

    ), whereas that used in an FBR blanket fuel is depleted uranium (0.25 atom percent 235U). The energy production in the FBR core is from the seed fuel subassemblies containing mixed oxides (MOX) of PuO2 and UO2. A plot of fast and thermal neutron energy... of the program involves a fleet of fast breeder reactors. The stage two fast breeder reactors, beginning with the PFBR, will be fueled with reactor-grade plutonium and depleted uranium from the reprocessed spent fuel of stage one reactors and will breed more...

  12. Ceramic breeder blanket development for fusion experimental reactor in JAERI

    SciTech Connect (OSTI)

    Kurasawa, T.; Takatsu, H.; Sato, S. [JAERI, Ibaraki-ken (Japan)] [and others

    1994-12-31T23:59:59.000Z

    Ceramic breeding blanket is a promising breeding blanket concept for the fusion experimental reactor, and world-wide efforts have been devoted to the design and R&D. Irradiation damages of both of breeding materials and neutron multipliers are one of the critical issues for this type of blanket, and usage of these materials as a form of small pebbles has been proposed so as to accommodate expected irradiation damages without degradation of breeding capability. The present paper outlines the progress of the design of layered pebble bed breeding blanket and also shows preliminary results of concept development related to higher fusion power accommodation and convertible blanket.

  13. Special topics reports for the reference tandem mirror fusion breeder. Volume 2. Reactor safety assessment

    SciTech Connect (OSTI)

    Maya, I.; Hoot, C.G.; Wong, C.P.C.; Schultz, K.R.; Garner, J.K.; Bradbury, S.J.; Steele, W.G.; Berwald, D.H.

    1984-09-01T23:59:59.000Z

    The safety features of the reference fission suppressed fusion breeder reactor are presented. These include redundancy and overcapacity in primary coolant system components to minimize failure probability, an improved valve location logic to provide for failed component isolation, and double-walled coolant piping and steel guard vessel protection to further limit the extent of any leak. In addition to the primary coolant and decay heat removal system, reactor safety systems also include an independent shield cooling system, the module safety/fuel transfer coolant system, an auxiliary first wall cooling system, a psssive dump tank cooling system based on the use of heat pipes, and several lithium fire suppression systems. Safety system specifications are justified based on the results of thermal analysis, event tree construction, consequence calculations, and risk analysis. The result is a reactor design concept with an acceptably low probability of a major radioactivity release. Dose consequences of maximum credible accidents appear to be below 10CFR100 regulatory limits.

  14. Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 11691181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M"

    E-Print Network [OSTI]

    Laughlin, Robert B.

    Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 39, No. 11, p. 1169­1181 (November 2002) Conceptual Design of a Modular Island Core Fast Breeder Reactor "RAPID-M" Mitsuru KAMBE Central Research Institute and accepted September 10, 2002) A metal fueled modular island core sodium cooled fast breeder reactor concept

  15. Review of uncertainty estimates associated with models for assessing the impact of breeder reactor radioactivity releases

    SciTech Connect (OSTI)

    Miller, C.; Little, C.A.

    1982-08-01T23:59:59.000Z

    The purpose is to summarize estimates based on currently available data of the uncertainty associated with radiological assessment models. The models being examined herein are those recommended previously for use in breeder reactor assessments. Uncertainty estimates are presented for models of atmospheric and hydrologic transport, terrestrial and aquatic food-chain bioaccumulation, and internal and external dosimetry. Both long-term and short-term release conditions are discussed. The uncertainty estimates presented in this report indicate that, for many sites, generic models and representative parameter values may be used to calculate doses from annual average radionuclide releases when these calculated doses are on the order of one-tenth or less of a relevant dose limit. For short-term, accidental releases, especially those from breeder reactors located in sites dominated by complex terrain and/or coastal meteorology, the uncertainty in the dose calculations may be much larger than an order of magnitude. As a result, it may be necessary to incorporate site-specific information into the dose calculation under these circumstances to reduce this uncertainty. However, even using site-specific information, natural variability and the uncertainties in the dose conversion factor will likely result in an overall uncertainty of greater than an order of magnitude for predictions of dose or concentration in environmental media following shortterm releases.

  16. Theory, design, and operation of liquid metal fast breeder reactors, including operational health physics

    SciTech Connect (OSTI)

    Adams, S.R.

    1985-10-01T23:59:59.000Z

    A comprehensive evaluation was conducted of the radiation protection practices and programs at prototype LMFBRs with long operational experience. Installations evaluated were the Fast Flux Test Facility (FFTF), Richland, Washington; Experimental Breeder Reactor II (EBR-II), Idaho Falls, Idaho; Prototype Fast Reactor (PFR) Dounreay, Scotland; Phenix, Marcoule, France; and Kompakte Natriumgekuhlte Kernreak Toranlange (KNK II), Karlsruhe, Federal Republic of Germany. The evaluation included external and internal exposure control, respiratory protection procedures, radiation surveillance practices, radioactive waste management, and engineering controls for confining radiation contamination. The theory, design, and operating experience at LMFBRs is described. Aspects of LMFBR health physics different from the LWR experience in the United States are identified. Suggestions are made for modifications to the NRC Standard Review Plan based on the differences.

  17. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    SciTech Connect (OSTI)

    Budd, W.A. (ed.)

    1986-03-01T23:59:59.000Z

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  18. Frequency response testing at Experimental Breeder Reactor II using discrete-level periodic signals

    SciTech Connect (OSTI)

    Rhodes, W.D.; Larson, H.A. (Idaho State Univ., Pocatello, ID (USA). Coll. of Engineering); Dean, E.M. (Argonne National Lab., Idaho Falls, ID (USA))

    1990-01-01T23:59:59.000Z

    The Experimental Breeder Reactor 2 (EBR-2) reactivity-to-power frequency-response function was measured with pseudo-random, discrete-level, periodic signals. The reactor power deviation was small with insignificant perturbation of normal operation and in-place irradiation experiments. Comparison of results with measured rod oscillator data and with theoretical predictions show good agreement. Moreover, measures of input signal quality (autocorrelation function and energy spectra) confirm the ability to enable this type of frequency response determination at EBR-2. Measurements were made with the pseudo-random binary sequence, quadratic residue binary sequence, pseudo-random ternary sequence, and the multifrequency binary sequence. 10 refs., 7 figs., 3 tabs.

  19. Argonne's rich scientific heritage Argonne's Experimental Breeder Reactor-I in Idaho lit this string of four

    E-Print Network [OSTI]

    Kemner, Ken

    Argonne's rich scientific heritage Argonne's Experimental Breeder Reactor-I in Idaho lit was December 20, 1951. Argonne National Laboratory is a U.S. Department of Energy laboratory managed by UChicago Argonne, LLC October 2010Argonne National Laboratory, 9700 South Cass Avenue, Lemont, IL 60439

  20. Tritium trapping in silicon carbide in contact with solid breeder under high flux isotope reactor irradiation

    SciTech Connect (OSTI)

    H. Katsui; Y. Katoh; A. Hasegawa; M. Shimada; Y. Hatano; T. Hinoki; S. Nogami; T. Tanaka; S. Nagata; T. Shikama

    2013-11-01T23:59:59.000Z

    The trapping of tritium in silicon carbide (SiC) injected from ceramic breeding materials was examined via tritium measurements using imaging plate (IP) techniques. Monolithic SiC in contact with ternary lithium oxide (lithium titanate and lithium aluminate) as a ceramic breeder was irradiated in the High Flux Isotope Reactor (HFIR) in Oak Ridge, Tennessee, USA. The distribution of photo-stimulated luminescence (PSL) of tritium in SiC was successfully obtained, which separated the contribution of 14C ß-rays to the PSL. The tritium incident from ceramic breeders was retained in the vicinity of the SiC surface even after irradiation at 1073 K over the duration of ~3000 h, while trapping of tritium was not observed in the bulk region. The PSL intensity near the SiC surface in contact with lithium titanate was higher than that obtained with lithium aluminate. The amount of the incident tritium and/or the formation of a Li2SiO3 phase on SiC due to the reaction with lithium aluminate under irradiation likely were responsible for this observation.

  1. Clinch River Breeder Reactor: an assessment of need for power and regulatory issues

    SciTech Connect (OSTI)

    Hamblin, D M; Tepel, R C; Bjornstad, D J; Hill, L J; Cantor, R A; Carroll, P J; Cohn, S M; Hadder, G R; Holcomb, B D; Johnson, K E

    1983-09-01T23:59:59.000Z

    The purpose of this report is to present the results of a research effort designed to assist the US Department of Energy in: (1) reviewing the need for power from the Clinch River Breeder Reactor (CRBR) in the Southeastern Electric Reliability Council (SERC) region, not including Florida, and (2) isolating specific regulatory and institutional issues and physical transmission capacities that may constrain the market for CRBR power. A review of existing electric power wheeling arrangements in the Southeast and specific federal and state regulatory obstacles that may affect power sales from the CRBR was undertaken. This review was a contributing factor to a decision to target the service territory to SERC-less Florida.

  2. Light Water Breeder Reactor fuel rod design and performance characteristics (LWBR Development Program)

    SciTech Connect (OSTI)

    Campbell, W.R.; Giovengo, J.F.

    1987-10-01T23:59:59.000Z

    Light Water Breeder Reactor (LWBR) fuel rods were designed to provide a reliable fuel system utilizing thorium/uranium-233 mixed-oxide fuel while simultaneously minimizing structural material to enhance fuel breeding. The fuel system was designed to be capable of operating successfully under both load follow and base load conditions. The breeding objective required thin-walled, low hafnium content Zircaloy cladding, tightly spaced fuel rods with a minimum number of support grid levels, and movable fuel rod bundles to supplant control rods. Specific fuel rod design considerations and their effects on performance capability are described. Successful completion of power operations to over 160 percent of design lifetime including over 200 daily load follow cycles has proven the performance capability of the fuel system. 68 refs., 19 figs., 44 tabs.

  3. Coated ceramic breeder materials

    DOE Patents [OSTI]

    Tam, Shiu-Wing (Downers Grove, IL); Johnson, Carl E. (Elk Grove, IL)

    1987-01-01T23:59:59.000Z

    A breeder material for use in a breeder blanket of a nuclear reactor is disclosed. The breeder material comprises a core material of lithium containing ceramic particles which has been coated with a neutron multiplier such as Be or BeO, which coating has a higher thermal conductivity than the core material.

  4. Studies on Plutonium Burning in the Prototype Fast Breeder Reactor Monju

    SciTech Connect (OSTI)

    Wehmann, Udo [Japan Nuclear Cycle Development Center (Japan); Kinjo, Hidehito [Japan Nuclear Cycle Development Center (Japan); Kageyama, Takeshi [Nuclear Energy System, Inc. (Japan)

    2002-03-15T23:59:59.000Z

    Studies have been performed on plutonium burning in the Japanese prototype fast breeder reactor Monju. The main aims of these studies were to illustrate the plutonium-burning capabilities of fast reactors and to investigate the consequences of the related core design measures on the main core characteristics of Monju. Burner cores with diluting pins, with diluting subassemblies (also called diluents), and with an internal slice of inert material have been investigated; these require an increased average plutonium enrichment and thus offer an enhanced plutonium-burning rate. On the other hand, the consequences of the elimination of the radial and/or axial blanket have been investigated.Among the burner concepts, the B{sub 4}C containing diluents have been found to be preferable because they cause the smallest maximum linear rating increase and offer the largest flexibility to adapt their reactivity via a modification of the B{sub 4}C content. They also do not require a new fuel subassembly concept.For the case of the blanket elimination, the replacement of the blankets by steel reflectors has been found to be the best solution. The main consequence of the elimination of both blankets is the increase of the maximum linear rating by up to 11%. Whether this increase may lead to problems will depend on the actual linear power level of the core.

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

    SciTech Connect (OSTI)

    Louison, R.; Boardman, C.E.

    1981-05-29T23:59:59.000Z

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

  6. Analysis of the conceptual shielding design for the upflow Gas-Cooled Fast Breeder Reactor

    SciTech Connect (OSTI)

    Slater, C.O.; Reed, D.A.; Cramer, S.N.; Emmett, M.B.; Tomlinson, E.T.

    1981-01-01T23:59:59.000Z

    Conceptual Shielding Configuration III for the Gas-Cooled Fast Breeder Reactor (GCFR) was analyzed by performing global calculations of neutron and gamma-ray fluences and correcting the results as appropriate with bias factors from localized calculations. Included among the localized calculations were the radial and axial cell streaming calculations, plus extensive preliminary calculations and three final confirmation calculations of the plenum flow-through shields. The global calculations were performed on the GCFR mid-level and the lower and upper plenum regions. Calculated activities were examined with respect to the design constraint, if any, imposed on the particular activity. The spatial distributions of several activities of interest were examined with the aid of isoplots (i.e., symbols are used to describe a surface on which the activity level is everywhere the same). In general the results showed that most activities were below the respective design constraints. Only the total neutron fluence in the core barrel appeared to be marginal with the present reactor design. Since similar results were obtained for an earlier design, it has been proposed that the core barrel be cooled with inlet plenum gas to maintain it at a temperature low enough that it can withstand a higher fluence limit. Radiation levels in the prestressed concrete reactor vessel (PCRV) and liner appeared to be sufficiently below the design constraint that expected results from the Radial Shield Heterogeneity Experiment should not force any levels above the design constraint. A list was also made of a number of issues which should be examined before completion of the final shielding design.

  7. Data management for the Clinch River Breeder Reactor Plant Project by use of document status and hold systems

    SciTech Connect (OSTI)

    Hunt, C S; Beck, A E; Akhtar, M S

    1982-01-01T23:59:59.000Z

    This paper describes the development, framework, and scope of the Document Status System and the Document Hold System for the Clinch River Breeder Reactor Plant Project. It shows how data are generated at five locations and transmitted to a central computer for processing and storage. The resulting computerized data bank provides reports needed to perform day-to-day management and engineering planning. Those reports also partially satisfy the requirements of the Project's Quality Assurance Program.

  8. Final report for the Light Water Breeder Reactor proof-of-breeding analytical support project

    SciTech Connect (OSTI)

    Graczyk, D.G.; Hoh, J.C.; Martino, F.J.; Nelson, R.E.; Osudar, J.; Levitz, N.M.

    1987-05-01T23:59:59.000Z

    The technology of breeding /sup 233/U from /sup 232/Th in a light water reactor is being developed and evaluated by the Westinghouse Bettis Atomic Power Laboratory (BAPL) through operation and examination of the Shippingport Light Water Breeder Reactor (LWBR). Bettis is determining the end-of-life (EOL) inventory of fissile uranium in the LWBR core by nondestructive assay of a statistical sample comprising approximately 500 EOL fuel rods. This determination is being made with an irradiated-fuel assay gauge based on neutron interrogation and detection of delayed neutrons from each rod. The EOL fissile inventory will be compared with the beginning-of-life fissile loading of the LWBR to determine the extent of breeding. In support of the BAPL proof-of-breeding (POB) effort, Argonne National Laboratory (ANL) carried out destructive physical, chemical, and radiometric analyses on 17 EOL LWBR fuel rods that were previously assayed with the nondestructive gauge. The ANL work included measurements on the intact rods; shearing of the rods into pre-designated contiguous segments; separate dissolution of each of the more than 150 segments; and analysis of the dissolver solutions to determine each segment's uranium content, uranium isotopic composition, and loading of selected fission products. This report describes the facilities in which this work was carried out, details operations involved in processing each rod, and presents a comprehensive discussion of uncertainties associated with each result of the ANL measurements. Most operations were carried out remotely in shielded cells. Automated equipment and procedures, controlled by a computer system, provided error-free data acquisition and processing, as well as full replication of operations with each rod. Despite difficulties that arose during processing of a few rod segments, the ANL destructive-assay results satisfied the demanding needs of the parent LWBR-POB program.

  9. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    OF LARGE FAST REACTORS Calculation examples A typicalMonte Carlo Reactor Physics Burnup Calculation Code. Tech.reactor core design from experience and coarse calculations

  10. Estimation of measured control rod worths in Fast Breeder Test Reactor -- Effect of different delayed neutron parameters

    SciTech Connect (OSTI)

    Mohanakrishnan, P.; Reddy, C.P.; Gopalakrishnan, V.; Arul, J. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India). Reactor Physics Div.

    1996-03-01T23:59:59.000Z

    Control rod worths have been measured by the inverse kinetics method in the small PuC-UC core of the Fast Breeder Test Reactor at Kalpakkam. Delayed neutron fractional yields based on Tuttle`s data, ENDF/B-VI data, and the full summation approach of Brady and England have been used to get measured control rod worths. Unreasonably large reductions in control rod worths are obtained by the ENDF/B-VI data. It is suspected that the procedure, of normalizing fractional yields obtained by the summation approach to earlier evaluated total yields, is inconsistent.

  11. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    SciTech Connect (OSTI)

    Not Available

    1980-05-01T23:59:59.000Z

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  12. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    fission gas plenum212 Conventional fast reactor core designGUPTA. “A Compact Gas-Cooled Fast Reactor with an Ultra-Longbreed and burn gas-cooled fast reactor”. Ph.D. Thesis. MIT,

  13. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    to large conventional sodium fast reactors (SFR). TerraPowerincrease for a typical sodium fast reactor fuel rod geometryof the new Russian sodium fast reactor BN-800 [111]. The

  14. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    AND-BURN REACTOR PHYSICS wave burnup principle. The CANDLEand physical principle Breed-and-burn reactors (B&B) areBURN REACTOR PHYSICS The FIMA burnup unit - principles and

  15. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    4 Reactivity feedback of large fast reactors 4.1temperature . . . . . . . . . . . . . . . . . . Fast reactorfission gas plenum212 Conventional fast reactor core design

  16. Gas-cooled fast breeder reactor. Quarterly progress report, November 1, 1979 through January 31, 1980

    SciTech Connect (OSTI)

    Not Available

    1980-02-01T23:59:59.000Z

    Information is presented concerning the nuclear steam supply system; reactor core; systems engineering; safety and reliability; and circulator test facility.

  17. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    type fast reactor of the IV generation for regional powerELECTRA-FCC: a centre for Generation IV system research andunder the framework of generation-IV nuclear pro- grams or

  18. Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor

    SciTech Connect (OSTI)

    Guidez, J.; Martin, L. [Commissariat a l'Energie Atomique - CEA (France); Dupraz, R. [AREVA NP (France)

    2006-07-01T23:59:59.000Z

    The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960's, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 deg. C, Inspection of the conical shell supporting the core dia-grid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remote inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960's, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant. (authors)

  19. Gas-Cooled Fast Breeder Reactor Preliminary Safety Information Document, Amendment 10. GCFR residual heat removal system criteria, design, and performance

    SciTech Connect (OSTI)

    Not Available

    1980-09-01T23:59:59.000Z

    This report presents a comprehensive set of safety design bases to support the conceptual design of the gas-cooled fast breeder reactor (GCFR) residual heat removal (RHR) systems. The report is structured to enable the Nuclear Regulatory Commission (NRC) to review and comment in the licensability of these design bases. This report also presents information concerning a specific plant design and its performance as an auxiliary part to assist the NRC in evaluating the safety design bases.

  20. Evaluation of the Initial Isothermal Physics Measurements at the Fast Flux Test Facility, a Prototypic Liquid Metal Fast Breeder Reactor

    SciTech Connect (OSTI)

    John D. Bess

    2010-03-01T23:59:59.000Z

    The Fast Flux Test Facility (FFTF) was a 400-MWt, sodium-cooled, low-pressure, high-temperature, fast-neutron flux, nuclear fission reactor plant designed for the irradiation testing of nuclear reactor fuels and materials for the development of liquid metal fast breeder reactors (LMFBRs). The FFTF was fueled with plutonium-uranium mixed oxide (MOX) and reflected by Inconel-600. Westinghouse Hanford Company operated the FFTF as part of the Hanford Engineering Development Laboratory (HEDL) for the U.S. Department of Energy on the Hanford Site near Richland, Washington. Although the FFTF was a testing facility not specifically designed to breed fuel or produce electricity, it did provide valuable information for LMFBR projects and base technology programs in the areas of plant system and component design, component fabrication, prototype testing, and site construction. The major objectives of the FFTF were to provide a strong, disciplined engineering base for the LMFBR program, provide fast flux testing for other U.S. programs, and contribute to the development of a viable self-sustaining competitive U.S. LMFBR industry. During its ten years of operation, the FFTF acted as a national research facility to test advanced nuclear fuels, materials, components, systems, nuclear power plant operating and maintenance procedures, and active and passive reactor safety technologies; it also produced a large number of isotopes for medical and industrial users, generated tritium for the U.S. fusion research program, and participated in cooperative, international research work. Prior to the implementation of the reactor characterization program, a series of isothermal physics measurements were performed; this acceptance testing program consisted of a series of control rod worths, critical rod positions, subcriticality measurements, maximum reactivity addition rates, shutdown margins, excess reactivity, and isothermal temperature coefficient reactivity. The results of these experiments were of particular importance because they provide extensive information which can be directly applied to the design of large LMFBR’s. It should be recognized that the data presented in the initial report were evaluated only to the extent necessary to ensure that adequate data were obtained. Later reports provided further interpretation and detailed comparisons with prediction techniques. The conclusion of the isothermal physics measurements was that the FFTF nuclear characteristics were essentially as designed and all safety requirements were satisfied. From a nuclear point of view, the FFTF was qualified to proceed into power operation mode. The FFTF was completed in 1978 and first achieved criticality on February 9, 1980. Upon completion of the isothermal physics and reactor characterization programs, the FFTF operated for ten years from April 1982 to April 1992. Reactor operations of the FFTF were terminated and the reactor facility was then defueled, deactivated, and placed into cold standby condition. Deactivation of the reactor was put on hold from 1996 to 2000 while the U.S. Department of Energy examined alternative uses for the FFTF but then announced the permanent deactivation of the FFTF in December 2001. Its core support basket was later drilled in May 2005, so as to remove all remaining sodium coolant. On April 17, 2006, the American Nuclear Society designated the FFTF as a “National Nuclear Historic Landmark”.

  1. A parametric study of the breeding ratio in sodium cooled fast breeder reactors

    E-Print Network [OSTI]

    Sobey, Thomas Milburn

    1969-01-01T23:59:59.000Z

    &et p ndu t ~. cn of glutnniu D fission of p!utonium ) reactor + F + L ? F ) rea. tol ''9 o 1 '2 Fbi s ef ni tie. . ro& elves 1't s "?. ' be Use a c "n, s ca 1n ~ lt d-:t n i. . dig fez en! 9 u-onion i ='' topee. 31 Fool Breedin Ratio 1...

  2. Worths and interactions of simulated fast breeder reactor control rods in ZEBRA and SNEAK assemblies

    SciTech Connect (OSTI)

    Giese, H.; Pilate, S.; Stevenson, J.M.

    1984-07-01T23:59:59.000Z

    Measurements of the worths of simulated control rods for fast power reactors have been made in the ZEBRA and SNEAK critical assemblies by the modified source multiplication method (MSMM). The assemblies used were the conventional and unconventional core arrangements from the BIZET program and a compacted version of a conventional core. The control rods were mainly natural B/sub 4/C, with some study of 40% /sup 10/B-enriched B/sub 4/C and of Eu/sub 2/O/sub 3/. Correction factors for the MSMM were obtained from eigenvalue and source-mode diffusion theory calculations in XY geometry. The measured rod worths and interactions are compared with calculated values from methods and data similar to those used by the different participants in the BIZET program to predict the corresponding parameters in fast power reactors. In general, acceptable agreement is found.

  3. Automated operator procedure prompting for startup of Experimental Breeder Reactor-2

    SciTech Connect (OSTI)

    Renshaw, A.W.; Ball, S.J.; Ford, C.E.

    1990-11-01T23:59:59.000Z

    This report describes the development of an operator procedure prompting aid for startup of a nuclear reactor. This operator aid is a preliminary design for a similar aid that eventually will be used with the Advanced Liquid Metal Reactor (ALMR) presently in the design stage. Two approaches were used to develop this operator procedure prompting aid. One method uses an expert system software shell, and the other method uses database software. The preliminary requirements strongly pointed toward features traditionally associated with both database and expert systems software. Database software usually provides data manipulation flexibility and user interface tools, and expert systems tools offer sophisticated data representation and reasoning capabilities. Both methods, including software and associated hardware, are described in this report. Proposals for future enhancements to improve the expert system approach to procedure prompting and for developing other operator aids are also offered. 25 refs., 14 figs.

  4. Optimization of a heterogeneous fast breeder reactor core with improved behavior during unprotected transients

    SciTech Connect (OSTI)

    Poumerouly, S.; Schmitt, D.; Massara, S.; Maliverney, B. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)

    2012-07-01T23:59:59.000Z

    Innovative Sodium-cooled Fast Reactors (SFRs) are currently being investigated by CEA, AREVA and EDF in the framework of a joint French collaboration, and the construction of a GEN IV prototype, ASTRID (Advanced Sodium Technical Reactor for Industrial Demonstration), is scheduled in the years 2020. Significant improvements are expected so as to improve the reactor safety: the goal is to achieve a robust safety demonstration of the mastering of the consequences of a Core Disruptive Accident (CDA), whether by means of prevention or mitigation features. In this framework, an innovative design was proposed by CEA in 2010. It aims at strongly reducing the sodium void effect, thereby improving the core behavior during unprotected loss of coolant transients. This design is strongly heterogeneous and includes, amongst others, a fertile plate, a sodium plenum associated with a B{sub 4}C upper blanket and a stepwise modulation of the fissile height of the core (onwards referred to as the 'diabolo shape'). In this paper, studies which were entirely carried out at EDF are presented: the full potential of this heterogeneous concept is thoroughly investigated using the SDDS methodology. (authors)

  5. Microstructure analysis for chemical interaction between cesium and SUS 316 steel in fast breeder reactor application

    SciTech Connect (OSTI)

    Sasaki, K.; Fukumoto, K. I.; Oshima, T.; Tanigaki, T.; Masayoshi, U. [RINE, Univ. of Fukui, 3-9-1, Bunkyo, Fukui, 910-8507 (Japan)

    2012-07-01T23:59:59.000Z

    In this study the corrosion products on a surface after cesium corrosion examination at 650 deg. C for 100 hrs were characterized by TEM observation around the corroded area on the surface in order to understand the corrosion mechanism of cesium fission product for cladding materials in fast reactor. The experimental results suggest the main corrosion mechanism occurred in the process of the separation of cesium chromate and metal (Fe, Ni). The main reaction of corrosion process was considered to be equation, 2Cs + 7/2 O{sub 2} + 2Cr {yields} Cs{sub 2}Cr{sub 2}O{sub 7}(L). (authors)

  6. Core loading pattern optimization of thorium fueled heavy water breeder reactor using genetic algorithm

    SciTech Connect (OSTI)

    Soewono, C. N.; Takaki, N. [Dept. of Applied Science Engineering, Faculty Tokai Univ., Kanagawa-ken, Hiratsuka-shi Kitakaname 4-1-1 (Japan)

    2012-07-01T23:59:59.000Z

    In this work genetic algorithm was proposed to solve fuel loading pattern optimization problem in thorium fueled heavy water reactor. The objective function of optimization was to maximize the conversion ratio and minimize power peaking factor. Those objectives were simultaneously optimized using non-dominated Pareto-based population ranking optimal method. Members of non-dominated population were assigned selection probabilities based on their rankings in a manner similar to Baker's single criterion ranking selection procedure. A selected non-dominated member was bred through simple mutation or one-point crossover process to produce a new member. The genetic algorithm program was developed in FORTRAN 90 while neutronic calculation and analysis was done by COREBN code, a module of core burn-up calculation for SRAC. (authors)

  7. Cost/performance comparison between pulse columns and centrifugal contactors designed to process Clinch River Breeder Reactor fuel

    SciTech Connect (OSTI)

    Ciucci, J.A. Jr.

    1983-12-01T23:59:59.000Z

    A comparison between pulse columns and centrifugal contactors was made to determine which type of equipment was more advantageous for use in the primary decontamination cycle of a remotely operated fuel reprocessing plant. Clinch River Breeder Reactor (CRBR) fuel was chosen as the fuel to be processed in the proposed 1 metric tonne/day reprocessing facility. The pulse columns and centrifugal contactors were compared on a performance and total cost basis. From this comparison, either the pulse columns or the centrifugal contactors will be recommended for use in a fuel reprocessing plant built to reprocess CRBR fuel. The reliability, solvent exposure to radiation, required time to reach steady state, and the total costs were the primary areas of concern for the comparison. The pulse column units were determined to be more reliable than the centrifugal contactors. When a centrifugal contactor motor fails, it can be remotely changed in less than one eight hour shift. Pulse columns expose the solvent to approximately five times as much radiation dose as the centrifugal contactor units; however, the proposed solvent recovery system adequately cleans the solvent for either case. The time required for pulse columns to reach steady state is many times longer than the time required for centrifugal contactors to reach steady state. The cost comparison between the two types of contacting equipment resulted in centrifugal contactors costing 85% of the total cost of pulse columns when the contactors were stacked on three levels in the module. If the centrifugal contactors were all positioned on the top level of a module with the unoccupied volume in the module occupied by other equipment, the centrifugal contactors cost is 66% of the total cost of pulse columns. Based on these results, centrifugal contactors are recommended for use in a remotely operated reprocessing plant built to reprocess CRBR fuel.

  8. Validation Work to Support the Idaho National Engineering and Environmental Laboratory Calculational Burnup Methodology Using Shippingport Light Water Breeder Reactor (LWBR) Spent Fuel Assay Data

    SciTech Connect (OSTI)

    J. W. Sterbentz

    1999-08-01T23:59:59.000Z

    Six uranium isotopes and fourteen fission product isotopes were calculated on a mass basis at end-of-life (EOL) conditions for three fuel rods from different Light Water Breeder Reactor (LWBR) measurements. The three fuel rods evaluated here were taken from an LWBR seed module, a standard blanket module, and a reflector (Type IV) module. The calculated results were derived using a depletion methodology previously employed to evaluate many of the radionuclide inventories for spent nuclear fuels at the Idaho National Engineering and Environmental Laboratory. The primary goal of the calculational task was to further support the validation of this particular calculational methodology and its application to diverse reactor types and fuels. Result comparisons between the calculated and measured mass concentrations in the three rods indicate good agreement for the three major uranium isotopes (U-233, U-234, U-235) with differences of less than 20%. For the seed and standard blanket rod, the U-233 and U-234 differences were within 5% of the measured values (these two isotopes alone represent greater than 97% of the EOL total uranium mass). For the major krypton and xenon fission product isotopes, differences of less than 20% and less than 30% were observed, respectively. In general, good agreement was obtained for nearly all the measured isotopes. For these isotopes exhibiting significant differences, possible explanations are discussed in terms of measurement uncertainty, complex transmutations, etc.

  9. Summary of the radiological assessment of the fuel cycle for a thorium-uranium carbide-fueled fast breeder reactor

    SciTech Connect (OSTI)

    Tennery, V.J.; Bomar, E.S.; Bond, W.D.; Meyer, H.R.; Morse, L.E.; Till, J.E.; Yalcintas, M.G.

    1980-01-01T23:59:59.000Z

    A large fraction of the potential fuel for nuclear power reactors employing fissionable materials exists as ores of thorium. In addition, certain characteristics of a fuel system based on breeding of the fissionable isotope {sup 233}U from thorium offer the possibility of a greater resistance to the diversion of fissionable material for the fabrication of nuclear weapons. This report consolidates into a single source the principal content of two previous reports which assess the radiological environmental impact of mining and milling of thorium ore and of the reprocessing and refabrication of spent FBR thorium-uranium carbide fuel.

  10. Processing and waste disposal representative for fusion breeder blanket systems

    SciTech Connect (OSTI)

    Finn, P.A.; Vogler, S.

    1987-01-01T23:59:59.000Z

    This study is an evaluation of the waste handling concepts applicable to fusion breeder systems. Its goal is to determine if breeder blanket waste can be disposed of in shallow land burial, the least restrictive method under US Nuclear Regulatory regulations. The radionuclides expected in the materials used in fusion reactor blankets are described, as are plans for reprocessing and disposal of the components of different breeder blankets. An estimate of the operating costs involved in waste disposal is made.

  11. Thermal breeder fuel enrichment zoning

    DOE Patents [OSTI]

    Capossela, Harry J. (Schenectady, NY); Dwyer, Joseph R. (Albany, NY); Luce, Robert G. (Schenectady, NY); McCoy, Daniel F. (Latham, NY); Merriman, Floyd C. (Rotterdam, NY)

    1992-01-01T23:59:59.000Z

    A method and apparatus for improving the performance of a thermal breeder reactor having regions of higher than average moderator concentration are disclosed. The fuel modules of the reactor core contain at least two different types of fuel elements, a high enrichment fuel element and a low enrichment fuel element. The two types of fuel elements are arranged in the fuel module with the low enrichment fuel elements located between the high moderator regions and the high enrichment fuel elements. Preferably, shim rods made of a fertile material are provided in selective regions for controlling the reactivity of the reactor by movement of the shim rods into and out of the reactor core. The moderation of neutrons adjacent the high enrichment fuel elements is preferably minimized as by reducing the spacing of the high enrichment fuel elements and/or using a moderator having a reduced moderating effect.

  12. MODELING TRITIUM TRANSPORT IN PBLI BREEDER BLANKETS UNDER STEADY STATE , M. Abdou1

    E-Print Network [OSTI]

    Abdou, Mohamed

    MODELING TRITIUM TRANSPORT IN PBLI BREEDER BLANKETS UNDER STEADY STATE H. Zhang1 , A. Ying1 , M breeder blankets under realistic reactor-like conditions in this paper. Tritium concentration. Tritium behavior in the liquid metal breeder blanket requires a thorough understanding of the sequence

  13. CELLULAR FOAMS: A POTENTIAL INNOVATIVE SOLID BREEDER MATERIAL FOR FUSION APPLICATIONS

    E-Print Network [OSTI]

    Ghoniem, Nasr M.

    the development of ceramic foams or cellular ceramics for solid breeders in fusion reactor blankets. A cellular breeder material has a number of thermo-mechanical advantages over pebble beds, which can enhance blanket, and improved breeder-wall contact would result in a reduction of blanket multiplier and structure volume

  14. Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

    SciTech Connect (OSTI)

    Shott, Gregory [NSTec

    2014-08-31T23:59:59.000Z

    The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229Th, 230Th, 232U, 233U, and 234U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

  15. Thermal control of ceramic breeder blankets

    SciTech Connect (OSTI)

    Raffray, A.R.; Tillack, M.S.; Abdou, M.A. (Univ. of California, Los Angeles, CA (United States))

    1993-05-01T23:59:59.000Z

    Thermal control is an important issue for ceramic breeder blankets since the breeder needs to operate within its temperature window for the tritium release and inventory to be acceptable. A thermal control region is applicable not only to situations where the coolant can be run at low temperature, such as for the International Thermonuclear Experimental Reactor (ITER) base blanket, but also to ITER test module and power reactor situations, where it would allow for ceramic breeder operation over a wide range of power densities in space and time. Four thermal control mechanisms applicable to ceramic breeder blanket designs are described: A helium gap, a beryllium sintered block region, a beryllium sintered block region with a metallic felt at the beryllium-cladding interface, and a beryllium packed-bed region. Key advantages and issues associated with each of these mechanisms are discussed. Experimental and modeling studies focusing on beryllium packed-bed thermal conductivity and wall conductance, and beryllium sintered block-stainless steel cladding contact resistance are then described. Finally, an assessment of the potential of the different mechanisms for both passive and active control is carried out based on example calculations for a given set of ITER-like conditions. 28 refs., 33 figs., 3 tabs.

  16. Accelerator breeders: will they replace liquid metal fast breeders

    SciTech Connect (OSTI)

    Grand, P.; Powell, J.R.; Steinberg, M.; Takahashi, H.

    1983-06-01T23:59:59.000Z

    Investigation of accelerator breeders at Brookhaven National Laboratory indicate that the AB-LWR fuel cycle is economically competitive with the LMFBR fuel cycle. The same can be said about the accelerator breeder-High Temperature Gas Reactor symbiosis. This system appears to be very competitive with the added real advantage of superior safety and proliferation resistance. This discussion would be incomplete if the real competitor to accelerator breeding was not mentioned, namely Fusion Hybrid Breeding (FHB). Fusion Hybrid Breeding is a nearer option than pure fusion, as the breakeven Q value requirements are much more modest. Fusion Hybrid Breeding, if successful and practical, has the potential for highly efficient fissile fuel breeding, leading to cheaper fuel. The system, however, has yet to be demonstrated scientifically and to be shown commercially feasible. This is in contrast with the AB system which is an extension of proven, state-of-the-art technology with implementation possible within twenty years. 25 references, 4 figures, 5 tables.

  17. Laser fusion driven breeder design study. Final report

    SciTech Connect (OSTI)

    Berwald, D.H.; Massey, J.V.

    1980-12-01T23:59:59.000Z

    The results of the Laser Fusion Breeder Design Study are given. This information primarily relates to the conceptual design of an inertial confinement fusion (ICF) breeder reactor (or fusion-fission hybrid) based upon the HYLIFE liquid metal wall protection concept developed at Lawrence Livermore National Laboratory. The blanket design for this breeder is optimized to both reduce fissions and maximize the production of fissile fuel for subsequent use in conventional light water reactors (LWRs). When the suppressed fission blanket is compared with its fast fission counterparts, a minimal fission rate in the blanket results in a unique reactor safety advantage for this concept with respect to reduced radioactive inventory and reduced fission product decay afterheat in the event of a loss-of-coolant-accident.

  18. Progress in tritium retention and release modeling for ceramic breeders

    SciTech Connect (OSTI)

    Raffray, A.R.; Federici, G. [Max-Planck-Institut fuer Plasmaphysik, Muenchen (Germany)] [and others

    1994-12-31T23:59:59.000Z

    An important aspect of the design and analysis of ceramic breeder blankets is the ability to predict the phenomenological behavior of tritium in the ceramic breeder under operating reactor conditions. By understanding the behavior of tritium in such materials, analysis and accurate predictions can be made regarding the blanket tritium release and inventory which are key design issues based on safety and fuel self-sufficiency considerations. This paper highlights the progress in tritium modeling over the last decade. Key tritium transport mechanisms are briefly described along with the more recent and sophisticated models developed to help understand them. Recent experimental data are highlighted and model calibration and validation discussed. Finally, example applications to blanket cases are shown as illustration of current predictions for ceramic breeder blanket tritium inventory.

  19. Helium-cooled solid breeder blanket for ITER

    SciTech Connect (OSTI)

    Raffray, A.R.; Abdou, M.A.; Chou, P.; Gorbis, Z.; Tillack, M.; Watanabe, Y.; Ying, A.

    1989-03-01T23:59:59.000Z

    This paper summarizes the latest results of a design study of a helium-cooled solid breeder blanket for ITER. Attractive features of this design include the following: (1) There is a significant design margin since only part of the allowable solid breeder temperature window needs to be used. (2) There is an expanding data base available from solid breeder experiments carried out internationally. (3) The solid breeder can be designed to operate at high reactor-relevant temperature, while the helium is kept at moderate temperature and pressure for safety and reliability. In addition, since helium is a gas, it can be run so as to optimize the structure temperature and accommodate long term power variation without incurring any substantial pressure penalty. (4) The use of helium, an inert gas minimizing any chemical reaction and corrosion, in combination with a low activation solid breeder, is a safety advantage. An extensive list of the blanket operating parameters is provided and key factors are discussed.

  20. Status of the database for solid breeder materials

    SciTech Connect (OSTI)

    NONE

    1996-12-31T23:59:59.000Z

    The databases for solid breeder ceramics (Li{sub 2}O, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3} and LiAl0{sub 2}) and beryllium multiplier material were critically reviewed and evaluated as part of the ITER/CDA design effort (1988-1990). The results have been documented in a detailed technical report which includes progress made in expanding the solid breeder and beryllium databases up through September 1993. Emphasis was placed on the physical, thermal, mechanical, chemical-stability/compatibility, tritium retention/release and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Materials properties correlations were selected for use in design analysis, and ranges of input parameters (e.g., temperature, porosity, etc.) were established. The need for updating the ceramic breeder database was discussed at the Third Ceramic Breeder Blanket Interactions (CBBI-3) workshop at UCLA in June 1994. Progress made in expanding the ceramic breeder database and plans for updating the database are discussed.

  1. Fusion-breeder program

    SciTech Connect (OSTI)

    Moir, R.W.

    1982-11-19T23:59:59.000Z

    The various approaches to a combined fusion-fission reactor for the purpose of breeding /sup 239/Pu and /sup 233/U are described. Design aspects and cost estimates for fuel production and electricity generation are discussed. (MOW)

  2. Water-cooled solid-breeder concept for ITER

    SciTech Connect (OSTI)

    Gohar, Y.; Baker, C.C.; Attaya, H.; Billone, M.; Clemmer, R.C.; Finn, P.A.; Hassanein, A.; Johnson, C.E.; Majumdar, S.; Mattas, R.F.

    1988-08-01T23:59:59.000Z

    A water-cooled solid-breeder blanket concept was developed for ITER. The main function of this blanket is to produce the necessary tritium for the ITER operation. Several design features are incorporated in this blanket concept to increase its attractiveness. It is assumed that the blanket operation at commercial power reactor conditions can be sacrificed to achieve a high tritium breeding ratio with minimum additional research and development, and minimal impact on reactor design and operation. Operating temperature limits are enforced for each material to insure a satisfactory blanket performance. In fact, the design was iterated to maximize the tritium breeding ratio and satisfy these temperature limits. The other design constraint is to permit a large increase in the neutron wall loading without exceeding the temperature limits for the different blanket materials. The blanket concept contains 1.8 cm of Li/sub 2/O and 22.5 cm of beryllium both with a 0.8 density factor. The water coolant is isolated from the breeder material by several zones which reduces the tritium buildup in the water by permeation, reduces the chance for water-breeder interaction, and permits the breeder to operate at high temperature with a low temperature coolant. This improves the safety and environmental aspects of the blanket and eliminates the costly process of the tritium recovery from the water. The key features and design analysis of this blanket are summarized in this paper. 11 refs., 2 figs., 3 tabs.

  3. Liquid Metal Fast Breeder Reactors: a bibliography

    SciTech Connect (OSTI)

    Raleigh, H.D. (ed.) [ed.

    1980-11-01T23:59:59.000Z

    This bibliography includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  4. Liquid Metal Fast Breeder Reactors: a bibliography

    SciTech Connect (OSTI)

    Raleigh, H.D. (ed.) [ed.

    1980-11-01T23:59:59.000Z

    This bibliogralphy includes 5465 selected citations on LMFBR development. The citations were compiled from the DOE Energy Data Base covering the period January 1978 (EDB File No. 78R1087) through August 1980 (EDB File No. 80C79142). The references are to reports from the Department of Energy and its contractors, reports from other government or private organizations, and journal articles, books, conference papers, and monographs from US originators. Report citations are arranged alphanumerically by report number; nonreport literature citations are arranged chronologically. Corporate, Personal Author, Subject, and Report Number Indexes are provided in Volume 2.

  5. Solid breeder/structure mechanical interaction and thermal stability

    SciTech Connect (OSTI)

    Liu, Y.Y.; Billone, M.C.; Taghavi, K.

    1985-04-01T23:59:59.000Z

    Solid breeder/structure mechanical interaction (BSMI) during fusion reactor blanket operation is a potential failure mode which could limit the lifetime of the blanket. The severity of BSMI will generally depend on the materials, specific blanket designs, and blanket operating conditions. Thermomechanical analyses performed for a helium-cooled blanket employing Li/sub 2/O/HT-9 plates indicate that BSMI could be a serious concern for this blanket.

  6. US solid breeder blanket design for ITER

    SciTech Connect (OSTI)

    Gohar, Y.; Attaya, H.; Billone, M.; Lin, C.; Johnson, C.; Majumdar, S.; Smith, D. (Argonne National Lab., IL (USA)); Goranson, P.; Nelson, B.; Williamson, D.; Baker, C. (Oak Ridge National Lab., TN (USA)); Raffray, A.; Badawi, A.; Gorbis, Z.; Ying, A.; Abdou, M. (California Univ., Los Angeles, CA (USA)); Sviatoslavsky, I.; Blanchard, J.; Mogahed, E.; Sawan, M.; Kulcinski, G. (Wisconsin Univ., Madison, WI (USA))

    1990-09-01T23:59:59.000Z

    The US blanket design activity has focused on the developments and the analyses of a solid breeder blanket concept for ITER. The main function of this blanket is to produce the necessary tritium required for the ITER operation and the test program. Safety, power reactor relevance, low tritium inventory, and design flexibility are the main reasons for the blanket selection. The blanket is designed to operate satisfactorily in the physics and the technology phases of ITER without the need for hardware changes. Mechanical simplicity, predictability, performance, minimum cost, and minimum R D requirements are the other criteria used to guide the design process. The design aspects of the blanket are summarized in this paper. 2 refs., 7 figs., 3 tabs.

  7. Ceramic breeder materials : status and needs.

    SciTech Connect (OSTI)

    Johnson, C.E.

    1998-02-02T23:59:59.000Z

    The tritium breeding blanket is one of the most important components of a fusion reactor because it directly involves both energy extraction and tritium production, both of which are critical to fusion power. Because of their overall desirable properties, lithium-containing ceramic solids are recognized as attractive tritium breeding materials for fusion reactor blankets. Indeed, their inherent thermal stability and chemical inertness are significant safety advantages. In numerous in-pile experiments, these materials have performed well, showing good thermal stability and good tritium release characteristics. Tritium release is particularly facile when an argon or helium purge gas containing hydrogen, typically at levels of about 0.1%, is used. However, the addition of hydrogen to the purge gas imposes a penalty when it comes to recovery of the tritium produced in the blanket. In particular, a large amount of hydrogen in the purge gas will necessitate a large multiple-stage tritium purification unit, which could translate into higher costs. Optimizing tritium release while minimizing the amount of hydrogen necessary in the purge gas requires a deeper understanding of the tritium release process, especially the interactions of hydrogen with the surface of the lithium ceramic. This paper reviews the status of ceramic breeder research and highlights several issues and data needs.

  8. ASSESSMENT OF FIRST WALL AND BLANKET OPTIONS WITH THE USE OF LIQUID BREEDER C.P.C. Wong, S. Malang,1 M. Sawan,2 S. Smolentsev,3 S. Majumdar,4 B. Merrill,5 D.K. Sze,6 N. Morley,3

    E-Print Network [OSTI]

    Abdou, Mohamed

    ASSESSMENT OF FIRST WALL AND BLANKET OPTIONS WITH THE USE OF LIQUID BREEDER C.P.C. Wong, S. Malang.S. blanket design team has proposed to focus on the self-cooled (SC) and dual- coolant (DC) liquid breeder of liquid breeder blanket options for an advanced reactor design. These options include self-cooled (SC

  9. airlift reactors influence: Topics by E-print Network

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

    some success in R. F. Gimpel 13 The Thorium Molten Salt Reactor Moving on from the MSBR CERN Preprints Summary: A re-evaluation of the Molten Salt Breeder Reactor concept has...

  10. Fusion Engineering and Design 81 (2006) 455460 Breeder foam: an innovative low porosity solid breeder material

    E-Print Network [OSTI]

    Ghoniem, Nasr M.

    2006-01-01T23:59:59.000Z

    ; Foam processing; Foam applications 1. Introduction Solid breeder blanket concepts are typically based structural and Be material requirements in a breeder blanket. However, high density Li-ceramic foams haveFusion Engineering and Design 81 (2006) 455­460 Breeder foam: an innovative low porosity solid

  11. Nuclear reactor composite fuel assembly

    DOE Patents [OSTI]

    Burgess, Donn M. (Richland, WA); Marr, Duane R. (West Richland, WA); Cappiello, Michael W. (Richland, WA); Omberg, Ronald P. (Richland, WA)

    1980-01-01T23:59:59.000Z

    A core and composite fuel assembly for a liquid-cooled breeder nuclear reactor including a plurality of elongated coextending driver and breeder fuel elements arranged to form a generally polygonal bundle within a thin-walled duct. The breeder elements are larger in cross section than the driver elements, and each breeder element is laterally bounded by a number of the driver elements. Each driver element further includes structure for spacing the driver elements from adjacent fuel elements and, where adjacent, the thin-walled duct. A core made up of the fuel elements can advantageously include fissile fuel of only one enrichment, while varying the effective enrichment of any given assembly or core region, merely by varying the relative number and size of the driver and breeder elements.

  12. A professional internship with Texas Breeders Service

    E-Print Network [OSTI]

    McKemie, Jack Furman

    1989-01-01T23:59:59.000Z

    and testing procedures are used to determine quality and evaluation of the final product. TEXAS BREEDERS SERVICE INC. offers Contract AI Programs for the commercial breeder to help select and locate semen from performance tested buQa to enhance marketing... House" donor management services are available. TEXAS BREKDERS SERVICE INC. assists clients in Marketing Genetics in both domestic and foreign markets. EXAS BREEDERS SERVICE INC P. O. Box l635 San Marcos, TX 78667 5 l 2-357-6 l 62 ACKNOWLEDGMENTS...

  13. A European proposal for a ITER water cooled solid breeder blanket

    SciTech Connect (OSTI)

    Lorenzetto, P. [NET, Garching (Germany); Gierszewski, P. [CFFTP, Mississauga, Ontario (Canada); Simbolotti, G. [ENEA, Frascati (Italy)

    1994-12-31T23:59:59.000Z

    The Water Cooled Solid Breeder Blanket concept here proposed is based on a conservative approach, involving well proven technologies and-qualified materials. 316 L type stainless steel has been selected as the structural material. The nominal performances are: 1 MW/m{sup 2} as the average neutron wall load which corresponds to a fusion power of about 1.5 GW, and 1 MWy/m{sup 2} as the average neutron fluence. The power margins of the proposed concept have been estimated. The proposed blanket concept is based on a Breeder Inside Tube (BIT) type blanket with poloidal breeding elements, whose dimensions are compatible with space available in test fission reactor core channels, that makes easier in-pile testing required for the blanket development and qualification. Each breeding element consists of two concentric tubes. 1.2 mm lithium metazirconate (Li{sub 2}ZrO{sub 3}) pebbles are filled into the inner tube, the water coolant flows in the annular channel between the two tubes, and 2 mm Beryllium pebbles are poured into the blanket box outside the outer tube. Lithium metazirconate has been selected as the breeder material because it presents today the best tritium release properties at low temperature. A helium purge gas flows through the breeder pebble bed for tritium recovery. A Shielding Blanket can be derived from the proposed Blanket concept by removing the breeder pebbles from the inner tube. In-situ convertibility issues are addressed.

  14. Assessment of First Wall and Blanket Options with the Use of Liquid Breeder

    SciTech Connect (OSTI)

    Wong, C.P.C.; Malang, S.; Sawan, M. (and others)

    2005-04-15T23:59:59.000Z

    As candidate blanket concepts for a U.S. advanced reactor power plant design, with consideration of the time frame for ITER development, we assessed first wall and blanket design concepts based on the use of reduced activation ferritic steel as structural material and liquid breeder as the coolant and tritium breeder. The liquid breeder choice includes the conventional molten salt Li{sub 2}BeF{sub 4} and the low melting point molten salts such as LiBeF{sub 3} and LiNaBeF{sub 4} (FLiNaBe). Both self-cooled and dual coolant molten salt options were evaluated. We have also included the dual coolant leadeutectic Pb-17Li design in our assessment. We take advantage of the molten salt low electrical and thermal conductivity to minimize impacts from the MHD effect and the heat losses from the breeder to the actively cooled steel structure. For the Pb-17Li breeder we employ flow channel inserts of SiC{sub f}/SiC composite with low electrical and thermal conductivity to perform respective insulation functions. We performed preliminary assessments of these design options in the areas of neutronics, thermal-hydraulics, safety, and power conversion system. Status of the R and D items of selected high performance blanket concepts is reported. Results from this study will form the technical basis for the formulation of the U.S. ITER test module program and corresponding test plan.

  15. Nuclear reactor control

    SciTech Connect (OSTI)

    Ingham, R.V.

    1980-01-01T23:59:59.000Z

    A liquid metal cooled fast breeder nuclear reactor has power setback means for use in an emergency. On initiation of a trip-signal a control rod is injected into the core in two stages, firstly, by free fall to effect an immediate power-set back to a safe level and, secondly, by controlled insertion. Total shut-down of the reactor under all emergencies is avoided. 4 claims.

  16. Reproductive performance of broiler breeders in cages

    E-Print Network [OSTI]

    Campos, Egladson Joao

    1971-01-01T23:59:59.000Z

    REPRODUCTIVE PERFORMANCE OF BROILER BREEDERS IN CAGES A Thesis by EGLADSON JOAO CAMPOS Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirement for the degree of MASTER OF SCIENCE December 1971... Major Subject: Poultry Science REPRODUCTIVE PERFORMANCE OF BROILER BREEDERS IN CAGES A Thesis by EGLADSON JOAO CAMPOS Approved as to style and content by: Chairman of Commi ee) ad of Department) (member) c', i- (member) member) (member...

  17. The Breeding Blanket Interface (BBI): Recent results for the solid breeder and the aqueous salt solution blanket concepts

    SciTech Connect (OSTI)

    Clemmer, R.G.; Finn, P.A.; Greenwood, L.R.; Sze, D.K.; Bartlit, J.R.; Sherman, R.; Anderson, J.L.; Yoshida, H.; Naruse, Y.; Enoeda, M.; Okuno, K. (Argonne National Lab., IL (USA); Los Alamos National Lab., NM (USA); Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan))

    1989-10-01T23:59:59.000Z

    The Tritium Systems Test Assembly (TSTA) at Los Alamos is a full-scale facility dedicated to testing tritium processing for fusion reactors. We are involved in a study of adding a Breeder Blanket Interface (BBI) to the TSTA. The BBI is to test the processing required for the tritium output streams for the various fusion reactor breeder blankets. In the current phase of the study, we are evaluating the characteristics of the output from various breeding blankets types. Emphasis is placed on defining the output stream with respect to H/T ratio, impurity content, and radionuclide content. Reported herein is an assessment for two blanket concepts: solid breeder blanket (ceramic, Li{sub 2}O), and aqueous salt solution. 24 refs., 2 figs., 2 tabs.

  18. Continuous production of tritium in an isotope-production reactor with a separate circulation system

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19T23:59:59.000Z

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium is allowed to flow through the reactor in separate loops in order to facilitate the production and removal of tritium.

  19. Proceedings of the NEACRP/IAEA Specialists meeting on the international comparison calculation of a large sodium-cooled fast breeder reactor at Argonne National Laboratory on February 7-9, 1978

    SciTech Connect (OSTI)

    LeSage, L.G.; McKnight, R.D.; Wade, D.C.; Freese, K.E.; Collins, P.J.

    1980-08-01T23:59:59.000Z

    The results of an international comparison calculation of a large (1250 MWe) LMFBR benchmark model are presented and discussed. Eight reactor configurations were calculated. Parameters included with the comparison were: eigenvalue, k/sub infinity/, neutron balance data, breeding reaction rate ratios, reactivity worths, central control rod worth, regional sodium void reactivity, core Doppler and effective delayed neutron fraction. Ten countries participated in the comparison, and sixteen solutions were contributed. The discussion focuses on the variation in parameter values, the degree of consistency among the various parameters and solutions, and the identification of unexpected results. The results are displayed and discussed both by individual participants and by groupings of participants (e.g., results from adjusted data sets versus non-adjusted data sets).

  20. Neutronics and thermal design analyses of US solid breeder blanket for ITER

    SciTech Connect (OSTI)

    Gohar, Y.; Billone, M.; Attaya, H. (Argonne National Lab., IL (USA)); Sawan, M. (Wisconsin Univ., Madison, WI (USA))

    1990-09-01T23:59:59.000Z

    The US Solid Breeder Blanket is designed to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Safety, low tritium inventory, reliability, flexibility cost, and minimum R D requirements are the other design criteria. To satisfy these criteria, the produced tritium is recovered continuously during operation and the blanket coolant operates at low pressure. Beryllium multiplier material is used to control the solid-breeder temperature. Neutronics and thermal design analyses were performed in an integrated manner to define the blanket configuration. The reference parameters of ITER including the operating scenarios, the neutron wall loading distribution and the copper stabilizer are included in the design analyses. Several analyses were performed to study the impact of the reactor parameters, blanket dimensions, material characteristics, and heat transfer coefficient at the material interfaces on the blanket performance. The design analyses and the results from the different studies are summarized. 6 refs., 3 figs., 3 tabs.

  1. Design of a helium-cooled molten salt fusion breeder

    SciTech Connect (OSTI)

    Moir, R.W.; Lee, J.D.; Fulton, F.J.; Huegel, F.; Neef, W.S. Jr.; Sherwood, A.E.; Berwald, D.H.; Whitley, R.H.; Wong, C.P.C.; DeVan, J.H.

    1985-02-01T23:59:59.000Z

    A new conceptual blanket design for a fusion reactor produces fissile material for fission power plants. Fission is suppressed by using beryllium, rather than uranium, to multiply neutrons and also by minimizing the fissile inventory. The molten-salt breeding media (LiF + BeF/sub 2/ + TghF/sub 4/) is circulated through the blanket and on to the online processing system where /sup 233/U and tritium are continuously removed. Helium cools the blanket including the steel pipes containing the molten salt. Austenitic steel was chosen because of its ease of fabrication, adequate radiation-damage lifetime, and low corrosion rate by molten salt. We estimate the breeder, having 3000 MW of fusion power, produces 6400 kg of /sup 233/U per year, which is enough to provide make up for 20 GWe of LWR per year (or 14 LWR plants of 4440 MWt) or twice that many HTGRs or CANDUs. Safety is enhanced because the afterheat is low and the blanket materials do not react with air or water. The fusion breeder based on a pre-MARS tandem mirror is estimated to cost $4.9B or 2.35 times an LWR of the same power. The estimated present value cost of the /sup 2/anumber/sup 3/U produced is $40/g if utility financed or $16/g if government financed.

  2. Homogeneous fast-flux isotope-production reactor

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19T23:59:59.000Z

    A method is described for producing tritium in a liquid metal fast breeder reactor. Lithium target material is dissolved in the liquid metal coolant in order to facilitate the production and removal of tritium.

  3. Thermal control of solid breeder blankets

    SciTech Connect (OSTI)

    Raffray, A.R.; Ying, A.; Gorbis, Z.; Tillack, M.S.; Abdou, M.A.

    1991-12-31T23:59:59.000Z

    An assessment of the thermal control mechanisms applicable to solid breeder blanket designs under ITER-like operating conditions is presented in this paper. Four cases are considered: a helium gap; a sintered block Be region; a sintered block helium region with a metallic felt at the Be/clad interface; and a Be packed bed region. For these cases, typical operating are explored to determine the ranges of wall load which can be accommodated while maintaining the breeder within its allowable operating temperature window. The corresponding region thicknesses are calculated to help identify practicality and design tolerances.

  4. Thermal control of solid breeder blankets

    SciTech Connect (OSTI)

    Raffray, A.R.; Ying, A.; Gorbis, Z.; Tillack, M.S.; Abdou, M.A.

    1991-01-01T23:59:59.000Z

    An assessment of the thermal control mechanisms applicable to solid breeder blanket designs under ITER-like operating conditions is presented in this paper. Four cases are considered: a helium gap; a sintered block Be region; a sintered block helium region with a metallic felt at the Be/clad interface; and a Be packed bed region. For these cases, typical operating are explored to determine the ranges of wall load which can be accommodated while maintaining the breeder within its allowable operating temperature window. The corresponding region thicknesses are calculated to help identify practicality and design tolerances.

  5. Experimental Studies of Active Temperature Control in Solid Breeder Blankets

    E-Print Network [OSTI]

    Tillack, Mark

    1 Experimental Studies of Active Temperature Control in Solid Breeder Blankets M. S. Tillack, A. R barrier regions for solid breeder blankets. In particular, particle beds have been studied because breeder blankets is thermomechanical behavior in the fusion environment. Stable and predictable

  6. Ceramic Breeder Blanket for ARIES-CS

    SciTech Connect (OSTI)

    Raffray, A.R. [University of California-San Diego (United States); Malang, S. [Fusion Nuclear Technology Consulting (United States); El-Guebaly, L. [University of Wisconsin (United States); Wang, X. [University of California-San Diego (United States)

    2005-05-15T23:59:59.000Z

    This paper describes the conceptual design of a ceramic breeder blanket considered as one of the candidate blankets in the first phase of the ARIES-CS study. The blanket is coupled to a Brayton power cycle to avoid the safety concern associated with the possibility of Be/steam reaction in case of accident.

  7. Modeling Tritium Transport in PbLi Breeder Blankets Under Steady State

    SciTech Connect (OSTI)

    H. Zhang; A. Ying; M. Abdou; B. Merrill

    2011-08-01T23:59:59.000Z

    Tritium behavior in the breeder/coolant plays a crucial role in keeping the tritium loss under an allowable limit and realizing high tritium recovery efficiency. In this paper, progress toward the development of a comprehensive 3D predictive capability is discussed and presented. The sequence of transport processes leading to tritium release includes diffusion and convection through the PbLi, transfer across the liquid/solid interface, diffusion of atomic tritium through the structure, and dissolution-recombination at the solid/gas interface. Numerical simulation of the coupled individual physics phenomena of tritium transport is performed for DCLL/HCLL type breeder blankets under realistic reactor-like conditions in this paper. Tritium concentration and permeation are presented and the MHD effects are evaluated. Preliminary results shows that the MHD velocity profile has the significant effect in preventing tritium permeation due to the higher convection effects near the wall.

  8. Tritium Recovery from Solid Breeder Blanket by Water Vapor Addition to Helium Sweep Gas

    SciTech Connect (OSTI)

    Kawamura, Yoshinori; Iwai, Yasunori; Nakamura, Hirofumi; Hayashi, Takumi; Yamanishi, Toshihiko; Nishi, Masataka [Japan Atomic Energy Research Institute (Japan)

    2005-07-15T23:59:59.000Z

    In the solid breeder blanket of fusion reactor, bred tritium is planned to be extracted from the blanket as HT by passing of H{sub 2}-added sweep gas in general. In that case, tritium leakage by permeation to coolant can not be ignored. So, the application of H{sub 2}O-added sweep gas is discussed, with which tritium leakage to coolant can be much reduced. As the result of discussion, H{sub 2}O-added sweep gas is probable method of tritium recovery. For the further detailed discussion, it is important to enrich the data correlated to the interaction of H{sub 2}, H{sub 2}O, breeder, multiplier and structures.

  9. Cellular Foams: A Potential Innovative Solid Breeder Material for Fusion Applications

    SciTech Connect (OSTI)

    Sharafat, S.; Ghoniem, N.; Williams, B.; Babcock, J. [University of California Los Angeles (United States)

    2005-05-15T23:59:59.000Z

    Ceramic foam and cellular materials are being used in a wide variety of industries and are finding ever growing number of applications. Over the past decade advances in manufacturing of cellular materials have resulted in ceramics with highly uniform interconnected porosities ranging in size from a few {mu}m to several mm. These relatively new ceramic foam materials have a unique set of thermo-mechanical properties, such as excellent thermal shock resistance and high surface to volume ratios. Based on new advances in processing ceramic foams, we suggest the development of ceramic foams or cellular ceramics for solid breeders in fusion reactor blankets. A cellular breeder material has a number of thermo-mechanical advantages over pebble beds, which can enhance blanket performance, improve operational stability, and reduce overall blanket costs.

  10. accelerator breeders: Topics by E-print Network

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

    against chicken anemia virus in unvaccinated broilers and broiler breeders in Croatia CiteSeer Summary: The prevalence of antibodies against infectious chicken anemia...

  11. adult breeder canaries: Topics by E-print Network

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

    responses of the breeder units under fusion-relevant conditions. Yixiang Gan; Francisco Hernandez; Dorian Hanaor; Ratna Annabattula; Marc Kamlah; Pavel Pereslavtsev 2014-06-17 47...

  12. advanced liquid breeder: Topics by E-print Network

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

    responses of the breeder units under fusion-relevant conditions. Yixiang Gan; Francisco Hernandez; Dorian Hanaor; Ratna Annabattula; Marc Kamlah; Pavel Pereslavtsev 2014-06-17 37...

  13. Solid breeder blanket option for the ITER conceptual design

    SciTech Connect (OSTI)

    Gohar, Y.; Attaya, H.; Billone, M.C.; Finn, P.; Majumdar, S.; Turner, L.R.; Baker, C.C.; Nelson, B.E.; Raffray, R. (Argonne National Lab., IL (USA); Oak Ridge National Lab., TN (USA); California Univ., Los Angeles, CA (USA))

    1989-10-01T23:59:59.000Z

    A solid-breeder water-cooled blanket option was developed for ITER based on a multilayer configuration. The blanket uses beryllium for neutron multiplication and lithium oxide for tritium breeding. The material forms are sintered products for both material with 0.8 density factor. The lithium-6 enrichment is 90%. This blanket has the capability to accommodate a factor of two change in the neutron wall loading without violating the different design guidelines. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. At the same time, the reliability and the safety aspects of the blanket are enhanced by the use of a low-pressure coolant and the separation of the tritium purge lines from the coolant system. The blanket modules are made by hot vacuum forming and diffusion bonding a double wall structure with integral cooling channels. The different aspects of the blanket design including tritium breeding, nuclear heat deposition, activation analyses, thermal-hydraulics, tritium inventory, structural analyses, and water coolant conditions are summarized in this paper. 12 refs., 2 figs., 1 tab.

  14. Assemblies with both target and fuel pins in an isotope-production reactor

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19T23:59:59.000Z

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins adjacent to fuel pins in order to increase the tritium production rate.

  15. Vented target elements for use in an isotope-production reactor. [LMFBR

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19T23:59:59.000Z

    A method is described for producing tritium gas in a fast breeder reactor cooled with liquid metal. Lithium target material is placed in pins equipped with vents, and tritium gas is recovered from the coolant.

  16. Fuel pins with both target and fuel pellets in an isotope-production reactor

    DOE Patents [OSTI]

    Cawley, W.E.; Omberg, R.P.

    1982-08-19T23:59:59.000Z

    A method is described for producing tritium in a fast breeder reactor cooled with liquid metal. Lithium target pellets are placed in close contact with fissile fuel pellets in order to increase the tritium production rate.

  17. Reactor materials study of EBR-II and BN350

    E-Print Network [OSTI]

    Yilmaz, Fatma

    2002-01-01T23:59:59.000Z

    The objective of this research is to go through the technical review of how the body of information relating to the in-reactor behavior of structural materials of Experimental Breeder Reactor-II (EBR-II) and BN350 are associated. Such an effort...

  18. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    SciTech Connect (OSTI)

    Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

    2006-02-01T23:59:59.000Z

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17LI is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Ph-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure.

  19. Neutronic Study of Slightly Modified Water Reactors and Application to Transition Scenarios Richard Chambon

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    conver- ting or full breeding cycles with technologies such as Fast Breeder Reactors (FBRs) or Molten conventional fuels are replaced by GenIV reactors as fast as the neces- sary fissile fuel stocks can allow itNeutronic Study of Slightly Modified Water Reactors and Application to Transition Scenarios Richard

  20. Development of Solid Breeder Blanket at JAERI

    SciTech Connect (OSTI)

    Enoeda, Mikio; Hatano, Toshihisa; Tsuchiya, Kunihiko; Ochiai, Kentaro; Kawamura, Yoshinori; Hayashi, Kimio; Nishitani, Takeo; Nishi, Masataka; Akiba, Masato [Japan Atomic Energy Research Institute (Japan)

    2005-05-15T23:59:59.000Z

    Japan Atomic Energy Research Institute (JAERI) has been performing blanket development based on the long-term research program of fusion blankets in Japan, which was approved by the Fusion Council of Japan in 1999. The blanket development consists of out-pile R and D, In-pile R and D, TBM Neutronics and TPR Tests and Tritium Recovery System R and D. Based on the achievements of element technology development, the R and D program is now stepping to the engineering testing phase, in which scalable mockup tests will be performed for obtaining engineering data unique to the specific structure of the components, with the objective to define the fabrication specification of test blanket modules for ITER. This paper presents the major achievements of the element technology development of solid breeder blanket in JAERI.

  1. Status of fusion reactor blanket design

    SciTech Connect (OSTI)

    Smith, D.L.; Sze, D.K.

    1986-11-01T23:59:59.000Z

    This paper provides a brief review of the Blanket Comparison and Selection Study (BCSS)/sup 1/ and an overview of more recent fusion reactor blanket design efforts. Specific areas covered include improvements in leading blanket concepts identified in the BCSS, viz., self-cooled liquid metal concepts, helium-cooled solid breeder concepts, and helium-cooled liquid breeder concepts. In addition, a summary of innovative blanket concepts and design features is presented. The key features and critical issues associated with these designs are identified.

  2. Nuclear reactor safety heat transfer

    SciTech Connect (OSTI)

    Jones, O.C.

    1982-07-01T23:59:59.000Z

    Reviewed is a book which has 5 parts: Overview, Fundamental Concepts, Design Basis Accident-Light Water Reactors (LWRs), Design Basis Accident-Liquid-Metal Fast Breeder Reactors (LMFBRs), and Special Topics. It combines a historical overview, textbook material, handbook information, and the editor's personal philosophy on safety of nuclear power plants. Topics include thermal-hydraulic considerations; transient response of LWRs and LMFBRs following initiating events; various accident scenarios; single- and two-phase flow; single- and two-phase heat transfer; nuclear systems safety modeling; startup and shutdown; transient response during normal and upset conditions; vapor explosions, natural convection cooling; blockages in LMFBR subassemblies; sodium boiling; and Three Mile Island.

  3. Computational Nuclear Forensics Analysis of Weapons-grade Plutonium Separated from Fuel Irradiated in a Thermal Reactor

    E-Print Network [OSTI]

    Coles, Taylor Marie

    2014-04-27T23:59:59.000Z

    in which the uranium is irradiated. It 4 is unusual for a power reactor to discharge fuel at that low of a burnup. During the normal operation of a Fast Breeder Reactor (FBR), depleted uranium used in the peripheral region of the reactor core...

  4. Lithium Ceramic Blankets for Russian Fusion Reactors and Influence of Breeding Operation Mode on Parameters of Reactor Tritium Systems

    SciTech Connect (OSTI)

    Kapyshev, Victor K.; Chernetsov, Mikhail Yu.; Zhevotov, Sergej I.; Kersnovskij, Alexandr Yu.; Kolbasov, Boris N.; Kovalenko, Victor G.; Paltusov, Nikolaj P.; Sernyaev, Georgeij A.; Sterebkov, Juri S.; Zyryanov, Alexej P. [A.A. Bochvar Institute of Inorganic Materials (Russian Federation)

    2005-07-15T23:59:59.000Z

    Russian controlled fusion program supposes development of a DEMO reactor design and participation in ITER Project. A solid breeder blanket of DEMO contains a ceramic lithium orthosilicate breeder and a beryllium multiplier. Test modules of the blanket are developed within the scope of ITER activities. Experimental models of module tritium breeding zones (TBZ), materials and fabrication technology of the TBZ, tritium reactor systems to analyse and process gas released from lithium ceramics are being developed. Two models of tritium breeding and neutron multiplying elements of the TBZ have been designed, manufactured and tested in IVV-2M nuclear reactor. Initial results of the in-pile experiments and outcome of lithium ceramics irradiation in a water-graphite nuclear reactor are considered to be a data base for development of the test modules and initial requirements for DEMO tritium system design. Influence of the tritium release parameters and hydrogen concentration in a purge gas on parameters of reactor system are discussed.

  5. Water-cooled solid-breeder blanket concept for ITER

    SciTech Connect (OSTI)

    Gohar, Y.; Baker, C.C.; Attaya, H.; Billone, M.; Clemmer, R.C.; Finn, P.A.; Hassanein, A.; Johnson, C.E.; Majumdar, S.; Mattas, R.F.

    1989-03-01T23:59:59.000Z

    A water cooled solid-breeder blanket concept was developed for ITER. The main function of this blanket is to produce the necessary tritium for the ITER operation. Several design features are incorporated in this blanket concept to increase its attractiveness. The main features are the following: (a) a multilayer concept which reduces fabrication cost; (b) a simple blanket configuration which results in reliability advantages; (c) a very small breeder volume is employed to reduce the tritium inventory and the blanket cost; (d) a high tritium breeding ratio eliminates the need for an outside tritium supply; (e) a low-pressure system decreases the required steel fraction for structural purposes; (f) a low-temperature operation reduces the swelling concerns for beryllium; and (g) the small fractions of structure and breeder materials used in the blanket reduce the decay heat source. The key features and design analyses of this blanket are summarized in this paper.

  6. Fusion Engineering and Design 81 (2006) 659664 Solid breeder test blanket module design and analysis

    E-Print Network [OSTI]

    Abdou, Mohamed

    2006-01-01T23:59:59.000Z

    Fusion Engineering and Design 81 (2006) 659­664 Solid breeder test blanket module design This paper presents the design and analysis for the US ITER solid breeder blanket test articles. Objectives of solid breeder blanket testing during the first phase of the ITER operation focus on exploration

  7. Reactor control rod timing system

    DOE Patents [OSTI]

    Wu, Peter T. K. (Clifton Park, NY)

    1982-01-01T23:59:59.000Z

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  8. Reactor control rod timing system

    SciTech Connect (OSTI)

    Wu, P.T.

    1982-02-09T23:59:59.000Z

    A fluid driven jet-edge whistle timing system for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (Above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  9. Thermal resistance gaps for solid breeder blankets using packed beds

    SciTech Connect (OSTI)

    Gorbis, Z.R.; Raffray, A.R.; Tillack, M.S.; Abdou, M.A.

    1989-03-01T23:59:59.000Z

    The main design features of a new concept for solid breeder blanket thermal resistance gaps are described and analysis is shown for the blanket thermal characteristics. The effective thermal conductivity of a helium-beryllium packed bed configuration is studied, including the effect of a purge stream. Possible applications of this concept to ITER blanket designs are stressed.

  10. The Thorium Molten Salt Reactor : Moving on from the MSBR

    E-Print Network [OSTI]

    L. Mathieu; D. Heuer; R. Brissot; C. Le Brun; E. Liatard; J. M. Loiseaux; O. Méplan; E. Merle-Lucotte; A. Nuttin; J. Wilson; C. Garzenne; D. Lecarpentier; E. Walle; the GEDEPEON Collaboration

    2005-06-02T23:59:59.000Z

    A re-evaluation of the Molten Salt Breeder Reactor concept has revealed problems related to its safety and to the complexity of the reprocessing considered. A reflection is carried out anew in view of finding innovative solutions leading to the Thorium Molten Salt Reactor concept. Several main constraints are established and serve as guides to parametric evaluations. These then give an understanding of the influence of important core parameters on the reactor's operation. The aim of this paper is to discuss this vast research domain and to single out the Molten Salt Reactor configurations that deserve further evaluation.

  11. The Thorium Molten Salt Reactor Moving on from the MSBR

    E-Print Network [OSTI]

    Mathieu, L; Brissot, R; Le Brun, C; Liatard, E; Loiseaux, J M; Méplan, O; Merle-Lucotte, E; Nuttin, A; Wilson, J; Garzenne, C; Lecarpentier, D; Walle, E

    2006-01-01T23:59:59.000Z

    A re-evaluation of the Molten Salt Breeder Reactor concept has revealed problems related to its safety and to the complexity of the reprocessing considered. A reflection is carried out anew in view of finding innovative solutions leading to the Thorium Molten Salt Reactor concept. Several main constraints are established and serve as guides to parametric evaluations. These then give an understanding of the influence of important core parameters on the reactor's operation. The aim of this paper is to discuss this vast research domain and to single out the Molten Salt Reactor configurations that deserve further evaluation.

  12. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    SciTech Connect (OSTI)

    Wong, C C; Malang, S; Sawan, M; Dagher, M; Smolentsev, S; Merrill, B; Youssef, M; Reyes, S; Sze, D D; Morley, N B; Sharafat, S; Calderoni, P; Sviatoslavsky, G; Kurtz, R; Fogarty, P; Zinkle, S; Abdou, M

    2005-05-13T23:59:59.000Z

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperature of 700 C. We have identified critical issues for the concept, some of which include the first wall design, the assessment of MHD effects with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

  13. An overview of dual coolant Pb-17Li breeder first wall and blanket concept development for the US ITER-TBM design

    SciTech Connect (OSTI)

    Wong, Clement; Malang, S.; Sawan, M.; Dagher, Mohamad; Smolentsev, S.; Merrill, Brad; Youssef, M.; Reyes, Susanna; Sze, Dai Kai; Morley, Neil B.; Sharafat, Shahran; Calderoni, P.; Sviatoslavsky, G.; Kurtz, Richard J.; Fogarty, Paul J.; Zinkle, Steven J.; Abdou, Mohamed A.

    2006-07-05T23:59:59.000Z

    An attractive blanket concept for the fusion reactor is the dual coolant Pb-17Li liquid (DCLL) breeder design. Reduced activation ferritic steel (RAFS) is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb-17Li is circulated for power conversion and for tritium breeding. A SiCf/SiC composite insert is used as the magnetohydrodynamic (MHD) insulation to reduce the impact from the MHD pressure drop of the circulating Pb-17Li and as the thermal insulator to separate the high temperature Pb-17Li from the helium cooled RAFS structure. For the reference tokamak power reactor design, this blanket concept has the potential of satisfying the design limits of RAFS while allowing the feasibility of having a high Pb-17Li outlet temperture of 700C. We have identified critical issues for the concept, some of which inlude the first wall design, the assessment of MHD effectrs with the SiC-composite flow coolant insert, and the extraction and control of the bred tritium from the Pb-17Li breeder. R&D programs have been proposed to address these issues. At the same time, we have proposed a test plan for the DCLL ITER-Test Blanket Module program.

  14. Reactivity control assembly for nuclear reactor. [LMFBR

    DOE Patents [OSTI]

    Bollinger, L.R.

    1982-03-17T23:59:59.000Z

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  15. Three dimensional neutronics calculation comparison study for a fusion breeder with large channels

    SciTech Connect (OSTI)

    Huang, J.H.; Xie, Z.Y.; You, C.L. [Southwestern Inst. of Physics, Chengdu (China)] [and others

    1994-12-31T23:59:59.000Z

    A tokamak reactor is characterized by a toroidal geometry with large ports and channels. The three dimensional calculation seems necessary for the prediction of the neutronics parameters of the calculation. People have been attempting to simulate the configuration by one dimensional model. Assuming that the neutronics parameters such as tritium breeding ratio depends almost only on the primary 14 MeV neutron number entering the blanket, the following approximate scheme was proposed: {Alpha} = {Sigma}W{sub i}{sm_bullet}{Alpha}{sub i}, where {Alpha} represents the total value of a parameter; {Alpha}{sub i} is a partial value contributed from the i`th part of the blanket and is calculated by a 1-D cylindrical model; W{sub i} is the fusion number entering the i`th part of the blanket through the first wall. This scheme seems reasonable for a pure fusion reactor with similar inboard and outboard blankets, since neutron flux angular distribution is strongly forward and the mutual influence between similar blankets is weak. A study on influence between inboard and outboard blankets showed a rather strong influence between blankets in a fusion breeder, where the partial blankets are quite different in neutronics characteristics. The explanation is that the neutron source from fission and n,2n reactions in the outboard blanket causes considerable neutron leakage through the inner surface when it faces an {open_quote}inferior{close_quote} inboard blanket.

  16. Advanced Safeguards Approaches for New Fast Reactors

    SciTech Connect (OSTI)

    Durst, Philip C.; Therios, Ike; Bean, Robert; Dougan, A.; Boyer, Brian; Wallace, Rick L.; Ehinger, Michael H.; Kovacic, Don N.; Tolk, K.

    2007-12-15T23:59:59.000Z

    This third report in the series reviews possible safeguards approaches for new fast reactors in general, and the ABR in particular. Fast-neutron spectrum reactors have been used since the early 1960s on an experimental and developmental level, generally with fertile blanket fuels to “breed” nuclear fuel such as plutonium. Whether the reactor is designed to breed plutonium, or transmute and “burn” actinides depends mainly on the design of the reactor neutron reflector and the whether the blanket fuel is “fertile” or suitable for transmutation. However, the safeguards issues are very similar, since they pertain mainly to the receipt, shipment and storage of fresh and spent plutonium and actinide-bearing “TRU”-fuel. For these reasons, the design of existing fast reactors and details concerning how they have been safeguarded were studied in developing advanced safeguards approaches for the new fast reactors. In this regard, the design of the Experimental Breeder Reactor-II “EBR-II” at the Idaho National Laboratory (INL) was of interest, because it was designed as a collocated fast reactor with a pyrometallurgical reprocessing and fuel fabrication line – a design option being considered for the ABR. Similarly, the design of the Fast Flux Facility (FFTF) on the Hanford Site was studied, because it was a successful prototype fast reactor that ran for two decades to evaluate fuels and the design for commercial-scale fast reactors.

  17. Progress in conceptual design of a tokamak engineering test breeder

    SciTech Connect (OSTI)

    Huang, J.; Sheng, G.

    1993-12-31T23:59:59.000Z

    A tokamak engineering test breeder, TETB, was proposed in 1988. It has a liquid lithium self-cooled blanket of the fast fission type. Since 1989, revisions have been made for an improved version, the TETB-II. A fission suppressed blanket was adopted and the lithium cooling pattern changed, resulting in a much lower MHD pressure drop. The emphasis of this report is on the component design and analysis using computer codes.

  18. Evaluation of organic moderator/coolants for fusion breeder blankets

    SciTech Connect (OSTI)

    Romero, J.B.

    1980-03-01T23:59:59.000Z

    Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process.

  19. Overview of EU activities on DEMO liquid metal breeder blanket

    SciTech Connect (OSTI)

    Giancarli, L.; Proust, E. [DRN/DMT/SERMA, Gif-sur-Yvette (France); Benamati, G. [CRE Brasimone, Camugnano (Italy)] [and others

    1994-12-31T23:59:59.000Z

    The European test-blanket development programme, started in 1988, is aiming at the selection by 1995 of two DEMO-relevant blanket lines to be tested in ITER. At present, four lines of blanket are under development, two of them using solid and the other two liquid breeder materials. As far as liquid breeders are concerned, two lines of blankets have been selected within the European Union, the water-cooled lithium-lead (the eutectic Pb-17Li) blankets and the dual-coolant Pb-17Li blankets. Designs have been developed considering an agreed set of DEMO specifications, such as, for instance, a fusion power of 2,200 MW, a neutron wall-loading of 2MW/m{sup 2}, a life-time of 20,000 hours, and the use of martensitic steel as a structural material. Moreover, an experimental program has been set up in order to address the main critical issues for each line. The present paper gives an overview of both design and experimental activities within the European Union concerning these two lines of liquid breeder blankets.

  20. Progress in tritium retention and release modeling for ceramic breeders

    SciTech Connect (OSTI)

    Raffray, A.R.; Federici, G. [ITER Joint Work Site, Garching (Germany); Billone, M.C. [Argonne National Lab., IL (United States); Tanaka, S. [Univ. of Tokyo (Japan). Dept. of Quantum Engineering and Systems Science

    1994-07-11T23:59:59.000Z

    Tritium behavior in ceramic breeder blankets is a key design issue for this class of blanket because of its impact on safety and fuel self-sufficiency. Over the past 10-15 years, substantial theoretical and experimental efforts have been dedicated world-wide to develop a better understanding of tritium transport in ceramic breeders. Models that are available today seem to cover reasonably well all the key physical transport and trapping mechanisms. They have allowed for reasonable interpretation and reproduction of experimental data and have helped in pointing out deficiencies in material property data base, in providing guidance for future experiments, and in analyzing blanket tritium behavior. This paper highlights the progress in tritium modeling over the last decade. Key tritium transport mechanisms are briefly described along with the more recent and sophisticated models developed to help understand them. Recent experimental data are highlighted and model calibration and validation discussed. Finally, example applications to blanket cases are shown as illustration of progress in the prediction of ceramic breeder blanket tritium inventory.

  1. STUDENT MENTAL HEALTH POLICY

    E-Print Network [OSTI]

    Martin, Ralph R.

    pastoral care as well as teaching Take positive steps to promote students well-being Ensure the health students to make contact with their GP or local mental health services. The Duty of Care and Negligence1 STUDENT MENTAL HEALTH POLICY Revised January 2013 #12;2 A. INTRODUCTION 1. Context Widening

  2. Helium-cooled, FLiBe-breeder, beryllium-multiplier blanket for MINIMARS

    SciTech Connect (OSTI)

    Moir, R.W.; Lee, J.D.

    1986-06-01T23:59:59.000Z

    We adapted the helium-cooled, FLiBe-breeder blanket to the commercial tandem-mirror fusion-reactor design, MINIMARS. Vanadium was used to achieve high performance from the high-energy-release neutron-capture reactions and from the high-temperature operation permitted by the refractory property of the material, which increases the conversion efficiency and decreases the helium-pumping power. Although this blanket had the highest performance among the MINIMARS blankets designs, measured by Mn/sub th/ (blanket energy multiplication times thermal conversion efficiency), it had a cost of electricity (COE) 18% higher than the University of Wisconsin (UW) blanket design (42.5 vs 35.9 mills/kW.h). This increased cost was due to using higher-cost blanket materials (beryllium and vanadium) and a thicker blanket, which resulted in higher-cost central-cell magnets and the need for more blanket materials. Apparently, the high efficiency does not substantially affect the COE. Therefore, in the future, we recommend lowering the helium temperature so that ferritic steel can be used. This will result in a lower-cost blanket, which may compensate for the lower performance resulting from lower efficiency.

  3. A review of experiments and results from the transient reactor test (TREAT) facility.

    SciTech Connect (OSTI)

    Deitrich, L. W.

    1998-07-28T23:59:59.000Z

    The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop.

  4. First-wall and blanket engineering development for magnetic-fusion reactors

    SciTech Connect (OSTI)

    Baker, C.; Herman, H.; Maroni, V.; Turner, L.; Clemmer, R.; Finn, P.; Johnson, C.; Abdou, M.

    1981-01-01T23:59:59.000Z

    A number of programs in the USA concerned with materials and engineering development of the first wall and breeder blanket systems for magnetic-fusion power reactors are described. Argonne National Laboratory has the lead or coordinating role, with many major elements of the research and engineering tests carried out by a number of organizations including industry and other national laboratories.

  5. Fabrication Technological Development of the Oxide Dispersion Strengthened Alloy MA957 for Fast Reactor Applications

    SciTech Connect (OSTI)

    Hamilton, Margaret L.; Gelles, David S.; Lobsinger, Ralph J.; Johnson, Gerald D.; Brown, W. F.; Paxton, Michael M.; Puigh, Raymond J.; Eiholzer, Cheryl R.; Martinez, C.; Blotter, M. A.

    2000-02-28T23:59:59.000Z

    A significant amount of effort has been devoted to determining the properties and understanding the behavior of the alloy MA957 to define its potential usefulness as a cladding material in the fast breeder reactor program. The numerous characterization and fabrication studies that were conducted are documented in this report.

  6. MENTAL ILLNESS in the CLASSROOM

    E-Print Network [OSTI]

    MENTAL ILLNESS in the CLASSROOM: How Educators Can Help Students Succeed Ingle International cares with parents or health care practitioners about a student's mental health issue. Make your first step informing to start: · Canadian Mental Health Association: http://bit.ly/bmDylH · Centre for Addiction and Mental

  7. Advanced reactor safety research quarterly report, October-December 1982. Volume 24

    SciTech Connect (OSTI)

    Not Available

    1984-04-01T23:59:59.000Z

    This report describes progress in a number of activities dealing with current safety issues relevant to both light water reactors (LWRs) and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  8. Loca study for a helium-cooled solid breeder design for ITER

    SciTech Connect (OSTI)

    Gorbis, Z.R.; Raffray, A.R.; Fujimura, K.; Jun, I.; Abdou, M.A.

    1989-03-01T23:59:59.000Z

    The analysis of thermal processes after a loss-of-coolant accident (LOCA) in a solid breeder blanket is important because of the first wall and solid breeder maximum allowable temperature constraints. The objective is to design for a LOCA so that following a LOCA, the maximum solid breeder and structure temperatures are less than the limit beyond which irreversible damage is done, which would lead to loss of investment. The temporal temperature profiles for the solid breeder and first wall regions of a helium-cooled solid breeder design for ITER were calculated based on afterheat values for adiabatic and non-adiabatic conditions and the results are presented in this paper. It is found that, for this design, even when excluding radiation to the cooled inboard, a LOCA can be recommended by energy removal through a flowing purge with a reasonable flow rate.

  9. Fusion Engineering and Design 81 (2006) 461467 An overview of dual coolant Pb17Li breeder first wall and blanket

    E-Print Network [OSTI]

    Abdou, Mohamed

    2006-01-01T23:59:59.000Z

    . Helium is used to cool the first wall and blanket structure, and the self-cooled breeder Pb­17Li.V. All rights reserved. Keywords: Blanket; Pb­17Li breeder; Dual-coolant; Helium-cooled; ITER coolant Pb­17Li liquid breeder (DCLL) blanket design, a con- cept that has been explored extensively

  10. Reactor control rod timing system. [LMFBR

    DOE Patents [OSTI]

    Wu, P.T.K.

    1980-03-18T23:59:59.000Z

    A fluid driven jet-edge whistle timing system is described for control rods of a nuclear reactor for producing real-time detection of the timing of each control rod in its scram operation. An important parameter in reactor safety, particularly for liquid metal fast breeder reactors (LMFBR), is the time deviation between the time the control rod is released and the time the rod actually reaches the down position. The whistle has a nearly pure tone signal with center frequency (above 100 kHz) far above the frequency band in which the energy of the background noise is concentrated. Each control rod can be fitted with a whistle with a different frequency so that there is no ambiguity in differentiating the signal from each control rod.

  11. Irradiation behavior of metallic fast reactor fuels

    SciTech Connect (OSTI)

    Pahl, R.G.; Porter, D.L.; Crawford, D.C.; Walters, L.C.

    1991-01-01T23:59:59.000Z

    Metallic fuels were the first fuels chosen for liquid metal cooled fast reactors (LMR's). In the late 1960's world-wide interest turned toward ceramic LMR fuels before the full potential of metallic fuel was realized. However, during the 1970's the performance limitations of metallic fuel were resolved in order to achieve a high plant factor at the Argonne National Laboratory's Experimental Breeder Reactor II. The 1980's spawned renewed interest in metallic fuel when the Integral Fast Reactor (IFR) concept emerged at Argonne National Laboratory. A fuel performance demonstration program was put into place to obtain the data needed for the eventual licensing of metallic fuel. This paper will summarize the results of the irradiation program carried out since 1985.

  12. AB INITIO STUDY OF ADVANCED METALLIC NUCLEAR FUELS FOR FAST BREEDER REACTORS

    SciTech Connect (OSTI)

    Landa, A; Soderlind, P; Grabowski, B; Turchi, P A; Ruban, A V; Vitos, L

    2012-04-23T23:59:59.000Z

    Density-functional formalism is applied to study the ground state properties of {gamma}-U-Zr and {gamma}-U-Mo solid solutions. Calculated heats of formation are compared with CALPHAD assessments. We discuss how the heat of formation in both alloys correlates with the charge transfer between the alloy components. The decomposition curves for {gamma}-based U-Zr and U-Mo solid solutions are derived from Ising-type Monte Carlo simulations. We explore the idea of stabilization of the {delta}-UZr{sub 2} compound against the {alpha}-Zr (hcp) structure due to increase of Zr d-band occupancy by the addition of U to Zr. We discuss how the specific behavior of the electronic density of states in the vicinity of the Fermi level promotes the stabilization of the U{sub 2}Mo compound. The mechanism of possible Am redistribution in the U-Zr and U-Mo fuels is also discussed.

  13. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    the ignition for a nuclear fission deflagration wave thatsteel, can tolerate. A nuclear fission deflagration wave caneither nuclear transmutation or another fission reaction.

  14. Seismic design technology for breeder reactor structures. Volume 4. Special topics in piping and equipment

    SciTech Connect (OSTI)

    Reddy, D.P.

    1983-04-01T23:59:59.000Z

    This volume is divided into five chapters: experimental verification of piping systems, analytical verification of piping restraint systems, seismic analysis techniques for piping systems with multisupport input, development of floor spectra from input response spectra, and seismic analysis procedures for in-core components. (DLC)

  15. Clinch River breeder reactor sodium fire protection system design and development

    SciTech Connect (OSTI)

    Foster, K.W.; Boasso, C.J.; Kaushal, N.N.

    1984-04-13T23:59:59.000Z

    To assure the protection of the public and plant equipment, improbable accidents were hypothesized to form the basis for the design of safety systems. One such accident is the postulated failure of the Intermediate Heat Transfer System (IHTS) piping within the Steam Generator Building (SGB), resulting in a large-scale sodium fire. This paper discusses the design and development of plant features to reduce the consequences of the accident to acceptable levels. Additional design solutions were made to mitigate the sodium spray contribution to the accident scenario. Sodium spill tests demonstrated that large sodium leaks can be safely controlled in a sodium-cooled nuclear power plant.

  16. Am. J. Phys. 51(1), Jan. 1983 Breeder reactors: A renewable energy source

    E-Print Network [OSTI]

    Laughlin, Robert B.

    in undiminished quantity at present costs for as long as the current relationship between the sun and Earth estimates of the world's uranium resources refer to quantities available at the current market price of about $40 per pound. At that price, uranium supply contributes about 0.2 cents/kW h to the cost

  17. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    the formation of solid lead oxides may lead to clogging ofexceeds this value, lead oxide (PbO) particles will form infor the formation of lead oxide is: Pb + O 2 ? PbO The

  18. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    of conventional LWR systems (PWR & BWRs), partly due to thethe margin to boiling in a PWR is ?15 ? C, while the coolantprimary heat exhangers of a PWR, in which borated water is

  19. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    Overview”. In: Proc. of ICAPP 2013, Seogwipo, Jeju-do, Southvoid worth”. In: Proc. of ICAPP 2011, Nice, France [118] A.System to JSFR”. In: Proc. of ICAPP 2008, Anaheim, CA, US (

  20. Piping support system for liquid-metal fast-breeder reactor

    DOE Patents [OSTI]

    Brussalis, Jr., William G. (Forward Township, Washington County, PA)

    1984-01-01T23:59:59.000Z

    A pipe support consisting of a rigid link pivotally attached to a pipe and an anchor, adapted to generate stress or strain in the link and pipe due to pipe thermal movement, which stress or strain can oppose further pipe movement and generally provides pipe support. The pipe support can be used in multiple combinations with other pipe supports to form a support system. This support system is most useful in applications in which the pipe is normally operated at a constant elevated or depressed temperature such that desired stress or strain can be planned in advance of pipe and support installation. The support system is therefore especially useful in steam stations and in refrigeration equipment.

  1. Thermal-performance study of liquid metal fast breeder reactor insulation

    SciTech Connect (OSTI)

    Shiu, Kelvin K.

    1980-09-01T23:59:59.000Z

    Three types of metallic thermal insulation were investigated analytically and experimentally: multilayer reflective plates, multilayer honeycomb composite, and multilayer screens. Each type was subjected to evacuated and nonevacuated conditions, where thermal measurements were made to determine thermal-physical characteristics. A variation of the separation distance between adjacent reflective plates of multilayer reflective plates and multilayer screen insulation was also experimentally studied to reveal its significance. One configuration of the multilayer screen insulation was further selected to be examined in sodium and sodium oxide environments. The emissivity of Type 304 stainless steel used in comprising the insulation was measured by employing infrared technology. A comprehensive model was developed to describe the different proposed types of thermal insulation. Various modes of heat transfer inherent in each type of insulation were addressed and their relative importance compared. Provision was also made in the model to allow accurate simulation of possible sodium and sodium oxide contamination of the insulation. The thermal-radiation contribution to heat transfer in the temperature range of interest for LMFBR's was found to be moderate, and the suppression of natural convection within the insulation was vital in preserving its insulating properties. Experimental data were compared with the model and other published results. Moreover, the three proposed test samples were assessed and compared under various conditions as viable LMFBR thermal insulations.

  2. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    Fuel assembly Inward thermal bow Flux shape Rigid support (High-gradient assembly Outward thermal bow Core center Low-gradient assembly Outward thermal bow Flux shape Rigid

  3. Contribution of Clinch River Breeder Reactor plant design and development to the LMFBR fuel cycle

    SciTech Connect (OSTI)

    Riley, D.R.; Dickson, P.W.

    1981-01-01T23:59:59.000Z

    This paper describes how the CRBRP development and CRBRP focus of the LMFBR base technology program have led to advances in the state of the art in physics, thermal-hydraulics, structural analysis, core restraint, seismic analysis, and analysis of hypothetical core-disruptive accident energetics, all of which have been incorporated through disciplined engineering into the final CRBRP design. The total development in the US of fuels and materials, the analytical advances made on CRBRP design, and the incorporation of the latest experimental results into that design have put the US technology in general and the CRBRP design in particular at the forefront of technology. This has placed the US in a position to develop the most favorable LMFBR fuel cycle.

  4. Liquid-metal fast-breeder reactors: Preliminary safety and environmental information document. Volume VI

    SciTech Connect (OSTI)

    Not Available

    1980-01-01T23:59:59.000Z

    Information is presented concerning LMFBR design characteristics; uranium-plutonium/uranium recycle homogeneous core; uranium-plutonium/uranium spiked recycle heterogeneous core; uranium-plutonium/uranium spiked recycle homogeneous core; uranium-plutonium/thorium spiked recycle heterogeneous core; uranium-plutonium/thorium spiked recycle homogeneous core; thorium-plutonium/thorium spiked recycle homogeneous core; denatured uranium-233/thorium cycle homogeneous core; safety consideration for the LMFBR; and environmental considerations.

  5. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    performance code”. In: Nuclear Technology 89 (1990), pp.feedback”. In: Nuclear Technology 160.3 (2007), pp. 257–Future BN-800 Based Nuclear Technology”. In: Atomic Energy

  6. Gas-cooled fast breeder reactor fuel element thermal-hydraulic investigations : final report

    E-Print Network [OSTI]

    Eaton, Thomas Eldon

    1975-01-01T23:59:59.000Z

    Experimental and analytical work was performed to determine the influence of rod surface roughening on the thermal-hydraulic behavior of rod array type, nuclear fuel elements. Experimental data was obtained using a ...

  7. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    Florida, USA (1997). [34] P. HEJZLAR et. al. “Traveling WaveLaboratory, 2013. [76] P. HEJZLAR and C. B. DAVIS. “studies on TWRs Yarsky, Hejzlar, Driscoll (MIT) Gas-cooled

  8. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    The nuclear Doppler effectcalled the (nuclear) Doppler effect. Cross-sections for allare subject to the Doppler effect. While the total cross-

  9. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    and F. HEIDET. “Energy sustainability and economic stabilityEnhanced Nuclear Energy Sustainability”. In: Sustainabilityfor nuclear energy sustainability”. In: Energy Conversion

  10. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    with burnup of a depleted-uranium fueled sodium-cooled B&Bwith burnup of a depleted-uranium fueled sodium-cooled B&Bbalance integral of a depleted-uranium fueled sodium-cooled

  11. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    by GHG-emissions, coal power plant emissions such as sulfur2]. In Europe, coal power plant emissions cause an estimated

  12. Safety and core design of large liquid-metal cooled fast breeder reactors

    E-Print Network [OSTI]

    Qvist, Staffan Alexander

    2013-01-01T23:59:59.000Z

    pitch 30 x Rod Diameter Radial reflectors 2 assembly rows (?mechanical movement of a radial reflector [16]. In the early

  13. Public comment sought on final end state of Experimental Breeder Reactor-II

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible forPortsmouth/Paducah47,193.70COMMUNITY AEROSOL INLET8, 2012Media Contacts: Danielle

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

    SciTech Connect (OSTI)

    Soldevilla, M.; Salmons, S.; Espinosa, B. [CEA-Saclay, CEA/DEN/DANS/DM2S/SERMA, 91191 Gif-sur-Yvette (France); Clanet, M.; Boudin, X. [CEA-Bruyeres-le-Chatel, 91297 Arpajon (France)

    2013-07-01T23:59:59.000Z

    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)

  15. A Solid breeder tokamak blanket designed for failure mode operation

    E-Print Network [OSTI]

    Chen, Franklin Fun Kun

    1977-01-01T23:59:59.000Z

    The objective of this study was to evaluate a new concept for a Tokamak type fusion reactor blanket. The design was based on using a packed bed of lithium aluminate as the breeding material with helium gas cooling. The ...

  16. Students & Mental Health Resource Pack

    E-Print Network [OSTI]

    Stevenson, Mark

    Students & Mental Health Resource Pack Produced by - www.rethink.org/at-ease/ SHEFFIELD EARLY://clik.to/eis NORTH EAST SHEFFIELD Northlands Community Health Centre, Southey Hill, Sheffield S5 8BE Tel: 0114 is severe mental illness? 1.4 Treatment and prognosis What is mental health awareness? 2.1 Introduction 2

  17. A Thermal Discrete Element Analysis of EU Solid Breeder Blanket subjected to Neutron Irradiation

    E-Print Network [OSTI]

    Yixiang Gan; Francisco Hernandez; Dorian Hanaor; Ratna Annabattula; Marc Kamlah; Pavel Pereslavtsev

    2014-06-17T23:59:59.000Z

    Due to neutron irradiation, solid breeder blankets are subjected to complex thermo-mechanical conditions. Within one breeder unit, the ceramic breeder bed is composed of spherical-shaped lithium orthosilicate pebbles, and as a type of granular material, it exhibits strong coupling between temperature and stress fields. In this paper, we study these thermo-mechanical problems by developing a thermal discrete element method (Thermal-DEM). This proposed simulation tool models each individual ceramic pebble as one element and considers grain-scale thermo-mechanical interactions between elements. A small section of solid breeder pebble bed in HCPB is modelled using thousands of individual pebbles and subjected to volumetric heating profiles calculated from neutronics under ITER-relevant conditions. We consider heat transfer at the grain-scale between pebbles through both solid-to-solid contacts and the interstitial gas phase, and we calculate stresses arising from thermal expansion of pebbles. The overall effective conductivity of the bed depends on the resulting compressive stress state during the neutronic heating. The thermal-DEM method proposed in this study provides the access to the grain-scale information, which is beneficial for HCPB design and breeder material optimization, and a better understanding of overall thermo-mechanical responses of the breeder units under fusion-relevant conditions.

  18. Fusion reactor blanket-main design aspects

    SciTech Connect (OSTI)

    Strebkov, Yu.; Sidorov, A.; Danilov, I. [Research and Development Inst. of Power Engineering, Moscow (Russian Federation)

    1994-12-31T23:59:59.000Z

    The main function of the fusion reactor blanket is ensuring tritium breeding and radiation shield. The blanket version depends on the reactor type (experimental, DEMO, commercial) and its parameters. Blanket operation conditions are defined with the heat flux, neutron load/fluence, cyclic operation, dynamic heating/force loading, MHD effects etc. DEMO/commercial blanket design is distinguished e.g. by rather high heat load and neutron fluence - up to 100 W/cm{sup 2} and 7 MWa/m{sup 2} accordingly. This conditions impose specific requirements for the materials, structure, maintenance of the blanket and its most loaded components - FW and limiter. The liquid Li-Pb eutectic is one of the possible breeder for different kinds of blanket in view of its advantages one of which is the blanket convertibility that allow to have shielding blanket (borated water) or breeding one (Li-Pb eutectic). Using Li-Pb eutectic for both ITER and DEMO blankets have been considered. In the conceptual ITER design the solid eutectic blanket was carried out. The liquid eutectic breeder/coolant is suggested also for the advanced (high parameter) blanket.

  19. Engineering Scaling Requirements for Solid Breeder Blanket Testing

    SciTech Connect (OSTI)

    Ying, A.; Sharafat, S.; Youssef, M.; An, J.; Hunt, R.; Rainsberry, P.; Abdou, M. [University of California, Los Angeles (United States)

    2005-05-15T23:59:59.000Z

    An engineering scaling process is applied to the solid breeder ITER TBM designs in accordance with the testing objectives of validating the design tools and the database, and evaluating blanket performance under prototypical operating conditions. The goal of scaling is to ensure that changes in structural response and performance caused by changes in size and operating conditions do not reduce the usefulness of the tests. Initially, constitutive equations are applied to lay out the basic operating and design parameters that dominate blanket phenomena. The suitability of these similarity criteria for the TBM design is then confirmed by comparing finite element predictions of prototype and scale model responses. The TBM design also takes into account the need to check the codes and data for future design use. Specifically, predictability of tritium production and nuclear heating rates in a complex geometry, tritium release and permeation characteristics under fusion environments belong to this category. We conclude that this engineering scaling design process has maximized the value of ITER testing.

  20. ISOTOPES

    E-Print Network [OSTI]

    Lederer, C. Michael

    2013-01-01T23:59:59.000Z

    §fissile materials in fast breeder reactors currently underwater reactor, FBR = fast breeder reactor. The band belowinc 1 udes heavy-water reactors, fast breeders, and 11

  1. Development of pyro-processing technology for thorium-fuelled molten salt reactor

    SciTech Connect (OSTI)

    Uhlir, J.; Straka, M.; Szatmary, L. [Nuclear Research Inst. ReZ Plc, ReZ 130, Husinec - CZ-250 68 (Czech Republic)

    2012-07-01T23:59:59.000Z

    The Molten Salt Reactor (MSR) is classified as the non-classical nuclear reactor type based on the specific features coming out from the use of liquid fuel circulating in the MSR primary circuit. Other uniqueness of the reactor type is based on the fact that the primary circuit of the reactor is directly connected with the on-line reprocessing technology, necessary for keeping the reactor in operation for a long run. MSR is the only reactor system, which can be effectively operated within the {sup 232}Th- {sup 233}U fuel cycle as thorium breeder with the breeding factor significantly higher than one. The fuel cycle technologies proposed as ford the fresh thorium fuel processing as for the primary circuit fuel reprocessing are pyrochemical and mainly fluoride. Although these pyrochemical processes were never previously fully verified, the present-day development anticipates an assumption for the successful future deployment of the thorium-fuelled MSR technology. (authors)

  2. New York City Department of Health & Mental Hygiene (DOHMH) Mental Health Scholarship Program

    E-Print Network [OSTI]

    Qiu, Weigang

    New York City Department of Health & Mental Hygiene (DOHMH) Mental Health Scholarship Program One and Mental Hygiene Bureau of Mental Health (BMH) or Bureau of Children, Youth and Families (CYF). Individuals

  3. Fusion Engineering and Design 41 (1998) 561567 Combination of a self-cooled liquid metal breeder blanket with

    E-Print Network [OSTI]

    1998-01-01T23:59:59.000Z

    , the cost of electricity. Self-cooled liquid metal breeder blankets have a high potential to meetFusion Engineering and Design 41 (1998) 561­567 Combination of a self-cooled liquid metal breeder blanket with a gas turbine power conversion system S. Malang a, *, H. Schnauder a , M.S. Tillack b

  4. "Territories for Mental Health": An empowerment process

    E-Print Network [OSTI]

    Boyer, Edmond

    Hospitals, Local Mental Health Departments, Social Care Department; Local Authorities) Azienda Ospedaliera"Territories for Mental Health": An empowerment process through peer collaboration among community district stakeholders Fabio Lucchi Department of Mental Health Azienda Ospedaliera Desenzano d/G (Brescia

  5. MENTAL HEALTH and INTERNATIONAL STUDENTS

    E-Print Network [OSTI]

    MENTAL HEALTH and INTERNATIONAL STUDENTS: What Educators Need to Know Ingle International cares not adequately researched, it is well accepted by health care professionals that early intervention could about you and your students www.studyinsured.com #12;www.studyinsured.comMental Health and International

  6. Mental Health Clinic Intake Assessment

    E-Print Network [OSTI]

    Weiblen, George D

    Mental Health Clinic Intake Assessment Welcome to the Mental Health Clinic at Boynton Health Clinic. The Medical Social Worker can provide you with resources if you require an evaluation-- 612 Clinic utilizes a short-term model of psychotherapy. This means that we are able to offer eleven

  7. Breeder-in-tube design for a helium-cooled Li/sub 2/O tokamak blanket

    SciTech Connect (OSTI)

    Billone, M.C.; Jung, J.; Liu, Y.Y.; Smith, D.L.

    1986-01-01T23:59:59.000Z

    Of the solid breeder designs considered in the Blanket Comparison and Selection Study (BCSS), the lithium oxide breeder with helium coolant and ferritic steel (HT-9) structural material received the highest overall ranking for both tokamak and tandem mirror systems in terms of engineering, economics, safety, and R and D requirements. The BCSS blanket surrounding the fusion plasma consists of a number of thin breeder plates externally cooled by flowing helium and internally purged of tritium by a separate helium stream. A detailed review of this design indicated that significant improvements would be realized in the areas of tritium breeding, blanket thickness, blanket energy multiplication, power-conversion efficiency, breeder temperature window, and geometrical integrity of the coolant and purge paths by using a neutron multiplier (beryllium), a higher temperature structural material (vanadium-based alloy), and a tube geometry. The neutronics, thermal-hydraulics, tritium recovery, and structural performance characteristics of this innovative solid breeder design are discussed in this paper.

  8. THERMAL ANALYSIS OF A SOLID BREEDER TBM UNDER ITER OPERATIONAL CONDITIONS A. Abou-Sena, A. Ying, M. Youssef, M. Abdou

    E-Print Network [OSTI]

    Abdou, Mohamed

    @engineering.ucla.edu The Quarter Port Submodule (QPS) was proposed as a Solid Breeder (SB) Test Blanket Module under the US program candidates to serve as tritium breeder for the SB blankets. The SB and Be, in form of pebble bedsTHERMAL ANALYSIS OF A SOLID BREEDER TBM UNDER ITER OPERATIONAL CONDITIONS A. Abou-Sena, A. Ying, M

  9. SOLID BREEDER BLANKET OPTION FOR THE ITER CONCEPTUAL DESIGN* Y. Cohar, H. Attaya, M.C. Billone, P. Finn, S. Majumdar, L.R. Turner

    E-Print Network [OSTI]

    Raffray, A. René

    SOLID BREEDER BLANKET OPTION FOR THE ITER CONCEPTUAL DESIGN* Y. Cohar, H. Attaya, M.C. Billone, P in this paper. lntroduction A solid-breeder water-cooled blanket option was developed for ITER based lithium-6 enrichment reduces the solid breeder volume required in the blanket and the total tritium

  10. A method for measurement of delayed neutron parameters for liquid-metal-cooled power reactors

    SciTech Connect (OSTI)

    Vilim, R.B. [Argonne National Lab., IL (United States); Brock, R.W. [Babcock and Wilcox, Lynchburg, VA (United States)

    1996-06-01T23:59:59.000Z

    The trend toward increased reliance on passive features for power reactor safety makes it important to obtain the characteristics of the reactor system from measurements on the system. A method is described for solving for the delayed neutron parameters in a liquid-metal power reactor by fitting an analytic solution of the point-kinetics equations to the flux die-away from a dropped rod in an initially critical core. The method includes treatment of those conditions found in a power reactor that depart from those in a critical assembly experiment. These include a comparatively long rod drop time and a detector signal that instead of providing an integrated count rate is a sampled data signal proportional to the instantaneous fission power. The delayed neutron parameter values calculated from a rod drop experiment in the Experimental Breeder Reactor II are in agreement with values calculated using first principles and knowledge of core material composition and nuclear cross sections.

  11. System for fuel rod removal from a reactor module

    DOE Patents [OSTI]

    Matchett, Richard L. (Bethel Park, PA); Roof, David R. (North Huntingdon, PA); Kikta, Thomas J. (Pittsburgh, PA); Wilczynski, Rosemarie (McKees Rocks, PA); Nilsen, Roy J. (Pittsburgh, PA); Bacvinskas, William S. (Bethel Park, PA); Fodor, George (Pittsburgh, PA)

    1990-01-01T23:59:59.000Z

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system.

  12. System for fuel rod removal from a reactor module

    DOE Patents [OSTI]

    Matchett, R.L.; Fodor, G.; Kikta, T.J.; Bacvinsicas, W.S.; Roof, D.R.; Nilsen, R.J.; Wilczynski, R.

    1988-07-28T23:59:59.000Z

    A robotic system for remote underwater withdrawal of the fuel rods from fuel modules of a light water breeder reactor includes a collet/grapple assembly for gripping and removing fuel rods in each module, which is positioned by use of a winch and a radial support means attached to a vertical support tube which is mounted over the fuel module. A programmable logic controller in conjunction with a microcomputer, provides control for the accurate positioning and pulling force of the rod grapple assembly. Closed circuit television cameras are provided which aid in operator interface with the robotic system. 7 figs.

  13. Accepted Manuscript Numerical Characterization of Thermo-mechanical Performance of Breeder

    E-Print Network [OSTI]

    Abdou, Mohamed

    -mechanical properties, mathematical and computational method. #12;ACCEPTED MANUSCRIPT 3 1. Introduction Ceramic breeder Mechanical & Aerospace Engineering Department, UCLA, Los Angeles, CA 90095-1597 Tel: (310) 794-4452 Fax: (310, and Mohamed Abdou Mechanical and Aerospace Engineering Dept., UCLA, Los Angeles, CA 90095-1597, USA an

  14. Responding to environmental change: plastic responses1 vary little in a synchronous breeder.2

    E-Print Network [OSTI]

    Nussey, Dan

    The impact of environmental change on animal populations is strongly influenced by the10 ability. 2002).5 Determining how individuals base key life-history decisions on environmental cues is6 therefore1 Responding to environmental change: plastic responses1 vary little in a synchronous breeder.2 3

  15. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1981-01-01T23:59:59.000Z

    Chapters are presented concerning energy from nuclear fission; nuclear reactions and radiations; diffusion and slowing-down of neutrons; principles of reactor analysis; nuclear reactor kinetics and control; energy removal; non-fuel reactor materials; the reactor fuel system; radiation protection and environmental effects; nuclear reactor shielding; nuclear reactor safety; and power reactor systems.

  16. Austenitic alloy and reactor components made thereof

    DOE Patents [OSTI]

    Bates, John F. (Ogden, UT); Brager, Howard R. (Richland, WA); Korenko, Michael K. (Wexford, PA)

    1986-01-01T23:59:59.000Z

    An austenitic stainless steel alloy is disclosed, having excellent fast neutron irradiation swelling resistance and good post irradiation ductility, making it especially useful for liquid metal fast breeder reactor applications. The alloy contains: about 0.04 to 0.09 wt. % carbon; about 1.5 to 2.5 wt. % manganese; about 0.5 to 1.6 wt. % silicon; about 0.030 to 0.08 wt. % phosphorus; about 13.3 to 16.5 wt. % chromium; about 13.7 to 16.0 wt. % nickel; about 1.0 to 3.0 wt. % molybdenum; and about 0.10 to 0.35 wt. % titanium.

  17. Mental content, holism and communication 

    E-Print Network [OSTI]

    Pollock, Joanna Katharine Mary

    2014-07-01T23:59:59.000Z

    In this project, I defend a holistic, internalist conceptual-role theory of mental content (‘Holism’, for short). The account of communicative success which must be adopted by the Holist is generally thought to be ...

  18. Rural Schools' Mental Health Needs

    E-Print Network [OSTI]

    Lee, Steven W.; Lohmeier, Jill H.; Niileksela, Christopher Robert; Oeth, Jessica

    2009-01-01T23:59:59.000Z

    Rural schools often can not provide the same access to mental health service as schools in larger population areas can.. Understanding the implications of these sometimes limited services is important in overcoming barriers to adequate services...

  19. Bioconversion reactor

    DOE Patents [OSTI]

    McCarty, Perry L. (Stanford, CA); Bachmann, Andre (Palo Alto, CA)

    1992-01-01T23:59:59.000Z

    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.

  20. Catalytic reactor

    DOE Patents [OSTI]

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

    2009-03-10T23:59:59.000Z

    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.

  1. New York City Department of Health & Mental Hygiene (DOHMH) Mental Health Scholarship Program

    E-Print Network [OSTI]

    Qiu, Weigang

    New York City Department of Health & Mental Hygiene (DOHMH) Mental Health Scholarship Program One Division of Mental Hygiene? (Ask your Supervisor how to find out this information for your program.) Yes

  2. Parental involvement in mental health services for diverse youth

    E-Print Network [OSTI]

    Liang, June

    2010-01-01T23:59:59.000Z

    M. (2002). Mental health services for Latino adolescentsfamilies to mental health services: The role of the familyoutpatient mental health services. New York: Human Sciences

  3. Psychoeducation and parental engagement in mental health services for youth

    E-Print Network [OSTI]

    Martinez, Jonathan

    2013-01-01T23:59:59.000Z

    1995). Children’s mental health service use across serviceinto and through mental health services for children andSchoolbased mental health services in the United States:

  4. Performance indicators for public mental healthcare: A systematic international inventory

    E-Print Network [OSTI]

    Lauriks, Steve; Buster, Marcel CA; de Wit, Matty AS; Arah, Onyebuchi A; Klazinga, Niek S

    2012-01-01T23:59:59.000Z

    for mental health services: values, accountability,Center for Mental Health Services, Substance Abuseand Mental Health Services Administration; 2003. 59. Edlund

  5. An evaluation of sesame oil meal and Peruvian fish for egg-strain and broiler-type breeder hens

    E-Print Network [OSTI]

    Smith, Edwin B

    1964-01-01T23:59:59.000Z

    meal as protein sources for egg-strain hens . . . . . . . . ~ ~ . ~ 16 Composition of. laying hen diets . . . . . . ~ . 17, 18 Calculated percent protein and certain amino acid content in laying hen diets from analyses of ingredients... . . . , . . . . . . . . . ~ ~ , 19 Experimental design . . . . . . . . . . . ~ . . 20 Composition of broiler-type breeder diets ~ . . 21 Calculated protein and certain amino acid con- tent in broiler-type breeder hen diets from analy- ses of ingredients...

  6. Neutronic reactor

    DOE Patents [OSTI]

    Wende, Charles W. J. (Augusta, GA); Babcock, Dale F. (Wilmington, DE); Menegus, Robert L. (Wilmington, DE)

    1983-01-01T23:59:59.000Z

    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.

  7. Development of a neutronics calculation method for designing commercial type Japanese sodium-cooled fast reactor

    SciTech Connect (OSTI)

    Takeda, T.; Shimazu, Y.; Hibi, K.; Fujimura, K. [Research Inst. of Nuclear Engineering, Univ. of Fukui, 1cho-me 2gaiku 4, Kanawa-cho, Tsuruga-shi, Fukui 914-0055 (Japan)

    2012-07-01T23:59:59.000Z

    Under the R and D project to improve the modeling accuracy for the design of fast breeder reactors the authors are developing a neutronics calculation method for designing a large commercial type sodium- cooled fast reactor. The calculation method is established by taking into account the special features of the reactor such as the use of annular fuel pellet, inner duct tube in large fuel assemblies, large core. The Verification and Validation, and Uncertainty Qualification (V and V and UQ) of the calculation method is being performed by using measured data from the prototype FBR Monju. The results of this project will be used in the design and analysis of the commercial type demonstration FBR, known as the Japanese Sodium fast Reactor (JSFR). (authors)

  8. The effect of water on tritium release behavior from solid breeder candidates

    SciTech Connect (OSTI)

    Suematsu, K.; Nishikawa, M.; Fukada, S.; Kinjyo, T.; Koyama, T.; Yamashita, N. [Graduate School of Engineering Science, Kyushu Univ., Fukuoka, 812-8581 (Japan)

    2008-07-15T23:59:59.000Z

    The authors have made a tritium release model to represent the release behavior of bred tritium from solid breeder materials using a series of studies. It has been observed that a large amount of adsorbed water and water produced by water formation reaction are released to the purge gas even though dry purge gas with hydrogen is introduced to solid breeder materials. According to our tritium release model, the presence of water in the purge gas and surface water on the material has a large effect on the tritium release behavior. In this study, the authors quantified the amount of adsorbed water and the capacity of the water formation reaction for various solid breeder materials (Li{sub 2}TiO{sub 3}, Li{sub 4}SiO{sub 4}, Li{sub 2}ZrO{sub 3}, LiAlO{sub 2}). The effect of surface water on the chemical form of tritium released from the LiAlO{sub 2} blanket is also discussed in this study. (authors)

  9. Thermodynamic considerations for the use of vanadium alloys with ceramic breeder materials

    SciTech Connect (OSTI)

    Johnson, C.E.; Johnson, I.; Kopasz, J.P.

    1995-12-31T23:59:59.000Z

    Fusion energy is considered to be an attractive energy form because of its minimal environmental impact. In order to maintain this favorable status, every effort needs to be made to use low activation materials wherever possible. The tritium breeder blanket is a focal point of system design engineers who must design environmentally attractive blankets through the use of low activation materials. Of the several candidate lithium-containing ceramics being considered for use in the breeder blanket, Li{sub 2}O, Li{sub 2}TiO{sub 3}, are attractive choices because of their low activation. Also, low activation materials like the vanadium alloys are being considered for use as structural materials in the blanket. The suitability of vanadium alloys for containment of lithium ceramics is the subject of this study. Thermodynamic evaluations are being used to estimate the compatibility and stability of candidate ceramic breeder materials (Li{sub 2}O, Li{sub 2}TiO{sub 3}, and Li{sub 2}ZrO{sub 3}) with vanadium and vanadium alloys. This thermodynamic evaluation will focus first on solid-solid interactions. As a tritium breeding blanket will use a purge gas for tritium recovery, gas-solid systems will also receive attention.

  10. The Application of Structural Materials Data From the BN-350 Fast Reactor to Life Extension of Light Water Reactors

    SciTech Connect (OSTI)

    Romanenko, O.G. [Nuclear Technology Safety Center, Liza Chaikina 4, Almaty 050020 (Kazakhstan); Kislitsin, S.B.; Maksimkin, O.P. [Institute of Nuclear Physics, 1 Ibragimova St., Almaty, 050032 (Kazakhstan); Shiganakov, Sh.B.; Chumakov, Ye.V. [Kazakhstan Atomic Energy Committee, Liza Chaikina 4, Almaty (Kazakhstan); Dumchev, I.V. [MAEC Kazatomprom, Aktau, 130000 (Kazakhstan)

    2006-07-01T23:59:59.000Z

    This paper describes the results of investigations of 08Cr16Ni11Mo3 (AISI 316 steel analogue) austenitic stainless steel irradiated in BN-350 breeder reactor at irradiation conditions close to that for Light Water Reactor (LWR) Internals. The pores were found in 08Cr16Ni11Mo3 steel irradiated at temperature 280 deg. C up to rather low damage 1.3 dpa and with dose rate 3.9 x 10{sup -9} dpa/s. There were obtained dose rate dependencies of yield strength, ultimate strength and ductility for 08Cr16Ni11Mo3 steel irradiated up to 7-13 dpa at 302-311 deg. C. These dependencies show a decrease in both yield strength and ultimate strength when dose rate decreases. There was observed an apparent decrease in total elongation when dose rate decreases, which was presumably connected with the pores formation in the steel at low dose rates. (authors)

  11. US ITER (International Thermonuclear Experimental Reactor) shield and blanket design activities

    SciTech Connect (OSTI)

    Baker, C.C.

    1988-08-01T23:59:59.000Z

    This paper summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. Primary tasks carried out during the past year include design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components, and issues regarding structural materials for an ITER device. The blanket concepts considered are the aqueous/Li salt solution, a water-cooled, solid breeder blanket, a helium-cooled, solid-breeder blanket, a blanket cooled by helium containing lithium-bearing particulates, and a blanket concept based on breeding tritium from He/sup 3/. 1 ref., 2 tabs.

  12. Advanced reactor safety research. Quarterly report, April-June 1982. Volume 22

    SciTech Connect (OSTI)

    none,

    1983-10-01T23:59:59.000Z

    Overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2) understanding risk-significant accident sequences, (3) developing and verifying models used in safety assessments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the United States without appropriate consideration being given to their effects on health and safety. This report describes progress in a number of activities dealing with current safety issues relevant to both light water and breeder reactors. The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and verification of the complex computer simulation models and codes used in accident analysis and licensing reviews. Such a program must include the development of analytical models, verified by experiment, which can be used to predict reactor and safety system performance under a broad variety of abnormal conditions. Current major emphasis is focused on providing information to NRC relevant to (1) its deliberations and decisions dealing with severe LWR accidents, and (2) its safety evaluation of the proposed Clinch River Breeder Reactor.

  13. Structure of processes in flow reactor and closed reactor: Flow reactor

    E-Print Network [OSTI]

    Greifswald, Ernst-Moritz-Arndt-Universität

    Structure of processes in flow reactor and closed reactor: Flow reactor Closed reactor Active Zone -- chemical quasi- equilibria, similarity principles and macroscopic kinetics", in: Lectures on Plasma Physics

  14. j\\TOMKI Report All ( 1983) DIREdr TEST OF THE TIME-INOCPENDENCE OF f1JNIlA.MENTAL

    E-Print Network [OSTI]

    Shlyakhter, Ilya

    .}.!TS USING THE OKLO NATURAL REACTOR* A. I. Shlyakhter Leningrad Nuclear Physics Institute, Gatchinaj\\TOMKI Report All ( 1983) DIREdr TEST OF THE TIME- INOCPENDENCE OF f1JNIlA.MENTAL NUCLEAR CONSTP the important natural cons tants of cosIOOlogy and atomic theory are connected by simple Imthematical rela

  15. Seismic design technology for breeder reactor structures. Volume 2. Special topics in soil/structure interaction analyses

    SciTech Connect (OSTI)

    Reddy, D.P.

    1983-04-01T23:59:59.000Z

    This volume is divided into six chapters: definition of seismic input ground motion, review of state-of-the-art procedures, analysis guidelines, rock/structure interaction analysis example, comparison of two- and three-dimensional analyses, and comparison of analyses using FLUSH and TRI/SAC Codes. (DLC)

  16. Numerical simulation of intermediate heat exchanger of the liquid metal fast breeder reactor using COMMIX-1B

    E-Print Network [OSTI]

    Saleh, Habeeb H.

    1990-01-01T23:59:59.000Z

    , conditioning, storing, installing in and removing from the core all routinely removable core components and test assemblies, and for storing irradiated fuel. Limited examination and packaging capabilities (for offsite shipment) are also provided. III. 2...

  17. Hybrid adsorptive membrane reactor

    DOE Patents [OSTI]

    Tsotsis, Theodore T. (Huntington Beach, CA); Sahimi, Muhammad (Altadena, CA); Fayyaz-Najafi, Babak (Richmond, CA); Harale, Aadesh (Los Angeles, CA); Park, Byoung-Gi (Yeosu, KR); Liu, Paul K. T. (Lafayette Hill, PA)

    2011-03-01T23:59:59.000Z

    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.

  18. Nuclear reactor engineering

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1982-07-01T23:59:59.000Z

    A book is reviewed which emphasizes topics directly related to the light water reactor power plant and the fast reactor power system. Current real-world problems are addressed throughout the text, and a chapter on safety includes much of the postThree Mile Island impact on operating systems. Topics covered include Doppler broadening, neutron resonances, multigroup diffusion theory, reactor kinetics, reactor control, energy removal, nonfuel materials, reactor fuel, radiation protection, environmental effects, and reactor safety.

  19. Measurement and monitoring of tritium and other critical issues in Lead Lithium Ceramic Breeder (LLCB)

    SciTech Connect (OSTI)

    Tangri, V. K.; Mohan, S. [Heavy Water Div., Bhabha Atomic Research Centre (India); Narayanan, A.; Narayan, K. K. [Radiation Safety Systems Div., Bhabha Atomic Research Centre (India)

    2008-07-15T23:59:59.000Z

    A new Indian concept involving a lead lithium ceramic breeder is being explored. LLCB based tritium blanket modules require tritium extraction from lead-lithium as well as from helium purge gas. This paper addresses the concept of efficiency enhancement using high surface area, low-pressure drop structured gas liquid contactors for tritium extraction from the lead lithium. Conceptual flow schemes for both loops are discussed and critical issues are highlighted. Tritium monitoring systems (TMS) for measurement and monitoring of tritium is also dealt. A fast responding tritium monitor has also been developed for in situ measurement of tritium in water or gas form. It has been tested for liquid effluents. (authors)

  20. Evaluation of Fluid Conduction and Mixing within a Subassembly of the Actinide Burner Test Reactor

    SciTech Connect (OSTI)

    Cliff B. Davis

    2007-09-01T23:59:59.000Z

    The RELAP5-3D code is being considered as a thermal-hydraulic system code to support the development of the sodium-cooled Actinide Burner Test Reactor as part of the Global Nuclear Energy Partnership. An evaluation was performed to determine whether the control system could be used to simulate the effects of non-convective mechanisms of heat transport in the fluid, including axial and radial heat conduction and subchannel mixing, that are not currently represented with internal code models. The evaluation also determined the relative importance of axial and radial heat conduction and fluid mixing on peak cladding temperature for a wide range of steady conditions and during a representative loss-of-flow transient. The evaluation was performed using a RELAP5-3D model of a subassembly in the Experimental Breeder Reactor-II, which was used as a surrogate for the Actinide Burner Test Reactor.

  1. Engineering Scaling Requirements for Solid Breeder Blanket Testing A. Ying, S. Sharafat, M. Youssef, J. An, R. Hunt, P. Rainsberry, M. Abdou

    E-Print Network [OSTI]

    Abdou, Mohamed

    Engineering Scaling Requirements for Solid Breeder Blanket Testing A. Ying, S. Sharafat, M. Youssef@fusion.ucla.edu An engineering scaling process is applied to the solid breeder ITER TBM designs in accordance with the testing objectives of validating the design tools and the database, and evaluating blanket performance under

  2. Advanced burner test reactor preconceptual design report.

    SciTech Connect (OSTI)

    Chang, Y. I.; Finck, P. J.; Grandy, C.; Cahalan, J.; Deitrich, L.; Dunn, F.; Fallin, D.; Farmer, M.; Fanning, T.; Kim, T.; Krajtl, L.; Lomperski, S.; Moisseytsev, A.; Momozaki, Y.; Sienicki, J.; Park, Y.; Tang, Y.; Reed, C.; Tzanos, C; Wiedmeyer, S.; Yang, W.; Chikazawa, Y.; JAEA

    2008-12-16T23:59:59.000Z

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand, to address nuclear waste management concerns and to promote non-proliferation. Implementation of the GNEP requires development and demonstration of three major technologies: (1) Light water reactor (LWR) spent fuel separations technologies that will recover transuranics to be recycled for fuel but not separate plutonium from other transuranics, thereby providing proliferation-resistance; (2) Advanced Burner Reactors (ABRs) based on a fast spectrum that transmute the recycled transuranics to produce energy while also reducing the long term radiotoxicity and decay heat loading in the repository; and (3) Fast reactor fuel recycling technologies to recover and refabricate the transuranics for repeated recycling in the fast reactor system. The primary mission of the ABR Program is to demonstrate the transmutation of transuranics recovered from the LWR spent fuel, and hence the benefits of the fuel cycle closure to nuclear waste management. The transmutation, or burning of the transuranics is accomplished by fissioning and this is most effectively done in a fast spectrum. In the thermal spectrum of commercial LWRs, some transuranics capture neutrons and become even heavier transuranics rather than being fissioned. Even with repeated recycling, only about 30% can be transmuted, which is an intrinsic limitation of all thermal spectrum reactors. Only in a fast spectrum can all transuranics be effectively fissioned to eliminate their long-term radiotoxicity and decay heat. The Advanced Burner Test Reactor (ABTR) is the first step in demonstrating the transmutation technologies. It directly supports development of a prototype full-scale Advanced Burner Reactor, which would be followed by commercial deployment of ABRs. The primary objectives of the ABTR are: (1) To demonstrate reactor-based transmutation of transuranics as part of an advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

  3. On mental privacy: the having of mental states

    E-Print Network [OSTI]

    Dembitzer, Simon David

    1998-01-01T23:59:59.000Z

    , according to Churchland, wishes to defend. This modalized version of the argument, as Churchland presents it (1992, pp. 59-60), says: (MEC) (1) My mental states are knowable by me by introspection. 24 (2) My brain states are not knowable by me...1) and (EC2), and begs the question in (MEC). My claim is that these charges are a result of Churchland's mischaracterization of Nagel's notion of privacy. The characterizations of privacy in (EC1), (EC2), and (MEC) are not Nagel...

  4. Change What? Identifying Quality Improvement Targets by Investigating Usual Mental Health Care

    E-Print Network [OSTI]

    Garland, Ann F.; Bickman, Leonard; Chorpita, Bruce F.

    2010-01-01T23:59:59.000Z

    satisfaction with mental health services and changes indon’t we have effective mental health services? [Editorial].Mental Health and Mental Health Services Research, 35, 437–

  5. On the Criticality Safety of Transuranic Sodium Fast Reactor Fuel Transport Casks

    SciTech Connect (OSTI)

    Samuel Bays; Ayodeji Alajo

    2010-05-01T23:59:59.000Z

    This work addresses the neutronic performance and criticality safety issues of transport casks for fuel pertaining to low conversion ratio sodium cooled fast reactors, conventionally known as Advanced Burner Reactors. The criticality of a one, three, seven and 19-assembly cask capacity is presented. Both dry “helium” and flooded “water” filled casks are considered. No credit for fuel burnup or fission products was assumed. As many as possible of the conservatisms used in licensing light water reactor universal transport casks were incorporated into this SFR cask criticality design and analysis. It was found that at 7-assemblies or more, adding moderator to the SFR cask increases criticality margin. Also, removal of MAs from the fuel increases criticality margin of dry casks and takes a slight amount of margin away for wet casks. Assuming credit for borated fuel tube liners, this design analysis suggests that as many as 19 assemblies can be loaded in a cask if limited purely by criticality safety. If no credit for boron is assumed, the cask could possibly hold seven assemblies if low conversion ratio fast reactor grade fuel and not breeder reactor grade fuel is assumed. The analysis showed that there is a need for new cask designs for fast reactors spent fuel transportation. There is a potential of modifying existing transportation cask design as the starting point for fast reactor spent fuel transportation.

  6. Analysis on burnup step effect for evaluating reactor criticality and fuel breeding ratio

    SciTech Connect (OSTI)

    Saputra, Geby; Purnama, Aditya Rizki; Permana, Sidik [Nuclear Physics and Biophysics Research Division, Physics Department, Institut Teknologi Bandung (Indonesia); Suzuki, Mitsutoshi [Department of Science and Technology for Nuclear Material Management (STNM), Japan Atomic Energy Agency (JAEA) (Japan)

    2014-09-30T23:59:59.000Z

    Criticality condition of the reactors is one of the important factors for evaluating reactor operation and nuclear fuel breeding ratio is another factor to show nuclear fuel sustainability. This study analyzes the effect of burnup steps and cycle operation step for evaluating the criticality condition of the reactor as well as the performance of nuclear fuel breeding or breeding ratio (BR). Burnup step is performed based on a day step analysis which is varied from 10 days up to 800 days and for cycle operation from 1 cycle up to 8 cycles reactor operations. In addition, calculation efficiency based on the variation of computer processors to run the analysis in term of time (time efficiency in the calculation) have been also investigated. Optimization method for reactor design analysis which is used a large fast breeder reactor type as a reference case was performed by adopting an established reactor design code of JOINT-FR. The results show a criticality condition becomes higher for smaller burnup step (day) and for breeding ratio becomes less for smaller burnup step (day). Some nuclides contribute to make better criticality when smaller burnup step due to individul nuclide half-live. Calculation time for different burnup step shows a correlation with the time consuming requirement for more details step calculation, although the consuming time is not directly equivalent with the how many time the burnup time step is divided.

  7. Neutronic and thermal calculation of blanket for high power operating condition of fusion reactor

    SciTech Connect (OSTI)

    Sagawa, H.; Shimakawa, S.; Kuroda, T. [Oarai Research Establishement of JAERI, Ibaraki (Japan)] [and others

    1994-12-31T23:59:59.000Z

    Internal (breeding region) structures of ceramic breeder blanket to accommodate high power operating conditions such as a DEMO reactor have been investigated. The conditions considered here are the maximum neutron wall load of 2.8 MW/m{sup 2} at outboard midplane corresponding to a fusion power of 3.0 GW and the coolant temperature of 200{degrees}C. Structure of a blanket is based on the layered pebble bed concept, which has been proposed by Japan since the ITER CDA. Lithium oxide with 50% enriched {sup 6}Li is used in a shape of small spherical pebbles which are filled in a 316SS can avoid its compatibility issue with Be. Beryllium around the breeder can is filled also in a shape of spherical pebbles which works not only as a neutron multiplier but also as a thermal resistant layer to maintain breeder temperature for effective in-situ tritium recovery. Diameters and packing fractions of both pebbles are {<=} 1 mm and 65%, respectively. A layer of block Be between cooling panels is introduced as a neutron multiplier (not as the thermal resistant layer) to enhance tritium breeding performance. Inlet temperature of water coolant is 200{degrees}C to meet the high temperature conditioning requirement to the first wall which is one of walls of the blanket vessel. Neutronics calculations have been carried out by one-dimensional transport code, and thermal calculations have also been carried out by one-dimensional slab code.

  8. SRS Small Modular Reactors

    SciTech Connect (OSTI)

    None

    2012-04-27T23:59:59.000Z

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

  9. SRS Small Modular Reactors

    ScienceCinema (OSTI)

    None

    2014-05-21T23:59:59.000Z

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

  10. Reactor safety method

    DOE Patents [OSTI]

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11T23:59:59.000Z

    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.

  11. Nuclear reactor

    DOE Patents [OSTI]

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16T23:59:59.000Z

    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.

  12. A version of this appeared in Current Science 75(6) 1998 India's Nuclear Breeders: Technology, Viability, and Options

    E-Print Network [OSTI]

    A version of this appeared in Current Science 75(6) 1998 India's Nuclear Breeders: Technology, Viability, and Options Rahul Tongia & V. S. Arunachalam Department of Engineering & Public Policy Carnegie tongia@andrew.cmu.edu; vsa@andrew.cmu.edu Abstract: India's nuclear power program is based on indigenous

  13. Activation analysis and characteristics of the European community water cooled ceramic breeder blanket design proposal for ITER

    SciTech Connect (OSTI)

    Petrizzi, L.; Rado, V. [ENEA-ERG-FUS, Frascati (Italy); Cepraga, D.G. [ENEA-INN-FIS, Bologna (Italy)

    1994-12-31T23:59:59.000Z

    The European Community (EC) Home Team has proposed various alternative blanket designs to the basic concept (essentially integrated first wall, cooled by liquid metal, with structures made by vanadium alloys). One of the EC proposal is the Water Cooled Ceramic Blanket developed on the basis of a common action between NET and ENEA. It is based on a more conservative approach, but involving well proven technologies and qualified materials: SS-316L as structural material, Li{sub 2}ZrO{sub 3} as first breeder material choice (50% Li{sup 6} enrichment) and low temperature water coolant (160/200{degrees}C). Beryllium has been chosen as multiplying material. The nominal performance are: 1 MW/m{sup 2} as average neutron wall load, corresponding to 1.5 GW fusion power, 1 MW-y/m{sup 2} beneath it has been proved to withstand power excursion till 5 GW. The proposed blanket concept is based on a Breeder Inside Tube (BIT) type technology, with poloidal breeding elements, each one consisting of two concentric tubes. Breeder pebbles are filled into the inner tube, the water coolant flows in the annular channel between the two tubes. Beryllium pebbles fill the space of the blanket box outside the outer tube. A helium purge gas flows through the breeder pebbles bed for tritium recovery. Alternative operating water temperature and pressure are proposed, considering also batch tritium recovery.

  14. Feather loss and egg production in broiler breeders A.D. MILLS, J.M. FAURE J.B. WILLIAMS

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    of three tier battery cage units. The broiler breeders and layers were housed in separate cage units to 18 weeks of age the birds were housed, in single strain groups, in floor pens. When the birds were 18, but both cage units were located in the same poultry house. The lighting regimen in the poultry house

  15. Nuclear reactor heat transport system component low friction support system

    DOE Patents [OSTI]

    Wade, Elman E. (Ruffs Dale, PA)

    1980-01-01T23:59:59.000Z

    A support column for a heavy component of a liquid metal fast breeder reactor heat transport system which will deflect when the pipes leading coolant to and from the heavy component expand or contract due to temperature changes includes a vertically disposed pipe, the pipe being connected to the heavy component by two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles and the pipe being supported through two longitudinally spaced cycloidal dovetail joints wherein the distal end of each of the dovetails constitutes a part of the surface of a large diameter cylinder and the centerlines of these large diameter cylinders intersect at right angles, each of the cylindrical surfaces bearing on a flat and horizontal surface.

  16. Atmospheric Pressure Reactor System | EMSL

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

    Atmospheric Pressure Reactor System Atmospheric Pressure Reactor System The atmospheric pressure reactor system is designed for testing the efficiency of various catalysts for the...

  17. A helium-cooled blanket design of the low aspect ratio reactor

    SciTech Connect (OSTI)

    Wong, C.P.; Baxi, C.B.; Reis, E.E. [General Atomics, San Diego, CA (United States); Cerbone, R.; Cheng, E.T. [TSI Research, Solana Beach, CA (United States)

    1998-03-01T23:59:59.000Z

    An aggressive low aspect ratio scoping fusion reactor design indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of > 4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 436 C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m(2), respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650 C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh.

  18. A PROFILE OF KENTUCKY MEDICAID MENTAL HEALTH

    E-Print Network [OSTI]

    Hayes, Jane E.

    can be advanced--among patients, health care providers, and the community at large. This workA PROFILE OF KENTUCKY MEDICAID MENTAL HEALTH DIAGNOSES, 2000-2010 #12; #12; i A Profile of Kentucky Medicaid Mental Health Diagnoses, 20002010 BY Michael T. Childress

  19. Electron cyclotron resonance charge breeder ion source simulation by MCBC and GEM

    SciTech Connect (OSTI)

    Kim, J. S.; Zhao, L.; Cluggish, B. P.; Bogatu, I. N.; Pardo, R. [FAR-TECH, Inc., 3550 General Atomics Court 15-155, San Diego, California 92121 (United States); Argonne National Laboratory, Argonne, Illinois 60439 (United States)

    2008-02-15T23:59:59.000Z

    Numerical simulation results by the GEM and MCBC codes are presented, along with a comparison with experiments for beam capture dynamics and parameter studies of charge state distribution (CSD) of electron cyclotron resonance charge breeder ion sources. First, steady state plasma profiles are presented by GEM with respect to key experimental parameters such as rf power and gas pressure. As rf power increases, electron density increases by a small amount and electron energy by a large amount. The central electrostatic potential dip also increased. Next, MCBC is used to trace injected beam ions to obtain beam capture profiles. Using the captured ion profiles, GEM obtains a CSD of beam ions. As backscattering can be significant, capturing the ions near the center of the device enhances the CSD. The effect of rf power on the beam CSD is mainly due to different steady states plasmas. Example cases are presented assuming that the beam ions are small enough not to affect the plasma.

  20. Remote State Preparation of Mental Information

    E-Print Network [OSTI]

    Patrizio E. Tressoldi; Andrei Khrennikov

    2012-07-26T23:59:59.000Z

    The aim of this paper is to define in theoretical terms and summarise the available experimental evidence that physical and mental "objects", if considered "information units", may present similar classical and quantum models of communication beyond their specific characteristics. Starting with the Remote State Preparation protocol, a variant of the Teleportation protocol, for which formal models and experimental evidence are already available in quantum mechanics, we outline a formal model applied to mental information we defined Remote State Preparation of Mental Information (RSPMI), and we summarise the experimental evidence supporting the feasibility of a RSPMI protocol. The available experimental evidence offers strong support to the possibility of real communication at distance of mental information promoting the integration between disciplines that have as their object of knowledge different aspects of reality, both physical and the mental, leading to a significant paradigm shift in cognitive and information science.

  1. The ReA electron-beam ion trap charge breeder for reacceleration of rare isotopes

    SciTech Connect (OSTI)

    Lapierre, A.; Schwarz, S.; Kittimanapun, K.; Fogleman, J.; Krause, S.; Nash, S.; Rencsok, R.; Tobos, L.; Perdikakis, G.; Portillo, M.; Rodriguez, J. A.; Wittmer, W.; Wu, X.; Bollen, G.; Leitner, D.; Syphers, M. [National Superconducting Cyclotron Laboratory (NSCL), Michigan State University (MSU), 640 South Shaw Lane, East Lansing, MI 48824 (United States); Collaboration: ReA Team

    2013-04-19T23:59:59.000Z

    ReA is a post-accelerator at the National Superconducting Cyclotron Laboratory at Michigan State University. ReA is designed to reaccelerate rare isotopes to energies of a few MeV/u following production by projectile fragmentation and thermalization in a gas cell. The facility consists of four main components: an electron-beam ion trap (EBIT) charge breeder, an achromatic charge-over-mass (Q/A) separator, a radio-frequency quadrupole accelerator, and a superconducting radio-frequency linear accelerator. The EBIT charge breeder was specifically designed to efficiently capture continuous beams of singly charged ions injected at low energy (<60 keV), charge breed in less than 50 ms, and extract highly charged ions to the Q/A separator for charge-state selection and reacceleration through the accelerator structures. The use of highly charged ions to reach high beam energies is a key aspect that makes ReA a compact and cost-efficient post-accelerator. The EBIT is characterized by a high-current electron gun, a long multi-electrode trap structure and a dual magnet to provide both the high electron-beam current density necessary for fast charge breeding of short-lived isotopes as well as the high capture probability of injected beams. This paper presents an overview and the status of the ReA EBIT, which has extracted for reacceleration tests stable {sup 20}Ne{sup 8+} ion beams produced from injected gas and more recently {sup 39}K{sup 16+} beams by injecting stable {sup 39,41}K{sup +} ions from an external ion source.

  2. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    SciTech Connect (OSTI)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30T23:59:59.000Z

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying/tuning of lower-fidelity models, which now require costly experimental qualification for each different type of design

  3. Women’s Pathways to Mental Health in India

    E-Print Network [OSTI]

    Sood, Anubha

    2008-01-01T23:59:59.000Z

    settings? How can India’s mental health policy frameworkhealth policy making and implementation in India in recentthat India’s current mental health policies are detrimental

  4. Performance indicators for public mental healthcare: A systematic international inventory

    E-Print Network [OSTI]

    Lauriks, Steve; Buster, Marcel CA; de Wit, Matty AS; Arah, Onyebuchi A; Klazinga, Niek S

    2012-01-01T23:59:59.000Z

    a systematic international inventory. BMC Public Health 2012Mental Health: National inventory of mental health qualitya systematic international inventory Steve Lauriks 1,2* ,

  5. Analysis of Assembly Bill 154: Mental Health Services

    E-Print Network [OSTI]

    California Health Benefits Review Program (CHBRP)

    2011-01-01T23:59:59.000Z

    drug abuse and mental health treatment on medical careabuse treatment Mental Health Services except substance related disorders, life transition problems, skilled nursing services, home health care,

  6. Analysis of Assembly Bill 154: Mental Health Services

    E-Print Network [OSTI]

    California Health Benefits Review Program (CHBRP)

    2011-01-01T23:59:59.000Z

    Bill 154:Mental Health Services. Report to California Statehealth specialty care. Health Services Research. 2004;39:mental health care needs? Health Services Research. 2007;42:

  7. Role of small lead-cooled fast reactors for international deployment in worldwide sustainable nuclear energy supply.

    SciTech Connect (OSTI)

    Sienicki, J. J.; Wade, D. C.; Moisseytsev, A.; Nuclear Engineering Division

    2008-01-01T23:59:59.000Z

    Most recently, the global nuclear energy partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium-sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the small secure transportable autonomous reactor (SSTAR) has been under ongoing development as part of the US advanced nuclear energy systems programs. Meeting future worldwide projected energy demands during this century (e.g., 1000 to 2000 GWe by 2050) in a sustainable manner while maintaining CO2 emissions at or below today's level will require massive deployments of nuclear reactors in non-fuel cycle states as well as fuel cycle states. The projected energy demands of non-fuel cycle states will not be met solely through the deployment of Light Water Reactors (LWRs) in those states without using up the world's resources of fissile material (e.g., known plus speculative virgin uranium resources = 15 million tonnes). The present U.S. policy is focused upon domestic deployment of large-scale LWRs and sodium-cooled fast spectrum Advanced Burner Reactors (ABRs) working in a symbiotic relationship that burns existing fissile material while destroying the actinides which are generated. Other major nuclear nations are carrying out the development and deployment of SFR breeders as witness the planning for SFR breeder deployments in France, Japan, China, India, and Russia. Small (less that 300 MWe) and medium (300 to 700 MWe) size reactors are better suited to the growing economies and infrastructures of many non-fuel cycle states and developing nations. For those deployments, fast reactor converters which are fissile self-sufficient by creating as much fissile material as they consume are preferred to breeders that create more fissile material than they consume. Thus, there is a need for small and medium size fast reactors in non-fuel cycle states operating in a converter mode as well as large sodium-cooled fast breeders in fuel cycle states. Desired attributes for exportable small fast reactors include: proliferation resistance features such as restricted access to fuel; long core life further restricting access by reducing or eliminating the need for refueling; restricted potential to be misused in a breeding mode; fuel form that is unattractive in the safeguards sense; and a conversion ratio of unity to self-generate as much fissile material as is consumed. Desired attributes for exportable small reactor deployments in developing nations and remote sites also include: a small power level to match the smaller demand of towns or sites that are off-grid or on immature local grids; low enough cost to be economically competitive with alternative energy sources available to developing nation customers (e.g. diesel generators in remote locations); readily transported and assembled from transportable modules; simple to operate and highly reliable reducing plant operating staff requirements; as well as high reliability and passive safety reducing the number of accident initiators and need for safety systems as well as reducing the size of the exclusion and emergency planning zones. The Lead-Cooled Fast Reactor (LFR) has the desired attributes. An example of a small exportable LFR concept is the 20 MWe (45 MWt) Small Secure Transportable Autonomous Reactor (SSTAR) incorporating proliferation resistance, fissile selfsufficiency, autonomous load following, a high degree of passive safety, and supercritical carbon dioxide Brayton cycle energy conversion for high plant efficiency and improved economic competitiveness.

  8. Gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, Kenny C. (Lemont, IL); Laug, Matthew T. (Idaho Falls, ID)

    1985-01-01T23:59:59.000Z

    The invention discloses the use of stable isotopes of neon and argon, that are grouped in preselected different ratios one to the other and are then sealed as tags in different cladded nuclear fuel elements to be used in a liquid metal fast breeder reactor. Failure of the cladding of any fuel element allows fission gases generated in the reaction and these tag isotopes to escape and to combine with the cover gas held in the reactor over the fuel elements. The isotopes specifically are Ne.sup.20, Ne.sup.21 and Ne.sup.22 of neon and Ar.sup.36, Ar.sup.38 and Ar.sup.40 of argon, and the cover gas is helium. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between approximately 0.degree. and -25.degree. C. operable to remove the fission gases from the cover gas and tags and the second or tag recovery system bed is held between approximately -170.degree. and -185.degree. C. operable to isolate the tags from the cover gas. Spectrometric analysis further is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be specifically determined.

  9. R and D of On-line Reprocessing Technology for Molten-Salt Reactor Systems

    SciTech Connect (OSTI)

    Uhlir, Jan; Tulackova, Radka; Chuchvalcova Bimova, Karolina [Nuclear Research Institute Rez plc CZ-250 68 Husinec - Rez 130 (Serbia and Montenegro)

    2006-07-01T23:59:59.000Z

    The Molten Salt Reactor (MSR) represents one of promising future nuclear reactor concept included in the Generation IV reactors family. The reactor can be operated as the thorium breeder or as the actinide transmuter. However, the future deployment of Molten-Salt Reactors will be significantly dependent on the successful mastering of advanced reprocessing technologies dedicated to their fuel cycle. Here the on-line reprocessing technology connected with the fuel circuit of MSR is of special importance because the reactor cannot be operated for a long run without the fuel salt clean-up. Generally, main MSR reprocessing technologies are pyrochemical, majority of them are fluoride technologies. The proposed flow-sheets of MSR on-line reprocessing are based on a combination of molten-salt / liquid metal extraction and electro-separation processes, which can be added to the gas extraction process already verified during the MSRE project in ORNL. The crucial separation method proposed for partitioning of actinides from fission products is based on successive Anodic dissolution and Cathodic deposition processes in molten fluoride media. (authors)

  10. Ethnic Differences in Children’s Entry into Public Mental Health Care via Emergency Mental Health Services

    E-Print Network [OSTI]

    Snowden, Lonnie R.; Masland, Mary C.; Fawley, Kya; Wallace, Neal

    2009-01-01T23:59:59.000Z

    A. W. (2003). Mental health services use among school-agedand care coordination. Health Services Research, 38,K. L. (2001). Mental health services for youths in foster

  11. Reactor Sharing Program

    SciTech Connect (OSTI)

    Vernetson, W.G.

    1993-01-01T23:59:59.000Z

    Progress achieved at the University of Florida Training Reactor (UFTR) facility through the US Department of Energy's University Reactor Sharing Program is reported for the period of 1991--1992.

  12. Undergraduate reactor control experiment

    SciTech Connect (OSTI)

    Edwards, R.M.; Power, M.A.; Bryan, M. (Pennsylvania State Univ., University Park (United States))

    1992-01-01T23:59:59.000Z

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise.

  13. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-01-01T23:59:59.000Z

    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.

  14. Attrition reactor system

    DOE Patents [OSTI]

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

    1993-09-28T23:59:59.000Z

    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.

  15. Implications of Fast Reactor Transuranic Conversion Ratio

    SciTech Connect (OSTI)

    Steven J. Piet; Edward A. Hoffman; Samuel E. Bays

    2010-11-01T23:59:59.000Z

    Theoretically, the transuranic conversion ratio (CR), i.e. the transuranic production divided by transuranic destruction, in a fast reactor can range from near zero to about 1.9, which is the average neutron yield from Pu239 minus 1. In practice, the possible range will be somewhat less. We have studied the implications of transuranic conversion ratio of 0.0 to 1.7 using the fresh and discharge fuel compositions calculated elsewhere. The corresponding fissile breeding ratio ranges from 0.2 to 1.6. The cases below CR=1 (“burners”) do not have blankets; the cases above CR=1 (“breeders”) have breeding blankets. The burnup was allowed to float while holding the maximum fluence to the cladding constant. We graph the fuel burnup and composition change. As a function of transuranic conversion ratio, we calculate and graph the heat, gamma, and neutron emission of fresh fuel; whether the material is “attractive” for direct weapon use using published criteria; the uranium utilization and rate of consumption of natural uranium; and the long-term radiotoxicity after fuel discharge. For context, other cases and analyses are included, primarily once-through light water reactor (LWR) uranium oxide fuel at 51 MWth-day/kg-iHM burnup (UOX-51). For CR<1, the heat, gamma, and neutron emission increase as material is recycled. The uranium utilization is at or below 1%, just as it is in thermal reactors as both types of reactors require continuing fissile support. For CR>1, heat, gamma, and neutron emission decrease with recycling. The uranium utilization exceeds 1%, especially as all the transuranic elements are recycled. exceeds 1%, especially as all the transuranic elements are recycled. At the system equilibrium, heat and gamma vary by somewhat over an order of magnitude as a function of CR. Isotopes that dominate heat and gamma emission are scattered throughout the actinide chain, so the modest impact of CR is unsurprising. Neutron emitters are preferentially found among the higher actinides, so the neutron emission varies much stronger with CR, about three orders of magnitude.

  16. Advanced Test Reactor Tour

    SciTech Connect (OSTI)

    Miley, Don

    2011-01-01T23:59:59.000Z

    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.

  17. High solids fermentation reactor

    DOE Patents [OSTI]

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

    1993-03-02T23:59:59.000Z

    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.

  18. High solids fermentation reactor

    DOE Patents [OSTI]

    Wyman, Charles E. (Lakewood, CO); Grohmann, Karel (Littleton, CO); Himmel, Michael E. (Littleton, CO); Richard, Christopher J. (Lakewood, CO)

    1993-01-01T23:59:59.000Z

    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.

  19. Improved vortex reactor system

    DOE Patents [OSTI]

    Diebold, James P. (Lakewood, CO); Scahill, John W. (Evergreen, CO)

    1995-01-01T23:59:59.000Z

    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.

  20. Advanced Test Reactor Tour

    ScienceCinema (OSTI)

    Miley, Don

    2013-05-28T23:59:59.000Z

    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.

  1. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    SciTech Connect (OSTI)

    Not Available

    1986-01-01T23:59:59.000Z

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant).

  2. Reactor vessel support system

    DOE Patents [OSTI]

    Golden, Martin P. (Trafford, PA); Holley, John C. (McKeesport, PA)

    1982-01-01T23:59:59.000Z

    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.

  3. Radiogenic lead with dominant content of {sup 208}Pb: New coolant and neutron moderator for innovative nuclear reactors

    SciTech Connect (OSTI)

    Shmelev, A. N.; Kulikov, G. G.; Kryuchkov, E. F.; Apse, V. A.; Kulikov, E. G. [National Research Nuclear Univ. MEPhI, Kashirskoe shosse, 31, 115409, Moscow (Russian Federation)

    2012-07-01T23:59:59.000Z

    The advantages of radiogenic lead with dominant content of {sup 208}Pb as a reactor coolant with respect to natural lead are caused by unique nuclear properties of {sup 208}Pb which is a double-magic nucleus with closed proton and neutron shells. This results in significantly lower micro cross section and resonance integral of radiative neutron capture by {sup 208}Pb than those for numerous light neutron moderators. The extremely weak ability of {sup 208}Pb to absorb neutrons results in the following effects. Firstly, neutron moderating factor (ratio of scattering to capture cross sections) is larger than that for graphite and light water. Secondly, age and diffusion length of thermal neutrons are larger than those for graphite, light and heavy water. Thirdly, neutron lifetime in {sup 208}Pb is comparable with that for graphite, beryllium and heavy water what could be important for safe reactor operation. The paper presents some results obtained in neutronics and thermal-hydraulics evaluations of the benefits from the use of radiogenic lead with dominant content of {sup 208}Pb instead of natural lead as a coolant of fast breeder reactors. The paper demonstrates that substitution of radiogenic lead for natural lead can offer the following benefits for operation of fast breeder reactors. Firstly, improvement of the reactor safety thanks to the better values of coolant temperature reactivity coefficient and, secondly, improvement of some thermal-hydraulic reactor parameters. Radiogenic lead can be extracted from thorium sludge without isotope separation as {sup 208}Pb is a final isotope in the decay chain of {sup 232}Th. (authors)

  4. NEW YORK CITY DEPARTMENT OF HEALTH AND MENTAL HYGIENE

    E-Print Network [OSTI]

    Qiu, Weigang

    NEW YORK CITY DEPARTMENT OF HEALTH AND MENTAL HYGIENE Thomas Farley, MD, MPH Commissioner New York City Department of Health & Mental Hygiene (DOHMH) 2014 Mental Health Scholarship Program One Year Hygiene (DOHMH) Bureau of Mental Health and Bureau of Children, Youth and Families invite qualified

  5. Reactor water cleanup system

    DOE Patents [OSTI]

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

    1994-12-20T23:59:59.000Z

    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.

  6. Spinning fluids reactor

    SciTech Connect (OSTI)

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

    2012-11-20T23:59:59.000Z

    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.

  7. Determining Reactor Neutrino Flux

    E-Print Network [OSTI]

    Jun Cao

    2012-03-08T23:59:59.000Z

    Flux is an important source of uncertainties for a reactor neutrino experiment. It is determined from thermal power measurements, reactor core simulation, and knowledge of neutrino spectra of fuel isotopes. Past reactor neutrino experiments have determined the flux to (2-3)% precision. Precision measurements of mixing angle $\\theta_{13}$ by reactor neutrino experiments in the coming years will use near-far detector configurations. Most uncertainties from reactor will be canceled out. Understanding of the correlation of uncertainties is required for $\\theta_{13}$ experiments. Precise determination of reactor neutrino flux will also improve the sensitivity of the non-proliferation monitoring and future reactor experiments. We will discuss the flux calculation and recent progresses.

  8. Testing and Commissioning of a Multifunctional Tool for the Dismantling of the Activated Internals of the KNK Reactor Shaft - 13524

    SciTech Connect (OSTI)

    Rothschmitt, Stefan; Graf, Anja [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein-Leopoldshafen (Germany)] [WAK Rueckbau- und Entsorgungs- GmbH, P.O.Box 12 63, 76339 Eggenstein-Leopoldshafen (Germany); Bauer, Stefan; Klute, Stefan; Koselowski, Eiko [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany)] [Siempelkamp Nukleartechnik GmbH, Am Taubenfeld 25/1, 69123 Heidelberg (Germany); Hendrich, Klaus [Ingenieurbuero Hendrich, Moerikeweg 14, 75015 Bretten (Germany)] [Ingenieurbuero Hendrich, Moerikeweg 14, 75015 Bretten (Germany)

    2013-07-01T23:59:59.000Z

    The Compact Sodium Cooled Reactor Facility Karlsruhe (KNK), a prototype reactor to demonstrate the Fast Breeder Reactor Technology in Germany, was in operation from 1971 to 1991. The dismantling activities started in 1991. The project aim is the green field in 2020. Most of the reactor internals as well as the primary and secondary cooling loops are already dismantled. The total contaminated sodium inventory has already been disposed of. Only the high activated reactor vessel shielding structures are remaining. Due to the high dose rates these structures must be dismantled remotely. For the dismantling of the primary shielding of the reactor vessel, 12 stacked cast iron blocks with a total mass of 90 Mg and single masses up to 15.5 Mg, a remote-controlled multifunctional dismantling device (HWZ) was designed, manufactured and tested in a mock-up. After successful approval of the test sequences by the authorities, the HWZ was implemented into the reactor building containment for final assembling of the auxiliary equipment and subsequent hot commissioning in 2012. Dismantling of the primary shielding blocks is scheduled for early 2013. (authors)

  9. The MSFR as a flexible CR reactor: the viewpoint of safety

    SciTech Connect (OSTI)

    Fiorina, C.; Cammi, A. [Politecnico di Milano, Via La Masa 34, 20136 Milan (Italy); Franceschini, F. [Westinghouse Electric Company LL, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States); Krepel, J. [Paul Scherrer Institut, PSI WEST, 5234 Villigen (Switzerland)

    2013-07-01T23:59:59.000Z

    In this paper, the possibility has first been discussed of using the liquid-fuelled Molten Salt Fast Reactor (MSFR) as a flexible conversion ratio (CR) reactor without design modification. By tuning the reprocessing rate it is possible to determine the content of fission products in the core, which in turn can significantly affect the neutron economy without incurring in solubility problems. The MSFR can thus be operated as U-233 breeder (CR>1), iso-breeder (CR=1) and burner reactor (CR<1). In particular a 40 year doubling time can be achieved, as well as a considerable Transuranics and MA (minor actinide) burning rate equal to about 150 kg{sub HN}/GWE-yr. The safety parameters of the MSFR have then been evaluated for different fuel cycle strategies. Th use and a softer spectrum combine to give a strong Doppler coefficient, one order of magnitude higher compared to traditional fast reactors (FRs). The fuel expansion coefficient is comparable to the Doppler coefficient and is only mildly affected by core compositions, thus assisting the fuel cycle flexibility of the MSFR. ?eff and generation time are comparable to the case of traditional FRs, if a static fuel is assumed. A notable reduction of ?eff is caused by salt circulation, but a low value of this parameter is a limited concern in the MSFR thanks to the lack of a burnup reactivity swing and of positive feedbacks. A simple approach has also been developed to evaluate the MSFR capabilities to withstand all typical double-fault accidents, for different fuel cycle options.

  10. Mental Representations Formed From Educational Website Formats

    SciTech Connect (OSTI)

    Elizabeth T. Cady; Kimberly R. Raddatz; Tuan Q. Tran; Bernardo de la Garza; Peter D. Elgin

    2006-10-01T23:59:59.000Z

    The increasing popularity of web-based distance education places high demand on distance educators to format web pages to facilitate learning. However, limited guidelines exist regarding appropriate writing styles for web-based distance education. This study investigated the effect of four different writing styles on reader’s mental representation of hypertext. Participants studied hypertext written in one of four web-writing styles (e.g., concise, scannable, objective, and combined) and were then administered a cued association task intended to measure their mental representations of the hypertext. It is hypothesized that the scannable and combined styles will bias readers to scan rather than elaborately read, which may result in less dense mental representations (as identified through Pathfinder analysis) relative to the objective and concise writing styles. Further, the use of more descriptors in the objective writing style will lead to better integration of ideas and more dense mental representations than the concise writing style.

  11. Fission-suppressed fusion breeder on the thorium cycle and nonproliferation

    SciTech Connect (OSTI)

    Moir, R. W. [Vallecitos Molten Salt Research, 607 E. Vallecitos Rd., Livermore, CA 94550 925-447-8804 (United States)

    2012-06-19T23:59:59.000Z

    Fusion reactors could be designed to breed fissile material while suppressing fissioning thereby enhancing safety. The produced fuel could be used to startup and makeup fuel for fission reactors. Each fusion reaction can produce typically 0.6 fissile atoms and release about 1.6 times the 14 MeV neutron's energy in the blanket in the fission-suppressed design. This production rate is 2660 kg/1000 MW of fusion power for a year. The revenues would be doubled from such a plant by selling fuel at a price of 60/g and electricity at $0.05/kWh for Q=P{sub fusion}/P{sub input}=4. Fusion reactors could be designed to destroy fission wastes by transmutation and fissioning but this is not a natural use of fusion whereas it is a designed use of fission reactors. Fusion could supply makeup fuel to fission reactors that were dedicated to fissioning wastes with some of their neutrons. The design for safety and heat removal and other items is already accomplished with fission reactors. Whereas fusion reactors have geometry that compromises safety with a complex and thin wall separating the fusion zone from the blanket zone where wastes could be destroyed. Nonproliferation can be enhanced by mixing {sup 233}U with {sup 238}U. Also nonproliferation is enhanced in typical fission-suppressed designs by generating up to 0.05 {sup 232}U atoms for each {sup 233}U atom produced from thorium, about twice the IAEA standards of 'reduced protection' or 'self protection.' With 2.4%{sup 232}U, high explosive material is predicted to degrade owing to ionizing radiation after a little over 1/2 year and the heat rate is 77 W just after separation and climbs to over 600 W ten years later. The fissile material can be used to fuel most any fission reactor but is especially appropriate for molten salt reactors (MSR) also called liquid fluoride thorium reactors (LFTR) because of the molten fuel does not need hands on fabrication and handling.

  12. Behavior of 241Am in fast reactor systems - a safeguards perspective

    SciTech Connect (OSTI)

    Beddingfield, David H [Los Alamos National Laboratory; Lafleur, Adrienne M [Los Alamos National Laboratory

    2009-01-01T23:59:59.000Z

    Advanced fuel-cycle developments around the world currently under development are exploring the possibility of disposing of {sup 241}Am from spent fuel recycle processes by burning this material in fast reactors. For safeguards practitioners, this approach could potentially complicate both fresh- and spent-fuel safeguards measurements. The increased ({alpha},n) production in oxide fuels from the {sup 241}Am increases the uncertainty in coincidence assay of Pu in MOX assemblies and will require additional information to make use of totals-based neutron assay of these assemblies. We have studied the behavior of {sup 241}Am-bearing MOX fuel in the fast reactor system and the effect on neutron and gamma-ray source-terms for safeguards measurements. In this paper, we will present the results of simulations of the behavior of {sup 241}Am in a fast breeder reactor system. Because of the increased use of MOX fuel in thermal reactors and advances in fuel-cycle designs aimed at americium disposal in fast reactors, we have undertaken a brief study of the behavior of americium in these systems to better understand the safeguards impacts of these new approaches. In this paper we will examine the behavior of {sup 241}Am in a variety of nuclear systems to provide insight into the safeguards implications of proposed Am disposition schemes.

  13. A Framework for Human Performance Criteria for Advanced Reactor Operational Concepts

    SciTech Connect (OSTI)

    Jacques V Hugo; David I Gertman; Jeffrey C Joe

    2014-08-01T23:59:59.000Z

    This report supports the determination of new Operational Concept models needed in support of the operational design of new reactors. The objective of this research is to establish the technical bases for human performance and human performance criteria frameworks, models, and guidance for operational concepts for advanced reactor designs. The report includes a discussion of operating principles for advanced reactors, the human performance issues and requirements for human performance based upon work domain analysis and current regulatory requirements, and a description of general human performance criteria. The major findings and key observations to date are that there is some operating experience that informs operational concepts for baseline designs for SFR and HGTRs, with the Experimental Breeder Reactor-II (EBR-II) as a best-case predecessor design. This report summarizes the theoretical and operational foundations for the development of a framework and model for human performance criteria that will influence the development of future Operational Concepts. The report also highlights issues associated with advanced reactor design and clarifies and codifies the identified aspects of technology and operating scenarios.

  14. Spin-On for the Renaissance? The Current State of China's Nuclear Industry

    E-Print Network [OSTI]

    Yuan, Jing-dong

    2010-01-01T23:59:59.000Z

    China Experimental Fast Reactor (CEFR) and high-temperaturehigh-temperature and fast breeder reactors appears to be atechnologies such as fast breeder reactors like the

  15. Advanced methods comparisons of reaction rates in the purdue fast breeder blanket facility

    SciTech Connect (OSTI)

    Hill, R.N. (Argonne National Lab., IL (USA). Engineering Div.); Ott, K.O. (Purdue Univ., Lafayette, IN (USA). School of Nuclear Engineering)

    1989-09-01T23:59:59.000Z

    The authors discuss how experiments in the large uniform Purdue University fast breeder blanket facility blanket yield an accurate quantification of this discrepancy. Using standard production code methods (diffusion theory with 50-group cross sections), a consistent calculated-to-experimental (C/E) drop-off is observed for various reaction rates. A 50% increase in the calculated results at the outer edge of the 51-cm blanket is necessary for agreement with experiments. The usefulness of refined group constant generation, utilizing speciaized weighting spectra, and transport theory methods in correcting this discrepancy is analyzed. Refined group constants reduce the discrepancy to half that observed using the standard method. The result is that transport methods have no effect on the blanket deviations; thus, the present multigroup transport theory does not constitute or even contribute to an explanation of the blanket discrepancies. The residual blanket C/E drop-off (about half the standard drop-off) using advanced methods must be caused by approximations that are applied in all current multigroup methods.

  16. Improved vortex reactor system

    DOE Patents [OSTI]

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

    1995-05-09T23:59:59.000Z

    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. Tokamak reactor first wall

    DOE Patents [OSTI]

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

    1984-11-20T23:59:59.000Z

    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.

  18. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, Juhani (Karhula, FI)

    1996-01-01T23:59:59.000Z

    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.

  19. Pressurized fluidized bed reactor

    DOE Patents [OSTI]

    Isaksson, J.

    1996-03-19T23:59:59.000Z

    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.

  20. Improved gas tagging and cover gas combination for nuclear reactor

    DOE Patents [OSTI]

    Gross, K.C.; Laug, M.T.

    1983-09-26T23:59:59.000Z

    The invention discloses the use of stable isotopes of neon and argon, sealed as tags in different cladding nuclear fuel elements to be used in a liquid metal fast breeder reactor. Cladding failure allows fission gases and these tag isotopes to escape and to combine with the cover gas. The isotopes are Ne/sup 20/, Ne/sup 21/ and Ne/sup 22/ and Ar/sup 36/, Ar/sup 38/ and Ar/sup 40/, and the cover gas is He. Serially connected cryogenically operated charcoal beds are used to clean the cover gas and to separate out the tags. The first or cover gas cleanup bed is held between 0 and -25/sup 0/C to remove the fission gases from the cover gas and tags, and the second or tag recovery system bed between -170 and -185/sup 0/C to isolate the tags from the cover gas. Spectrometric analysis is used to identify the specific tags that are recovered, and thus the specific leaking fuel element. By cataloging the fuel element tags to the location of the fuel elements in the reactor, the location of the leaking fuel element can then be determined.

  1. Expert system driven fuzzy control application to power reactors

    SciTech Connect (OSTI)

    Tsoukalas, L.H.; Berkan, R.C.; Upadhyaya, B.R.; Uhrig, R.E.

    1990-01-01T23:59:59.000Z

    For the purpose of nonlinear control and uncertainty/imprecision handling, fuzzy controllers have recently reached acclaim and increasing commercial application. The fuzzy control algorithms often require a supervisory'' routine that provides necessary heuristics for interface, adaptation, mode selection and other implementation issues. Performance characteristics of an on-line fuzzy controller depend strictly on the ability of such supervisory routines to manipulate the fuzzy control algorithm and enhance its control capabilities. This paper describes an expert system driven fuzzy control design application to nuclear reactor control, for the automated start-up control of the Experimental Breeder Reactor-II. The methodology is verified through computer simulations using a valid nonlinear model. The necessary heuristic decisions are identified that are vitally important for the implemention of fuzzy control in the actual plant. An expert system structure incorporating the necessary supervisory routines is discussed. The discussion also includes the possibility of synthesizing the fuzzy, exact and combined reasoning to include both inexact concepts, uncertainty and fuzziness, within the same environment.

  2. Expert system driven fuzzy control application to power reactors

    SciTech Connect (OSTI)

    Tsoukalas, L.H.; Berkan, R.C.; Upadhyaya, B.R.; Uhrig, R.E.

    1990-12-31T23:59:59.000Z

    For the purpose of nonlinear control and uncertainty/imprecision handling, fuzzy controllers have recently reached acclaim and increasing commercial application. The fuzzy control algorithms often require a ``supervisory`` routine that provides necessary heuristics for interface, adaptation, mode selection and other implementation issues. Performance characteristics of an on-line fuzzy controller depend strictly on the ability of such supervisory routines to manipulate the fuzzy control algorithm and enhance its control capabilities. This paper describes an expert system driven fuzzy control design application to nuclear reactor control, for the automated start-up control of the Experimental Breeder Reactor-II. The methodology is verified through computer simulations using a valid nonlinear model. The necessary heuristic decisions are identified that are vitally important for the implemention of fuzzy control in the actual plant. An expert system structure incorporating the necessary supervisory routines is discussed. The discussion also includes the possibility of synthesizing the fuzzy, exact and combined reasoning to include both inexact concepts, uncertainty and fuzziness, within the same environment.

  3. The Eye of the Beholder: Youths and Parents Differ on What Matters in Mental Health Services

    E-Print Network [OSTI]

    2010-01-01T23:59:59.000Z

    and adolescent mental health services. Journal of Child andwith mental health services. Psychiatric Services, 50(7),regarding mental health services: Race, age, and gender

  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. REACTOR OPERATIONS AND CONTROL

    E-Print Network [OSTI]

    Pázsit, Imre

    REACTOR OPERATIONS AND CONTROL KEYWORDS: core calculations, neural networks, control rod elevation of a control rod, or a group of control rods, is an important parameter from the viewpoint of reactor control DETERMINATION OF PWR CONTROL ROD POSITION BY CORE PHYSICS AND NEURAL NETWORK METHODS NINOS S. GARIS* and IMRE

  6. Reed Reactor Facility. Final report

    SciTech Connect (OSTI)

    Frantz, S.G.

    1994-12-31T23:59:59.000Z

    This report discusses the operation and maintenance of the Reed Reactor Facility. The Reed reactor is mostly used for education and train purposes.

  7. Reactor & Nuclear Systems Publications | ORNL

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

    Science Home | Science & Discovery | Nuclear Science | Publications and Reports | Reactor and Nuclear Systems Publications SHARE Reactor and Nuclear Systems Publications The...

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

    SciTech Connect (OSTI)

    Higinbotham, W.A.

    1994-11-07T23:59:59.000Z

    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. DC Students Flex Their Mental Muscles in Regional Science Bowl...

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

    DC Students Flex Their Mental Muscles in Regional Science Bowl Competition DC Students Flex Their Mental Muscles in Regional Science Bowl Competition February 23, 2015 - 3:12pm...

  10. The effect of added biotin and vitamin E in broiler breeder diets on hatchability and early livability in broilers

    E-Print Network [OSTI]

    Brown, Ronald Blyn

    1987-01-01T23:59:59.000Z

    of embryonic mortality: the first week of incubation and the last three days of incubation. When breeder hens are fed a low biotin diet their embryos develop congenital per osi s, atoxia, and characteristic. skeletal deformities. The deformities can... and 2 males per pen. Each pen was littered with 8 to 10 cm of fresh wood shav1ngs. Nine nests littered with fresh shavings for nest1ng mater1al were utilized per pen. Water was suppl1ed ad lib1tum by 2 Plasson waterers per pen. Feed was restricted...

  11. Waste Stream Generated and Waste Disposal Plans for Molten Salt Reactor Experiment at Oak Ridge National Laboratory

    SciTech Connect (OSTI)

    Haghighi, M. H.; Szozda, R. M.; Jugan, M. R.

    2002-02-26T23:59:59.000Z

    The Molten Salt Reactor Experiment (MSRE) site is located in Tennessee, on the U.S. Department of Energy (DOE) Oak Ridge Reservation (ORR), south of the Oak Ridge National Laboratory (ORNL) main plant across Haw Ridge in Melton Valley. The MSRE was run by ORNL to demonstrate the desirable features of the molten-salt concept in a practical reactor that could be operated safely and reliably. It introduced the idea of a homogeneous reactor using fuel salt media and graphite moderation for power and breeder reactors. The MSRE reactor and associated components are located in cells beneath the floor in the high-bay area of Building 7503 (Figure 1). The reactor was operated from June 1965 to December 1969. When the reactor was shut down, fuel salt was drained from the reactor circuit to two drain tanks. A ''clean'' salt was then circulated through the reactor as a decontamination measure and drained to a third drain tank. When operations ceased, the fuel and flush salts were allowed t o cool and solidify in the drain tanks. At shutdown, the MSRE facility complex was placed in a surveillance and maintenance program. As a result of the S&M program, it was discovered in 1994 that gaseous uranium (233U/232U) hexafluoride (UF6) had moved throughout the MSRE process systems. The UF6 was generated when radiolysis of the fluorine salts caused the individual constituents to dissociate to their component atoms, including free fluorine.Some of the free fluorine combined with uranium fluorides (UF4) in the salt to form UF6. UF6 is gaseous at slightly above ambient temperatures; thus, periodic heating of the fuel salts (which was intended to remedy the radiolysis problems) and simple diffusion had allowed the UF6 to move out of the salt and into the process systems of MSRE.

  12. Th/U-233 multi-recycle in pressurized water reactors : feasibility study of multiple homogeneous and heterogeneous assembly designs.

    SciTech Connect (OSTI)

    Yun, D.; Taiwo, T. A.; Kim, T. K.; Mohamed, A.; Nuclear Engineering Division

    2010-10-01T23:59:59.000Z

    The use of thorium in current or advanced light water reactors (LWRs) has been of interest in recent years. These interests have been associated with the need to increase nuclear fuel resources and the perceived non-proliferation advantages of the utilization of thorium in the fuel cycle. Various options have been considered for the use of thorium in the LWR fuel cycle. The possibility for thorium utilization in a multi-recycle system has also been considered in past literature, primarily because of the potential for near breeders with Th/U-233 in the thermal energy range. The objective of this study is to evaluate the potential of Th/U-233 fuel multi-recycle in current LWRs, focusing on pressurized water reactors (PWRs). Approaches for sustainable multi-recycle without the need for external fissile material makeup have been investigated. The intent is to obtain a design that allows existing PWRs to be used with minimal modifications.

  13. Current status of the Run-Beyond-Cladding Breach (RBCB) tests for the Integral Fast Reactor (IFR). Metallic Fuels Program

    SciTech Connect (OSTI)

    Batte, G.L.; Pahl, R.G. [Argonne National Lab., Idaho Falls, ID (United States); Hofman, G.L. [Argonne National Lab., IL (United States)

    1993-09-01T23:59:59.000Z

    This paper describes the results from the Integral Fast Reactor (IFR) metallic fuel Run-Beyond-Cladding-Breach (RBCB) experiments conducted in the Experimental Breeder Reactor II (EBR-II). Included in the report are scoping test results and the data collected from the prototypical tests as well as the exam results and discussion from a naturally occurring breach of one of the lead IFR fuel tests. All results showed a characteristic delayed neutron and fission gas release pattern that readily allows for identification and evaluation of cladding breach events. Also, cladding breaches are very small and do not propagate during extensive post breach operation. Loss of fuel from breached cladding was found to be insignificant. The paper will conclude with a brief description of future RBCB experiments planned for irradiation in EBR-II.

  14. Nuclear reactor control column

    SciTech Connect (OSTI)

    Bachovchin, D.M.

    1982-08-10T23:59:59.000Z

    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 crosssectional 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.

  15. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Edler, S. K.

    1981-07-01T23:59:59.000Z

    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.

  16. Nuclear reactor control column

    DOE Patents [OSTI]

    Bachovchin, Dennis M. (Plum Borough, PA)

    1982-01-01T23:59:59.000Z

    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.

  17. Mental Models of Physical Mechanisms and Their

    E-Print Network [OSTI]

    de Kleer, Johan

    is that of mechanistic devices, including physical machines, electronic and hydraulic systems, and even hybrids such as electro--mechanical systems. Our top-level goals are: (1) to investigate what it means for a person to understand a complex system, in particular, the mental models that experts form of how a system functions

  18. Code qualification of structural materials for AFCI advanced recycling reactors.

    SciTech Connect (OSTI)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31T23:59:59.000Z

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP) and the Power Reactor Innovative Small Module (PRISM), the NRC/Advisory Committee on Reactor Safeguards (ACRS) raised numerous safety-related issues regarding elevated-temperature structural integrity criteria. Most of these issues remained unresolved today. These critical licensing reviews provide a basis for the evaluation of underlying technical issues for future advanced sodium-cooled reactors. Major materials performance issues and high temperature design methodology issues pertinent to the ARR are addressed in the report. The report is organized as follows: the ARR reference design concepts proposed by the Argonne National Laboratory and four industrial consortia were reviewed first, followed by a summary of the major code qualification and licensing issues for the ARR structural materials. The available database is presented for the ASME Code-qualified structural alloys (e.g. 304, 316 stainless steels, 2.25Cr-1Mo, and mod.9Cr-1Mo), including physical properties, tensile properties, impact properties and fracture toughness, creep, fatigue, creep-fatigue interaction, microstructural stability during long-term thermal aging, material degradation in sodium environments and effects of neutron irradiation for both base metals and weld metals. An assessment of modified versions of Type 316 SS, i.e. Type 316LN and its Japanese version, 316FR, was conducted to provide a perspective for codification of 316LN or 316FR in Subsection NH. Current status and data availability of four new advanced alloys, i.e. NF616, NF616+TMT, NF709, and HT-UPS, are also addressed to identify the R&D needs for their code qualification for ARR applications. For both conventional and new alloys, issues related to high temperature design methodology are described to address the needs for improvements for the ARR design and licensing. Assessments have shown that there are significant data gaps for the full qualification and licensing of the ARR structural materials. Development and evaluation of structural materials require a variety of experimental facilities that have been seriously degraded

  19. Internship Handbook For Mental Health and College Counseling

    E-Print Network [OSTI]

    Internship Handbook For Mental Health and College Counseling Internship in Mental Health Counseling COUN 667 Internship in College Counseling COUN 666 Graduate Counseling Program Old Dominion University January 2014 #12;Mental Health & College Counselor Internship Handbook January 2014 2 CONTENTS I

  20. Extracting Team Mental Models Through Textual Analysis Kathleen M. Carley

    E-Print Network [OSTI]

    Sadeh, Norman M.

    Extracting Team Mental Models Through Textual Analysis Kathleen M. Carley Associate Professor of this paper appears in: Kathleen M. Carley, 1997, "Extracting Team Mental Models Through Textual Analysis for this paper, but for future work in this area. #12;-- 1 -- Extracting Team Mental Models Through Textual

  1. National Institutes of Health National Institute of Mental Health

    E-Print Network [OSTI]

    Baker, Chris I.

    National Institutes of Health National Institute of Mental Health Department of Health and HumanNational Institute of Mental Health Division of Intramural Research Programs http://intramural.nimh.nih.gov/ [NIMH of Fellowship Training] National Institutes of Health National Institute of Mental Health Department of Health

  2. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Martin Peng, Y.K.M.

    1985-10-03T23:59:59.000Z

    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.

  3. Nuclear reactor reflector

    DOE Patents [OSTI]

    Hopkins, Ronald J. (Pensacola, FL); Land, John T. (Pensacola, FL); Misvel, Michael C. (Pensacola, FL)

    1994-01-01T23:59:59.000Z

    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.

  4. Nuclear reactor reflector

    DOE Patents [OSTI]

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

    1994-06-07T23:59:59.000Z

    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.

  5. Microfluidic electrochemical reactors

    DOE Patents [OSTI]

    Nuzzo, Ralph G. (Champaign, IL); Mitrovski, Svetlana M. (Urbana, IL)

    2011-03-22T23:59:59.000Z

    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.

  6. TECHNIQUES FOR MONITORING PLUTONIUM IN THE ENVIRONMENT

    E-Print Network [OSTI]

    Nero Jr., A.V.

    2011-01-01T23:59:59.000Z

    Liquid Metal Fast Breeder Reactor Program" (3 vols. ), U.S.8) Ci Liquid-metal fast breeder reactor fueled with deEletedof water-cooled reactors, the In fast breeder In amount of

  7. DAMAGE TO MITOCHONDRIAL ELECTRON TRANSPORT AND ENERGY COUPLING BY VISIBLE LIGHT

    E-Print Network [OSTI]

    Aggarwal, B.B.

    2011-01-01T23:59:59.000Z

    Liquid Metal Fast Breeder Reactor Program" (3 vols. ), U.S.Ci Liquid-metal fast breeder reactor fueled with depletedof water-cooled reactors, the In fast breeder In amount of

  8. Hypothetical Reactor Accident Study

    E-Print Network [OSTI]

    POPULATIONS; IODINE 131; MELTDOWN; METEOROLOGY; NUCLEAR POWER PLANTS; P CODES; PWR TYPE REACTORS; RADIATION in a Typical BWR and in a typical PWR. Comparison with WASH-1400 by C F . Højerup 202 APPENDIX 3. Calculation

  9. P Reactor Grouting

    SciTech Connect (OSTI)

    None

    2010-01-01T23:59:59.000Z

    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.

  10. Polymerization reactor control

    SciTech Connect (OSTI)

    Ray, W.H.

    1985-01-01T23:59:59.000Z

    The principal difficulties in achieving good control of polymerization reactors are related to inadequate on-line measurement, a lack of understanding of the dynamics of the process, the highly sensitive and nonlinear behavior of these reactors, and the lack of well-developed techniques for the control of nonlinear processes. Some illustrations of these problems and a discussion of potential techniques for overcoming some of these difficulties is provided.

  11. Molten metal reactors

    DOE Patents [OSTI]

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

    2013-11-05T23:59:59.000Z

    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.

  12. Reactor- Nuclear Science Center 

    E-Print Network [OSTI]

    Unknown

    2011-08-17T23:59:59.000Z

    A COMPARISON OF NUCLEAR REACTOR CONTROL ROOM DISPLAY PANELS A Thesis by FRANCES RENAE BOWERS Submitted to the Graduate College of Texas A&M University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE May 1988... Major Subject: Industrial Engineering A COMPARISON OF NUCLEAR REACTOR CONTROL ROOM DISPLAY PANELS A Thesis by FRANCES RENAE BOWERS Approved as to style and content by: Rod er . oppa (Cha' of 'ttee) R. Quinn Brackett (Member) rome . Co gleton...

  13. Power Burst Facility (PBF) Reactor Reactor Decommissioning

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level:Energy: Grid Integration Redefining What's Possible forPortsmouth/Paducah Project Office PressPostdoctoraldecadal7Powder DropperReactor

  14. Reactor Safety Research Programs

    SciTech Connect (OSTI)

    Dotson, CW

    1980-08-01T23:59:59.000Z

    This document summarizes the work performed by Pacific Northwest laboratory from October 1 through December 31, 1979, for the Division of Reactor Safety Research within the Nuclear Regulatory Commission. Evaluation of nondestructive examination (NDE) techniques and instrumentation are reported; areas of investigation include demonstrating the feasibilty of determining structural graphite strength, evaluating the feasibilty of detecting and analyzing flaw growth in reactor pressure boundary systems, examining NDE reliability and probabilistic fracture mechanics, and assessing the remaining integrity of pressurized water reactor steam generator tubes where service-induced degradation has been indicated. Test assemblies and analytical support are being provided for experimental programs at other facilities. These programs include the loss-of-coolant accident simulation tests at the NRU reactor, Chalk River, Canada; the fuel rod deformation and post-accident coolability tests for the ESSOR Test Reactor Program, lspra, Italy; the blowdown and reflood tests in the test facility at Cadarache, France; the instrumented fuel assembly irradiation program at Halden, Norway; and the experimental programs at the Power Burst Facility, Idaho National Engineering Laboratory. These programs will provide data for computer modeling of reactor system and fuel performance during various abnormal operating conditions.

  15. F Reactor Inspection

    SciTech Connect (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29T23:59:59.000Z

    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."

  16. F Reactor Inspection

    ScienceCinema (OSTI)

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-11-24T23:59:59.000Z

    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."

  17. Argonne, KAERI to develop prototype nuclear reactor | Argonne...

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

    that incorporates an innovative metal fuel developed at Argonne. The fuel's inherent safety potential was demonstrated in landmark tests conducted on the Experimental Breeder...

  18. Overview of Fusion-Fission Hybrid Reactor Design Study in China

    SciTech Connect (OSTI)

    Huang Jinhua [Southwestern Institute of Physics (China); Feng Kaiming [Southwestern Institute of Physics (China); Deng Baiquan [Southwestern Institute of Physics (China); Deng, P.Zh. [Southwestern Institute of Physics (China); Zhang Guoshu [Southwestern Institute of Physics (China); Hu Gang [Southwestern Institute of Physics (China); He Kaihui [Southwestern Institute of Physics (China); Wu Yican [Institute of Plasma Physics (China); Qiu Lijian [Institute of Plasma Physics (China); Huang Qunying [Institute of Plasma Physics (China); Xiao Bingjia [Institute of Plasma Physics (China); Liu Xiaoping [Institute of Plasma Physics (China); Chen Yixue [Institute of Plasma Physics (China); Kong, M.H. [Institute of Plasma Physics (China)

    2002-07-15T23:59:59.000Z

    The motivation for developing fusion-fission hybrid reactors is discussed in the context of electricity power requirements by 2050 in China. A detailed conceptual design of the Fusion Experimental Breeder (FEB) was developed from 1986-1995. The FEB has a subignited tokamak fusion core with a major radius of 4.0 m, a fusion power of 145 MW, and a fusion energy gain Q of 3. Based on this, an engineering outline design study of the FEB, FEB-E, has been performed. This design study is a transition from conceptual to engineering design in this research. The main results beyond that given in the detailed conceptual design are included in this paper, namely, the design studies of the blanket, divertor, test blanket, and tritium and environment issues. In-depth analyses have been performed to support the design. Studies of related advanced concepts such as the waste transmutation blanket concept and the spherical tokamak core concept are also presented.

  19. Failed fuel monitoring and surveillance techniques for liquid metal cooled fast reactors

    SciTech Connect (OSTI)

    Lambert, J.D.B.; Mikaili, R.; Gross, K.C.; Strain, R.V. [Argonne National Lab., IL (United States); Aoyama, T.; Ukai, S.; Nomura, S.; Nakae, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan)

    1995-05-01T23:59:59.000Z

    The Experimental Breeder Reactor II (EBR-II) has been used as a facility for irradiation of LMR fuels and components for thirty years. During this time many tests of experimental fuel were continued to cladding breach in order to study modes of element failure; the methods used to identify such failures are described in a parallel paper. This paper summarizes experience of monitoring the delayed-neutron (DN) and fission-gas (FG) release behavior of a smaller number of elements that continued operation in the run-beyond-cladding-breach (RBCB) mode. The scope of RBCB testing, the methods developed to characterize failures on-line, and examples of DN/FG behavior are described.

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

    SciTech Connect (OSTI)

    P. Delmolino

    2005-05-06T23:59:59.000Z

    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.

  1. Zero Power Reactor simulation | Argonne National Laboratory

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

    Zero Power Reactor simulation Share Description Ever wanted to see a nuclear reactor core in action? Here's a detailed simulation of the Zero Power Reactor experiment, run by...

  2. Methanation assembly using multiple reactors

    DOE Patents [OSTI]

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24T23:59:59.000Z

    A methanation assembly for use with a water supply and a gas supply containing gas to be methanated in which a reactor assembly has a plurality of methanation reactors each for methanating gas input to the assembly and a gas delivery and cooling assembly adapted to deliver gas from the gas supply to each of said methanation reactors and to combine water from the water supply with the output of each methanation reactor being conveyed to a next methanation reactor and carry the mixture to such next methanation reactor.

  3. A TEN MEGAWATT BOILING HETEROGENEOUS PACKAGE POWER REACTOR. Reactor...

    Office of Scientific and Technical Information (OSTI)

    A reactor and associated power plant designed to produce 1.05 Mwh and 3.535 Mwh of steam for heating purposes are described. The total thermal output of the reactor is 10 Mwh....

  4. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, Jr., Lawrence A. (Bellaire, TX); Hearn, Dennis (Houston, TX); Jones, Jr., Edward M. (Friendswood, TX)

    1993-01-01T23:59:59.000Z

    A liquid phase process for oligomerization of C.sub.4 and C.sub.5 isoolefins or the etherification thereof with C.sub.1 to C.sub.6 alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120.degree. to 300.degree. F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  5. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-08-01T23:59:59.000Z

    Paper presented at the 29th IECEC in Monterey, CA in August 1994. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their (thermionic reactor) performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling.

  6. Reactor for exothermic reactions

    DOE Patents [OSTI]

    Smith, L.A. Jr.; Hearn, D.; Jones, E.M. Jr.

    1993-03-02T23:59:59.000Z

    A liquid phase process is described for oligomerization of C[sub 4] and C[sub 5] isoolefins or the etherification thereof with C[sub 1] to C[sub 6] alcohols wherein the reactants are contacted in a reactor with a fixed bed acid cation exchange resin catalyst at an LHSV of 5 to 20, pressure of 0 to 400 psig and temperature of 120 to 300 F. Wherein the improvement is the operation of the reactor at a pressure to maintain the reaction mixture at its boiling point whereby at least a portion but less than all of the reaction mixture is vaporized. By operating at the boiling point and allowing a portion of the reaction mixture to vaporize, the exothermic heat of reaction is dissipated by the formation of more boil up and the temperature in the reactor is controlled.

  7. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, Ernest (Wilmette, IL)

    1986-01-01T23:59:59.000Z

    A safety device is disclosed for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of an upward thermal excursion. Such safety device comprises a laminated helical ribbon configured as a tube-like helical coil having contiguous helical turns with slidably abutting edges. The helical coil is disclosed as a portion of a drive member connected axially to the control rod. The laminated ribbon is formed of outer and inner laminae. The material of the outer lamina has a greater thermal coefficient of expansion than the material of the inner lamina. In the event of an upward thermal excursion, the laminated helical coil curls inwardly to a smaller diameter. Such inward curling causes the total length of the helical coil to increase by a substantial increment, so that the control rod is axially repositioned by a corresponding amount to reduce the power output of the reactor.

  8. Fusion reactor control

    SciTech Connect (OSTI)

    Plummer, D.A.

    1995-12-31T23:59:59.000Z

    The plasma kinetic temperature and density changes, each per an injected fuel density rate increment, control the energy supplied by a thermonuclear fusion reactor in a power production cycle. This could include simultaneously coupled control objectives for plasma current, horizontal and vertical position, shape and burn control. The minimum number of measurements required, use of indirect (not plasma parameters) system measurements, and distributed control procedures for burn control are to be verifiable in a time dependent systems code. The International Thermonuclear Experimental Reactor (ITER) has the need to feedback control both the fusion output power and the driven plasma current, while avoiding damage to diverter plates. The system engineering of fusion reactors must be performed to assure their development expeditiously and effectively by considering reliability, availability, maintainability, environmental impact, health and safety, and cost.

  9. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    1987-01-01T23:59:59.000Z

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extends from the metal base plate downwardly and outwardly into the earth.

  10. Heat dissipating nuclear reactor

    DOE Patents [OSTI]

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21T23:59:59.000Z

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  11. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID); Berry, Ray A. (Idaho Falls, ID)

    1999-01-01T23:59:59.000Z

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream.

  12. Fast quench reactor method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.; Berry, R.A.

    1999-08-10T23:59:59.000Z

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a means of rapidly expanding a reactant stream, such as a restrictive convergent-divergent nozzle at its outlet end. Metal halide reactants are injected into the reactor chamber. Reducing gas is added at different stages in the process to form a desired end product and prevent back reactions. The resulting heated gaseous stream is then rapidly cooled by expansion of the gaseous stream. 8 figs.

  13. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, Daniel L. (Princeton, NJ)

    1988-01-01T23:59:59.000Z

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam.

  14. Reactor Neutrino Experiments

    E-Print Network [OSTI]

    Jun Cao

    2007-12-06T23:59:59.000Z

    Precisely measuring $\\theta_{13}$ is one of the highest priority in neutrino oscillation study. Reactor experiments can cleanly determine $\\theta_{13}$. Past reactor neutrino experiments are reviewed and status of next precision $\\theta_{13}$ experiments are presented. Daya Bay is designed to measure $\\sin^22\\theta_{13}$ to better than 0.01 and Double Chooz and RENO are designed to measure it to 0.02-0.03. All are heading to full operation in 2010. Recent improvements in neutrino moment measurement are also briefed.

  15. Perspectives on reactor safety

    SciTech Connect (OSTI)

    Haskin, F.E. [New Mexico Univ., Albuquerque, NM (United States). Dept. of Chemical and Nuclear Engineering; Camp, A.L. [Sandia National Labs., Albuquerque, NM (United States)

    1994-03-01T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) maintains a technical training center at Chattanooga, Tennessee to provide appropriate training to both new and experienced NRC employees. This document describes a one-week course in reactor, safety concepts. The course consists of five modules: (1) historical perspective; (2) accident sequences; (3) accident progression in the reactor vessel; (4) containment characteristics and design bases; and (5) source terms and offsite consequences. The course text is accompanied by slides and videos during the actual presentation of the course.

  16. Innovative design of uranium startup fast reactors

    E-Print Network [OSTI]

    Fei, Tingzhou

    2012-01-01T23:59:59.000Z

    Sodium Fast Reactors are one of the three candidates of GEN-IV fast reactors. Fast reactors play an important role in saving uranium resources and reducing nuclear wastes. Conventional fast reactors rely on transuranic ...

  17. Reactor operation environmental information document

    SciTech Connect (OSTI)

    Haselow, J.S.; Price, V.; Stephenson, D.E.; Bledsoe, H.W.; Looney, B.B.

    1989-12-01T23:59:59.000Z

    The Savannah River Site (SRS) produces nuclear materials, primarily plutonium and tritium, to meet the requirements of the Department of Defense. These products have been formed in nuclear reactors that were built during 1950--1955 at the SRS. K, L, and P reactors are three of five reactors that have been used in the past to produce the nuclear materials. All three of these reactors discontinued operation in 1988. Currently, intense efforts are being extended to prepare these three reactors for restart in a manner that protects human health and the environment. To document that restarting the reactors will have minimal impacts to human health and the environment, a three-volume Reactor Operations Environmental Impact Document has been prepared. The document focuses on the impacts of restarting the K, L, and P reactors on both the SRS and surrounding areas. This volume discusses the geology, seismology, and subsurface hydrology. 195 refs., 101 figs., 16 tabs.

  18. Power calibrations for TRIGA reactors

    SciTech Connect (OSTI)

    Whittemore, W.L.; Razvi, J.; Shoptaugh, J.R. Jr. [General Atomics, San Diego, CA (United States)

    1988-07-01T23:59:59.000Z

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than {+-} 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to {+-} 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  19. Reactor operation safety information document

    SciTech Connect (OSTI)

    Not Available

    1990-01-01T23:59:59.000Z

    The report contains a reactor facility description which includes K, P, and L reactor sites, structures, operating systems, engineered safety systems, support systems, and process and effluent monitoring systems; an accident analysis section which includes cooling system anomalies, radioactive materials releases, and anticipated transients without scram; a summary of onsite doses from design basis accidents; severe accident analysis (reactor core disruption); a description of operating contractor organization and emergency planning; and a summary of reactor safety evolution. (MB)

  20. Reed Reactor Facility Annual Report

    SciTech Connect (OSTI)

    Frantz, Stephen G.

    2000-09-01T23:59:59.000Z

    This is the report of the operations, experiments, modifications, and other aspects of the Reed Reactor Facility for the year.

  1. ataxia mental retardation: Topics by E-print Network

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

    syndromes have helped (more) Vaidyanathan, Umamaheswari 2011-01-01 399 FERN CREEK MENTAL AND BEHAVIORAL HEALTH HOSPITAL DESIGN. Open Access Theses and Dissertations...

  2. Performance indicators for public mental healthcare: A systematic international inventory

    E-Print Network [OSTI]

    Lauriks, Steve; Buster, Marcel CA; de Wit, Matty AS; Arah, Onyebuchi A; Klazinga, Niek S

    2012-01-01T23:59:59.000Z

    Lauriks et al. : Performance indicators for public mentaland properties of unique performance indicators for publicOA, Klazinga NS: Performance indicators for public mental

  3. acute inpatient mental: Topics by E-print Network

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

    University of Chicago Diagnosis and Management in Internal Medicine 1. Applies clinical knowledge in conducting a thorough history Issa, Naoum 11 Mental Games University of...

  4. Analysis of Assembly Bill 1600: Mental Health Services

    E-Print Network [OSTI]

    California Health Benefits Review Program (CHBRP)

    2010-01-01T23:59:59.000Z

    and outpatient mental health services use. Social Sciencegroup insurers, nonprofit health service plans, and HMOs toA. Utilization of health services among patients referred to

  5. Analysis of Senate Bill 572: Mental Health Services

    E-Print Network [OSTI]

    California Health Benefits Review Program (CHBRP)

    2005-01-01T23:59:59.000Z

    health specialty care. Health Services Research. 39 (5)Journal of Behavioral Health Services and Research. 26 (4)Abuse and Mental Health Services Administration, Center for

  6. adult mental health: Topics by E-print Network

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

    A&M University - TxSpace Summary: Knowledge concerning the current utilization of mental health care by the nation's geriatric population is greatly lacking. Research conducted...

  7. adolescent mental health: Topics by E-print Network

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

    including posttraumatic (more) Nickerson, Angela Marissa 2009-01-01 45 Health Care Reform: What School Mental Health Professionals Need to Know CiteSeer Summary: care...

  8. A point-kinetics approach to sensitivity study of fast reactor dynamic behavior and delayed neutron spectra

    SciTech Connect (OSTI)

    Das, S. [Bhabha Atomic Research Centre, Bombay (India). Theoretical Physics Div.

    1996-03-01T23:59:59.000Z

    The method of point reactor kinetics in conjunction with the new concepts of delayed spectrum factor and beta growth factor is used to calculate the sensitivity of the dynamic behavior of a fast breeder reactor to large changes in delayed neutron energies following postulated reactivity accidents. The positive ramp rates are introduced not to simulate physical possibilities but solely to test the sensitivity to delayed neutron spectral changes under different conditions. A limited number of transient calculations are made using the point-kinetics code SENSTVTY, six precursor groups, and Doppler feedback. The calculational method and the reactor model are described. Delayed neutron requirements in reactor dynamics are discussed, and a brief review of the sensitivity studies is presented. The results of the sensitivity calculations indicate that the relative power, the peak power, and the accident energy release are sensitive to changes in {beta}{sub eff} resulting from uncertainty in the delayed spectral data, but the sensitivity of the relative power is much greater than the peak power and the accident energy release. The spread in the maximum reactivity reached is found to be {approximately}18%, and the time spread in the melting of fuel and cladding is in milliseconds.

  9. Nuclear Reactors and Technology

    SciTech Connect (OSTI)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01T23:59:59.000Z

    This publication Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency`s Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, is available to DOE and its contractors through the DOE Integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user`s needs.

  10. NETL - Chemical Looping Reactor

    ScienceCinema (OSTI)

    None

    2014-06-26T23:59:59.000Z

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  11. Thermal Reactor Safety

    SciTech Connect (OSTI)

    Not Available

    1980-06-01T23:59:59.000Z

    Information is presented concerning fire risk and protection; transient thermal-hydraulic analysis and experiments; class 9 accidents and containment; diagnostics and in-service inspection; risk and cost comparison of alternative electric energy sources; fuel behavior and experiments on core cooling in LOCAs; reactor event reporting analysis; equipment qualification; post facts analysis of the TMI-2 accident; and computational methods.

  12. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, Perng-Fei (Saratoga, CA); Townsend, Harold E. (Campbell, CA); Barbanti, Giancarlo (Sirtori, IT)

    1994-01-01T23:59:59.000Z

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

  13. Nuclear reactor building

    DOE Patents [OSTI]

    Gou, P.F.; Townsend, H.E.; Barbanti, G.

    1994-04-05T23:59:59.000Z

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed there above. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define there between an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin. 4 figures.

  14. Fossil fuel furnace reactor

    DOE Patents [OSTI]

    Parkinson, William J. (Los Alamos, NM)

    1987-01-01T23:59:59.000Z

    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.

  15. NETL - Chemical Looping Reactor

    SciTech Connect (OSTI)

    None

    2013-07-24T23:59:59.000Z

    NETL's Chemical Looping Reactor unit is a high-temperature integrated CLC process with extensive instrumentation to improve computational simulations. A non-reacting test unit is also used to study solids flow at ambient temperature. The CLR unit circulates approximately 1,000 pounds per hour at temperatures around 1,800 degrees Fahrenheit.

  16. Reactor Monitoring with Neutrinos

    E-Print Network [OSTI]

    M. Cribier

    2007-04-06T23:59:59.000Z

    The fundamental knowledge on neutrinos acquired in the recent years open the possibility of applied neutrino physics. Among it the automatic and non intrusive monitoring of nuclear reactor by its antineutrino signal could be very valuable to IAEA in charge of the control of nuclear power plants. Several efforts worldwide have already started.

  17. Reactor vessel support system. [LMFBR

    DOE Patents [OSTI]

    Golden, M.P.; Holley, J.C.

    1980-05-09T23:59:59.000Z

    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.

  18. China's Nuclear Industry After Fukushima

    E-Print Network [OSTI]

    YUAN, Jingdong

    2013-01-01T23:59:59.000Z

    China Experimen- tal Fast Reactor, or CEFR) and highpressurized water reactor—fast reactors—fusion reactor,”experimental fast breeder reactors, and heavy water

  19. Activation Analysis for Two Molten Salt Dual-Coolant Blanket Concepts for the U.S. Demo Reactor

    SciTech Connect (OSTI)

    Youssef, M.Z. [University of California-Los Angeles (United States); Sawan, M.E. [University of Wisconsin-Madison (United States)

    2005-04-15T23:59:59.000Z

    The US has considered, among other options, two blanket concepts for Demo reactor in which helium is primarily used to cool the first wall (FW) and structure whereas molten salt (MS) is used as both coolant and breeder. Conventional reduced activation ferritic steel (RAFS, F82H) is used as the structural material in both blanket concepts. The low melting point Flibe ({approx}380 deg. C) is used in the first option while the Flinabe ({approx}305 deg. C) is used in the second option. In this paper, we present the results for assessing the radioactivity and decay heat. This assessment is performed separately for the structural material, the Be multiplier and the breeder (Flibe/Flinabe). The Class C waste disposal rating (WDR) was estimated for each material. For Flibe, Flinabe and Be the WDR is much lower than unity. However, the WDR for F82H is {approx}0.6-1.3. They are attributed to reactions with Mo and Nb present in F82H with levels of 70 wppm and 4 wppm, respectively. To ensure that F82H qualifies for shallow land burial, it is suggested to reduce these two impurities to {approx}50 and {approx}3 wppm, respectively.

  20. Nuclear reactor construction with bottom supported reactor vessel

    DOE Patents [OSTI]

    Sharbaugh, John E. (Bullskin Township, Fayette County, PA)

    1987-01-01T23:59:59.000Z

    An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core within the pool. The reactor vessel has an open top end, a closed flat bottom end wall and a continuous cylindrical closed side wall interconnecting the top end and bottom end wall. The reactor also has a generally cylindrical concrete containment structure surrounding the reactor vessel and being formed by a cylindrical side wall spaced outwardly from the reactor vessel side wall and a flat base mat spaced below the reactor vessel bottom end wall. A central support pedestal is anchored to the containment structure base mat and extends upwardly therefrom to the reactor vessel and upwardly therefrom to the reactor core so as to support the bottom end wall of the reactor vessel and the lower end of the reactor core in spaced apart relationship above the containment structure base mat. Also, an annular reinforced support structure is disposed in the reactor vessel on the bottom end wall thereof and extends about the lower end of the core so as to support the periphery thereof. In addition, an annular support ring having a plurality of inward radially extending linear members is disposed between the containment structure base mat and the bottom end of the reactor vessel wall and is connected to and supports the reactor vessel at its bottom end on the containment structure base mat so as to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event. The reactor construction also includes a bed of insulating material in sand-like granular form, preferably being high density magnesium oxide particles, disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall on the containment structure base mat so as to insulate the reactor vessel bottom end wall from the containment structure base mat and allow the reactor vessel bottom end wall to freely expand as it heats up while providing continuous support thereof. Further, a deck is supported upon the side wall of the containment structure above the top open end of the reactor vessel, and a plurality of serially connected extendible and retractable annular bellows extend between the deck and the top open end of the reactor vessel and flexibly and sealably interconnect the reactor vessel at its top end to the deck. An annular guide ring is disposed on the containment structure and extends between its side wall and the top open end of the reactor vessel for providing lateral support of the reactor vessel top open end by limiting imposition of lateral loads on the annular bellows by the occurrence of a lateral seismic event.

  1. Adult Mental Health Needs in California: Findings from the 2007 California Health Interview Survey

    E-Print Network [OSTI]

    Grant, David; Padilla-Frausto, Imelda; Aydin, May; Streja, Leanne; Aguilar-Gaxiola, Sergio; Caldwell, Julia

    2011-01-01T23:59:59.000Z

    Health (DMH). Mental Health Services Act. Accessed June 29,prop_63/mhsa/docs/Mental_Health_Services_Act_Ful l_ Text.pdfand Use of Mental Health Services by California Adults. UCLA

  2. Using a Mental Measurements Yearbook Review to Evaluate a Test

    E-Print Network [OSTI]

    Abolmaesumi, Purang

    1 Using a Mental Measurements Yearbook Review to Evaluate a Test Anthony J. Nitko Professor, University of Arizona Introduction Once you have located a test, you will want to read its Mental Measurements Yearbook (MMY) review. You need to use the review to make judgments about the quality of the test

  3. Interactive Dimensions in the Construction of Mental Representations for Text

    E-Print Network [OSTI]

    Patel, Aniruddh D.

    Interactive Dimensions in the Construction of Mental Representations for Text David N. Rapp be as critical to the construction of complex mental models as the discrete dimensions themselves. In the present a bead on Specify again. Incredibly, the horse was still rolling along. A pang of fear went through Woolf

  4. Nuclear reactor safety device

    DOE Patents [OSTI]

    Hutter, E.

    1983-08-15T23:59:59.000Z

    A safety device is described for use in a nuclear reactor for axially repositioning a control rod with respect to the reactor core in the event of a thermal excursion. It comprises a laminated strip helically configured to form a tube, said tube being in operative relation to said control rod. The laminated strip is formed of at least two materials having different thermal coefficients of expansion, and is helically configured such that the material forming the outer lamina of the tube has a greater thermal coefficient of expansion than the material forming the inner lamina of said tube. In the event of a thermal excursion the laminated strip will tend to curl inwardly so that said tube will increase in length, whereby as said tube increases in length it exerts a force on said control rod to axially reposition said control rod with respect to said core.

  5. Fusion reactor pumped laser

    DOE Patents [OSTI]

    Jassby, D.L.

    1987-09-04T23:59:59.000Z

    A nuclear pumped laser capable of producing long pulses of very high power laser radiation is provided. A toroidal fusion reactor provides energetic neutrons which are slowed down by a moderator. The moderated neutrons are converted to energetic particles capable of pumping a lasing medium. The lasing medium is housed in an annular cell surrounding the reactor. The cell includes an annular reflecting mirror at the bottom and an annular output window at the top. A neutron reflector is disposed around the cell to reflect escaping neutrons back into the cell. The laser radiation from the annular window is focused onto a beam compactor which generates a single coherent output laser beam. 10 figs.

  6. Spherical torus fusion reactor

    DOE Patents [OSTI]

    Peng, Yueng-Kay M. (Oak Ridge, TN)

    1989-01-01T23:59:59.000Z

    A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

  7. Nuclear reactor sealing system

    DOE Patents [OSTI]

    McEdwards, James A. (Calabasas, CA)

    1983-01-01T23:59:59.000Z

    A liquid metal-cooled nuclear reactor sealing system. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel.

  8. The Effect of Placement Change on Foster Children's Utilization of Emergency Mental Health Services

    E-Print Network [OSTI]

    Fawley-King, Kya

    2010-01-01T23:59:59.000Z

    Journal of Behavioral Health Services Research, 23(4), 389-Outpatient mental health services for children in fosteroutpatient mental health service use. Child Abuse and

  9. Analysis of Assembly Bill 423: Health Care Coverage: Mental Health Services

    E-Print Network [OSTI]

    California Health Benefits Review Program (CHBRP)

    2007-01-01T23:59:59.000Z

    health specialty care. Health Services Research. 2004;39:Abuse and Mental Health Services Administration, 2004.Abuse and Mental Health Services Administration, Office of

  10. Evaluation of lysine deficient grower diets for heavy breed replacement pullets and a comparison of sorghum grains and corn as a carbohydrate source for broiler-breeder hens

    E-Print Network [OSTI]

    Tolan, Alan

    1967-01-01T23:59:59.000Z

    EVALUATION OF LYSINE DEFICIENT GROWER DIETS FOR HEAVY BREED REPLACEMENT PULLETS AND A COMPARISON OF SORGHUM GRAINS AND CORN AS A CARBOHYDRATE SOURCE FOR BROILER-BREEDER HENS A Thesis ALAN TOLAN Submitted to the Graduate College of the Texas... ARM University in partial fulfillment of the requirements for the degree of MASTER OF SCIENCE January 1967 Major Subject: Animal Nutrition EVALUATION OF LYSINE DEFICIENT GROWER DIETS FOR HEAVY BREED REPLACEMENT PULLETS AND A COMPARISON OF SORGHUM...

  11. Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future.

  12. Modular Pebble Bed Reactor High Temperature Gas Reactor

    E-Print Network [OSTI]

    Built · Site Assembled · On--line Refueling · Modules added to meet demand. · No Reprocessing · High Experiments · Non-Proliferation · Safeguards · Waste Disposal · Reactor Research/ Demonstration Facility

  13. Thermionic Reactor Design Studies

    SciTech Connect (OSTI)

    Schock, Alfred

    1994-06-01T23:59:59.000Z

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic losses in the electrodes of full-core-height diodes. Moreover, placing the fuel on the outside of the diode makes possible reactors with much higher fuel volume fractions, which enable power-flattened fast reactors scalable to very low power levels without the need for life-limiting hydride moderators or the use of efficiency-limiting driver fuel. In addition, with the fuel on the outside its swelling does not increase the emitter diameter or reduce the interelectrode gap. This should permit long lifetimes even with closer spacings, which can significantly improve the system efficiences. This was confirmed by coupled neutronic, thermal, thermionic, and electrical system analyses - some of which are presented in this paper - and by subsequent experiments. A companion paper presented next describes the fabrication and testing of full-scale converter elements, both fueled and unfueled, and summarizes the test results obtained. There is a duplicate copy in the file.

  14. Designing Reactors to Facilitate Decommissioning

    SciTech Connect (OSTI)

    Richard H. Meservey

    2006-06-01T23:59:59.000Z

    Critics of nuclear power often cite issues with tail-end-of-the-fuel-cycle activities as reasons to oppose the building of new reactors. In fact, waste disposal and the decommissioning of large nuclear reactors have proven more challenging than anticipated. In the early days of the nuclear power industry the design and operation of various reactor systems was given a great deal of attention. Little effort, however, was expended on end-of-the-cycle activities, such as decommissioning and disposal of wastes. As early power and test reactors have been decommissioned difficulties with end-of-the-fuel-cycle activities have become evident. Even the small test reactors common at the INEEL were not designed to facilitate their eventual decontamination, decommissioning, and dismantlement. The results are that decommissioning of these facilities is expensive, time consuming, relatively hazardous, and generates large volumes of waste. This situation clearly supports critics concerns about building a new generation of power reactors.

  15. Fast quench reactor and method

    DOE Patents [OSTI]

    Detering, B.A.; Donaldson, A.D.; Fincke, J.R.; Kong, P.C.

    1998-05-12T23:59:59.000Z

    A fast quench reactor includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This ``freezes`` the desired end product(s) in the heated equilibrium reaction stage. 7 figs.

  16. Fuel Breeding and Core Behavior Analyses on In Core Fuel Management of Water Cooled Thorium Reactors

    SciTech Connect (OSTI)

    Permana, Sidik [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Sekimoto, Hiroshi [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-17, O-okayama, Meguro-ku, Tokyo 152-8550 (Japan); Waris, Abdul; Subhki, Muhamad Nurul [Department of Physics, Bandung Institute of Technology, Gedung Fisika, Jl. Ganesha 10, Bandung 40132 (Indonesia); Ismail, [BAPETEN (Indonesia)

    2010-12-23T23:59:59.000Z

    Thorium fuel cycle with recycled U-233 has been widely recognized having some contributions to improve the water-cooled breeder reactor program which has been shown by a feasible area of breeding and negative void reactivity which confirms that fissile of 233U contributes to better fuel breeding and effective for obtaining negative void reactivity coefficient as the main fissile material. The present study has the objective to estimate the effect of whole core configuration as well as burnup effects to the reactor core profile by adopting two dimensional model of fuel core management. About more than 40 months of cycle period has been employed for one cycle fuel irradiation of three batches fuel system for large water cooled thorium reactors. All position of fuel arrangement contributes to the total core conversion ratio which gives conversion ratio less than unity of at the BOC and it contributes to higher than unity (1.01) at the EOC after some irradiation process. Inner part and central part give the important part of breeding contribution with increasing burnup process, while criticality is reduced with increasing the irradiation time. Feasibility of breeding capability of water-cooled thorium reactors for whole core fuel arrangement has confirmed from the obtained conversion ratio which shows higher than unity. Whole core analysis on evaluating reactivity change which is caused by the change of voided condition has been employed for conservative assumption that 100% coolant and moderator are voided. It obtained always a negative void reactivity coefficient during reactor operation which shows relatively more negative void coefficient at BOC (fresh fuel composition), and it becomes less negative void coefficient with increasing the operation time. Negative value of void reactivity coefficient shows the reactor has good safety properties in relation to the reactivity profile which is the main parameter in term of criticality safety analysis. Therefore, this evaluation has confirmed that breeding condition and negative coefficient can be obtained simultaneously for water-cooled thorium reactor obtains based on the whole core fuel arrangement.

  17. Neutrino Oscillation Studies with Reactors

    E-Print Network [OSTI]

    Petr Vogel; Liangjian Wen; Chao Zhang

    2015-03-03T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  18. Neutrino Oscillation Studies with Reactors

    E-Print Network [OSTI]

    Petr Vogel; Liangjian Wen; Chao Zhang

    2015-04-27T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  19. Neutrino Oscillation Studies with Reactors

    E-Print Network [OSTI]

    Vogel, Petr; Zhang, Chao

    2015-01-01T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective, and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavors are quantum mechanical mixtures. Over the past several decades reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle $\\theta_{13}$. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  20. Progress Update: Reactor Disassembly Grouting

    ScienceCinema (OSTI)

    Cody, Tom

    2012-06-14T23:59:59.000Z

    Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

  1. Progress Update: Reactor Disassembly Grouting

    SciTech Connect (OSTI)

    Cody, Tom

    2010-01-01T23:59:59.000Z

    Grouting the P&R reactors in order to remove these basins as an environmental threat. This will end the Cold War legacy and end the environmental footprint.

  2. Neutrino oscillation studies with reactors

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

    Vogel, P. [California Inst. of Technology (CalTech), Pasadena, CA (United States). Kellog Radiation Lab.; Wen, L.J. [Chinese Academy of Sciences (CAS), Beijing (China). Inst. of High Energy Physics (IHEP); Zhang, C. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-04-27T23:59:59.000Z

    Nuclear reactors are one of the most intense, pure, controllable, cost-effective and well-understood sources of neutrinos. Reactors have played a major role in the study of neutrino oscillations, a phenomenon that indicates that neutrinos have mass and that neutrino flavours are quantum mechanical mixtures. Over the past several decades, reactors were used in the discovery of neutrinos, were crucial in solving the solar neutrino puzzle, and allowed the determination of the smallest mixing angle ?13. In the near future, reactors will help to determine the neutrino mass hierarchy and to solve the puzzling issue of sterile neutrinos.

  3. Thermonuclear Reflect AB-Reactor

    E-Print Network [OSTI]

    Alexander Bolonkin

    2008-03-26T23:59:59.000Z

    The author offers a new kind of thermonuclear reflect reactor. The remarkable feature of this new reactor is a three net AB reflector, which confines the high temperature plasma. The plasma loses part of its energy when it contacts with the net but this loss can be compensated by an additional permanent plasma heating. When the plasma is rarefied (has a small density), the heat flow to the AB reflector is not large and the temperature in the triple reflector net is lower than 2000 - 3000 K. This offered AB-reactor has significantly less power then the currently contemplated power reactors with magnetic or inertial confinement (hundreds-thousands of kW, not millions of kW). But it is enough for many vehicles and ships and particularly valuable for tunnelers, subs and space apparatus, where air to burn chemical fuel is at a premium or simply not available. The author has made a number of innovations in this reactor, researched its theory, developed methods of computation, made a sample computation of typical project. The main point of preference for the offered reactor is its likely cheapness as a power source. Key words: Micro-thermonuclear reactor, Multi-reflex AB-thermonuclear reactor, Self-magnetic AB-thermonuclear reactor, aerospace thermonuclear engine.

  4. Reactor coolant pump flywheel

    DOE Patents [OSTI]

    Finegan, John Raymond; Kreke, Francis Joseph; Casamassa, John Joseph

    2013-11-26T23:59:59.000Z

    A flywheel for a pump, and in particular a flywheel having a number of high density segments for use in a nuclear reactor coolant pump. The flywheel includes an inner member and an outer member. A number of high density segments are provided between the inner and outer members. The high density segments may be formed from a tungsten based alloy. A preselected gap is provided between each of the number of high density segments. The gap accommodates thermal expansion of each of the number of segments and resists the hoop stress effect/keystoning of the segments.

  5. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, J.E.; Meuschke, R.E.

    1995-05-02T23:59:59.000Z

    A method of refueling a nuclear reactor is disclosed whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced. 2 figs.

  6. Nuclear reactor control assembly

    SciTech Connect (OSTI)

    Negron, S.B.

    1991-06-11T23:59:59.000Z

    This patent describes an assembly for providing global power control in a nuclear reactor having the core split into two halves. It comprises a disk assembly formed from at least two disks each machined with an identical surface hole pattern such that rotation of one disk relative to the other causes the hole pattern to open or close, the disk assembly being positioned substantially at the longitudinal center of and coaxial with the core halves; and means for rotating at least one of the disks relative to the other.

  7. Nuclear reactor control apparatus

    SciTech Connect (OSTI)

    Sridhar, B.N.

    1981-08-28T23:59:59.000Z

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additonal magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  8. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1996-02-27T23:59:59.000Z

    A fluidized bed reactor system is described which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves. 3 figs.

  9. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN)

    1993-01-01T23:59:59.000Z

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

  10. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

    1995-01-01T23:59:59.000Z

    A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves.

  11. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, Charles D. (Oak Ridge, TN); Marasco, Joseph A. (Kingston, TN)

    1996-01-01T23:59:59.000Z

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary and tertiary particulate phases, continuously introduced and removed simultaneously in the cocurrent and countercurrent mode, act in a role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Means for introducing and removing the sorbent phases include feed screw mechanisms and multivane slurry valves.

  12. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.

    1993-12-14T23:59:59.000Z

    A fluidized bed reactor system which utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase is described. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figures.

  13. Reactor refueling containment system

    DOE Patents [OSTI]

    Gillett, James E. (Greensburg, PA); Meuschke, Robert E. (Pittsburgh, PA)

    1995-01-01T23:59:59.000Z

    A method of refueling a nuclear reactor whereby the drive mechanism is disengaged and removed by activating a jacking mechanism that raises the closure head. The area between the barrier plate and closure head is exhausted through the closure head penetrations. The closure head, upper drive mechanism, and bellows seal are lifted away and transported to a safe area. The barrier plate acts as the primary boundary and each drive and control rod penetration has an elastomer seal preventing excessive tritium gases from escaping. The individual instrumentation plugs are disengaged allowing the corresponding fuel assembly to be sealed and replaced.

  14. Biparticle fluidized bed reactor

    DOE Patents [OSTI]

    Scott, C.D.; Marasco, J.A.

    1995-04-25T23:59:59.000Z

    A fluidized bed reactor system utilizes a fluid phase, a retained fluidized primary particulate phase, and a migratory second particulate phase. The primary particulate phase is a particle such as a gel bead containing an immobilized biocatalyst. The secondary particulate phase, continuously introduced and removed in either cocurrent or countercurrent mode, acts in a secondary role such as a sorbent to continuously remove a product or by-product constituent from the fluid phase. Introduction and removal of the sorbent phase is accomplished through the use of feed screw mechanisms and multivane slurry valves. 3 figs.

  15. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-11-01T23:59:59.000Z

    Nuclear reactor core safety rod release apparatus comprises a control rod having a detent notch in the form of an annular peripheral recess at its upper end, a control rod support tube for raising and lowering the control rod under normal conditions, latches pivotally mounted on the control support tube with free ends thereof normally disposed in the recess in the control rod, and cam means for pivoting the latches out of the recess in the control rod when a scram condition occurs. One embodiment of the invention comprises an additional magnetically-operated latch for releasing the control rod under two different conditions, one involving seismic shock.

  16. N Reactor - Hanford Site

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

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645U.S. DOE Office of Science (SC)Integrated Codes |IsLove Your1 SECTIONES2008-54174MoreMuseum Day atMyNERSCVanN Reactor

  17. Fast Reactor Fuel Type and Reactor Safety Performance

    SciTech Connect (OSTI)

    R. Wigeland; J. Cahalan

    2009-09-01T23:59:59.000Z

    Fast Reactor Fuel Type and Reactor Safety Performance R. Wigeland , Idaho National Laboratory J. Cahalan, Argonne National Laboratory The sodium-cooled fast neutron reactor is currently being evaluated for the efficient transmutation of the highly-hazardous, long-lived, transuranic elements that are present in spent nuclear fuel. One of the fundamental choices that will be made is the selection of the fuel type for the fast reactor, whether oxide, metal, carbide, nitride, etc. It is likely that a decision on the fuel type will need to be made before many of the related technologies and facilities can be selected, from fuel fabrication to spent fuel reprocessing. A decision on fuel type should consider all impacts on the fast reactor system, including safety. Past work has demonstrated that the choice of fuel type may have a significant impact on the severity of consequences arising from accidents, especially for severe accidents of low probability. In this paper, the response of sodium-cooled fast reactors is discussed for both oxide and metal fuel types, highlighting the similarities and differences in reactor response and accident consequences. Any fast reactor facility must be designed to be able to successfully prevent, mitigate, or accommodate all consequences of potential events, including accidents. This is typically accomplished by using multiple barriers to the release of radiation, including the cladding on the fuel, the intact primary cooling system, and most visibly the reactor containment building. More recently, this has also included the use of ‘inherent safety’ concepts to reduce or eliminate the potential for serious damage in some cases. Past experience with oxide and metal fuel has demonstrated that both fuel types are suitable for use as fuel in a sodium-cooled fast reactor. However, safety analyses for these two fuel types have also shown that there can be substantial differences in accident consequences due to the neutronic and thermophysical properties of the fuel and their compatibility with the reactor coolant, with corresponding differences in the challenges presented to the reactor developers. Accident phenomena are discussed for the sodium-cooled fast reactor based on the mechanistic progression of conditions from accident initiation to accident termination, whether a benign state is achieved or more severe consequences are expected. General principles connecting accident phenomena and fuel properties are developed from the oxide and metal fuel safety analyses, providing guidelines that can be used as part of the evaluation for selection of fuel type for the sodium-cooled fast reactor.

  18. Granular Dynamics in Pebble Bed Reactor Cores

    E-Print Network [OSTI]

    Laufer, Michael Robert

    2013-01-01T23:59:59.000Z

    in a pebble-bed nuclear reactor,” Phys. Rev. E, vol. 74, no.cycles of the pebble bed reactor,” Nuclear Engineering andoptimization of pebble-bed reactors,” Annals of Nuclear

  19. Proceedings of IMECE 2004 2004 ASME International Mechanical Engineering Congress

    E-Print Network [OSTI]

    Bahrami, Majid

    recovery processes, heat exchangers, heat storage sys- tems, the breeder blanket about fusion reactors [1

  20. Gaseous reactor control system

    SciTech Connect (OSTI)

    Abdel-Khalik, S.

    1991-09-03T23:59:59.000Z

    This paper describes a nuclear reactor control system for controlling the reactivity of the core of a nuclear reactor. It includes a control gas having a high neutron cross-section; a first tank containing a first supply of the control gas; a first conduit providing a first fluid passage extending into the core, the first conduit being operatively connected to communicate with the first tank; a first valve operatively connected to regulate the flow of the control gas between the first tank and the first conduit; a second conduit concentrically disposed around the first conduit such that a second fluid passage is defined between the outer surface of the first conduit and the inner surface of the second conduit; a second tank containing a second supply of the control gas, the second tank being operatively connected to communicate with the second fluid passage; a second supply valve operatively connected to regulate the flow of the control gas between the second tank and the second fluid passage.

  1. Solvent refined coal reactor quench system

    DOE Patents [OSTI]

    Thorogood, R.M.

    1983-11-08T23:59:59.000Z

    There is described an improved SRC reactor quench system using a condensed product which is recycled to the reactor and provides cooling by evaporation. In the process, the second and subsequent reactors of a series of reactors are cooled by the addition of a light oil fraction which provides cooling by evaporation in the reactor. The vaporized quench liquid is recondensed from the reactor outlet vapor stream. 1 fig.

  2. Analysis of Assembly Bill 423: Health Care Coverage: Mental Health Services

    E-Print Network [OSTI]

    California Health Benefits Review Program (CHBRP)

    2007-01-01T23:59:59.000Z

    drug abuse and mental health treatment on medical caresubstance abuse treatment, or behavioral health care (Table

  3. Reactor Neutrino Physics -- An Update

    E-Print Network [OSTI]

    Felix Boehm

    1999-06-18T23:59:59.000Z

    We review the status and the results of reactor neutrino experiments. Long baseline oscillation experiments at Palo Verde and Chooz have provided limits for the oscillation parameters while the recently proposed Kamland experiment at a baseline of more than 100km is now in the planning stage. We also describe the status of neutrino magnetic moment experiments at reactors.

  4. Radiation Shielding for Fusion Reactors

    SciTech Connect (OSTI)

    Santoro, R.T.

    1999-10-01T23:59:59.000Z

    Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel.

  5. Granular Dynamics in Pebble Bed Reactor Cores

    E-Print Network [OSTI]

    Laufer, Michael Robert

    2013-01-01T23:59:59.000Z

    gas reactors with the effective heat transfer of a molten salt coolant and the passive natural circulation safety systems of sodium fast reactors.

  6. MOOSE simulating nuclear reactor CRUD buildup

    ScienceCinema (OSTI)

    None

    2014-07-21T23:59:59.000Z

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  7. MOOSE simulating nuclear reactor CRUD buildup

    SciTech Connect (OSTI)

    None

    2014-02-06T23:59:59.000Z

    This simulation uses multiple physical models to show how the buildup of boron deposits on reactor fuel can affect performance and the reactor's power profile.

  8. The Endurance Bioenergy Reactor | Argonne National Laboratory

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

    The Endurance Bioenergy Reactor Share Description Argonne biophysicist Dr. Philip Laible and Air Force Major Matt Michaud talks about he endurance bioenergy reactor-a device that...

  9. aboriginal mental health: Topics by E-print Network

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

    and ... Chen, Wan-Lee 2012-06-27 475 The Costs of Covering Mental Health and Substance Abuse Care at the Same Level as Medical Care in Private Insurance Plans CiteSeer Summary:...

  10. Older adult consumers of Texas public mental health services:

    E-Print Network [OSTI]

    Karlin, Bradley Eric

    2002-01-01T23:59:59.000Z

    at disproportionately low rates. Although recent changes in public policies and perceptions portend improved access to mental health services, we cannot assume that older adults are finding their way into the therapy room and receiving treatment. The present study...

  11. Fast reactors and nuclear nonproliferation

    SciTech Connect (OSTI)

    Avrorin, E.N. [Russian Federal Nuclear Center - Zababakhin Institute of Applied Physics, Snezhinsk (Russian Federation); Rachkov, V.I.; Chebeskov, A.N. [State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering, Bondarenko Square, 1, Obninsk, Kaluga region, 249033 (Russian Federation)

    2013-07-01T23:59:59.000Z

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (authors)

  12. Automatic reactor power control for a pressurized water reactor

    SciTech Connect (OSTI)

    Jungin Choi (Kyungwon Univ. (Korea, Republic of)); Yungjoon Hah (Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)); Unchul Lee (Seoul National Univ. (Korea, Republic of))

    1993-05-01T23:59:59.000Z

    An automatic reactor power control system is presented for a pressurized water reactor (PWR). The associated reactor control strategy is called mode K.' The new system implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial shape change, which allows automatic control of the axial power distribution. Thus, the mode K enables precise regulation of both the reactivity and the power distribution, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load-follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1,000-MW (electric) PWR. The simulation results illustrate that the mode K would be a practical reactor power control strategy for the increased automation of nuclear plants.

  13. Investigation of the Feasibility of a Small Scale Transmutation Device

    E-Print Network [OSTI]

    Sit, Roger Carson

    2009-01-01T23:59:59.000Z

    27], and the Gas-Cooled Fast Transmutation Reactor (GCFTR)fast breeder reactors (LMFBRs) and high temperature gas

  14. Patterns and predictors of mental health service use and serious mental illness among community-dwelling elderly

    E-Print Network [OSTI]

    Karlin, Bradley Eric

    2006-10-30T23:59:59.000Z

    Older adults have historically utilized mental health services at substantially low rates. Unfortunately, though professional, policy, and other recent developments portend an increase in service use, there has been scant empirical attention devoted...

  15. Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

    SciTech Connect (OSTI)

    Glasstone, S.; Sesonske, A.

    1994-12-31T23:59:59.000Z

    This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials.

  16. University Reactor Conversion Lessons Learned Workshop for Purdue University Reactor

    SciTech Connect (OSTI)

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01T23:59:59.000Z

    The Department of Energy’s Idaho National Laboratory, under its programmatic responsibility for managing the University Research Reactor Conversions, has completed the conversion of the reactor at Purdue University Reactor. With this work completed and in anticipation of other impending conversion projects, the INL convened and engaged the project participants in a structured discussion to capture the lessons learned. The lessons learned process has allowed us to capture gaps, opportunities, and good practices, drawing from the project team’s experiences. These lessons will be used to raise the standard of excellence, effectiveness, and efficiency in all future conversion projects.

  17. Nuclear reactor control

    DOE Patents [OSTI]

    Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

    1982-01-01T23:59:59.000Z

    1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

  18. Nuclear reactor control apparatus

    DOE Patents [OSTI]

    Sridhar, Bettadapur N. (Cupertino, CA)

    1983-10-25T23:59:59.000Z

    Nuclear reactor safety rod release apparatus comprises a ring which carries detents normally positioned in an annular recess in outer side of the rod, the ring being held against the lower end of a drive shaft by magnetic force exerted by a solenoid carried by the drive shaft. When the solenoid is de-energized, the detent-carrying ring drops until the detents contact a cam surface associated with the lower end of the drive shaft, at which point the detents are cammed out of the recess in the safety rod to release the rod from the drive shaft. In preferred embodiments of the invention, an additional latch is provided to release a lower portion of a safety rod under conditions that may interfere with movement of the entire rod.

  19. Nuclear reactor control

    SciTech Connect (OSTI)

    Cawley, W.E.; Warnick, R.F.

    1982-03-30T23:59:59.000Z

    In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column.

  20. Thermodynamic Investigations of Aqueous Ternary Complexes for Am/Cm Separation

    E-Print Network [OSTI]

    Leggett, Christina Joy

    2012-01-01T23:59:59.000Z

    reactors, as opposed to fast reactors, which use fasturanium nuclei. Many fast reactors are breeder reactorsin specially-designed fast reactors. Clearly, the recovery

  1. Feasibility of Ground Testing a Moon and Mars Surface Power Reactor in EBR-II

    SciTech Connect (OSTI)

    Sheryl Morton; Carl Baily; Tom Hill; Jim Werner

    2006-02-01T23:59:59.000Z

    Ground testing of a surface fission power system would be necessary to verify the design and validate reactor performance to support safe and sustained human exploration of the Moon and Mars. The Idaho National Laboratory (INL) has several facilities that could be adapted to support a ground test. This paper focuses on the feasibility of ground testing at the Experimental Breeder Reactor II (EBR-II) facility and using other INL existing infrastructure to support such a test. This brief study concludes that the INL EBR-II facility and supporting infrastructure are a viable option for ground testing the surface power system. It provides features and attributes that offer advantages to locating and performing ground testing at this site, and it could support the National Aeronautics and Space Administration schedules for human exploration of the Moon. This study used the initial concept examined by the U.S. Department of Energy Inter-laboratory Design and Analysis Support Team for surface power, a lowtemperature, liquid-metal, three-loop Brayton power system. With some facility modification, the EBR-II can safely house a test chamber and perform long-term testing of the space reactor power system. The INL infrastructure is available to receive and provide bonded storage for special nuclear materials. Facilities adjacent to EBR-II can provide the clean room environment needed to assemble and store the test article assembly, disassemble the power system at the conclusion of testing, and perform posttest examination. Capability for waste disposal is also available at the INL.

  2. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect (OSTI)

    Douglas Morrell

    2011-03-01T23:59:59.000Z

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  3. Reactor protection system design alternatives for sodium fast reactors

    E-Print Network [OSTI]

    DeWitte, Jacob D. (Jacob Dominic)

    2011-01-01T23:59:59.000Z

    Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a ...

  4. Reactor physics design of supercritical CO?-cooled fast reactors

    E-Print Network [OSTI]

    Pope, Michael A. (Michael Alexander)

    2004-01-01T23:59:59.000Z

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

  5. Nuclear reactor downcomer flow deflector

    DOE Patents [OSTI]

    Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

    2011-02-15T23:59:59.000Z

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  6. Web: www.MentalHealthFirstAid.org Email: BetsyS@TheNationalCouncil.org Phone: 202.684.7457 Mental Health First Aid: Military Members, Veterans and their Families

    E-Print Network [OSTI]

    Amin, S. Massoud

    connected to care. WHAT IS MENTAL HEALTH FIRST AID Mental Health First Aid USA is a live training program service employees, school and college staff, clergy and caring citizens. WHY MENTAL HEALTH FIRST AIDWeb: www.MentalHealthFirstAid.org Email: BetsyS@TheNationalCouncil.org Phone: 202.684.7457 Mental

  7. 2012 Annual Report Research Reactor Infrastructure Program

    SciTech Connect (OSTI)

    Douglas Morrell

    2012-11-01T23:59:59.000Z

    The content of this report is the 2012 Annual Report for the Research Reactor Infrastructure Program.

  8. How far is a Fusion Power Reactor from an Experimental Reactor?

    E-Print Network [OSTI]

    1 How far is a Fusion Power Reactor from an Experimental Reactor? R. Toschi(1) , P. Barabaschi(2 September 2000 Madrid, Spain #12;2 How far is a fusion power reactor from an experimental reactor? R. Toschi To support a request of very substantial resources to build and operate an experimental reactor such as ITER

  9. Distribution ICategory: General Reactor Technology

    E-Print Network [OSTI]

    Shlyakhter, Ilya

    --- Distribution ICategory: General Reactor Technology (UC-520) ANl-92/23 AR(;ONNE NATIONAL, progressively by Huygens, Maxwell and Roentgen, mankind has learned to observe it, measure it, control it

  10. Spectral shift reactor control method

    SciTech Connect (OSTI)

    Impink, A.J. Jr.

    1987-08-18T23:59:59.000Z

    The method is described of closely controlling the reactor water coolant temperature of an operating spectral-shift nuclear reactor, the reactor comprising a core formed of fuel assemblies through which the reactor water coolant flows; different types of elongated elements operable to be controllably moved into and out of the core; one type of the elongated elements comprising control rods formed of neutron absorbing material and operable to decrease reactivity through neutron absorption when inserted into the core; another of the types of elongated elements comprising displacer rods formed of material which has a low absorption for neutrons and which have overall neutron-absorbing and moderating characteristics essentially not exceeding those of hollow tubular Zircaloy members with a filling zirconium oxide or aluminum oxide, the displacer rods operating to displace an equivalent volume of water coolant fluid from the core when inserted therein to decrease reactivity and to increase reactivity when moved from the core.

  11. University Reactor Matching Grants Program

    SciTech Connect (OSTI)

    John Valentine; Farzad Rahnema; Said Abdel-Khalik

    2003-02-14T23:59:59.000Z

    During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given.

  12. Reactor physics project final report

    E-Print Network [OSTI]

    Driscoll, Michael J.

    1970-01-01T23:59:59.000Z

    This is the final report in an experimental and theoretical program to develop and apply single- and few-element methods for the determination of reactor lattice parameters. The period covered by the report is January 1, ...

  13. Combustion synthesis continuous flow reactor

    DOE Patents [OSTI]

    Maupin, G.D.; Chick, L.A.; Kurosky, R.P.

    1998-01-06T23:59:59.000Z

    The present invention is a reactor for combustion synthesis of inorganic powders. The reactor includes a reaction vessel having a length and a first end and a second end. The reaction vessel further has a solution inlet and a carrier gas inlet. The reactor further has a heater for heating both the solution and the carrier gas. In a preferred embodiment, the reaction vessel is heated and the solution is in contact with the heated reaction vessel. It is further preferred that the reaction vessel be cylindrical and that the carrier gas is introduced tangentially into the reaction vessel so that the solution flows helically along the interior wall of the reaction vessel. As the solution evaporates and combustion produces inorganic material powder, the carrier gas entrains the powder and carries it out of the reactor. 10 figs.

  14. Advanced Catalytic Hydrogenation Retrofit Reactor

    SciTech Connect (OSTI)

    Reinaldo M. Machado

    2002-08-15T23:59:59.000Z

    Industrial hydrogenation is often performed using a slurry catalyst in large stirred-tank reactors. These systems are inherently problematic in a number of areas, including industrial hygiene, process safety, environmental contamination, waste production, process operability and productivity. This program proposed the development of a practical replacement for the slurry catalysts using a novel fixed-bed monolith catalyst reactor, which could be retrofitted onto an existing stirred-tank reactor and would mitigate many of the minitations and problems associated with slurry catalysts. The full retrofit monolith system, consisting of a recirculation pump, gas/liquid ejector and monolith catalyst, is described as a monolith loop reactor or MLR. The MLR technology can reduce waste and increase raw material efficiency, which reduces the overall energy required to produce specialty and fine chemicals.

  15. Teaching About Nature's Nuclear Reactors

    E-Print Network [OSTI]

    Herndon, J M

    2005-01-01T23:59:59.000Z

    Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

  16. Nuclear Reactors and Technology; (USA)

    SciTech Connect (OSTI)

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01T23:59:59.000Z

    Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

  17. Reactor core isolation cooling system

    DOE Patents [OSTI]

    Cooke, Franklin E. (San Jose, CA)

    1992-01-01T23:59:59.000Z

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom.

  18. Reactor core isolation cooling system

    DOE Patents [OSTI]

    Cooke, F.E.

    1992-12-08T23:59:59.000Z

    A reactor core isolation cooling system includes a reactor pressure vessel containing a reactor core, a drywell vessel, a containment vessel, and an isolation pool containing an isolation condenser. A turbine is operatively joined to the pressure vessel outlet steamline and powers a pump operatively joined to the pressure vessel feedwater line. In operation, steam from the pressure vessel powers the turbine which in turn powers the pump to pump makeup water from a pool to the feedwater line into the pressure vessel for maintaining water level over the reactor core. Steam discharged from the turbine is channeled to the isolation condenser and is condensed therein. The resulting heat is discharged into the isolation pool and vented to the atmosphere outside the containment vessel for removing heat therefrom. 1 figure.

  19. Interfacial effects in fast reactors

    E-Print Network [OSTI]

    Saidi, Mohammad Said

    1979-01-01T23:59:59.000Z

    The problem of increased resonance capture rates near zone interfaces in fast reactor media has been examined both theoretically and experimentally. An interface traversing assembly was designed, constructed and employed ...

  20. Solid State Reactor Final Report

    SciTech Connect (OSTI)

    Mays, G.T.

    2004-03-10T23:59:59.000Z

    The Solid State Reactor (SSR) is an advanced reactor concept designed to take advantage of Oak Ridge National Laboratory's (ORNL's) recently developed graphite foam that has enhanced heat transfer characteristics and excellent high-temperature mechanical properties, to provide an inherently safe, self-regulated, source of heat for power and other potential applications. This work was funded by the U.S. Department of Energy's Nuclear Energy Research Initiative (NERI) program (Project No. 99-064) from August 1999 through September 30, 2002. The initial concept of utilizing the graphite foam as a basis for developing an advanced reactor concept envisioned that a suite of reactor configurations and power levels could be developed for several different applications. The initial focus was looking at the reactor as a heat source that was scalable, independent of any heat removal/power conversion process. These applications might include conventional power generation, isotope production and destruction (actinides), and hydrogen production. Having conducted the initial research on the graphite foam and having performed the scoping parametric analyses from neutronics and thermal-hydraulic perspectives, it was necessary to focus on a particular application that would (1) demonstrate the viability of the overall concept and (2) require a reasonably structured design analysis process that would synthesize those important parameters that influence the concept the most as part of a feasible, working reactor system. Thus, the application targeted for this concept was supplying power for remote/harsh environments and a design that was easily deployable, simplistic from an operational standpoint, and utilized the new graphite foam. Specifically, a 500-kW(t) reactor concept was pursued that is naturally load following, inherently safe, optimized via neutronic studies to achieve near-zero reactivity change with burnup, and proliferation resistant. These four major areas of research were undertaken: (1) establishing the design and safety-related basis via neutronic and reactor control assessments with the graphite foam as heat transfer medium; (2) evaluating the thermal performance of the graphite foam for heat removal, reactor stability, reactor operations, and overall core thermal characteristics; (3) characterizing the physical properties of the graphite foam under normal and irradiated conditions to determine any effects on structure, dimensional stability, thermal conductivity, and thermal expansion; and (4) developing a power conversion system design to match the reactor operating parameters.

  1. What Does It Take? California County Funding Requests for Recovery-Oriented Full Service Partnerships Under the Mental Health Services Act

    E-Print Network [OSTI]

    Felton, Mistique C.; Cashin, Cheryl E.; Brown, Timothy T.

    2010-01-01T23:59:59.000Z

    Health and Mental Health Services Research. 2009 Novemberof Regulations. Mental Health Services Act. (2010). Cashin,2007). Fact Sheet: Mental Health Services Act (Prop 63).

  2. Alternate-fuel reactor studies

    SciTech Connect (OSTI)

    Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

    1983-02-01T23:59:59.000Z

    A number of studies related to improvements and/or greater understanding of alternate-fueled reactors is presented. These studies cover the areas of non-Maxwellian distributions, materials and lifetime analysis, a /sup 3/He-breeding blanket, tritium-rich startup effects, high field magnet support, and reactor operation spanning the range from full D-T operation to operation with no tritium breeding.

  3. Automatic safety rod for reactors

    DOE Patents [OSTI]

    Germer, John H. (San Jose, CA)

    1988-01-01T23:59:59.000Z

    An automatic safety rod for a nuclear reactor containing neutron absorbing material and designed to be inserted into a reactor core after a loss-of-core flow. Actuation is based upon either a sudden decrease in core pressure drop or the pressure drop decreases below a predetermined minimum value. The automatic control rod includes a pressure regulating device whereby a controlled decrease in operating pressure due to reduced coolant flow does not cause the rod to drop into the core.

  4. Fast quench reactor and method

    DOE Patents [OSTI]

    Detering, Brent A. (Idaho Falls, ID); Donaldson, Alan D. (Idaho Falls, ID); Fincke, James R. (Idaho Falls, ID); Kong, Peter C. (Idaho Falls, ID)

    1998-01-01T23:59:59.000Z

    A fast quench reaction includes a reactor chamber having a high temperature heating means such as a plasma torch at its inlet and a restrictive convergent-divergent nozzle at its outlet end. Reactants are injected into the reactor chamber. The resulting heated gaseous stream is then rapidly cooled by passage through the nozzle. This "freezes" the desired end product(s) in the heated equilibrium reaction stage.

  5. Propellant actuated nuclear reactor steam depressurization valve

    DOE Patents [OSTI]

    Ehrke, Alan C. (San Jose, CA); Knepp, John B. (San Jose, CA); Skoda, George I. (Santa Clara, CA)

    1992-01-01T23:59:59.000Z

    A nuclear fission reactor combined with a propellant actuated depressurization and/or water injection valve is disclosed. The depressurization valve releases pressure from a water cooled, steam producing nuclear reactor when required to insure the safety of the reactor. Depressurization of the reactor pressure vessel enables gravity feeding of supplementary coolant water through the water injection valve to the reactor pressure vessel to prevent damage to the fuel core.

  6. When Do Commercial Reactors Permanently Shut Down?

    Reports and Publications (EIA)

    2011-01-01T23:59:59.000Z

    For those wishing to obtain current data, the following resources are available: U.S. reactors, go to the Energy Information Administration's nuclear reactor shutdown list. (Note: As of April 30, 2010, the last U.S. reactor to permanently shut down was Big Rock Point in 1997.) Foreign Reactors, go to the Power Reactor Information System (PRIS) on the International Atomic Energy Agency's website.

  7. Nuclear reactor control rod

    SciTech Connect (OSTI)

    Cearley, J.E.; Izzo, K.R.

    1987-06-30T23:59:59.000Z

    This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured.

  8. Reactor pressure vessel nozzle

    DOE Patents [OSTI]

    Challberg, R.C.; Upton, H.A.

    1994-10-04T23:59:59.000Z

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough. 2 figs.

  9. Reactor shroud joint

    DOE Patents [OSTI]

    Ballas, G.J.; Fife, A.B.; Ganz, I.

    1998-04-07T23:59:59.000Z

    A shroud for a nuclear reactor is described. In one embodiment, the shroud includes first and second shroud sections, and each shroud section includes a substantially cylindrical main body having a first end and a second end. With respect to each shroud section, a flange is located at the main body first end, and the flange has a plurality of bolt openings therein and a plurality of scalloped regions. The first shroud section is welded to the second shroud section, and at least some of the bolt openings in the first shroud section flange align with respective bolt openings in the second shroud section flange. In the event that the onset of inter-granular stress corrosion cracking is ever detected in the weld between the shroud section, bolts are inserted through bolt openings in the first shroud section flange and through aligned bolt openings the second shroud section flange. Each bolt, in one embodiment, has a shank section and first and second threaded end sections. Nuts are threadedly engaged to the threaded end sections and tightened against the respective flanges. 4 figs.

  10. Solar solids reactor

    DOE Patents [OSTI]

    Yudow, B.D.

    1986-02-24T23:59:59.000Z

    A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

  11. Solar solids reactor

    DOE Patents [OSTI]

    Yudow, Bernard D. (Chicago, IL)

    1987-01-01T23:59:59.000Z

    A solar powered kiln is provided, that is of relatively simple design and which efficiently uses solar energy. The kiln or solids reactor includes a stationary chamber with a rearward end which receives solid material to be reacted and a forward end through which reacted material is disposed of, and a screw conveyor extending along the bottom of the chamber for slowly advancing the material between the chamber ends. Concentrated solar energy is directed to an aperture at the forward end of the chamber to heat the solid material moving along the bottom of the chamber. The solar energy can be reflected from a mirror facing at an upward incline, through the aperture and against a heat-absorbing material near the top of the chamber, which moves towards the rear of the chamber to distribute heat throughout the chamber. Pumps at the forward and rearward ends of the chamber pump heated sweep gas through the length of the chamber, while minimizing the flow of gas through an open aperture through which concentrated sunlight is received.

  12. Novel Catalytic Membrane Reactors

    SciTech Connect (OSTI)

    Stuart Nemser, PhD

    2010-10-01T23:59:59.000Z

    There are many industrial catalytic organic reversible reactions with amines or alcohols that have water as one of the products. Many of these reactions are homogeneously catalyzed. In all cases removal of water facilitates the reaction and produces more of the desired chemical product. By shifting the reaction to right we produce more chemical product with little or no additional capital investment. Many of these reactions can also relate to bioprocesses. Given the large number of water-organic compound separations achievable and the ability of the Compact Membrane Systems, Inc. (CMS) perfluoro membranes to withstand these harsh operating conditions, this is an ideal demonstration system for the water-of-reaction removal using a membrane reactor. Enhanced reaction synthesis is consistent with the DOE objective to lower the energy intensity of U.S. industry 25% by 2017 in accord with the Energy Policy Act of 2005 and to improve the United States manufacturing competitiveness. The objective of this program is to develop the platform technology for enhancing homogeneous catalytic chemical syntheses.

  13. Analysis of mechanical effects caused by plasma disruptions in the European BOT solid breeder blanket design with MANET as structural material

    SciTech Connect (OSTI)

    Boccaccini, L.V.; Ruatto, P. [Institut fuer Neutronenphysik und Reaktortechnik, Karlsruhe (Germany)

    1994-12-31T23:59:59.000Z

    The Karlsruhe Nuclear Center is developing, through design and experimental work, a BOT (Breeder Out of Tube) Helium Cooled Solid Breeder Blanket for a DEMO application. One of the crucial problems in the blanket design is to demonstrate the capability of the structure to withstand the mechanical effects of a major plasma disruption as extrapolated to DEMO from the experience of present machines. In this paper the results of the assessment work are presented; the acceptability of the design is discussed on the basis of a stress analysis of the structure under combined thermal and electromagnetic loads. The martensitic steel MANET has been chosen as structural material, because it is able to withstand the high neutron fluence in Demo (70 dpa) without appreciably swelling and has good thermal-mechanical properties - lower thermal expansion and higher strength - in comparison to AISI 316L steel. As far as it concerns the mechanical effects of plasma disruptions, MANET presents two important features which have been carefully investigated in the assessment: the magnetic properties of the material and the degradation of the fracture toughness behavior under irradiation.

  14. Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation

    E-Print Network [OSTI]

    Heidet, Florent

    2010-01-01T23:59:59.000Z

    selected as part of the Generation IV reactors .. - 4 -The development of Generation IV fast reactors can make aconcepts selected for the Generation IV reactors, three,

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

    E-Print Network [OSTI]

    Djurcic, Zelimir

    2009-01-01T23:59:59.000Z

    neutrino Production at Nuclear Reactors Z. Djurcic 1 , ?emission rates from nuclear reactors are determined fromlarge commercial nuclear reactors are playing an important

  16. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01T23:59:59.000Z

    Fundamental aspects of nuclear reactor fuel elements.Unlike permanent nuclear reactor core components, nuclearof the first nuclear reactors, commercial nuclear fuel still

  17. PIA - Advanced Test Reactor National Scientific User Facility...

    Energy Savers [EERE]

    Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor National Scientific User Facility Users Week 2009 PIA - Advanced Test Reactor...

  18. acid immunoaffinity reactor: Topics by E-print Network

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

    sludge blanket reactors (more) Ning, Zuojun. 2009-01-01 8 Inexpensive Mini Thermonuclear Reactor CiteSeer Summary: This proposed design for a mini thermonuclear reactor...

  19. MULTIPHASE REACTOR MODELING FOR ZINC CHLORIDE CATALYZED COAL LIQUEFACTION

    E-Print Network [OSTI]

    Joyce, Peter James

    2011-01-01T23:59:59.000Z

    II. c. High Yield Batch Reactor Results. . . • .Objectives •. Slurry Reactors . . . . . . .A. StudiesSystems. D. Slurry Reactor Theory. General. . Application to

  20. Best Practices of Print Journalists Who Have Won Awards for Mental-Health Reporting: A Qualitative Interview Study

    E-Print Network [OSTI]

    Subramanian, Roma

    2012-02-14T23:59:59.000Z

    that it would be valuable to investigate in more detail how journalists' personal attitudes toward mental illness influence their reporting. Also, guidelines for mental-health reporting should be created with the collaboration of journalists and mental-health...

  1. Prospects for improved fusion reactors

    SciTech Connect (OSTI)

    Krakowski, R.A.; Miller, R.L.; Hagenson, R.L.

    1986-01-01T23:59:59.000Z

    Ideally, a new energy source must be capable of displacing old energy sources while providing both economic opportunities and enhanced environmental benefits. The attraction of an essentially unlimited fuel supply has generated a strong impetus to develop advanced fission breeders and, even more strongly, the exploitation of nuclear fusion. Both fission and fusion systems trade a reduced fuel charge for a more capital-intensive plant needed to utilize a cheaper and more abundant fuel. Results from early conceptual designs of fusion power plants, however, indicated a capital intensiveness that could override cost savings promised by an inexpensive fuel cycle. Early warnings of these problems appeared, and generalized routes to more economically attractive systems have been suggested; specific examples have also recently been given. Although a direct reduction in the cost (and mass) of the fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils, and primary structure) most directly reduces the overall cost of fusion power, with the mass power density (MPD, ratio of net electric power to FPC mass, kWe/tonne) being suggested as a figure-of-merit in this respect, other technical, safety/environmental, and institutional issues also enter into the definition of and direction for improved fusion concepts. These latter issues and related tradeoffs are discussed.

  2. International Research Reactor Decommissioning Project

    SciTech Connect (OSTI)

    Leopando, Leonardo [Philippine Nuclear Research Institute, Quezon City (Philippines); Warnecke, Ernst [International Atomic Energy Agency, Vienna (Austria)

    2008-01-15T23:59:59.000Z

    Many research reactors have been or will be shut down and are candidates for decommissioning. Most of the respective countries neither have a decommissioning policy nor the required expertise and funds to effectively implement a decommissioning project. The IAEA established the Research Reactor Decommissioning Demonstration Project (R{sup 2}D{sup 2}P) to help answer this need. It was agreed to involve the Philippine Research Reactor (PRR-1) as model reactor to demonstrate 'hands-on' experience as it is just starting the decommissioning process. Other facilities may be included in the project as they fit into the scope of R{sup 2}D{sup 2}P and complement to the PRR-1 decommissioning activities. The key outcome of the R{sup 2}D{sup 2}P will be the decommissioning of the PRR-1 reactor. On the way to this final goal the preparation of safety related documents (i.e., decommissioning plan, environmental impact assessment, safety analysis report, health and safety plan, cost estimate, etc.) and the licensing process as well as the actual dismantling activities could provide a model to other countries involved in the project. It is expected that the R{sup 2}D{sup 2}P would initiate activities related to planning and funding of decommissioning activities in the participating countries if that has not yet been done.

  3. Rapid starting methanol reactor system

    DOE Patents [OSTI]

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

    1984-01-01T23:59:59.000Z

    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.

  4. Imaging Fukushima Daiichi reactors with muons

    SciTech Connect (OSTI)

    Miyadera, Haruo; Borozdin, Konstantin N.; Greene, Steve J.; Milner, Edward C.; Morris, Christopher L. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Lukic, Zarija [Computational Cosmology Center, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Masuda, Koji [University of New Mexico, Albuquerque, NM 87131 (United States); Perry, John O. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); University of New Mexico, Albuquerque, NM 87131 (United States)

    2013-05-15T23:59:59.000Z

    A study of imaging the Fukushima Daiichi reactors with cosmic-ray muons to assess the damage to the reactors is presented. Muon scattering imaging has high sensitivity for detecting uranium fuel and debris even through thick concrete walls and a reactor pressure vessel. Technical demonstrations using a reactor mockup, detector radiation test at Fukushima Daiichi, and simulation studies have been carried out. These studies establish feasibility for the reactor imaging. A few months of measurement will reveal the spatial distribution of the reactor fuel. The muon scattering technique would be the best and probably the only way for Fukushima Daiichi to make this determination in the near future.

  5. Attitudes toward child mental health services: adaptation and development of an attitude scale

    E-Print Network [OSTI]

    Turner, Erlanger A

    2006-10-30T23:59:59.000Z

    on problems in adulthood, it is important to examine the influence of attitudes on seeking mental health service for children. Currently, no known measure exists to measure attitudes toward mental health services for children. Building on previous research...

  6. autosomal-recessive pitt-hopkins-like mental: Topics by E-print...

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

    the role of culture in the definition and maintenance of mental health and "mental illness." We will be exploring what our culture and various cultures of the world have to...

  7. abacus-based mental calculations: Topics by E-print Network

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

    the role of culture in the definition and maintenance of mental health and "mental illness." We will be exploring what our culture and various cultures of the world have to...

  8. Parental and youth attributions, acculturation, and treatment engagement of Latino families in youth mental health services : a preliminary examination

    E-Print Network [OSTI]

    Ho, Judy Keeching

    2007-01-01T23:59:59.000Z

    1995). Children's mental health service use across serviceJournal of Behavioral Health Services and Research, 30, 393-satisfaction with mental health services: Development of a

  9. Depression and the Ironic Effects of Thought Suppression: Therapeutic Strategies for ImprovingMental Control

    E-Print Network [OSTI]

    Beevers, Christopher

    are espc- cially likely to engage in thought suppression in an at- tempt to achieve mental control over

  10. Integral fast reactor safety features

    SciTech Connect (OSTI)

    Cahalan, J.E.; Kramer, J.M.; Marchaterre, J.F.; Mueller, C.J.; Pedersen, D.R.; Sevy, R.H.; Wade, D.C.; Wei, T.Y.C.

    1988-01-01T23:59:59.000Z

    The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The two major goals of the IFR development effort are improved economics and enhanced safety. In addition to liquid metal cooling, the principal design features that distinguish the IFR are: (1) a pool-type primary system, (2) an advanced ternary alloy metallic fuel, and (3) an integral fuel cycle with on-site fuel reprocessing and fabrication. This paper focuses on the technical aspects of the improved safety margins available in the IFR concept. This increased level of safety is made possible by (1) the liquid metal (sodium) coolant and pool-type primary system layout, which together facilitate passive decay heat removal, and (2) a sodium-bonded metallic fuel pin design with thermal and neutronic properties that provide passive core responses which control and mitigate the consequences of reactor accidents.

  11. Director's Report to the National Advisory Mental Health Council

    E-Print Network [OSTI]

    Baker, Chris I.

    , which I share with you in this report. NIH-Wide Update National Institutes of Health (NIH) Roadmap The NIH Roadmap is an integrated vision to deepen the understanding of biology, stimulate. The Roadmap is organized into three themes, and much activity involving National Institute of Mental Health

  12. MR-byMIT-MH-1011 Mental Health Service

    E-Print Network [OSTI]

    Polz, Martin

    treatment for mental illness, alcohol or drug abuse/treatment, domestic/sexual assault, or AIDS testing or 3 for release of information on care provided after the date of the patient's signature, unless you (the patient not fax records. c. There is no fee for records released directly to other health care providers. However

  13. Motor processes 1 Motor Processes in Mental Rotation1

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Motor processes 1 Motor Processes in Mental Rotation1 1 M.W. wishes to thank the LPPA for its are at least in part guided by motor processes, even in the case of images of abstract objects rather than of a specific motor action. We directly test the hypothesis by means of a dual-task paradigm in which subjects

  14. Mental Health & College Counselor Internship Handbook January 2014 1 INTERNSHIP HANDBOOK

    E-Print Network [OSTI]

    Mental Health & College Counselor Internship Handbook January 2014 1 INTERNSHIP HANDBOOK COUN 666 or 667: Internship in Counseling (College Counseling or Mental Health Counseling) MSEd Counseling, Old Dominion University (May 2014) #12;Mental Health & College Counselor Internship Handbook January 2014 2

  15. Human Mental Models of Humanoid Robots* Sau-lai Lee Sara Kiesler

    E-Print Network [OSTI]

    Kiesler, Sara

    Human Mental Models of Humanoid Robots* Sau-lai Lee Sara Kiesler Human Computer Interaction ground of understanding between the two. In two experiments modelled after human-human studies we robot emits a human's voice), their mental model of the system's behavior may approach their mental

  16. DOE's way-out reactors

    SciTech Connect (OSTI)

    Marshall, E.

    1986-03-21T23:59:59.000Z

    The SP-100 reactor, envisioned long before Star Wars, was to power civilian structures such as the space station and orbiting commercial labs. According to the SDI Organization, it will be the cornerstone for SDI, used as a no-maintenance, general source of energy for the military's infrastructure - weapons scale power will come later. DOE wants to spend $72 in FY 1977 to design and build these reactors. Funding problems with Congress, as well as some of the technology and timetables are discussed here.

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

    E-Print Network [OSTI]

    Djurcic, Zelimir

    2009-01-01T23:59:59.000Z

    reactors are determined from thermal power measure- ments and ?ssion rate calculations.of a reactor’s ther- mal power is given by a calculation ofCALCULATIONS During the power cycle of a nuclear reactor,

  18. Maximum Fuel Utilization in Advanced Fast Reactors without Actinides Separation

    E-Print Network [OSTI]

    Heidet, Florent

    2010-01-01T23:59:59.000Z

    Gas Expansion Module Gas-cooled Fast Reactor High Enrichedfast reactors: gas-cooled fast reactor (GFR), sodium-cooledderived from the Gas cooled Fast Reactor (GFR). This core

  19. Computational Analysis of Fluid Flow in Pebble Bed Modular Reactor

    E-Print Network [OSTI]

    Gandhir, Akshay

    2012-10-19T23:59:59.000Z

    High Temperature Gas-cooled Reactor (HTGR) is a Generation IV reactor under consideration by Department of Energy and in the nuclear industry. There are two categories of HTGRs, namely, Pebble Bed Modular Reactor (PBMR) and Prismatic reactor. Pebble...

  20. Reactor Operations informal monthly report September 1994

    SciTech Connect (OSTI)

    Junker, L.

    1994-09-01T23:59:59.000Z

    This paper presents operations at the MRR and HFBR reactors at Brookhaven National Laboratory for September 1994. Reactor run-times, instrumentation, mechanical maintenance, occurrence reports and safety information are listed. Irradiation summaries are included.

  1. Reactivity control assembly for nuclear reactor

    DOE Patents [OSTI]

    Bollinger, Lawrence R. (Schenectady, NY)

    1984-01-01T23:59:59.000Z

    Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

  2. Microfluidic reactors for the synthesis of nanocrystals

    E-Print Network [OSTI]

    Yen, Brian K. H

    2007-01-01T23:59:59.000Z

    Several microfluidic reactors were designed and applied to the synthesis of colloidal semiconductor nanocrystals (NCs). Initially, a simple single-phase capillary reactor was used for the synthesis of CdSe NCs. Precursors ...

  3. Stability analysis of supercritical water cooled reactors

    E-Print Network [OSTI]

    Zhao, Jiyun, Ph. D. Massachusetts Institute of Technology

    2005-01-01T23:59:59.000Z

    The Supercritical Water-Cooled Reactor (SCWR) is a concept for an advanced reactor that will operate at high pressure (25MPa) and high temperature (500°C average core exit). The high coolant temperature as it leaves the ...

  4. Auxiliary reactor for a hydrocarbon reforming system

    DOE Patents [OSTI]

    Clawson, Lawrence G.; Dorson, Matthew H.; Mitchell, William L.; Nowicki, Brian J.; Bentley, Jeffrey M.; Davis, Robert; Rumsey, Jennifer W.

    2006-01-17T23:59:59.000Z

    An auxiliary reactor for use with a reformer reactor having at least one reaction zone, and including a burner for burning fuel and creating a heated auxiliary reactor gas stream, and heat exchanger for transferring heat from auxiliary reactor gas stream and heat transfer medium, preferably two-phase water, to reformer reaction zone. Auxiliary reactor may include first cylindrical wall defining a chamber for burning fuel and creating a heated auxiliary reactor gas stream, the chamber having an inlet end, an outlet end, a second cylindrical wall surrounding first wall and a second annular chamber there between. The reactor being configured so heated auxiliary reactor gas flows out the outlet end and into and through second annular chamber and conduit which is disposed in second annular chamber, the conduit adapted to carry heat transfer medium and being connectable to reformer reaction zone for additional heat exchange.

  5. Challenges in the Development of Advanced Reactors

    SciTech Connect (OSTI)

    P. Sabharwall; M.C. Teague; S.M. Bragg-Sitton; M.W. Patterson

    2012-08-01T23:59:59.000Z

    Past generations of nuclear reactors have been successively developed and the next generation is currently being developed, demonstrating the constant progress and technical and industrial vitality of nuclear energy. In 2000 US Department of Energy launched Generation IV International Forum (GIF) which is one of the main international frameworks for the development of future nuclear systems. The six systems that were selected were: sodium cooled fast reactor, lead cooled fast reactor, supercritical water cooled reactor, very high temperature gas cooled reactor (VHTR), gas cooled fast reactor and molten salt reactor. This paper discusses some of the proposed advanced reactor concepts that are currently being researched to varying degrees in the United States, and highlights some of the major challenges these concepts must overcome to establish their feasibility and to satisfy licensing requirements.

  6. MULTIPHASE REACTOR MODELING FOR ZINC CHLORIDE CATALYZED COAL LIQUEFACTION

    E-Print Network [OSTI]

    Joyce, Peter James

    2011-01-01T23:59:59.000Z

    for the Coal Slurry Reactor Calculations are shown here for= Total reactor pressure, psi. The calculation is iterative,

  7. Incorporation of Hydride Nuclear Fuels in Commercial Light Water Reactors

    E-Print Network [OSTI]

    Terrani, Kurt Amir

    2010-01-01T23:59:59.000Z

    target subassemblies in a fast reactor core for effectiveof minor actinides in fast reactors [79]. In order to

  8. Granular Dynamics in Pebble Bed Reactor Cores

    E-Print Network [OSTI]

    Laufer, Michael Robert

    2013-01-01T23:59:59.000Z

    a simulant fluid to match the dynamics of fuel pebbles andfuel pebbles through reactor cores with and without coupled fluid

  9. Digital computer operation of a nuclear reactor

    DOE Patents [OSTI]

    Colley, Robert W. (Richland, WA)

    1984-01-01T23:59:59.000Z

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  10. Digital computer operation of a nuclear reactor

    DOE Patents [OSTI]

    Colley, R.W.

    1982-06-29T23:59:59.000Z

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  11. Liquid metal cooled nuclear reactor plant system

    DOE Patents [OSTI]

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

    1993-01-01T23:59:59.000Z

    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.

  12. Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories

    Office of Legacy Management (LM)

    Radiological Condition of the Westinghouse Advanced Reactors Division Plutonium Fuel Laboratories Cheswick, Pennsylvania -. -, -- AGENCY: Office of Operational Safety, Department...

  13. Adolescents leaving mental health or social care services: predictors of mental health and psychosocial outcomes one year later

    E-Print Network [OSTI]

    Memarzia, Jessica; St Clair, Michelle C.; Owens, Matt; Goodyer, Ian M.; Dunn, Valerie J.

    2015-05-02T23:59:59.000Z

    scales and psychiatric diagnoses: Responses to GHQ-12, K-10 and CIDI across the lifespan. J Affect Disorders. 2010;121:263–7. 29. Tennant R, Hiller L, Fishwick R, Platt S, Joseph S, Weich S, et al. The Warwick- Edinburgh mental well-being scale (WEMWBS...

  14. Interdisciplinary Institute for Innovation Nuclear reactors' construction

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    Interdisciplinary Institute for Innovation Nuclear reactors' construction costs: The role of lead@mines-paristech.fr hal-00956292,version1-6Mar2014 #12;hal-00956292,version1-6Mar2014 #12;Nuclear reactors' construction reactor construction costs in France and the United States. Studying the cost of nuclear power has often

  15. Laminar Entrained Flow Reactor (Fact Sheet)

    SciTech Connect (OSTI)

    Not Available

    2014-02-01T23:59:59.000Z

    The Laminar Entrained Flow Reactor (LEFR) is a modular, lab scale, single-user reactor for the study of catalytic fast pyrolysis (CFP). This system can be employed to study a variety of reactor conditions for both in situ and ex situ CFP.

  16. BROOKHAVEN NATIONAL LABORATORY'S HIGH FLUX BEAM REACTOR

    E-Print Network [OSTI]

    Ohta, Shigemi

    1 BROOKHAVEN NATIONAL LABORATORY'S HIGH FLUX BEAM REACTOR Compiled by S. M. Shapiro I. PICTORIAL with fiberglass insulation and a protective aluminum skin. The reactor vessel is shaped somewhat like a very large at the spherical end. It is located at the center of the reactor building and is surrounded by a lead and steel

  17. The Reactor An ObjectOriented Framework

    E-Print Network [OSTI]

    Schmidt, Douglas C.

    The Reactor An Object­Oriented Framework for Event Demultiplexing and Event Handler Dispatching Douglas C. Schmidt 1 Overview ffl The Reactor is an object­oriented frame­ work that encapsulates OS event demul­ tiplexing mechanisms -- e.g., the Reactor API runs transparently atop both Wait

  18. REACTOR SAFETY KEYWORDS: best estimate plus

    E-Print Network [OSTI]

    Hoppe, Fred M.

    REACTOR SAFETY KEYWORDS: best estimate plus uncertainty analysis, epistemic error and aleatory phe- nomena that underlie the safety analyses. The use of BE codes within the reactor technology in advance and that result from a variety of operating conditions or states. These arise because the reactor

  19. Chemical Reactor Analysis and Optimal Digestion

    E-Print Network [OSTI]

    Jumars, Pete

    J 310 Chemical Reactor Analysis and Optimal Digestion An optimal digestion theory can be readily derived from basic principles o f chemical reactor analysis and design Deborah L. Penry and Peter a reactor and an operating strategy that maximize the yield or yield rate of desired reaction products

  20. BioElectrochemically Assisted Microbial Reactor

    E-Print Network [OSTI]

    Lee, Dongwon

    BioElectrochemically Assisted Microbial Reactor (BEAMR) The BEAMR reactor uses only >0.2 V needed of combustion) >85% based on energy in electricity + acetate, at a rate of 1 m3 -H2 /m3 -reactor volume/d (100 A