Sample records for godiva reactor

  1. Source-Term and building-Wake Consequence Modeling for the Godiva IV Reactor at Los Alamos National Laboratory

    SciTech Connect

    Letellier, B.C.; McClure, P.; Restrepo, L.

    1999-06-13

    The objectives of this work were to evaluate the consequences of a postulated accident to onsite security personnel stationed near the facility during operations of the Godiva IV critical assembly and to identify controls needed to protect these personnel in case of an extreme criticality excursion equivalent to the design-basis accident (DBA). This paper presents the methodology and results of the source-term calculations, building ventilation rates, air concentrations, and consequence calculations that were performed using a multidisciplinary approach with several phenomenology models. Identification of controls needed to mitigate the consequences to near-field receptors is discussed.

  2. Godiva and Juliet Diagnostics CED-1 (IER-176)

    SciTech Connect

    Scorby, J C

    2011-12-21

    A suite of diagnostics are being proposed for use in the Juliet experiment (IER-128). In order to calibrate and test the diagnostics prior to use, the LLNL calibration facility and Godiva pulsed reactor will be used to provide intense sources of neutrons and gammas. Due to the similarities of the Godiva and Juliet radiation fields, the diagnostics being developed and tested for Juliet can also play an on-going role in diagnostics for Godiva as well as, perhaps, other critical assembly experiments. Similar work is also being conducted for IER-147 for the purpose of characterizing the Godiva radiation field in support of an upcoming international nuclear accident dosimetry exercise. Diagnostics developed and fielded under IER-147 can provide valuable data with respect to the neutron and gamma energy spectrums in the vicinity of Godiva which is relevant to the calibration of Juliet diagnostics.

  3. Determination of Godiva`s effective delayed neutron fraction using newly calculated delayed neutron spectra

    SciTech Connect

    Spriggs, G.D.; Campbell, J.M. [Los Alamos National Lab., NM (United States); Busch, R.D. [Univ. of New Mexico, Albuquerque, NM (United States)

    1999-09-01

    When calculating the effective delayed neutron fraction {beta}{sub eff} for a given reactor system, the assumed delayed neutron group spectra and the assumed number of delayed neutrons born per fission {nu}{sub d} can have a major impact on the final value. Over the years, the recommended values for the delayed neutron spectra and for {nu}{sub d} have slowly changed. To ascertain whether or not these changes have increased the accuracy of {beta}{sub eff} calculations in fast {sup 235}U systems, the authors have reevaluated {beta}{sub eff} for the benchmark system Godiva-I using newly calculated delayed neutron spectra and Tuttle`s recommended values of {nu}{sub d} for both {sup 235}U and {sup 238}U.

  4. Beta- and gamma-dose measurements of the Godiva IV critical assembly.

    PubMed

    Hankins, D E

    1984-03-01

    To aid in the re-evaluation of an exposure that occurred in 1963, information was required on the response of film badges to the beta- and gamma-ray doses from a critical assembly. Of particular interest was the beta spectra from the assembly. The techniques used and the results obtained in this study are of interest to health physicists at facilities where exposures to betas occur. The dose rates from the Los Alamos National Laboratory Godiva IV Critical Assembly were measured at numerous distances from the assembly four and 12 days following a burst. Information was obtained on the beta-particle spectra using absorption curve studies. The beta/gamma dose-rate ratio as a function of distance from the assembly was determined. Shielding provided by various metals, gloves and clothing was measured. The beta- and gamma-ray doses measured were compared with a film packet used in the past at the Nevada Test Site with two types of current TLD personnel badges. Measurements made with a commercial thin-window ion chamber instrument are compared with the dose rates obtained using other dosimeters. PMID:6698784

  5. Neutron initiation probability in fast burst reactor

    SciTech Connect

    Liu, X.; Du, J.; Xie, Q.; Fan, X. [Inst. of Nuclear Physics and Chemistry, China Academy of Engineering Physics, No.64, Mianshan Road, Mianyang, Sichuan (China)

    2012-07-01

    Based on the probability balance of neutron random events in multiply system, the four random process of neutron in prompt super-critical is described and then the equation of neutron initiation probability W(r,E,{Omega},t) is deduced. On the assumption of static, slightly prompt super-critical and the two factorial approximation, the formula of the average probability of 'one' neutron is derived which is the same with the result derived from the point model. The MC simulation using point model is applied in Godiva- II and CFBR-II, and the simulation result of one neutron initiation is well consistent with the theory that the initiation probability of Godiva- II inverted commas CFBR-II burst reactor are 0.00032, 0.00027 respectively on the ordinary burst operation. (authors)

  6. Godiva Rim Member: A new stratigraphic unit of the Green River Formation in southwest Wyoming and northwest Colorado. Geology of the Eocene Wasatch, Green River, and Bridger (Washakie) Formations, Greater Green River Basin, Wyoming, Utah, and Colorado. Professional paper

    SciTech Connect

    Roehler, H.W.

    1991-01-01

    The report names and describes the Godiva Rim Member of the Green River Formation in the eastern part of the Washakie basin in southwest Wyoming and the central part of the Sand Wash basin in northwest Colorado. The Godiva Rim Member comprises lithofacies of mixed mudflat and lacustrine origin situated between the overlying lacustrine Laney Member of the Green River Formation and the underlying fluvial Cathedral Bluffs Tongue of the Wasatch Formation. The Godiva Rim Member is laterally equivalent to and grades westward into the LaClede Bed of the Laney Member. The Godiva Rim Member of the Green River Formation was deposited along the southeast margins of Lake Gosiute and is correlated to similar lithologic units that were deposited along the northeast margins of Lake Uinta in the Parachute Creek Member of the Green River Formation. The stratigraphic data presented provide significant evidence that the two lakes were periodically connected around the east end of the Uinta Mountains during the middle Eocene.

  7. Reevaluation of an individual's radiation exposure at NTS in 1963-64. [FRAN reactor

    SciTech Connect

    Myers, D.S.

    1983-02-25

    The FRAN prompt burst reactor began operation at NTS on November 1, 1962 and continued in use until April 1965. From January 2, 1963 to August 12, 1964, an individual periodically performed maintenance and troubleshooting functions on various components of the FRAN reactor system. In June, 1980, the individual requested a review of the radiation dose that he received from his involvement with the FRAN reactor. An evaluation of the individual's radiation dose associated with the FRAN reactor operation was performed. This report details the reevaluation of the individual's estimated radiation dose from the FRAN reactor assembly, as derived from computer calculations, GODIVA-IV measurements, personnel dosimetry results, and a reconstruction of work scenarios.

  8. Shielding study for fast-burst reactor building

    SciTech Connect

    Rees, B. G. (Brian G.); Malenfant, R. E. (Richard E.)

    2002-01-01

    A study was conducted to evaluate the radiation levels and the response of various diagnostic components in and around the building that houses the Godiva IV fast-burst reactor assembly. In a typical operation of 1 MW-s (3.6{approx}10f'is{approx}s ions) in a 50 ps fwhm burst the peak power approaches 100,000 MW. The unshielded dose at 3 m is about 500 Rem. The results will be used to evaluate the radiation levels and shielding requirements for a new facility, and to anticipate problems with new safety, security, and diagnostic instrumentation. The study was required because of the difficulty of making accurate calculations, the intensity of the radiation, the mixed neutron and gamnta-ray source terms, and the complex nature of the structure. In addition to detailed dosimetry, attention was paid to the evaluation of spurious electromagnetic and radiofrequency signals produced in detectors, cables, and conduits.

  9. Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

    DOE PAGESBeta

    None

    2014-11-04

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with themore »GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.« less

  10. Reactor Physics Measurements and Benchmark Specifications for Oak Ridge Highly Enriched Uranium Sphere (ORSphere)

    DOE PAGESBeta

    None

    2014-11-04

    In the early 1970s Dr. John T. Mihalczo (team leader), J.J. Lynn, and J.R. Taylor performed experiments at the Oak Ridge Critical Experiments Facility (ORCEF) with highly enriched uranium (HEU) metal (called Oak Ridge Alloy or ORALLOY) in an effort to recreate GODIVA I results with greater accuracy than those performed at Los Alamos National Laboratory in the 1950s. The purpose of the Oak Ridge ORALLOY Sphere (ORSphere) experiments was to estimate the unreflected and unmoderated critical mass of an idealized sphere of uranium metal corrected to a density, purity, and enrichment such that it could be compared with the GODIVA I experiments. Additionally, various material reactivity worths, the surface material worth coefficient, the delayed neutron fraction, the prompt neutron decay constant, relative fission density, and relative neutron importance were all measured. The critical assembly, material reactivity worths, the surface material worth coefficient, and the delayed neutron fraction were all evaluated as benchmark experiment measurements. The reactor physics measurements are the focus of this paper; although for clarity the critical assembly benchmark specifications are briefly discussed.

  11. Reactor technology. Progress report, July-September 1980

    SciTech Connect

    Breslow, M. (ed.)

    1980-12-01

    Progress in the Space Power Advanced Reactor (SPAR) Program includes indications that revision of the BeO reflector configuration can reduce system weight. Observed boiling limit restrictions on the performance of the annular-wick core heat pipe have accelerated transition to the development of the target-design arterial heat pipe. Successful bends of core heat pipes have been made with sodium as the mandrel material. With the phasing out of the GCFR Program, work on the Low Power Safety Experiments Program is now concentrated on completion of the third 37-rod Full Length Subgroup test. In the Reactor Safety/Structural Analysis area, effort on the Category I Structures Program is toward developing an experimental test plan focusing on a specific structural design. Buckling experiments on thin-walled cylindrical shells with circular cutouts are reported. Results of a three-dimensional analysis of thermal stresses in the Fort St. Vrain core support block are presented. Materials investigations and operation of a molybdenum-core SiC heat pipe are reported. Entrainment limits for gravity-assisted heat pipes and heat pipe configurations for application to energy conservation are being investigated. The new solution critical assembly, SHEBA, was completed. Godiva IV was temporarily relocated at TA-15. Influence of scattered radiations in the test vault on InRad measurements was determined from detector scans of the vault produced by /sup 252/Cf neutron and /sup 137/Cs gamma sources.

  12. Verification of Unstructured Mesh Capabilities in MCNP6 for Reactor Physics Problems

    SciTech Connect

    Burke, Timothy P. [Los Alamos National Laboratory; Martz, Roger L. [Los Alamos National Laboratory; Kiedrowski, Brian C. [Los Alamos National Laboratory; Martin, William R. [Los Alamos National Laboratory

    2012-08-22

    New unstructured mesh capabilities in MCNP6 (developmental version during summer 2012) show potential for conducting multi-physics analyses by coupling MCNP to a finite element solver such as Abaqus/CAE[2]. Before these new capabilities can be utilized, the ability of MCNP to accurately estimate eigenvalues and pin powers using an unstructured mesh must first be verified. Previous work to verify the unstructured mesh capabilities in MCNP was accomplished using the Godiva sphere [1], and this work attempts to build on that. To accomplish this, a criticality benchmark and a fuel assembly benchmark were used for calculations in MCNP using both the Constructive Solid Geometry (CSG) native to MCNP and the unstructured mesh geometry generated using Abaqus/CAE. The Big Ten criticality benchmark [3] was modeled due to its geometry being similar to that of a reactor fuel pin. The C5G7 3-D Mixed Oxide (MOX) Fuel Assembly Benchmark [4] was modeled to test the unstructured mesh capabilities on a reactor-type problem.

  13. Compact Reactor

    SciTech Connect

    Williams, Pharis E. [Williams Research, P.O. Box 554, Los Alamos, NM87544 (United States)

    2007-01-30

    Weyl's Gauge Principle of 1929 has been used to establish Weyl's Quantum Principle (WQP) that requires that the Weyl scale factor should be unity. It has been shown that the WQP requires the following: quantum mechanics must be used to determine system states; the electrostatic potential must be non-singular and quantified; interactions between particles with different electric charges (i.e. electron and proton) do not obey Newton's Third Law at sub-nuclear separations, and nuclear particles may be much different than expected using the standard model. The above WQP requirements lead to a potential fusion reactor wherein deuterium nuclei are preferentially fused into helium nuclei. Because the deuterium nuclei are preferentially fused into helium nuclei at temperatures and energies lower than specified by the standard model there is no harmful radiation as a byproduct of this fusion process. Therefore, a reactor using this reaction does not need any shielding to contain such radiation. The energy released from each reaction and the absence of shielding makes the deuterium-plus-deuterium-to-helium (DDH) reactor very compact when compared to other reactors, both fission and fusion types. Moreover, the potential energy output per reactor weight and the absence of harmful radiation makes the DDH reactor an ideal candidate for space power. The logic is summarized by which the WQP requires the above conditions that make the prediction of DDH possible. The details of the DDH reaction will be presented along with the specifics of why the DDH reactor may be made to cause two deuterium nuclei to preferentially fuse to a helium nucleus. The presentation will also indicate the calculations needed to predict the reactor temperature as a function of fuel loading, reactor size, and desired output and will include the progress achieved to date.

  14. Thirty-five years at Pajarito Canyon Site

    SciTech Connect

    Paxton, H.C.

    1981-05-01

    A history of the research activities performed at the Pajarito Canyon Site from 1946 to 1981 is presented. Critical assemblies described include: the Topsy assembly; Lady Godiva; Godiva 2; Jezebel; Flattop; the Honeycomb assembly for Rover studies; Kiwi-TNT; PARKA reactor; Big Ten; and Plasma Cavity Assembly.

  15. Disposition of fuel elements from the Aberdeen and Sandia pulse reactor (SPR-II) assemblies

    SciTech Connect

    Mckerley, Bill [Los Alamos National Laboratory; Bustamante, Jacqueline M [Los Alamos National Laboratory; Costa, David A [Los Alamos National Laboratory; Drypolcher, Anthony F [Los Alamos National Laboratory; Hickey, Joseph [Los Alamos National Laboratory

    2010-01-01

    We describe the disposition of fuel from the Aberdeen (APR) and the Sandia Pulse Reactors (SPR-II) which were used to provide intense neutron bursts for radiation effects testing. The enriched Uranium - 10% Molybdenum fuel from these reactors was shipped to the Los Alamos National Laboratory (LANL) for size reduction prior to shipment to the Savannah River Site (SRS) for final disposition in the H Canyon facility. The Shipper/Receiver Agreements (SRA), intra-DOE interfaces, criticality safety evaluations, safety and quality requirements and key materials management issues required for the successful completion of this project will be presented. This work is in support of the DOE Consolidation and Disposition program. Sandia National Laboratories (SNL) has operated pulse nuclear reactor research facilities for the Department of Energy since 1961. The Sandia Pulse Reactor (SPR-II) was a bare metal Godiva-type reactor. The reactor facilities have been used for research and development of nuclear and non-nuclear weapon systems, advanced nuclear reactors, reactor safety, simulation sources and energy related programs. The SPR-II was a fast burst reactor, designed and constructed by SNL that became operational in 1967. The SPR-ll core was a solid-metal fuel enriched to 93% {sup 235}U. The uranium was alloyed with 10 weight percent molybdenum to ensure the phase stabilization of the fuel. The core consisted of six fuel plates divided into two assemblies of three plates each. Figure 1 shows a cutaway diagram of the SPR-II Reactor with its decoupling shroud. NNSA charged Sandia with removing its category 1 and 2 special nuclear material by the end of 2008. The main impetus for this activity was based on NNSA Administrator Tom D'Agostino's six focus areas to reenergize NNSA's nuclear material consolidation and disposition efforts. For example, the removal of SPR-II from SNL to DAF was part of this undertaking. This project was in support of NNSA's efforts to consolidate the locations of special nuclear material (SNM) to reduce the cost of securing many SNM facilities. The removal of SPR-II from SNL was a significant accomplishment in SNL's de-inventory efforts and played a key role in reducing the number of locations requiring the expensive security measures required for category 1 and 2 SNM facilities. A similar pulse reactor was fabricated at the Y-12 National Security Complex beginning in the late 1960's. This Aberdeen Pulse Reactor (APR) was operated at the Army Pulse Radiation Facility (APRF) located at the Aberdeen Test Center (ATC) in Maryland. When the APRF was shut down in 2003, a portion of the DOE-owned Special Nuclear Material (SNM) was shipped to an interim facility for storage. Subsequently, the DOE determined that the material from both the SPR-II and the APR would be processed in the H-Canyon at the Savannah River Site (SRS). Because of the SRS receipt requirements some of the material was sent to the Los Alamos National Laboratory (LANL) for size-reduction prior to shipment to the SRS for final disposition.

  16. Catalytic reactor

    DOEpatents

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

    2009-03-10

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

  17. Bioconversion reactor

    DOEpatents

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

    1992-01-01

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

  18. (Reactor dosimetry)

    SciTech Connect

    West, C.D.

    1990-09-13

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

  19. Fast reactor programme

    Microsoft Academic Search

    Srinivasan

    1973-01-01

    India's fast reactor programme is described in detail. A 15 MW(e) Fast ; Breeder Test Reactor (FBTR) under construction in the Reactor Research Centre at ; Kalpakkam, will provide experience in construction and operation of a sodium ; cooled fast reactor. Fuel and material testing is an essential aspect of fast ; reactor development. For this purpose, FBTR will serve

  20. Nuclear Reactors. Revised.

    ERIC Educational Resources Information Center

    Hogerton, John F.

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: How Reactors Work; Reactor Design; Research, Teaching, and Materials Testing; Reactors (Research, Teaching and Materials); Production Reactors; Reactors for Electric Power…

  1. Photocatalytic reactor

    DOEpatents

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

    1999-01-19

    A photocatalytic reactor is described for processing selected reactants from a fluid medium comprising at least one permeable photocatalytic membrane having a photocatalytic material. The material forms an area of chemically active sites when illuminated by light at selected wavelengths. When the fluid medium is passed through the illuminated membrane, the reactants are processed at these sites separating the processed fluid from the unprocessed fluid. A light source is provided and a light transmitting means, including an optical fiber, for transmitting light from the light source to the membrane. 4 figs.

  2. Hybrid adsorptive membrane reactor

    NASA Technical Reports Server (NTRS)

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

    2011-01-01

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

  3. Hybrid adsorptive membrane reactor

    DOEpatents

    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)

    2011-03-01

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

  4. Nuclear reactor engineering

    Microsoft Academic Search

    S. Glasstone; A. Sesonske

    1982-01-01

    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,

  5. Reactor safety method

    DOEpatents

    Vachon, Lawrence J. (Clairton, PA)

    1980-03-11

    This invention relates to safety means for preventing a gas cooled nuclear reactor from attaining criticality prior to start up in the event the reactor core is immersed in hydrogenous liquid. This is accomplished by coating the inside surface of the reactor coolant channels with a neutral absorbing material that will vaporize at the reactor's operating temperature.

  6. Nuclear reactor

    DOEpatents

    Thomson, Wallace B. (Severna Park, MD)

    2004-03-16

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

  7. Evaluation of LLNL's Nuclear Accident Dosimeters at the CALIBAN Reactor September 2010

    SciTech Connect

    Hickman, D P; Wysong, A R; Heinrichs, D P; Wong, C T; Merritt, M J; Topper, J D; Gressmann, F A; Madden, D J

    2011-06-21

    The Lawrence Livermore National Laboratory uses neutron activation elements in a Panasonic TLD holder as a personnel nuclear accident dosimeter (PNAD). The LLNL PNAD has periodically been tested using a Cf-252 neutron source, however until 2009, it was more than 25 years since the PNAD has been tested against a source of neutrons that arise from a reactor generated neutron spectrum that simulates a criticality. In October 2009, LLNL participated in an intercomparison of nuclear accident dosimeters at the CEA Valduc Silene reactor (Hickman, et.al. 2010). In September 2010, LLNL participated in a second intercomparison of nuclear accident dosimeters at CEA Valduc. The reactor generated neutron irradiations for the 2010 exercise were performed at the Caliban reactor. The Caliban results are described in this report. The procedure for measuring the nuclear accident dosimeters in the event of an accident has a solid foundation based on many experimental results and comparisons. The entire process, from receiving the activated NADs to collecting and storing them after counting was executed successfully in a field based operation. Under normal conditions at LLNL, detectors are ready and available 24/7 to perform the necessary measurement of nuclear accident components. Likewise LLNL maintains processing laboratories that are separated from the areas where measurements occur, but contained within the same facility for easy movement from processing area to measurement area. In the event of a loss of LLNL permanent facilities, the Caliban and previous Silene exercises have demonstrated that LLNL can establish field operations that will very good nuclear accident dosimetry results. There are still several aspects of LLNL's nuclear accident dosimetry program that have not been tested or confirmed. For instance, LLNL's method for using of biological samples (blood and hair) has not been verified since the method was first developed in the 1980's. Because LLNL and the other DOE participants were limited in what they were allowed to do at the Caliban and Silene exercises and testing of various elements of the nuclear accident dosimetry programs cannot always be performed as guests at other sites, it has become evident that DOE needs its own capability to test nuclear accident dosimeters. Angular dependence determination and correction factors for NADs desperately need testing as well as more evaluation regarding the correct determination of gamma doses. It will be critical to properly design any testing facility so that the necessary experiments can be performed by DOE laboratories as well as guest laboratories. Alternate methods of dose assessment such as using various metals commonly found in pockets and clothing have yet to be evaluated. The DOE is planning to utilize the Godiva or Flattop reactor for testing nuclear accident dosimeters. LLNL has been assigned the primary operational authority for such testing. Proper testing of nuclear accident dosimeters will require highly specific characterization of the pulse fields. Just as important as the characterization of the pulsed fields will be the design of facilities used to process the NADs. Appropriate facilities will be needed to allow for early access to dosimeters to test and develop quick sorting techniques. These facilities will need appropriate laboratory preparation space and an area for measurements. Finally, such a facility will allow greater numbers of LLNL and DOE laboratory personnel to train on the processing and interpretation of nuclear accident dosimeters and results. Until this facility is fully operational for test purposes, DOE laboratories may need to continue periodic testing as guests of other reactor facilities such as Silene and Caliban.

  8. Ultimate Safe (US) Reactor

    SciTech Connect

    Gat, U.; Daugherty, S.R.

    1985-01-01

    The Ultimate Safe (US) Reactor is a reactor that eliminates the traditional safety concerns of nuclear fission reactors. The US reactor has an insignificant source term and no reasonable criticality accident. Furthermore, the negligible residual after-heat in the reactor renders its shutdown capability comparable or superior to conventional power sources. Fission products are continuously removed at the rate they are produced. The reactor is operated with no excess criticality, hence no criticality accident is reasonably possible. The reactor is controlled safely by its negative temperature coeffiient. The reactor maintains criticality by an internal breeding ratio that is trimmed to be exactly one. The US reactor requires a fluid fuel and on-line, continuous fuel processing. Molten salt fuel was selected for its low vapor pressure at high temperature; adequate solubility of uranium and thorium as fluorides; good compatibility with structural materials; absence of irradiation damage; high negative temperature coefficient and amply developed technology and experience.

  9. Attrition reactor system

    SciTech Connect

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

    1993-09-28

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

  10. Attrition reactor system

    SciTech Connect

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

    1993-01-01

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

  11. Hybrid plasmachemical reactor

    NASA Astrophysics Data System (ADS)

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

    2015-04-01

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

  12. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2013-05-28

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

  13. Advanced Test Reactor Tour

    SciTech Connect

    Miley, Don

    2011-01-01

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

  14. Silver reactor reclamation

    Microsoft Academic Search

    C. W. Malody; T. R. McKenzie; C. W. Pollock

    1959-01-01

    Two spent Redox reactors were transported to U-Plant for regeneration. The old saddle coating material was removed by flushing with water and sodium thiosulfate. The reactors were dried by pulling air through them and regenerated by spraying the Berl saddles with a 5 molar solution of silver nitrate. When the reclamation process was completed and the first reactor was removed

  15. High solids fermentation reactor

    DOEpatents

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

    1993-01-01

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

  16. High solids fermentation reactor

    DOEpatents

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

    1993-03-02

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

  17. Reactor vessel support system

    DOEpatents

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

    1982-01-01

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

  18. Spinning fluids reactor

    DOEpatents

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

    2012-11-20

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

  19. Reactor water cleanup system

    DOEpatents

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

    1994-12-20

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

  20. High temperature reactors

    NASA Astrophysics Data System (ADS)

    Dulera, I. V.; Sinha, R. K.

    2008-12-01

    With the advent of high temperature reactors, nuclear energy, in addition to producing electricity, has shown enormous potential for the production of alternate transport energy carrier such as hydrogen. High efficiency hydrogen production processes need process heat at temperatures around 1173-1223 K. Bhabha Atomic Research Centre (BARC), is currently developing concepts of high temperature reactors capable of supplying process heat around 1273 K. These reactors would provide energy to facilitate combined production of hydrogen, electricity, and drinking water. Compact high temperature reactor is being developed as a technology demonstrator for associated technologies. Design has been also initiated for a 600 MWth innovative high temperature reactor. High temperature reactor development programme has opened new avenues for research in areas like advanced nuclear fuels, high temperature and corrosion resistant materials and protective coatings, heavy liquid metal coolant technologies, etc. The paper highlights design of these reactors and their material related requirements.

  1. Fission reactors and materials

    SciTech Connect

    Frost, B.R.T.

    1981-12-01

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

  2. University Reactor Sharing Program

    SciTech Connect

    Dr. W.D. Reece

    1999-09-01

    The University Reactor Sharing Program provides funding for reactor experimentation to institutions that do not normally have access to a research reactor. Research projects supported by the program include items such as dating geological material to producing high current super conducting magnets. The funding also gives small colleges and universities the opportunity to use the facility for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy.

  3. The Integral Fast Reactor

    SciTech Connect

    Chang, Y.I.

    1988-01-01

    The Integral Fast Reactor (IFR) is an innovative liquid metal reactor concept being developed at Argonne National Laboratory. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system. This paper describes the key features and potential advantages of the IFR concept, with emphasis on its safety characteristics. 3 refs., 4 figs., 1 tab.

  4. Reactor service life extension program

    Microsoft Academic Search

    G. R. Caskey; R. L. Sindelar; R. S. Ondrejcin; E. W. Baumann

    1990-01-01

    A review of the Savannah River Site production reactor systems was initiated in 1980 and led to implementation of the Reactor Materials Program in 1984 to assess reactor safety and reactor service life. The program evaluated performance of the reactor tanks, primary coolant piping, and thermal shields, components of welded construction that were fabricated from Type 304 stainless steel. The

  5. Nerva Nuclear Reactor Instrumentation

    Microsoft Academic Search

    F. S. Malick; R. L. Ramp; G. A. Gilmour

    1966-01-01

    The environmental conditions existing at many of the points at which measurements are required in the ground testing of the NERVA reactors are beyond the capabilities of commercially available transducers. The measurement of the temperatures inside the reactor core was accomplished by pushing thermocouples to the high temperature limit of this sensing means. Tungsten and tungsten alloys have shown the

  6. EBT reactor analysis

    Microsoft Academic Search

    N. A. Uckan; E. F. Jaeger; R. T. Santoro; D. A. Spong; T. Uckan; L. W. Owen; J. M. Barnes; J. B. McBride

    1983-01-01

    This report summarizes the results of a recent ELMO Bumpy Torus (EBT) reactor study that includes ring and core plasma properties with consistent treatment of coupled ring-core stability criteria and power balance requirements. The principal finding is that constraints imposed by these coupling and other physics and technology considerations permit a broad operating window for reactor design optimization. Within this

  7. Nuclear Reactor Safety

    Microsoft Academic Search

    J. D. Bales; R. Boshears

    1996-01-01

    Nuclear Reactor Safety (NRS), published monthly, is a collection of abstracts of worldwide information available on all safety-related aspects of reactors, including accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures. 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

  8. Fast breeder reactors--1

    Microsoft Academic Search

    W. Marshall; L. M. Davies

    1979-01-01

    These four lectures describe the development and construction of fast breeder reactors, their safety, the fuel cycle and the economics of the system. Lastly, the concerns are considered that have been expressed with regard to the introduction of fast breeder reactors on a commercial scale.

  9. Fast breeder reactor safety

    Microsoft Academic Search

    1976-01-01

    The current state of knowledge with respect to the analysis of postulated accident sequences in fast breeder reactors, as well as some implications to fast reactor design, are reviewed. The accidents considered include loss of coolant flow and transient overpower, both with a postulated failure to scram. The associated accident phenomena considered include fuel, clad, and coolant motions during the

  10. Fast Breeder Reactor studies

    Microsoft Academic Search

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

    2010-01-01

    This report is a compilation of Fast Breeder Reactor (FBR) resource documents prepared to provide the technical basis for the US contribution to the International Nuclear Fuel Cycle Evaluation. The eight separate parts deal with the alternative fast breeder reactor fuel cycles in terms of energy demand, resource base, technical potential and current status, safety, proliferation resistance, deployment, and nuclear

  11. Fast breeder reactors

    Microsoft Academic Search

    A. E. Waltar; A. B. Reynolds

    1981-01-01

    This book describes the major design features of fast breeder reactors and the methods used for their design and analysis. The foremost objective of this book is to fulfill the need for a textbook on Fast Breeder Reactor (FBR) technology at the graduate level or the advanced undergraduate level. It is assumed that the reader has an introductory understanding of

  12. The Integral Fast Reactor

    SciTech Connect

    Till, C.E.; Chang, Y.I. (Argonne National Lab., IL (USA)); Lineberry, M.J. (Argonne National Lab., Idaho Falls, ID (USA))

    1990-01-01

    Argonne National Laboratory, since 1984, has been developing the Integral Fast Reactor (IFR). This paper will describe the way in which this new reactor concept came about; the technical, public acceptance, and environmental issues that are addressed by the IFR; the technical progress that has been made; and our expectations for this program in the near term. 5 refs., 3 figs.

  13. Advanced spheromak fusion reactor

    Microsoft Academic Search

    T. K. Fowler; E. B. Hooper

    1996-01-01

    The spheromak has no toroidal magnetic field coils or other structure along its geometric axis, and is thus more attractive than the leading magnetic fusion reactor concept, the tokamak. As a consequence of this and other attributes, the spheromak reactor may be compact and produce a power density sufficiently high to warrant consideration of a liquid `blanket` that breeds tritium,

  14. Spherical torus fusion reactor

    Microsoft Academic Search

    Martin Peng; Y. K. M

    1985-01-01

    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

  15. Natural fission reactor

    Microsoft Academic Search

    L G. A. Cowan

    1976-01-01

    The evidence supporting the establishment of the zones of a natural fission reactor that occurred about 2 billion years ago is presented. The reactor evidence was found in an open-pit uranium mine in the southeastern part of the Gabon Republic, near the Equator on the coast of West Africa. The history of the probable sequence of events leading to the

  16. Nuclear power technology. Volume 1: Reactor technology

    Microsoft Academic Search

    1984-01-01

    This book, first in series on nuclear power technology, presents papers on reactor technology. The topics covered are: how reactors work; reactor physics; gas-cooled reactors; light water reactors; fast reactors; a review of the IKAEA interest in heavy water reactors; novel reactor concepts; and prospects for fusion.

  17. Reactor physics of natriun-cooled fast breeder reactors

    Microsoft Academic Search

    M. Rajamaeki

    1980-01-01

    Reactor properties of liquid metal fast breeder reactors and pertinent computational methods with their applications and results are presented. Designs of reactor core and fuel are outlined. Main characteristics and layout of the homogenous and the heterogenous core are described. Reactor safety and the possibility of a hypothetical core descriptive accident is discussed. The influence of the neutron flux on

  18. Nuclear reactor control column

    DOEpatents

    Bachovchin, Dennis M. (Plum Borough, PA)

    1982-01-01

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

  19. Reactor Safety Research Programs

    SciTech Connect

    Edler, S. K.

    1981-07-01

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

  20. Nuclear reactor reflector

    DOEpatents

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

    1994-01-01

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

  1. Nuclear reactor reflector

    DOEpatents

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

    1994-06-07

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

  2. Reactor safety assessment system

    SciTech Connect

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

    1987-01-01

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

  3. Environmental aspects of fusion reactors

    Microsoft Academic Search

    F. E. Coffman; J. M. Williams

    1975-01-01

    Potential environmental impacts of commercial fusion reactors are discussed and compared with those of fission reactors. It is shown that the environmental impact of fusion reactors will be quite small, with the main contribution coming from thermal discharges. Some attractive safety and environmental characteristics of fusion reactors are described, including an effectively infinite low-cost fuel supply, their inherent incapacity for

  4. Moving-Medium Biofilm Reactors

    Microsoft Academic Search

    M. Rodgers; X.-M. Zhan

    2003-01-01

    Four moving-medium biofilm reactors treating wastewater were reviewed in this paper: the rotating biological contactor (RBC), the moving bed biofilm reactor (MBBR), the vertically moving biofilm reactor (VMBR) and the fluidized-bed reactor (FBR). The RBC process has been applied widely. MBBR is a good process for upgrading current wastewater treatment systems. VMBR is suitable for treating small wastewater flows. FBR

  5. Fusion reactor radioactive waste management

    Microsoft Academic Search

    J. D. Kaser; A. K. Postma; D. J. Bradley

    1976-01-01

    Quantities and compositions of non-tritium radioactive waste are estimated for some current conceptual fusion reactor designs, and disposal of large amounts of radioactive waste appears necessary. Although the initial radioactivity of fusion reactor and fission reactor wastes are comparable, the radionuclides in fusion reactor wastes are less hazardous and have shorter half-lives. Areas requiring further research are discussed.

  6. The Endurance Bioenergy Reactor

    SciTech Connect

    Laible, Philip

    2012-01-01

    Argonne biophysicist Dr. Philip Laible and Air Force Major Matt Michaud talks about he endurance bioenergy reactor—a device that contains bacteria that can convert energy from the sun into fuel molecules.

  7. Reactor hot spot analysis

    SciTech Connect

    Vilim, R.B.

    1985-08-01

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

  8. Toroidal fusion reactors

    Microsoft Academic Search

    Furth

    1983-01-01

    Toroidal reactor geometry offers the attractive features of compact size and low recirculating power. The tokamak approach (where a toroidally directed current circulates in the plasma) has proved to be a particularly cost-effective way of entering the reactor plasma regime. Temperatures above 7 keV and n\\/TAU\\/ \\/SUB E\\/ -values above 4 . 10¹³ cm⁻³ sec - already achieved in individual

  9. A Tokamak experimental power reactor

    Microsoft Academic Search

    W. M. Stacey Jr.; V. A. Maroni; J. R. Purcell; M. A. Abdou; P. J. Bertoncini; J. N. Brooks; J. B. Darby Jr.; K. Evans Jr.; J. A. Fasolo; R. L. Kustom

    1976-01-01

    A detailed report on status and objectives of the Tokamak experimental power reactor (TEPR) project to begin operation about 1985-87. This system should consitute the next step beyond the projected Tokamak fusion reactor, in achieving deuterium-tritium (D-T) plasma confinement for power reactors. Plasma physics, power reactor engineering feasibility, technology demonstration, and test facility utilization are the major problem definitions. Reactor

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

    E-print Network

    Greifswald, Ernst-Moritz-Arndt-Universität

    Structure of processes in flow reactor and closed reactor: Flow reactor Closed reactor Active Zone / Phase Te >> T · dissociation, excitation etc. · plasma wall interaction Passive Zone / Phase Te T · recombination, relaxation etc. · plasma wall interaction Feed gas Stable products A2 Ak Pel Pel Passive Zone

  11. F Reactor Inspection

    SciTech Connect

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-10-29

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

  12. Reactor Safety Research Programs

    SciTech Connect

    Dotson, CW

    1980-08-01

    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.

  13. F Reactor Inspection

    ScienceCinema

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2014-11-24

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

  14. Moon base reactor system

    NASA Technical Reports Server (NTRS)

    Chavez, H.; Flores, J.; Nguyen, M.; Carsen, K.

    1989-01-01

    The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.

  15. Nuclear power technology. Volume 1. Reactor technology

    Microsoft Academic Search

    1983-01-01

    This book presents papers on the nuclear industry. Topics considered include the basics of reactor operation, reactor physics, gas cooled reactors, light water reactors, fast reactors, the UKAEA interest in heavy water reactors, novel reactor concepts, prospects for fusion, fuel recycling, uranium and thorium raw materials, uranium supply, uranium enrichment, fuel design and fabrication, nuclear fuel reprocessing in the UK,

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

    SciTech Connect

    P. Delmolino

    2005-05-06

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

  17. REACTOR GROUT THERMAL PROPERTIES

    SciTech Connect

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

    2011-01-28

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

  18. Methanation assembly using multiple reactors

    DOEpatents

    Jahnke, Fred C.; Parab, Sanjay C.

    2007-07-24

    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.

  19. Looking Southwest at Reactor Box Furnaces With Reactor Boxes and ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Looking Southwest at Reactor Box Furnaces With Reactor Boxes and Repossessed Uranium in Recycle Recovery Building - Hematite Fuel Fabrication Facility, Recycle Recovery Building, 3300 State Road P, Festus, Jefferson County, MO

  20. Nuclear reactor safety device

    DOEpatents

    Hutter, Ernest (Wilmette, IL)

    1986-01-01

    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.

  1. Heat dissipating nuclear reactor

    DOEpatents

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

    1987-01-01

    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.

  2. Reactor for exothermic reactions

    DOEpatents

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

    1993-03-02

    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.

  3. Heat dissipating nuclear reactor

    DOEpatents

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    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.

  4. Compact reactor design automation

    NASA Technical Reports Server (NTRS)

    Nassersharif, Bahram; Gaeta, Michael J.

    1991-01-01

    A conceptual compact reactor design automation experiment was performed using the real-time expert system G2. The purpose of this experiment was to investigate the utility of an expert system in design; in particular, reactor design. The experiment consisted of the automation and integration of two design phases: reactor neutronic design and fuel pin design. The utility of this approach is shown using simple examples of formulating rules to ensure design parameter consistency between the two design phases. The ability of G2 to communicate with external programs even across networks provides the system with the capability of supplementing the knowledge processing features with conventional canned programs with possible applications for realistic iterative design tools.

  5. Reactor for exothermic reactions

    DOEpatents

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

    1993-01-01

    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.

  6. Aging degradation of BWR (Boiling Water Reactor) reactor internals

    Microsoft Academic Search

    Ware

    1989-01-01

    Researchers at the Idaho National Engineering Laboratory recently performed an assessment of the aging of the reactor internals in Boiling Water Reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. Several failures in BWR reactor internals have been caused by a combination of susceptible materials, environment, preload stress, and flow-induced vibration. ASME Code Section

  7. Fast quench reactor method

    DOEpatents

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

    1999-08-10

    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.

  8. Particle bed reactor modeling

    NASA Technical Reports Server (NTRS)

    Sapyta, Joe; Reid, Hank; Walton, Lew

    1993-01-01

    The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

  9. Reactor Neutrino Experiments

    E-print Network

    Jun Cao

    2007-12-06

    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.

  10. Fusion reactor pumped laser

    DOEpatents

    Jassby, Daniel L. (Princeton, NJ)

    1988-01-01

    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.

  11. Diagnostics for hybrid reactors

    NASA Astrophysics Data System (ADS)

    Orsitto, Francesco Paolo

    2012-06-01

    The Hybrid Reactor(HR) can be considered an attractive actinide-burner or a fusion assisted transmutation for destruction of transuranic(TRU) nuclear waste. The hybrid reactor has two important subsystems: the tokamak neutron source and the blanket which includes a fuel zone where the TRU are placed and a tritium breeding zone. The diagnostic system for a HR must be as simple and robust as possible to monitor and control the plasma scenario, guarantee the protection of the machine and monitor the transmutation.

  12. Inertial confinement fusion reactor systems

    SciTech Connect

    Frank, T.G.; Bohachevsky, I.O.; Pendergrass, J.H.

    1980-01-01

    A variety of reactor cavity concepts, drivers, and energy conversion mechanisms are being considered to realize commercial applications of ICF. Presented in this paper are: (1) a review of reactor concepts with estimates of practically achievable pulse repetition rates; (2) a survey of drivers with estimates of the requirements on reactor conditions imposed by beam propagation characteristics; and (3) an assessment of compatible driver-reactor combinations.

  13. Reactor operation environmental information document

    Microsoft Academic Search

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

    1989-01-01

    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

  14. Transient analysis of space reactors

    Microsoft Academic Search

    H. Q. Giap; F. Best

    1985-01-01

    Nuclear reactors are being considered as the primary energy source for an increasing number of space missions including the National Aeronautics and Space Administration's space station and the Strategic Defense Initiative's SP-100. Conceptual reactor designs include fast and thermal spectrum reactors with liquid-metal or gas cooling delivering thermal energy to Brayton, Stirling, thermoelectric, thermoionic, or Rankine cycle power conversion systems.

  15. Recent advances in reactor systems

    Microsoft Academic Search

    Insch

    1973-01-01

    New developments in reactor systems must aim to secure improvements ; which are greater than the advantage of replicating systems which are already ; proven. Features of light-water reactors which today establish criteria for ; future developments are discussed. For comparative purposes, developments with ; heavywater and high temperature reactors are described. (GE);

  16. Argonne National Laboratory's Reactor Performance

    E-print Network

    Kemner, Ken

    power reactors. Those older codes, while well calibrated for evaluating the safety of next to a specific reactor application. As a result, SHARP is largely technology neutral and can be appliedSHARP Argonne National Laboratory's Reactor Performance and Safety Simulation Suite #12

  17. Fast Breeder Reactors in France

    Microsoft Academic Search

    MYCLE SCHNEIDER

    2009-01-01

    France is the only country in the world ever to operate a commercial scale (1,200 MWe) sodium cooled, plutonium fuelled fast breeder reactor, the Superphénix at Creys-Malville. However, the French fast breeder reactor program turned out to be too costly and could never compete with light water reactor technology. Numerous technical problems, low uranium prices and massive opposition exacerbated the

  18. REACTOR SAFETY INSTRUMENTATION

    Microsoft Academic Search

    Schultz

    1963-01-01

    A review is presented of the development of reactor safety ; instrumentation. The reliability of multiple-channel instrumentation is ; discussed, and various checking systems for extending reliability are reviewed. ; The influence of transistorization on circuit design and the future trends in ; safety system design (particularly digitalization) are discussed. (D.L.C.);

  19. REACTORS NEED SUPERLATIVE WELDING

    Microsoft Academic Search

    1962-01-01

    Discussion is given on Tig, Mig, stick, and electron beam welding of ; atomic reactors using austenitic and ferritic stainless steel, stainless clad ; steel, Inconel, carbon steel, Zr, the precipitation hardening steels, and others. ; Testing by x ray, dye penetration, and magnetic particles is also included. ; (P.C.H.);

  20. Plasma core reactor applications

    NASA Technical Reports Server (NTRS)

    Latham, T. S.; Rodgers, R. J.

    1976-01-01

    Analytical and experimental investigations were conducted to demonstrate the feasibility of fissioning uranium plasma core reactors and to characterize space and terrestrial applications for such reactors. Uranium hexafluoride fuel is injected into core cavities and confined away from the surface by argon buffer gas injected tangentially from the peripheral walls. Radiant heat transfer calculations were performed for a six-cavity reactor configuration. Axial working fluid channels are located along a fraction of each cavity peripheral wall. Results of calculations for outward-directed radiant energy fluxes corresponding to radiating temperatures of 2000 to 5000 K indicate total operating pressures from 80 to 650 atm, centerline temperatures from 6900 to 30,000 K, and total radiated powers from 25 to 2500 MW, respectively. Applications are described for this type of reactor such as (1) high-thrust, high specific impulse space propulsion, (2) highly efficient systems for generation of electricity, and (3) hydrogen or synthetic fuel production systems using the intense radiant energy fluxes.

  1. ZIRCONIUM ALLOYS FOR REACTORS

    Microsoft Academic Search

    Nishihara

    1960-01-01

    The general characteristics of the reactor-grade Zr alloys are revewed, ; including the physical properties of pure Zr and of the pertinent alloys, ; manufacturing processes of Zr sponge and of Zircaloy, electric arc welding ; methods for the pure metal and the ingot, hot and cold working, annealing and ; welding. The high temperatare behavior of Zircaloy, Ozhennite and

  2. Nuclear reactor building

    DOEpatents

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

    1994-01-01

    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.

  3. Nuclear Reactor Safety; (USA)

    Microsoft Academic Search

    D. L. Cason; S. C. Hicks

    1991-01-01

    This publication announces on an monthly basis the current worldwide information available on all safety-related aspects of reactors, including: accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures. 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.

  4. NETL - Chemical Looping Reactor

    ScienceCinema

    None

    2014-06-26

    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.

  5. Nuclear reactor decontamination

    Microsoft Academic Search

    Torok

    1981-01-01

    Heat transfer and associated surfaces in nuclear reactors are decontaminated by treating the surface with ozone to oxidize acid -insoluble metal oxides to a more soluble state, removing oxidized solubilized metal oxides, and removing other surface oxides using low concentrations of decontaminating reagents. Ozone treatment has been found very effective with alloys having surface metal oxides rendered more easily dissolved

  6. Materials for fusion reactors

    Microsoft Academic Search

    1978-01-01

    Material problems for fusion reactor consist of finding materials to be used for the first wall, limiters, electrical insulation, neutron moderators, shielding, and component for superconducting magnets. Materials being considered for the first wall include austenitic stainless steel, nickel, titanium, refractory metals and aluminium alloys. TiC, graphite, copper, and refractory metals are being evaluated for limiters which are used to

  7. Stabilized Spheromak Fusion Reactors

    SciTech Connect

    Fowler, T

    2007-04-03

    The U.S. fusion energy program is focused on research with the potential for studying plasmas at thermonuclear temperatures, currently epitomized by the tokamak-based International Thermonuclear Experimental Reactor (ITER) but also continuing exploratory work on other plasma confinement concepts. Among the latter is the spheromak pursued on the SSPX facility at LLNL. Experiments in SSPX using electrostatic current drive by coaxial guns have now demonstrated stable spheromaks with good heat confinement, if the plasma is maintained near a Taylor state, but the anticipated high current amplification by gun injection has not yet been achieved. In future experiments and reactors, creating and maintaining a stable spheromak configuration at high magnetic field strength may require auxiliary current drive using neutral beams or RF power. Here we show that neutral beam current drive soon to be explored on SSPX could yield a compact spheromak reactor with current drive efficiency comparable to that of steady state tokamaks. Thus, while more will be learned about electrostatic current drive in coming months, results already achieved in SSPX could point to a productive parallel development path pursuing auxiliary current drive, consistent with plans to install neutral beams on SSPX in the near future. Among possible outcomes, spheromak research could also yield pulsed fusion reactors at lower capital cost than any fusion concept yet proposed.

  8. Spherical torus fusion reactor

    Microsoft Academic Search

    Peng; Yueng-Kay M

    1989-01-01

    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

  9. DT Beam Fusion Reactor

    Microsoft Academic Search

    F. J. Wessel; N. Rostoker

    1998-01-01

    The Beam Fusion Reactor (BFR) is based on a field-reversed configuration and confined ion energies in the range of hundreds of keV. Repetitively pulsed, intense ion beams sustain the ion distributions and provide current drive. In the BFR the ion orbit size is comparable to the dimensions of the confined plasma and the expectation is for classical transport of the

  10. ICF tritium production reactor

    SciTech Connect

    Meier, W.R.; McCarville, T.J.; Berwald, D.H.; Gordon, J.D.; Steele, W.G.

    1985-02-28

    The conceptual design of an ICF tritium production reactor is described. The chamber design uses a beryllium multiplier and a liquid lithium breeder to achieve a tritium breeding ratio of 2.08. The annual net tritium production of this 532 MW/sub t/ plant is 16.9 kg, and the estimated cost of tritium is $8100/g.

  11. Decontaminating reactor coolant systems

    Microsoft Academic Search

    R. Whitaker; C. Wood

    1984-01-01

    Chemical agents and processes that reduce radioactivity levels where people must work make repairs in the coolant circuits of nuclear power reactors faster and less costly. The low-oxidation-state metal ions (LOMI) chemical process quickly dissolves tough corrosion films that hold radioactive isotopes. Radiation fields on pipe surfaces can be cut by a factor of 20 before repair crews go to

  12. Thermal Reactor Safety

    SciTech Connect

    Not Available

    1980-06-01

    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.

  13. Nuclear reactor building

    DOEpatents

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

    1994-04-05

    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. In and Out Reactor

    NSDL National Science Digital Library

    Integrated Teaching and Learning Program, College of Engineering,

    Students learn about material balances, a fundamental concept of chemical engineering. They use stoichiometry to predict the mass of carbon dioxide that escapes after reacting measured quantities of sodium bicarbonate with dilute acetic acid. Students then produce the reactions of the chemicals in a small reactor made from a plastic water bottle and balloon.

  15. NETL - Chemical Looping Reactor

    SciTech Connect

    None

    2013-07-24

    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. Inertial fusion reactors (review)

    Microsoft Academic Search

    A V Kalinin

    1984-01-01

    The current status of the work on engineering problems of inertial fusion is reviewed. The state of laboratory physics research, aspects of driver choice, and pellet designs are discussed. Particular attention is paid to the development of inertial fusion reactors and their energy applications. It is shown that lasers with output energies in the 0.5-5.0 MJ range must be acknowledged

  17. Alternative approaches to fusion. [reactor design and reactor physics for Tokamak fusion reactors

    NASA Technical Reports Server (NTRS)

    Roth, R. J.

    1976-01-01

    The limitations of the Tokamak fusion reactor concept are discussed and various other fusion reactor concepts are considered that employ the containment of thermonuclear plasmas by magnetic fields (i.e., stellarators). Progress made in the containment of plasmas in toroidal devices is reported. Reactor design concepts are illustrated. The possibility of using fusion reactors as a power source in interplanetary space travel and electric power plants is briefly examined.

  18. Hybrid space nuclear reactor concept

    NASA Astrophysics Data System (ADS)

    Begg, Lester L.; Choong, Phillip T.; Teofilo, Vincent L.; Dunn, Charles; Bhattacharyya, Sam

    A hybrid space nuclear reactor which can provide both electrical power and thermal propulsion has been found to have significant benefits for near term military satellites; it combines the STAR-C thermionic reactor with NERVA rocket propulsion technology. This hybrid reactor would use a W/UO2 cermet core which was developed in the nuclear rocket program of the 1960s. Both the reactor core and the thermionic converters can be tested prior to launch. The design of a 10 kW(e) reactor capable of producing over 1300 N of thrust is described.

  19. Reactor vessel support system. [LMFBR

    DOEpatents

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

    1980-05-09

    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.

  20. Nuclear reactor construction with bottom supported reactor vessel

    DOEpatents

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

    1987-01-01

    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. Spherical torus fusion reactor

    DOEpatents

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

    1989-01-01

    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.

  2. Nuclear reactor safety device

    DOEpatents

    Hutter, E.

    1983-08-15

    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.

  3. Fusion reactor pumped laser

    DOEpatents

    Jassby, D.L.

    1987-09-04

    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.

  4. Fissioning Plasma Core Reactor

    NASA Technical Reports Server (NTRS)

    Albright, Dennis; Butler, Carey; West, Nicole; Cole, John W. (Technical Monitor)

    2002-01-01

    Institute for Scientific Research, Inc. (ISR) research program consist of: 1.Study core physics by adapting existing codes: MCNP4C - Monte Carlo code; COMBINE/VENTURE - diffusion theory; SCALE4 - Monte Carlo, with many utility codes. 2. Determine feasibility and study major design parameters: fuel selection, temperature and reflector sizing. 3. Study reactor kinetics: develop QCALC1 to model point kinetics; study dynamic behavior of the power release.

  5. Nuclear reactor decontamination

    SciTech Connect

    Torok, J.

    1981-09-01

    Heat transfer and associated surfaces in nuclear reactors are decontaminated by treating the surface with ozone to oxidize acid -insoluble metal oxides to a more soluble state, removing oxidized solubilized metal oxides, and removing other surface oxides using low concentrations of decontaminating reagents. Ozone treatment has been found very effective with alloys having surface metal oxides rendered more easily dissolved by ozone oxidation especially with chromium or chromium-nickel containing alloys.

  6. Tokamak fusion power reactors

    Microsoft Academic Search

    W. M. Stacey Jr.; M. A. Abdou

    1978-01-01

    The major parameters and corresponding economic characteristics of a representative class of commercial Tokamak fusion power reactors are examined as a function of four major design parameters: plasma beta-t, toroidal magnetic field strength, first-wall lifetime, and power output. It is shown that for beta-t greater than or equal to 0.06, the minimum cost of energy is obtained for toroidal field

  7. In situ reactor

    DOEpatents

    Radtke, Corey William; Blackwelder, David Bradley

    2004-01-27

    An in situ reactor for use in a geological strata, is described and which includes a liner defining a centrally disposed passageway and which is placed in a borehole formed in the geological strata; and a sampling conduit is received within the passageway defined by the liner and which receives a geological specimen which is derived from the geological strata, and wherein the sampling conduit is in fluid communication with the passageway defined by the liner.

  8. In-reactor performance of pressure tubes in CANDU reactors

    Microsoft Academic Search

    D. K. Rodgers; C. E. Coleman; M. Griffiths; G. A. Bickel; J. R. Theaker; I. Muir; A. A. Bahurmuz; S. St. Lawrence; M. Resta Levi

    2008-01-01

    The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr–2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS,

  9. 78 FR 71675 - Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor Correspondence

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-11-29

    ...NRC-2013-0260] Update of the Office of Nuclear Reactor Regulation's Electronic Operating Reactor...Evans, Director, Division of Operating Reactor Licensing, Office of Nuclear Reactor Regulation. [FR Doc. 2013-28699...

  10. Acceptability of reactors in space

    SciTech Connect

    Buden, D.

    1981-04-01

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it dies not appear that reactors add measurably to the risk associated with the Space Transportation System.

  11. Acceptability of reactors in space

    SciTech Connect

    Buden, D.

    1981-01-01

    Reactors are the key to our future expansion into space. However, there has been some confusion in the public as to whether they are a safe and acceptable technology for use in space. The answer to these questions is explored. The US position is that when reactors are the preferred technical choice, that they can be used safely. In fact, it does not appear that reactors add measurably to the risk associated with the Space Transportation System.

  12. URSULA reactor vessel examination system

    SciTech Connect

    NONE

    1996-09-01

    A system for ultrasonic inspection of reactor vessel welds is described. The modular system has a robotic arm; when equipped with dual robots, it can perform a vessel examination in four days. Its use at the Catawba and Crystal River-3 nuclear power plants, both pressurized water reactors, is briefly described. A comparison is made to the Automated Reactor Inspection System (ARIS) robot, and the inspection sequence is outlined.

  13. Fast quench reactor and method

    DOEpatents

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

    1998-05-12

    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.

  14. Standard Operating Procedure (Polypropylene Reactor System)

    E-print Network

    Choi, Kyu Yong

    1 Standard Operating Procedure (Polypropylene Reactor System) Facility: Polymer Reaction Scope: This SOP details the use of the Polypropylene Reactor system for work in the Polymer Reaction. Place reactor in KOH solution to clean the reactor for next experiment. Polypropylene Reactor System

  15. POWER REACTOR TECHNOLOGY. Volume 6, No. 3

    Microsoft Academic Search

    Zinn

    1963-01-01

    Reviews of selected reports in the fields of powerreactor research and ; development, power-reactor applications, design practice, and operating ; experience are presented. Topics covered include: economics and applications; ; fluid and thermal technology; materials; containment, radiation control, and ; siting; design practice; operating experience; gas-cooled reactors; pressurized-; water reactors; organic-cooled and organic-moderated reactors; supercritical ; water reactors; and physics.

  16. Thermonuclear Reflect AB-Reactor

    E-print Network

    Alexander Bolonkin

    2008-03-26

    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.

  17. Neutrino oscillation studies with reactors.

    PubMed

    Vogel, P; Wen, L J; Zhang, C

    2015-01-01

    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. PMID:25913819

  18. Mirror Advanced Reactor Study (MARS)

    SciTech Connect

    Logan, B.G.

    1983-03-28

    Progress in a two year study of a 1200 MWe commercial tandem mirror reactor (MARS - Mirror Advanced Reactor Study) has reached the point where major reactor system technologies are identified. New design features of the magnets, blankets, plug heating systems and direct converter are described. With the innovation of radial drift pumping to maintain low plug density, reactor recirculating power fraction is reduced to 20%. Dominance of radial ion and impurity losses into the halo permits gridless, circular direct converters to be dramatically reduced in size. Comparisons of MARS with the Starfire tokamak design are made.

  19. Neutrino Oscillation Studies with Reactors

    E-print Network

    Vogel, Petr; Zhang, Chao

    2015-01-01

    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. Neutrino Oscillation Studies with Reactors

    E-print Network

    Petr Vogel; Liangjian Wen; Chao Zhang

    2015-04-27

    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.

  1. Neutrino oscillation studies with reactors

    NASA Astrophysics Data System (ADS)

    Vogel, P.; Wen, L. J.; Zhang, C.

    2015-04-01

    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.

  2. Neutrino oscillation studies with reactors

    PubMed Central

    Vogel, P.; Wen, L.J.; Zhang, C.

    2015-01-01

    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. PMID:25913819

  3. Rotating reactor studies

    NASA Technical Reports Server (NTRS)

    Roberts, Glyn O.

    1991-01-01

    Undesired gravitational effects such as convection or sedimentation in a fluid can sometimes be avoided or decreased by the use of a closed chamber uniformly rotated about a horizontal axis. In a previous study, the spiral orbits of a heavy or buoyant particle in a uniformly rotating fluid were determined. The particles move in circles, and spiral in or out under the combined effects of the centrifugal force and centrifugal buoyancy. A optimization problem for the rotation rate of a cylindrical reactor rotated about its axis and containing distributed particles was formulated and solved. Related studies in several areas are addressed. A computer program based on the analysis was upgraded by correcting some minor errors, adding a sophisticated screen-and-printer graphics capability and other output options, and by improving the automation. The design, performance, and analysis of a series of experiments with monodisperse polystyrene latex microspheres in water were supported to test the theory and its limitations. The theory was amply confirmed at high rotation rates. However, at low rotation rates (1 rpm or less) the assumption of uniform solid-body rotation of the fluid became invalid, and there were increasingly strong secondary motions driven by variations in the mean fluid density due to variations in the particle concentration. In these tests the increase in the mean fluid density due to the particles was of order 0.015 percent. To a first approximation, these flows are driven by the buoyancy in a thin crescent-shaped depleted layer on the descending side of the rotating reactor. This buoyancy distribution is balanced by viscosity near the walls, and by the Coriolis force in the interior. A full analysis is beyond the scope of this study. Secondary flows are likely to be stronger for buoyant particles, which spiral in towards the neutral point near the rotation axis under the influence of their centrifugal buoyancy. This is because the depleted layer is thicker and extends all the way around the reactor.

  4. Transport Reactor Facility

    SciTech Connect

    Berry, D.A.; Shoemaker, S.A.

    1996-12-31

    The Morgantown Energy Technology Center (METC) is currently evaluating hot gas desulfurization (HGD)in its on-site transport reactor facility (TRF). This facility was originally constructed in the early 1980s to explore advanced gasification processes with an entrained reactor, and has recently been modified to incorporate a transport riser reactor. The TRF supports Integrated Gasification Combined Cycle (IGCC) power systems, one of METC`s advanced power generation systems. The HGD subsystem is a key developmental item in reducing the cost and increasing the efficiency of the IGCC concept. The TRF is a unique facility with high-temperature, high-pressure, and multiple reactant gas composition capability. The TRF can be configured for reacting a single flow pass of gas and solids using a variety of gases. The gas input system allows six different gas inputs to be mixed and heated before entering the reaction zones. Current configurations allow the use of air, carbon dioxide, carbon monoxide, hydrogen, hydrogen sulfide, methane, nitrogen, oxygen, steam, or any mixture of these gases. Construction plans include the addition of a coal gas input line. This line will bring hot coal gas from the existing Fluidized-Bed Gasifier (FBG) via the Modular Gas Cleanup Rig (MGCR) after filtering out particulates with ceramic candle filters. Solids can be fed either by a rotary pocket feeder or a screw feeder. Particle sizes may range from 70 to 150 micrometers. Both feeders have a hopper that can hold enough solid for fairly lengthy tests at the higher feed rates, thus eliminating the need for lockhopper transfers during operation.

  5. Fast neutron nuclear reactor

    SciTech Connect

    Cabrillat, M. Th.; Lions, N.

    1985-01-08

    The invention relates to a fast neutron nuclear reactor of the integrated type comprising a cylindrical inner vessel. The inner vessel comprises two concentric ferrules and the connection between the hot collector defined within this vessel and the inlet port of the exchangers is brought about by a hot structure forming a heat baffle and supported by the inner ferrule and by a cold structure surrounding the hot structure, supported by the outer ferrule and sealingly connected to the exchanger. Application to the generation of electric power in nuclear power stations.

  6. High flux reactor

    DOEpatents

    Lake, James A. (Idaho Falls, ID); Heath, Russell L. (Idaho Falls, ID); Liebenthal, John L. (Idaho Falls, ID); DeBoisblanc, Deslonde R. (Summit, NJ); Leyse, Carl F. (Idaho Falls, ID); Parsons, Kent (Idaho Falls, ID); Ryskamp, John M. (Idaho Falls, ID); Wadkins, Robert P. (Idaho Falls, ID); Harker, Yale D. (Idaho Falls, ID); Fillmore, Gary N. (Idaho Falls, ID); Oh, Chang H. (Idaho Falls, ID)

    1988-01-01

    A high flux reactor is comprised of a core which is divided into two symetric segments housed in a pressure vessel. The core segments include at least one radial fuel plate. The spacing between the plates functions as a coolant flow channel. The core segments are spaced axially apart such that a coolant mixing plenum is formed between them. A channel is provided such that a portion of the coolant bypasses the first core section and goes directly into the mixing plenum. The outlet coolant from the first core segment is mixed with the bypass coolant resulting in a lower inlet temperature to the lower core segment.

  7. Reactor refueling containment system

    DOEpatents

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

    1995-05-02

    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.

  8. Reactor coolant pump flywheel

    DOEpatents

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

    2013-11-26

    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.

  9. Reactor refueling containment system

    DOEpatents

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

    1995-01-01

    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.

  10. Sequencing Batch Reactors

    Microsoft Academic Search

    Lawrence K. Wang; Yan Li

    \\u000a A sequencing batch reactor (SBR) can be either a biological SBR (BIO-SBR) or a physicochemical SBR (PC-SBR). BIO-SBR includes\\u000a traditional sedimentation biological SBR, innovative flotation biological SBR (BIO-DAF-SBR), innovative membrane biological\\u000a SBR (MBR-SBR), aerobic digestion SBR (AD-SBR), etc. All PC-SBR are innovative processes including at least sedimentation PC-SBR\\u000a (PC-SED-SBR), flotation PC-SBR (PC-DAF-SBR), membrane PC-SBR (PC-membrance-SBR), granular activated carbon PC-SBR (PC-GAC-SBR),

  11. Reactor instrumentation renewal of the TRIGA reactor Vienna, Austria

    Microsoft Academic Search

    H. Boeck; H. Weiss; W. E. Hood; W. K. Hyde

    1992-01-01

    The TRIGA Mark-II reactor at the Atominstitut in Vienna, Austria is replacing its twenty-four year old instrumentation system with a microprocessor based control system supplied by General Atomics. Ageing components, new governmental safety requirements and a need for state of the art instrumentation for training students has spurred the demand for new reactor instrumentation. In Austria a government appointed expert

  12. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1991-04-01

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

  13. Sequencing batch biofilm reactor: from support design to reactor operation

    Microsoft Academic Search

    M. Matos; C. Alves; J. L. Campos; A. G. Brito; R. Nogueira

    2011-01-01

    The aim of this work was to improve the overall understanding of sequencing batch biofilm reactors (SBBRs) from support selection (biofilm formation) to reactor operation (carbon and nitrogen removal). Supports manufactured with different materials and geometries were tested in 2.5 L SBBRs and it was observed that biofilm accumulation was favoured on the supports that presented a higher internal surface

  14. SIMULATED BOILING WATER REACTOR AND SUPERHEAT REACTOR CORROSION FACILITY

    Microsoft Academic Search

    M. D. Fitzsimmons; W. L. Pearl; M. Siegler

    1963-01-01

    A versatile boiling water and superheated steam facility was developed ; for out-of-pile corrosion testing of materials that are being used and considered ; for application in boiling water reactor (BWR) and superheat reactor (SHR) ; systems. The following capabilities were achieved: simulation of the various ; environments (other than irradiation) that would contact materials in a Dresden ; type

  15. Cascade ICF power reactor

    SciTech Connect

    Hogan, W.J.; Pitts, J.H.

    1986-05-20

    The double-cone-shaped Cascade reaction chamber rotates at 50 rpm to keep a blanket of ceramic granules in place against the wall as they slide from the poles to the exit slots at the equator. The 1 m-thick blanket consists of layers of carbon, beryllium oxide, and lithium aluminate granules about 1 mm in diameter. The x rays and debris are stopped in the carbon granules; the neutrons are multiplied and moderated in the BeO and breed tritium in the LiAlO/sub 2/. The chamber wall is made up of SiO tiles held in compression by a network of composite SiC/Al tendons. Cascade operates at a 5 Hz pulse rate with 300 MJ in each pulse. The temperature in the blanket reaches 1600 K on the inner surface and 1350 K at the outer edge. The granules are automatically thrown into three separate vacuum heat exchangers where they give up their energy to high pressure helium. The helium is used in a Brayton cycle to obtain a thermal-to-electric conversion efficiency of 55%. Studies have been done on neutron activation, debris recovery, vaporization and recondensation of blanket material, tritium control and recovery, fire safety, and cost. These studies indicate that Cascade appears to be a promising ICF reactor candidate from all standpoints. At the 1000 MWe size, electricity could be made for about the same cost as in a future fission reactor.

  16. Solvent refined coal reactor quench system

    DOEpatents

    Thorogood, R.M.

    1983-11-08

    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.

  17. Aerosol reactor for silicon processing

    NASA Technical Reports Server (NTRS)

    Flagan, R. C.; Alam, M. K.; Johnson, B. E.; Wu, J. J.

    1984-01-01

    Silane pyrolysis, homogeneous nucleation; and an aerosol reactor for the growth of large silicon particles by silane pyrolysis are discussed. Graphs show data on temperature profile and reaction kinetics in the aerosol reactor, as well as particle growth, total clearance volume fraction, mass distribution of silicon aerosol in high temperatures, and silicon losses as a function of input silane concentration.

  18. DR Reactor VSR channel damage

    Microsoft Academic Search

    F. J. Kempf; J. K. Rawlins

    1961-01-01

    On July 11, 1961 the Ball 3X System at DR Reactor was inadventently tripped. All vertical safety rods dropped and all channels were filled with balls. This report has the twofold purpose of documenting borescope observations of ten vertical rod channels at DR Reactor and recording the estimated extent of graphite damage resulting from the above incident. Channel damage data

  19. India and Fast Breeder Reactors

    Microsoft Academic Search

    M. V. RAMANA

    2009-01-01

    India has long pursued a fast breeder program, motivated in part by the availability of only poor quality uranium resources within the country. But progress so far has been disappointing, with only one test reactor having been constructed and having a chequered operating history. The larger Prototype Fast Breeder Reactor that is being constructed has a design that compromises safety

  20. Proton Collimators for Fusion Reactors

    NASA Technical Reports Server (NTRS)

    Miley, George H.; Momota, Hiromu

    2003-01-01

    Proton collimators have been proposed for incorporation into inertial-electrostatic-confinement (IEC) fusion reactors. Such reactors have been envisioned as thrusters and sources of electric power for spacecraft and as sources of energetic protons in commercial ion-beam applications.

  1. Prospects for spheromak fusion reactors

    Microsoft Academic Search

    T. K. Fowler; D. D. Hua

    1995-01-01

    The reactor study of Hagenson and Krakowski demonstrated the attractiveness of the spheromak as a compact fusion reactor, based on physics principles confirmed in CTX experiments in many respects. Most uncertain was the energy confinement time and the role of magnetic turbulence inherent in the concept. In this paper, a one-dimensional model of heat confinement, calibrated by CTX, predicts negligible

  2. Existing reactor expansion study basis

    Microsoft Academic Search

    Heacock

    1959-01-01

    The latest HAPO Five Year Program review, HW-59633, forecasts substantial increases in Pu production from the eight existing Hanford reactors over the next several years. These production increases would be attained by a combination of several methods which include increased reactor power levels resulting from higher process water flow rates and coolant bulk outlet temperatures, improved time operated efficiency, higher

  3. Existing reactor expansion study basis

    Microsoft Academic Search

    Heacock

    1959-01-01

    The latest HAPO Five Year Program review, indicates that significant increases in Pu production from the eight existing Hanford reactors may be achieved. These production increases would be attained by a combination of several methods including increased process water flow rates, reactor coolant outlet temperature, improved time operated efficiency, conversion ratio and induced transient reactivity looses. In order to provide

  4. Radiation Shielding for Fusion Reactors

    SciTech Connect

    Santoro, R.T.

    1999-10-01

    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. Reactor-Produced Medical Radionuclides

    NASA Astrophysics Data System (ADS)

    Mirzadeh, S.; Mausner, L. F.; Garland, M. A.

    The therapeutic use of radionuclides in nuclear medicine, oncology, and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by ?- emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chapter is concluded by providing a number of examples encompassing the various possible reaction routes for the production of a number of medical radionuclides in a reactor.

  6. Fast reactors and nuclear nonproliferation

    SciTech Connect

    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)

    2013-07-01

    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)

  7. Reactor-Produced Medical Radionuclides

    SciTech Connect

    Mirzadeh, Saed [ORNL; Mausner, Leonard [Brookhaven National Laboratory (BNL); Garland, Marc A [ORNL

    2011-01-01

    The therapeutic use of radionuclides in nuclear medicine, oncology and cardiology is the most rapidly growing use of medical radionuclides. Since most therapeutic radionuclides are neutron rich and decay by beta emission, they are reactor-produced. This chapter deals mainly with production approaches with neutrons. Neutron interactions with matter, neutron transmission and activation rates, and neutron spectra of nuclear reactors are discussed in some detail. Further, a short discussion of the neutron-energy dependence of cross sections, reaction rates in thermal reactors, cross section measurements and flux monitoring, and general equations governing the reactor production of radionuclides are presented. Finally, the chapter is concluded by providing a number of examples encompassing the various possible reaction routes for production of a number of medical radionuclides in a reactor.

  8. Advanced Reactors Around the World

    SciTech Connect

    Majumdar, Debu

    2003-09-01

    At the end of 2002, 441 nuclear power plants were operating around the globe and providing 17% of the world's electricity. Although the rate of population growth has slowed, recent United Nations data suggest that two billion more people will be added to the world by 2050. A special report commissioned by the Intergovernmental Panel on Climate Change estimated that electricity demand would grow almost eight-fold from 2000 to 2050 in a high economic grown scenario and more than double in a low-growth scenario. There is also a global aspiration to keep the environment pristine. Because of these reasons, it is expected that a large number of new nuclear reactors may be operating by 2050. Realization of this has created an impetus for the development of a new generation of reactors in several countries. The goal is to make nuclear power cost-competitive with other resources and to enhance safety to a level that no evacuation outside a plant site would be necessary. It should also generate less waste, prevent materials diversion for weapons production, and be sustainable. This article discusses the status of next-generation reactors under development around the world. Specifically highlighted are efforts related to the Generation IV International Forum (GIF) and its six reactor concepts for research and development: Very High Temperature Reactor (VHTR); Gas-Cooled Fast Reactor (GFR); Supercritical Water-Cooled Reactor (SCWR); Sodium-Cooled Fast Reactor (SFR); Lead-Cooled Fast Reactor (LFR); and Molten Salt Reactor (MSR). Also highlighted are nuclear activities specific to Russia and India.

  9. Licensed reactor nuclear safety criteria applicable to DOE reactors

    SciTech Connect

    Not Available

    1993-11-01

    This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards.

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

    SciTech Connect

    Eric C. Woolstenhulme; Dana M. Hewit

    2008-09-01

    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.

  11. University Reactor Sharing Program

    SciTech Connect

    W.D. Reese

    2004-02-24

    Research projects supported by the program include items such as dating geological material and producing high current super conducting magnets. The funding continues to give small colleges and universities the valuable opportunity to use the NSC for teaching courses in nuclear processes; specifically neutron activation analysis and gamma spectroscopy. The Reactor Sharing Program has supported the construction of a Fast Neutron Flux Irradiator for users at New Mexico Institute of Mining and Technology and the University of Houston. This device has been characterized and has been found to have near optimum neutron fluxes for A39/Ar 40 dating. Institution final reports and publications resulting from the use of these funds are on file at the Nuclear Science Center.

  12. Nuclear reactor control

    DOEpatents

    Cawley, William E. (Phoenix, AZ); Warnick, Robert F. (Pasco, WA)

    1982-01-01

    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.

  13. Advances in reactor physics education: Visualization of reactor parameters

    SciTech Connect

    Snoj, L.; Kromar, M.; Zerovnik, G. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2012-07-01

    Modern computer codes allow detailed neutron transport calculations. In combination with advanced 3D visualization software capable of treating large amounts of data in real time they form a powerful tool that can be used as a convenient modern educational tool for reactor operators, nuclear engineers, students and specialists involved in reactor operation and design. Visualization is applicable not only in education and training, but also as a tool for fuel management, core analysis and irradiation planning. The paper treats the visualization of neutron transport in different moderators, neutron flux and power distributions in two nuclear reactors (TRIGA type research reactor and a typical PWR). The distributions are calculated with MCNP and CORD-2 computer codes and presented using Amira software. (authors)

  14. Reactor physics design of supercritical CO?-cooled fast reactors

    E-print Network

    Pope, Michael A. (Michael Alexander)

    2004-01-01

    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 ...

  15. Reactor protection system design alternatives for sodium fast reactors

    E-print Network

    DeWitte, Jacob D. (Jacob Dominic)

    2011-01-01

    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 ...

  16. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    SciTech Connect

    Douglas Morrell

    2011-03-01

    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.

  17. Nuclear reactor downcomer flow deflector

    DOEpatents

    Gilmore, Charles B. (Greensburg, PA); Altman, David A. (Pittsburgh, PA); Singleton, Norman R. (Murrysville, PA)

    2011-02-15

    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.

  18. Fueling of tandem mirror reactors

    SciTech Connect

    Gorker, G.E.; Logan, B.G.

    1985-01-01

    This paper summarizes the fueling requirements for experimental and demonstration tandem mirror reactors (TMRs), reviews the status of conventional pellet injectors, and identifies some candidate accelerators that may be needed for fueling tandem mirror reactors. Characteristics and limitations of three types of accelerators are described; neutral beam injectors, electromagnetic rail guns, and laser beam drivers. Based on these characteristics and limitations, a computer module was developed for the Tandem Mirror Reactor Systems Code (TMRSC) to select the pellet injector/accelerator combination which most nearly satisfies the fueling requirements for a given machine design.

  19. Human Factors Aspects of Operating Small Reactors

    Microsoft Academic Search

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

    2010-01-01

    The nuclear-power community has reached the stage of proposing advanced reactor designs to support power generation for decades to come. They are considering small modular reactors (SMRs) as one approach to meet these energy needs. While the power output of individual reactor modules is relatively small, they can be grouped to produce reactor sites with different outputs. Also, they can

  20. Novel photocatalytic reactor for water purification

    Microsoft Academic Search

    Ajay K. Ray; Antonie A. C. M. Beenackers

    1998-01-01

    A novel photocatalytic reactor design for water treatment is characterized by the use of new extremely narrow diameter lamps, thus allowing for much higher surface area for catalyst coating per unit reactor volume and consequently for much higher specific reactor capacity. Experiments in a reactor containing 21 novel U-shaped lamps coated with catalyst showed a 695% increase in efficiency of

  1. Conceptual design study of JSFR reactor building

    SciTech Connect

    Yamamoto, T.; Katoh, A.; Chikazawa, Y. [Japan Atomic Energy Agency (JAEA), 4002 Narita, Oarai, Ibaraki 311-1393 (Japan); Ohya, T.; Iwasaki, M.; Hara, H.; Akiyama, Y. [Mitsubishi FBR Systems, Inc. MFBR, 34-17, Jingumae 2-chome, Shibuya, Tokyo 150-0001 (Japan)

    2012-07-01

    Japan Sodium-cooled Fast Reactor (JSFR) is planning to adopt the new concepts of reactor building. One is that the steel plate reinforced concrete is adopted for containment vessel and reactor building. The other is the advanced seismic isolation system. This paper describes the detail of new concepts for JSFR reactor building and engineering evaluation of the new concepts. (authors)

  2. Pebble Flow Experiments For Pebble Bed Reactors

    E-print Network

    of Technology 2nd International Topical Meeting on High Temperature Reactor Technology Institute of Nuclear of pebbles in the reactor is well understood. In addition, to increase the core power of pebble bed reactors the core power distributions for pebble bed reactors. This flow model is shown in Figure 1 and has been

  3. Evolution of the Hanford graphite reactor technology

    Microsoft Academic Search

    Toffer

    1989-01-01

    The Hanford Site (Hanford) has played an important part in the development of nuclear energy and reactors in the first 50 yrs. This paper follows the evolution of the graphite-moderated, light-water cooled reactor technology at Hanford through the various generations of reactor types. Nine reactors were built and operated successfully between 1944 and 1987. Design improvements, safety feature advances, and

  4. Reactor core isolation cooling system

    DOEpatents

    Cooke, F.E.

    1992-12-08

    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.

  5. Reactor Cost Analysis Brian James

    E-print Network

    /Linde (fluidized bed) With WGS With or Without WGS H2 Purification PSA Separate Membrane Section Membrane/WGS Membrane Reactor OTM/ Water-Splitting ANL With WGS #12;Directed Technologies, Inc. 6-7 November 2007 BILIWG

  6. Nuclear Reactors and Technology; (USA)

    SciTech Connect

    Cason, D.L.; Hicks, S.C. (eds.)

    1991-01-01

    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.

  7. Combustion synthesis continuous flow reactor

    DOEpatents

    Maupin, G.D.; Chick, L.A.; Kurosky, R.P.

    1998-01-06

    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.

  8. Combustion synthesis continuous flow reactor

    DOEpatents

    Maupin, Gary D. (Richland, WA); Chick, Lawrence A. (West Richland, WA); Kurosky, Randal P. (Maple Valley, WA)

    1998-01-01

    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.

  9. Overview of fast breeder reactors

    Microsoft Academic Search

    Massoud Simnad

    1998-01-01

    A brief overview is presented of the history and status of advanced fast breeder reactors (FBRs). The first eight papers of this special issue, from seven countries where FBRs have been built and operated, are summarized.

  10. Solid State Reactor Final Report

    SciTech Connect

    Mays, G.T.

    2004-03-10

    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.

  11. Fast quench reactor and method

    DOEpatents

    Detering, Brent A.; Donaldson, Alan D.; Fincke, James R.; Kong, Peter C.

    2002-09-24

    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.

  12. Fast quench reactor and method

    DOEpatents

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

    2002-01-01

    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.

  13. Fast quench reactor and method

    DOEpatents

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

    1998-01-01

    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.

  14. Automatic safety rod for reactors

    DOEpatents

    Germer, John H. (San Jose, CA)

    1988-01-01

    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.

  15. Alternate-fuel reactor studies

    SciTech Connect

    Evans, K. Jr.; Ehst, D.A.; Gohar, Y.; Jung, J.; Mattas, R.F.; Turner, L.R.

    1983-02-01

    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.

  16. Directions for improved fusion reactors

    Microsoft Academic Search

    R. A. Krakowski; R. L. Miller; J. G. Delene

    1986-01-01

    Conceptual fusion reactor studies over the past 10 to 15 years have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points towards smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion

  17. Microchannel Reactors for ISRU Applications

    NASA Astrophysics Data System (ADS)

    Carranza, Susana; Makel, Darby B.; Blizman, Brandon; Ward, Benjamin J.

    2005-02-01

    Affordable planning and execution of prolonged manned space missions depend upon the utilization of local resources and the waste products which are formed in manned spacecraft and surface bases. Successful in-situ resources utilization (ISRU) will require component technologies which provide optimal size, weight, volume, and power efficiency. Microchannel reactors enable the efficient chemical processing of in situ resources. The reactors can be designed for the processes that generate the most benefit for each mission. For instance, propellants (methane) can be produced from carbon dioxide from the Mars atmosphere using the Sabatier reaction and ethylene can be produced from the partial oxidation of methane. A system that synthesizes ethylene could be the precursor for systems to synthesize ethanol and polyethylene. Ethanol can be used as a nutrient for Astrobiology experiments, as well as the production of nutrients for human crew (e.g. sugars). Polyethylene can be used in the construction of habitats, tools, and replacement parts. This paper will present recent developments in miniature chemical reactors using advanced Micro Electro Mechanical Systems (MEMS) and microchannel technology to support ISRU of Mars and lunar missions. Among other applications, the technology has been demonstrated for the Sabatier process and for the partial oxidation of methane. Microchannel reactors were developed based on ceramic substrates as well as metal substrates. In both types of reactors, multiple layers coated with catalytic material are bonded, forming a monolithic structure. Such reactors are readily scalable with the incorporation of extra layers. In addition, this reactor structure minimizes pressure drop and catalyst settling, which are common problems in conventional packed bed reactors.

  18. Biological sludge stabilization reactor evaluations

    SciTech Connect

    Corbitt, R.A.; Bowen, P.T.; Smith, P.E.

    1998-07-01

    Anaerobic digestion was chosen as the means to stabilize primary and thickened waste activated sludge for a 0.88 m{sup 3}/s (20 mgd) advanced wastewater reclamation facility. Two stage digestion was proposed to produce Class B sludge. Reactor shape was an important variable in design of the first stage digestion. Evaluation of conventional and egg shaped anaerobic digesters was performed. Based on the economic and non-economic criteria analysis, egg shaped reactors were selected.

  19. MARS: Mirror Advanced Reactor Study

    SciTech Connect

    Logan, B.G.

    1984-09-10

    A recently completed two-year study of a commercial tandem mirror reactor design (Mirror Advanced Reactor Study (MARS)) is briefly reviewed. The end plugs are designed for trapped particle stability, MHD ballooning, balanced geodesic curvature, and small radial electric fields in the central cell. New technologies such as lithium-lead blankets, 24T hybrid coils, gridless direct converters and plasma halo vacuum pumps are highlighted.

  20. When Do Commercial Reactors Permanently Shut Down?

    EIA Publications

    2011-01-01

    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.

  1. National strategies for nuclear power reactor development

    Microsoft Academic Search

    R. K. Lester; M. J. Driscoll; M. W. Golay; D. D. Lanning; L. M. Lidsky

    1985-01-01

    The document assesses the potential for design innovation in three areas of nuclear power plant technology: light-water reactor systems; liquid-metal reactor systems; and high-temperature-gas reactor systems. The question of how capital costs scale with unit size in nuclear power reactor systems is addressed. Small modular high temperature gas reactor designs are reviewed, and an electric-power-system capacity-planning model that allows estimates

  2. Propellant actuated nuclear reactor steam depressurization valve

    DOEpatents

    Ehrke, Alan C. (San Jose, CA); Knepp, John B. (San Jose, CA); Skoda, George I. (Santa Clara, CA)

    1992-01-01

    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.

  3. Power supply with nuclear reactor

    SciTech Connect

    Cook, B.M.

    1984-02-28

    Each parameter of the processes of the nuclear reactor and of the components of a power supply which convert the thermal energy generated by the reactor into electrical power is monitored by a set of four like sensors. One each of the unlike sensors which monitor the different parameters is contained in a reactor-trip logic channel. Each such unlike sensor is referred to here as a ''local sensor''. Each channel is interlocked with the other three channels and receives the signals sensed by the other three sensors, herein called ''remote sensors''. Each channel also includes means for processing the signals from the local and remote sensors. The apparatus also includes means for tripping the reactor to deenergize or trip the control rod drive and insert the control rods fully into the core so that the reactor stops supplying power. The apparatus normally operates on a ''two out of four'' configuration. This assumes that all sensors are in normal operating condition. To achieve this purpose, eight circuit breakers are provided. Two breaker contactors are controlled by each channel. The control is through the undervoltage relays of the respective breakers. To trip the reactor, it is necessary that at least four breaker contactors be opened by operation of two channels.

  4. Reactor pressure vessel nozzle

    SciTech Connect

    Challberg, Roy C. (Livermore, CA); Upton, Hubert A. (Morgan Hill, CA)

    1994-01-01

    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.

  5. Reactor pressure vessel nozzle

    DOEpatents

    Challberg, R.C.; Upton, H.A.

    1994-10-04

    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.

  6. Solar solids reactor

    DOEpatents

    Yudow, B.D.

    1986-02-24

    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.

  7. Solar solids reactor

    DOEpatents

    Yudow, Bernard D. (Chicago, IL)

    1987-01-01

    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.

  8. Reactor Simulator Testing

    NASA Technical Reports Server (NTRS)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise J..

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz.

  9. The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

    SciTech Connect

    Seifritz, W.

    1983-11-01

    A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase.

  10. Reactor operations: Brookhaven Medical Research Reactor, Brookhaven High Flux Beam Reactor. Informal report, July 1995

    SciTech Connect

    NONE

    1995-07-01

    Part one of this report gives the operating history for the Brookhaven Medical Research Reactor for the month of July. Also included are the BMRR technical safety surveillance requirements record and the summary of BMRR irradiations for the month. Part two gives the operating histories for the Brookhaven High Flux Beam Reactor and the Cold Neutron Source Facility for the month of July. Also included are the HFBR technical safety surveillance requirements record and the summary of HFBR irradiations for the month.

  11. Method of Experimental Power Reactor Noise Analysis without Use of High Power Reactor

    Microsoft Academic Search

    Hideki KODAIRA; Hidetoshi SHIMOTONO; Shunsuke KONDO; Yasumasa TOGO

    1977-01-01

    A method is proposed for conducting power reactor noise analysis without recourse to an actual high power reactor. The basic concept is to simulate the power reactor noise by integrating the different elements constituting the actual reactor noise, such as the random noise-generating force, the zero-power reactor transfer function, and feed- back loops between the reactor power and the noise-generating

  12. Reactor Simulator Testing

    NASA Technical Reports Server (NTRS)

    Schoenfeld, Michael P.; Webster, Kenny L.; Pearson, Boise Jon

    2013-01-01

    As part of the Nuclear Systems Office Fission Surface Power Technology Demonstration Unit (TDU) project, a reactor simulator test loop (RxSim) was design & built to perform integrated testing of the TDU components. In particular, the objectives of RxSim testing was to verify the operation of the core simulator, the instrumentation and control system, and the ground support gas and vacuum test equipment. In addition, it was decided to include a thermal test of a cold trap purification design and a pump performance test at pump voltages up to 150 V since the targeted mass flow rate of 1.75 kg/s was not obtained in the RxSim at the originally constrained voltage of 120 V. This paper summarizes RxSim testing. The gas and vacuum ground support test equipment performed effectively in NaK fill, loop pressurization, and NaK drain operations. The instrumentation and control system effectively controlled loop temperature and flow rates or pump voltage to targeted settings. The cold trap design was able to obtain the targeted cold temperature of 480 K. An outlet temperature of 636 K was obtained which was lower than the predicted 750 K but 156 K higher than the cold temperature indicating the design provided some heat regeneration. The annular linear induction pump (ALIP) tested was able to produce a maximum flow rate of 1.53 kg/s at 800 K when operated at 150 V and 53 Hz. Keywords: fission, space power, nuclear, liquid metal, NaK.

  13. UCLA program in reactor studies: The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1991-01-01

    The ARIES research program is a multi-institutional effort to develop several visions of tokamak reactors with enhanced economic, safety, and environmental features. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Four ARIES visions are currently planned for the ARIES program. The ARIES-1 design is a DT-burning reactor based on modest'' extrapolations from the present tokamak physics database and relies on either existing technology or technology for which trends are already in place, often in programs outside fusion. ARIES-2 and ARIES-4 are DT-burning reactors which will employ potential advances in physics. The ARIES-2 and ARIES-4 designs employ the same plasma core but have two distinct fusion power core designs; ARIES-2 utilize the lithium as the coolant and breeder and vanadium alloys as the structural material while ARIES-4 utilizes helium is the coolant, solid tritium breeders, and SiC composite as the structural material. Lastly, the ARIES-3 is a conceptual D-{sup 3}He reactor. During the period Dec. 1, 1990 to Nov. 31, 1991, most of the ARIES activity has been directed toward completing the technical work for the ARIES-3 design and documenting the results and findings. We have also completed the documentation for the ARIES-1 design and presented the results in various meetings and conferences. During the last quarter, we have initiated the scoping phase for ARIES-2 and ARIES-4 designs.

  14. Simplifying microbial electrosynthesis reactor design.

    PubMed

    Giddings, Cloelle G S; Nevin, Kelly P; Woodward, Trevor; Lovley, Derek R; Butler, Caitlyn S

    2015-01-01

    Microbial electrosynthesis, an artificial form of photosynthesis, can efficiently convert carbon dioxide into organic commodities; however, this process has only previously been demonstrated in reactors that have features likely to be a barrier to scale-up. Therefore, the possibility of simplifying reactor design by both eliminating potentiostatic control of the cathode and removing the membrane separating the anode and cathode was investigated with biofilms of Sporomusa ovata. S. ovata reduces carbon dioxide to acetate and acts as the microbial catalyst for plain graphite stick cathodes as the electron donor. In traditional 'H-cell' reactors, where the anode and cathode chambers were separated with a proton-selective membrane, the rates and columbic efficiencies of microbial electrosynthesis remained high when electron delivery at the cathode was powered with a direct current power source rather than with a potentiostat-poised cathode utilized in previous studies. A membrane-less reactor with a direct-current power source with the cathode and anode positioned to avoid oxygen exposure at the cathode, retained high rates of acetate production as well as high columbic and energetic efficiencies. The finding that microbial electrosynthesis is feasible without a membrane separating the anode from the cathode, coupled with a direct current power source supplying the energy for electron delivery, is expected to greatly simplify future reactor design and lower construction costs. PMID:26029199

  15. Heat-generating nuclear reactor

    SciTech Connect

    Dupuy, G.; Fajeau, M.; Labrousse, M.; Lerouge, B.; Minguet, J.

    1981-01-20

    A reactor vessel filled with coolant fluid is divided by a wall into an upper region and a lower region which contains the reactor core, part of the coolant fluid in the upper region being injected into the lower region. The injection flow rate is regulated as a function of the variations in pressure in the lower region by means of a baffle-plate container which communicates with a leak-tight chamber and with a storage reservoir, a flow of fluid from the chamber to the reservoir being established only at the time of a reduction in the rate of injection into the container. The reactor can be employed for the production of hot water which is passed through a heat exchanger and supplied to a heating installation.

  16. Reactor control rod timing system

    DOEpatents

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

    1982-01-01

    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.

  17. Integral fast reactor safety features

    SciTech Connect

    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.

    1988-01-01

    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.

  18. Gaseous fuel nuclear reactor research

    NASA Technical Reports Server (NTRS)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  19. On fast reactor kinetics studies

    SciTech Connect

    Seleznev, E. F.; Belov, A. A. [Nuclear Safety Inst. of the Russian Academy of Sciences IBRAE (Russian Federation); Matveenko, I. P.; Zhukov, A. M.; Raskach, K. F. [Inst. for Physics and Power Engineering IPPE (Russian Federation)

    2012-07-01

    The results and the program of fast reactor core time and space kinetics experiments performed and planned to be performed at the IPPE critical facility is presented. The TIMER code was taken as computation support of the experimental work, which allows transient equations to be solved in 3-D geometry with multi-group diffusion approximation. The number of delayed neutron groups varies from 6 to 8. The code implements the solution of both transient neutron transfer problems: a direct one, where neutron flux density and its derivatives, such as reactor power, etc, are determined at each time step, and an inverse one for the point kinetics equation form, where such a parameter as reactivity is determined with a well-known reactor power time variation function. (authors)

  20. The Very High Temperature Reactor

    SciTech Connect

    Hans D. Gougar; David A. Petti

    2011-06-01

    The High Temperature Reactor (HTR) and Very High Temperature Reactor (VHTR) are types of nuclear power plants that, as the names imply, operate at temperatures above those of the conventional nuclear power plants that currently generate electricity in the US and other countries. Like existing nuclear plants, heat generated from the fission of uranium or plutonium atoms is carried off by a working fluid and can be used generate electricity. The very hot working fluid also enables the VHTR to drive other industrial processes that require high temperatures not achievable by conventional nuclear plants (Figure 1). For this reason, the VHTR is being considered for non-electrical energy applications. The reactor and power conversion system are constructed using special materials that make a core meltdown virtually impossible.

  1. Space reactor preliminary mechanical design

    SciTech Connect

    Meier, K.L.

    1983-01-01

    An analysis was performed on the SABRE reactor space power system to determine the effect of the number and size of heat pipes on the design parameters of the nuclear subsystem. Small numbers of thin walled heat pipes were found to give a lower subsystem mass, but excessive fuel swelling resulted. The SP-100 preliminary design uses 120 heat pipes because of acceptable fuel swelling and a minimum nuclear subsystem mass of 1875 kg. Salient features of the reactor preliminary design are: individual fuel modules, ZrO/sub 2/ block core mounts, bolted collar fuel module restraints, and a BeO central plug.

  2. (International Thermonuclear Experimental Reactor support)

    SciTech Connect

    Dean, S.O.

    1990-10-15

    This report summarizes the activities under LLNL Purchase Order B089367, the purpose of which is to support the University/Lawrence Livermore National Laboratory Magnetic Fusion Program by evaluating the status of research relative to other national and international programs and assist in long-range plans and development strategies for magnetic fusion in general and for ITER in particular.'' Two specific subtasks are included: to review the LLNL Magnet Technology Development Program in the context of the International Thermonuclear Experimental Reactor Design Study'' and to assist LLNL to organize and prepare materials for an International Thermonuclear Experimental Reactor Design Study information meeting.''

  3. SP-100 space reactor safety

    SciTech Connect

    Not Available

    1987-05-01

    The SP-100 space reactor power system is being developed to meet the large electrical power requirements of civilian and military missions planned for the 1990's and beyond. It will remove the restrictions on electrical power generation that have tended to limit missions and will enable the fuller exploration and utilization of space. This booklet describes the SP-100 space reactor power system and its development. Particular emphasis is given to safety. The design aand operational features as well as the design and safety review process that will assure that the SP-100 can be launched nd operated safely are described.

  4. Actinide Burning in CANDU Reactors

    SciTech Connect

    Hyland, B.; Dyck, G.R. [Atomic Energy of Canada Limited, Chalk River, Ontario, K0J 1J0 (Canada)

    2007-07-01

    Actinide burning in CANDU reactors has been studied as a method of reducing the actinide content of spent nuclear fuel from light water reactors, and thereby decreasing the associated long term decay heat load. In this work simulations were performed of actinides mixed with natural uranium to form a mixed oxide (MOX) fuel, and also mixed with silicon carbide to form an inert matrix (IMF) fuel. Both of these fuels were taken to a higher burnup than has previously been studied. The total transuranic element destruction calculated was 40% for the MOX fuel and 71% for the IMF. (authors)

  5. EPRI reactor analysis support package

    SciTech Connect

    Agee, L.J.; Rose, S.C.

    1986-01-01

    The Electric Power Research Institute's Reactor Analysis Support Package (RASP) project was initiated to provide utilities with computer programs and analysis guidelines for describing nuclear power plant behavior and performance. The objective of the RASP project is to produce an integrated, interlinked, and validated code package (including analysis guidelines) that can be used by utilities for both fuel reload and plant safety analysis. This paper briefly describes the RASP project and its status for developing an analysis methodology that utilities can use to support their own integrated reactor analysis capabilities.

  6. From CANDLE reactor to pebble-bed reactor

    SciTech Connect

    Chen, X. N.; Maschek, W. [Inst. for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe, P.O.B. 3640, D-76021 Karlsruhe (Germany)

    2006-07-01

    This paper attempts to reveal theoretically, by studying a diffusion-burn-up coupled neutronic model, that a so-called CANDLE reactor and a pebble-bed type reactor have a common burn-up feature. As already known, a solitary burn-up wave that can develop in the common U-Pu and Th-U conversion processes is the basic mechanism of the CANDLE reactor. In this paper it is demonstrated that a family of burn-up wave solution exists in the boundary value problem characterizing a pebble bed reactor, in which the fuel is loaded from above into the core and unloaded from bottom. Among this solution family there is a particular case, namely, a partial solitary wave solution, which begins from the fuel entrance side and extends into infinity on the exit side, and has a maximal bum-up rate in this family. An example dealing with the {sup 232}Th-{sup 233}U conversion chain is studied and the solutions are presented in order to show the mechanism of the burn-up wave. (authors)

  7. Improved reactor trip breaker reliability for pressurized water reactors

    Microsoft Academic Search

    J. R. Redmon; D. J. Borchart

    1985-01-01

    Reactor trip breaker (RTB) reliability has been a major concern of the industry and the US Nuclear Regulatory Commission because of the incident with Westinghouse RTBs at Salem Unit I. A similar problem occurred at San Onofre Units 2 and 3 with General Electric (GE) RTBs. An investigation concluded that the reliability of GE AK-2-25 low voltage breakers used as

  8. Reactor Operations informal monthly report September 1994

    SciTech Connect

    Junker, L.

    1994-09-01

    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.

  9. Reactivity control assembly for nuclear reactor

    DOEpatents

    Bollinger, Lawrence R. (Schenectady, NY)

    1984-01-01

    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.

  10. Nuclear Reactor Safety: a current awareness bulletin

    Microsoft Academic Search

    1985-01-01

    Nuclear Reactor Safety announces on a semimonthly basis the current worldwide information available on all safety-related aspects of fission reactors, including: accident analysis, safety systems, radiation protection, decommissioning and dismantling, and security measures.

  11. Microfluidic reactors for the synthesis of nanocrystals

    E-print Network

    Yen, Brian K. H

    2007-01-01

    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 ...

  12. RACEWAY REACTOR FOR MICROALGAL BIODIESEL PRODUCTION

    EPA Science Inventory

    The proposed mathematical model incorporating mass transfer, hydraulics, carbonate/aquatic chemistry, biokinetics, biology and reactor design will be calibrated and validated using the data to be generated from the experiments. The practical feasibility of the proposed reactor...

  13. Performance of the PRISM (power reactor - innovative, small module) reactor's passive decay heat removal system

    Microsoft Academic Search

    P. M. Magee; A. Hunsbedt

    1989-01-01

    The PRISM (power reactor - innovative, small module) modular reactor concept has a totally passive decay heat removal system referred to as the reactor vessel auxiliary cooling system (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling

  14. Reactor Coolant Pump seal issues and their applicability to new reactor designs

    Microsoft Academic Search

    C. J. Ruger; J. C. Higgins

    1993-01-01

    Reactor Coolant Pumps (RCPs) of various types are used to circulate the primary coolant through the reactor in most reactor designs. RCPs generally contain mechanical seals to limit the leakage of pressurized reactor coolant along the pump drive shaft into the containment. The relatively large number of RCP seal and seal auxiliary system failures experienced at US operating plants during

  15. Innovative fast breeder reactor concept ‘RAPID’ for improvement of reactor performance and proliferation resistance

    Microsoft Academic Search

    Mitsuru Kambe; Masaki Uotani

    1997-01-01

    The 60 MWe metal fueled fast breeder reactor concept ‘RAPID’ to improve reactor performance and proliferation resistance has been demonstrated. The reactor can be operated without refueling for up to 5 years. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly (IFA) instead of conventional fuel subassemblies. RAPID concept enables quick and

  16. REACTOR: An Expert System for Diagnosis and Treatment of Nuclear Reactor Accidents

    Microsoft Academic Search

    William R. Nelson

    1982-01-01

    REACTOR is an expert system under development at EG&G Idaho, Inc., that will assist operators in the diagnosis and treatment of nuclear reactor accidents. This paper covers the background of the nuclear industry and why expert system tech- nology may prove valuable in the reactor control room. Some of the basic features of the REACTOR system are discussed, and future

  17. Evaluation of Impurity Behavior in the Reactor Water of a Boiling Water Reactor

    Microsoft Academic Search

    Hiroo Igarashi; Nobuyuki Ohta; Naoshi Usui; Katsumi Ohsumi; Shunsuke Uchida; Tsuneo Matsui

    2001-01-01

    A system has been developed for evaluating the effect of impurity intrusion on reactor water as part of the chemistry diagnostic system for boiling water reactor water. Past records of anomalous changes noted in reactor water quality were first surveyed to identify the substances presenting the highest probability of intrusion into reactor water. The ions and their concentrations were derived,

  18. Integrated modular water reactor (IMR) design

    Microsoft Academic Search

    Koki Hibi; Hitoi Ono; Takashi Kanagawa

    2004-01-01

    Integrated modular water reactor (IMR) has been developed as one of the advanced small-scale light water reactors, with a thermal output of 1000MW. The IMR adopts natural circulation and self-pressurization in the primary cooling system, and a reactor vessel built-in steam generators. The core design has been performed using the current light water reactor technology. Thermal-hydraulic sensitivity analyses have been

  19. Liquid metal cooled nuclear reactor plant system

    DOEpatents

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

    1993-01-01

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

  20. Oxidative dehydrogenation of butane using membrane reactors

    Microsoft Academic Search

    C. Tellez; M. Menendez; J. Santamaria

    1997-01-01

    The oxidative dehydrogenation of butane was studied at temperatures between 450 and 550 C using a conventional fixed-bed reactor and inert membrane catalytic reactors with different feed arrangements. When inert membrane reactors were employed, a ceramic membrane was used to distribute oxygen to a fixed bed of V-Mg-O catalyst. The membrane reactor was found to be more efficient than a

  1. Fission energy: The integral fast reactor

    SciTech Connect

    Chang, Yoon I.

    1989-01-01

    The Integral Fast Reactor (IFR) is an innovative reactor concept being developed at Argonne National Laboratory as a such next- generation reactor concept. The IFR concept has a number of specific technical advantages that collectively address the potential difficulties facing the expansion of nuclear power deployment. In particular, the IFR concept can meet all three fundamental requirements needed in a next-generation reactor as discussed below. This document discusses these requirements.

  2. Transmutation of actinides in power reactors.

    PubMed

    Bergelson, B R; Gerasimov, A S; Tikhomirov, G V

    2005-01-01

    Power reactors can be used for partial short-term transmutation of radwaste. This transmutation is beneficial in terms of subsequent storage conditions for spent fuel in long-term storage facilities. CANDU-type reactors can transmute the main minor actinides from two or three reactors of the VVER-1000 type. A VVER-1000-type reactor can operate in a self-service mode with transmutation of its own actinides. PMID:16604724

  3. Digital computer operation of a nuclear reactor

    DOEpatents

    Colley, R.W.

    1982-06-29

    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.

  4. Flexible conversion ratio fast reactors: Overview

    Microsoft Academic Search

    Neil E. Todreas; Pavel Hejzlar; Anna Nikiforova; Robert Petroski; Eugene Shwageraus; C. J. Fong; Michael J. Driscoll; M. A. Elliott; George Apostolakis

    2009-01-01

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. The performance achievable by the unity conversion ratio cores of these reactors was compared to an existing supercritical carbon dioxide-cooled (S-CO2) fast reactor design and an uprated version of an existing sodium-cooled fast reactor. All concepts have cores rated at 2400MWt. The cores of the liquid-cooled

  5. The modular high-temperature reactor

    Microsoft Academic Search

    H. Reutler; G. H. Lohnert

    1983-01-01

    Nearly all problems encountered in large high-temperature reactor power plants with respect to design and safety are related to the mere physical size of a larger reactor core. Our analyses show that it is feasible to subdivide a larger reactor core into modular units, analogous to the common practice of using several smaller units instead of one large unit. In

  6. Advanced Safeguards Approaches for New Fast Reactors

    Microsoft Academic Search

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

    2007-01-01

    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

  7. CLOSED-CYCLE WATER-BOILER REACTOR

    Microsoft Academic Search

    1962-01-01

    The design and characteristics of a liquid homogeneous reactor ; providing ultimate protection of the surrounding area against radioactive ; contamination are described. The reactor system is comprised of a water boiler ; reactor having a water solution of enriched uranyl nitrate for a core, hydrogen-; oxygen recombiner system, nitrogen-oxygen recombiner system, sealed closed cycle ; gas recirculating system operating

  8. D-D Tokamak reactor studies

    Microsoft Academic Search

    K. E. Evans Jr.; C. C. Baker; J. N. Brooks; D. A. Ehst; P. A. Finn; J. Jung; R. F. Mattas; B. Misra; D. L. Smith; H. C. Stevens

    1980-01-01

    A Tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Plasma engineering, first-wall\\/blanket\\/shield design, magnet design, and tritium\\/fuel\\/vacuum requirements are discussed. Conclusions concerning D-D Tokamak reactors are stated.

  9. Photocatalytic reactors: design for effective air purification

    Microsoft Academic Search

    Michael Birnie; Saffa Riffat; Mark Gillott

    2006-01-01

    This report is intended to review existing photocatalytic reactors as used in research and commercial applications. The guiding principles of effective reactor design are discussed, stressing the challenges faced by design teams. An account of specific reactor designs is given, highlighting innovative work and novel concepts. A concerted effort has been made to draw information from a wide variety of

  10. Search for other natural fission reactors

    Microsoft Academic Search

    K. E. Apt; J. P. Balagna; E. A. Bryant; G. A. Cowan; W. R. Daniels; R. J. Vidale

    1977-01-01

    Precambrian uranium ores have been surveyed for evidence of other natural fission reactors. The requirements for formation of a natural reactor direct investigations to uranium deposits with large, high-grade ore zones. Massive zones with volumes approximately greater than 1 m³ and concentrations approximately greater than 20 percent uranium are likely places for a fossil reactor if they are approximately greater

  11. Advances by the Integral Fast Reactor Program

    SciTech Connect

    Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

    1991-01-01

    The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs.

  12. Interdisciplinary Institute for Innovation Nuclear reactors' construction

    E-print Network

    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

  13. Irradiation Facilities at the Advanced Test Reactor

    SciTech Connect

    S. Blaine Grover

    2005-12-01

    The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC – formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world’s data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities1. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens.

  14. CONFERENCES AND SYMPOSIA FUSION REACTOR DESIGN IV

    E-print Network

    Abdou, Mohamed

    Reactor Design and Technology at Yalta, USSR, from 26 May -- 6 June 1986. This report contains all of fusion power reactor development is to bring to the world a new source of unlimited energy. While reactor experiments, design and technology -- identify the areas in which further work needs to be done

  15. Pebble Flow Experiments For Pebble Bed Reactors

    E-print Network

    Bazant, Martin Z.

    of Technology 2nd International Topical Meeting on High Temperature Reactor Technology Institute of Nuclear Student 2nd International Topical Meeting on HIGH TEMPERATURE REACTOR TECHNOLOGY Beijing, CHINA, September power of pebble bed reactors, proposals for the inclusion of a dynamic (graphite pebble) central column

  16. Figure of merit of magnetic fusion reactor

    Microsoft Academic Search

    Kazawa

    1986-01-01

    This paper proposes and defines ''the figure of merit of fusion reactor'' F which can indicate the economics of fusion reactor comparatively simply and easily by using every important evaluation factor. An attempt to express it synthetically in a numerical formula as to magnetic confined fusion reactor is made. By means of this evaluation formula F, some evaluation studies are

  17. Studies of conceptual spheromak fusion reactors

    Microsoft Academic Search

    M. Katsurai; M. Yamada

    1982-01-01

    Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain the characteristic relations among various parameters of the spheromak configuration for an aspect ratio of A ? 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) the DT reactor with two-component-type

  18. Fusion reactors as a future energy source

    Microsoft Academic Search

    W. Seifritz

    1975-01-01

    A detailed update of fusion research concepts is given. Discussions are given for the following areas: (1) the magnetic confinement principle, (2) UWMAK I: conceptual design for a fusion reactor, (3) the inertial confinement principle, (4) the laser fusion power plant, (5) electron-induced fusion, (6) the long-term development potential of fusion reactors, (7) the symbiosis between fusion and fission reactors,

  19. SWR 1000: The Innovative Boiling Water Reactor

    Microsoft Academic Search

    Werner Brettschuh; Greg Hudson

    2004-01-01

    Framatome ANP has developed the boiling water reactor SWR 1000 in close cooperation with German nuclear utilities and with support from various European partners. This advanced reactor design marks a new era in the successful tradition of boiling water reactor technology and, with a gross electric output of between 1290 and 1330 MW, is aimed at assuring competitive power generating

  20. Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina)

    E-print Network

    Gratta, Giorgio

    Nuclear reactor safeguards and monitoring with antineutrino detectors A. Bernsteina) Sandia of nuclear reactor types, including power reactors, research reactors, and plutonium production reactors-understood principles that govern the core's evolution in time, can be used to determine whether the reactor is being

  1. Computer aided nuclear reactor modeling 

    E-print Network

    Warraich, Khalid Sarwar

    1995-01-01

    Nuclear reactor modeling is an important activity that lets us analyze existing as well as proposed systems for safety, correct operation, etc. The quality of a analysis is directly proportional to the quality of the model used. In this work we look...

  2. Nozzle for electric dispersion reactor

    DOEpatents

    Sisson, W.G.; Harris, M.T.; Scott, T.C.; Basaran, O.A.

    1998-06-02

    A nozzle for an electric dispersion reactor includes two coaxial cylindrical bodies, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 5 figs.

  3. Nozzle for electric dispersion reactor

    DOEpatents

    Sisson, W.G.; Basaran, O.A.; Harris, M.T.

    1998-04-14

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

  4. PLUTONIUM TECHNOLOGY FOR REACTOR SYSTEMS

    Microsoft Academic Search

    M. B. Waldron; A. G. Adwick; H. Lloyd; M. J. Notley; D. M. Poole; L. E. Russell; J. B. Sayers

    1959-01-01

    The possibility exists of using plutonium in both thermal and fast ; reactors as solid metal, ceramic, or cermet or as liquid metal solutions or ; suspensions. Technological studies were undertaken to evaluate the ; practicability of the various concepts. For solid metallic fuels, powder ; metallurgy and extrusion experiments are reported; binary alloys with iron, ; thorium, uranium, zirconium;

  5. Catalytic reactor with disposable cartridge

    NASA Technical Reports Server (NTRS)

    Mccullough, C. M.

    1973-01-01

    Catalytic reactor, disposable cartridge enclosing iron catalyst, acts as container for solid carbon formed by decomposition of carbon monoxide. Deposition of carbon in other parts of oxygen recovery system does not occur because of lack of catalytic activity; filters trap carbon particles and prevent their being transported outside reaction zone.

  6. Turbulence may sink titanic reactor

    SciTech Connect

    Glanz, J.

    1996-12-06

    The $10 billion International Thermonuclear Experimental Reactor project is meant to show that fusion is a practical energy source, but a new set of calculations says ITER will fizzle. This article describes the expectations and the projections about its future, as well as the challenges recently mounted using new calculations. 3 figs.

  7. Nuclear Reactions and Reactor Safety

    E-print Network

    Onuchic, José

    Nuclear Reactions and Reactor Safety DO NOT LICK We haven't entirely nailed down what element nuclear chain reaction, 1938 #12;Nuclear Chain Reactions Do nuclear chain reactions lead to runaway explosions? or ? -Controlled nuclear chain reactions possible: control energy release/sec -> More

  8. Liquid metal fusion reactor systems

    Microsoft Academic Search

    B. G. Karasev; A. V. Tananaev

    1990-01-01

    The main conceptual designs of liquid metal blanket, methods of pressure drop decrease in the strong magnetic field of the fusion reactor are being discussed. Special features of the flows of electrically conductive fluids in the strong magnetic fields (N? M? 1) are examined. The approximate limits of the transition to the linear (Stokes) flow in the characteristic elements of

  9. INVESTIGATIONS WITH MERCURY FLOW REACTOR

    EPA Science Inventory

    The objective of the research performed in the Mercury Flow Reactor is to investigate short residence-time (seconds) adsorption of mercury species using different sorbents. Emphasis is placed on the effects of mercury concentration, flow rates, reaction temperatures, exposure ti...

  10. Control of reactor coolant flow path during reactor decay heat removal

    DOEpatents

    Hunsbedt, Anstein N. (Los Gatos, CA)

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  11. Tandem Mirror Reactor Systems Code (Version I)

    SciTech Connect

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.; Barrett, R.J.; Gorker, G.E.; Spampinaton, P.T.; Bulmer, R.H.; Dorn, D.W.; Perkins, L.J.; Ghose, S.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost.

  12. Reactor monitoring and safeguards using antineutrino detectors

    SciTech Connect

    Bowden, N S

    2008-09-07

    Nuclear reactors have served as the antineutrino source for many fundamental physics experiments. The techniques developed by these experiments make it possible to use these very weakly interacting particles for a practical purpose. The large flux of antineutrinos that leaves a reactor carries information about two quantities of interest for safeguards: the reactor power and fissile inventory. Measurements made with antineutrino detectors could therefore offer an alternative means for verifying the power history and fissile inventory of a reactors, as part of International Atomic Energy Agency (IAEA) and other reactor safeguards regimes. Several efforts to develop this monitoring technique are underway across the globe.

  13. Refurbishment of existing research reactors for BNCT

    SciTech Connect

    Jatuff, F.E.; Gessaghi, V. [INVAP S.E., de Bariloche (Argentina)

    1997-12-01

    Some research reactors have been selected for the development of boron neutron capture therapy (BNCT) in the United States like the Massachusetts Institute of Technology research reactor, the University of Missouri research reactor 2 or the Brookhaven Medical Research Reactor, among others. These reactors have received excellent analyses and designs to accommodate extremely optimized beam shaping assemblies (BSAs) for the proper tuning of neutron spectra and absorption of undesired particles such as photons and fast neutrons. Due to the importance of BNCT in these facilities, the physicists and engineers have used many degrees of freedom for the optimization process.

  14. Self isolating high frequency saturable reactor

    DOEpatents

    Moore, James A. (Powell, TN)

    1998-06-23

    The present invention discloses a saturable reactor and a method for decoupling the interwinding capacitance from the frequency limitations of the reactor so that the equivalent electrical circuit of the saturable reactor comprises a variable inductor. The saturable reactor comprises a plurality of physically symmetrical magnetic cores with closed loop magnetic paths and a novel method of wiring a control winding and a RF winding. The present invention additionally discloses a matching network and method for matching the impedances of a RF generator to a load. The matching network comprises a matching transformer and a saturable reactor.

  15. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, Edward F. (Naperville, IL); Olson, Arne P. (Western Springs, IL); Wade, David C. (Naperville, IL); Robinson, Bryan W. (Oak Lawn, IL)

    1984-01-01

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion.

  16. Shutdown system for a nuclear reactor

    DOEpatents

    Groh, E.F.; Olson, A.P.; Wade, D.C.; Robinson, B.W.

    1984-06-05

    An ultimate shutdown system is provided for termination of neutronic activity in a nuclear reactor. The shutdown system includes bead chains comprising spherical containers suspended on a flexible cable. The containers are comprised of mating hemispherical shells which provide a ruggedized enclosure for reactor poison material. The bead chains, normally suspended above the reactor core on storage spools, are released for downward travel upon command from an external reactor monitor. The chains are capable of horizontal movement, so as to flow around obstructions in the reactor during their downward motion. 8 figs.

  17. Research Program of a Super Fast Reactor

    SciTech Connect

    Oka, Yoshiaki; Ishiwatari, Yuki; Liu, Jie; Terai, Takayuki; Nagasaki, Shinya; Muroya, Yusa; Abe, Hiroaki [Nuclear Professional School / Department of Nuclear Engineering and Management, The University of Tokyo, Tokaimura, Naka-gun, Ibaraki, 319-1188 (Japan); Mori, Hideo [Department of Mechanical Engineering, Kyushu University (Japan); Akiba, Masato; Akimoto, Hajime; Okumura, Keisuke; Akasaka, Naoaki [Japan Atomic Energy Agency (Japan); GOTO, Shoji [Tokyo Electric Power Company (Japan)

    2006-07-01

    Research program of a supercritical-pressure light water cooled fast reactor (Super Fast Reactor) is funded by MEXT (Ministry of Education, Culture, Sports, Science and Technology) in December 2005 as one of the research programs of Japanese NERI (Nuclear Energy Research Initiative). It consists of three programs. (1) development of Super Fast Reactor concept; (2) thermal-hydraulic experiments; (3) material developments. The purpose of the concept development is to pursue the advantage of high power density of fast reactor over thermal reactors to achieve economic competitiveness of fast reactor for its deployment without waiting for exhausting uranium resources. Design goal is not breeding, but maximizing reactor power by using plutonium from spent LWR fuel. MOX will be the fuel of the Super Fast Reactor. Thermal-hydraulic experiments will be conducted with HCFC22 (Hydro chlorofluorocarbons) heat transfer loop of Kyushu University and supercritical water loop at JAEA. Heat transfer data including effect of grid spacers will be taken. The critical flow and condensation of supercritical fluid will be studied. The materials research includes the development and testing of austenitic stainless steel cladding from the experience of PNC1520 for LMFBR. Material for thermal insulation will be tested. SCWR (Supercritical-Water Cooled Reactor) of GIF (Generation-4 International Forum) includes both thermal and fast reactors. The research of the Super Fast Reactor will enhance SCWR research and the data base. The research period will be until March 2010. (authors)

  18. Decommissioning a nuclear reactor. [Water Boiler Reactor Project

    SciTech Connect

    Montoya, G.M.

    1991-01-01

    The process of decommissioning a facility such as a nuclear reactor or reprocessing plant presents many waste management options and concerns. Waste minimization is a primary consideration, along with protecting a personnel and the environment. Waste management is complicated in that both radioactive and chemical hazardous wastes must be dealt with. This paper presents the general decommissioning approach of a recent project at Los Alamos. Included are the following technical objectives: site characterization work that provided a thorough physical, chemical, and radiological assessment of the contamination at the site; demonstration of the safe and cost-effective dismantlement of a highly contaminated and activated nuclear-fuelded reactor; and techniques used in minimizing radioactive and hazardous waste. 12 figs.

  19. Safety control circuit for a neutronic reactor

    DOEpatents

    Ellsworth, Howard C. (Richland, WA)

    2004-04-27

    A neutronic reactor comprising an active portion containing material fissionable by neutrons of thermal energy, means to control a neutronic chain reaction within the reactor comprising a safety device and a regulating device, a safety device including means defining a vertical channel extending into the reactor from an aperture in the upper surface of the reactor, a rod containing neutron-absorbing materials slidably disposed within the channel, means for maintaining the safety rod in a withdrawn position relative to the active portion of the reactor including means for releasing said rod on actuation thereof, a hopper mounted above the active portion of the reactor having a door disposed at the bottom of the hopper opening into the vertical channel, a plurality of bodies of neutron-absorbing materials disposed within the hopper, and means responsive to the failure of the safety rod on actuation thereof to enter the active portion of the reactor for opening the door in the hopper.

  20. Nuclear reactor vessel fuel thermal insulating barrier

    DOEpatents

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  1. Reactor instrumentation renewal of the TRIGA reactor Vienna, Austria

    SciTech Connect

    Boeck, H. [Atominstitut, Vienna (Austria); Weiss, H. [Technical University, Graz (Austria); Hood, W.E.; Hyde, W.K. [General Atomics, TRIGA Division, San Diego, CA (United States)

    1992-07-01

    The TRIGA Mark-II reactor at the Atominstitut in Vienna, Austria is replacing its twenty-four year old instrumentation system with a microprocessor based control system supplied by General Atomics. Ageing components, new governmental safety requirements and a need for state of the art instrumentation for training students has spurred the demand for new reactor instrumentation. In Austria a government appointed expert is assigned the responsibility of reviewing the proposed installation and verifying all safety aspects. After a positive review, final assembly and checkout of the instrumentation system may commence. The instrumentation system consists of three basic modules: the control system console, the data acquisition console and the NH-1000 wide range channel. Digital communications greatly reduce interwiring requirements. Hardwired safety channels are independent of computer control, thus, the instrumentation system in no way relies on any computer intervention for safety function. In addition, both the CSC and DAC computers are continuously monitored for proper operation via watchdog circuits which are capable of shutting down the reactor in the event of computer malfunction. Safety channels include two interlocked NMP-1000 multi-range linear channels for steady state mode, an NPP-1000 linear safety channel for pulse mode and a set of three independent fuel temperature monitoring channels. The microprocessor controlled wide range NM- 1000 digital neutron monitor (fission chamber based) functions as a startup/operational channel, and provides all power level related Interlocks. The Atominstitut TRIGA reactor is configured for four modes of operation: manual mode, automatic mode (servo control), pulsing mode and square wave mode. Control of the standard control rods is via stepping motor control rod drives, which offers the operator the choice of which control rods are operated by the servo system in automatic and square wave model. (author)

  2. Fission fragment assisted reactor concept for space propulsion: Foil reactor

    NASA Technical Reports Server (NTRS)

    Wright, Steven A.

    1991-01-01

    The concept is to fabricate a reactor using thin films or foils of uranium, uranium oxide and then to coat them on substrates. These coatings would be made so thin as to allow the escaping fission fragments to directly heat a hydrogen propellant. The idea was studied of direct gas heating and direct gas pumping in a nuclear pumped laser program. Fission fragments were used to pump lasers. In this concept two substrates are placed opposite each other. The internal faces are coated with thin foil of uranium oxide. A few of the advantages of this technology are listed. In general, however, it is felt that if one look at all solid core nuclear thermal rockets or nuclear thermal propulsion methods, one is going to find that they all pretty much look the same. It is felt that this reactor has higher potential reliability. It has low structural operating temperatures, very short burn times, with graceful failure modes, and it has reduced potential for energetic accidents. Going to a design like this would take the NTP community part way to some of the very advanced engine designs, such as the gas core reactor, but with reduced risk because of the much lower temperatures.

  3. Low power unattended defense reactor

    SciTech Connect

    Kirchner, W.L.; Meier, K.L.

    1984-01-01

    A small, low power, passive, nuclear reactor electric power supply has been designed for unattended defense applications. Through innovative utilization of existing proven technologies and components, a highly reliable, walk-away safe design has been obtained. Operating at a thermal power level of 200 kWt, the reactor uses low enrichment uranium fuel in a graphite block core to generate heat that is transferred through heat pipes to a thermoelectric (TE) converter. Waste heat is removed from the TEs by circulation of ambient air. Because such a power supply offers the promise of minimal operation and maintenance (O and M) costs as well as no fuel logistics, it is particularly attractive for remote, unattended applications such as the North Warning System.

  4. Fluidized bed coal combustion reactor

    NASA Technical Reports Server (NTRS)

    Moynihan, P. I.; Young, D. L. (inventors)

    1981-01-01

    A fluidized bed coal reactor includes a combination nozzle-injector ash-removal unit formed by a grid of closely spaced open channels, each containing a worm screw conveyor, which function as continuous ash removal troughs. A pressurized air-coal mixture is introduced below the unit and is injected through the elongated nozzles formed by the spaces between the channels. The ash build-up in the troughs protects the worm screw conveyors as does the cooling action of the injected mixture. The ash layer and the pressure from the injectors support a fluidized flame combustion zone above the grid which heats water in boiler tubes disposed within and/or above the combustion zone and/or within the walls of the reactor.

  5. Operate Your Own Tokamak Reactor!

    NSDL National Science Digital Library

    Stotler, Daren.

    Princeton University's Plasma Physics Laboratory's Interactive Plasma Physics Education Experience Web site has updated its interactive Virtual Tokamak. The Java applet is designed to illustrate the basic principles of magnetically confined fusion, and users can now type in the three parameters that include the heating power, magnetic field, and plasma density. Although the applet doesn't work on older PCs, older browsers, and on most Macs, it's worth finding a newer PC to interactively learn about these specific types of reactors.

  6. Reactor vessel lower head integrity

    SciTech Connect

    Rubin, A.M.

    1997-02-01

    On March 28, 1979, the Three Mile Island Unit 2 (TMI-2) nuclear power plant underwent a prolonged small break loss-of-coolant accident that resulted in severe damage to the reactor core. Post-accident examinations of the TMI-2 reactor core and lower plenum found that approximately 19,000 kg (19 metric tons) of molten material had relocated onto the lower head of the reactor vessel. Results of the OECD TMI-2 Vessel Investigation Project concluded that a localized hot spot of approximately 1 meter diameter had existed on the lower head. The maximum temperature on the inner surface of the reactor pressure vessel (RPV) in this region reached 1100{degrees}C and remained at that temperature for approximately 30 minutes before cooling occurred. Even under the combined loads of high temperature and high primary system pressure, the TMI-2 RPV did not fail. (i.e. The pressure varied from about 8.5 to 15 MPa during the four-hour period following the relocation of melt to the lower plenum.) Analyses of RPV failure under these conditions, using state-of-the-art computer codes, predicted that the RPV should have failed via local or global creep rupture. However, the vessel did not fail; and it has been hypothesized that rapid cooling of the debris and the vessel wall by water that was present in the lower plenum played an important role in maintaining RPV integrity during the accident. Although the exact mechanism(s) of how such cooling occurs is not known, it has been speculated that cooling in a small gap between the RPV wall and the crust, and/or in cracks within the debris itself, could result in sufficient cooling to maintain RPV integrity. Experimental data are needed to provide the basis to better understand these phenomena and improve models of RPV failure in severe accident codes.

  7. Reactor safeguards against insider sabotage

    SciTech Connect

    Bennett, H.A.

    1982-03-01

    A conceptual safeguards system is structured to show how both reactor operations and physical protection resources could be integrated to prevent release of radioactive material caused by insider sabotage. Operational recovery capabilities are addressed from the viewpoint of both detection of and response to disabled components. Physical protection capabilities for preventing insider sabotage through the application of work rules are analyzed. Recommendations for further development of safeguards system structures, operational recovery, and sabotage prevention are suggested.

  8. Nuclear power-reactor decommissioning

    Microsoft Academic Search

    LaGuardia

    2009-01-01

    The article summarizes the major findings of an evaluation of several alternatives for decommissioning 1100-MW(e) nuclear power reactors. The evaluation included the technical feasibility of decommissioning and the costs, schedule, environmental impacts, and occupational exposures for three decommissioning alternatives: mothballing, entombment, and prompt removal of radioactive components and dismantling. In addition, two combinations of these alternatives were evaluated: mothballing--delayed removal

  9. Solid oxide electrochemical reactor science

    Microsoft Academic Search

    Neal P. Sullivan; Ellen Beth Stechel; Connor J. Moyer; Andrea Ambrosini; Robert J. Key

    2010-01-01

    Solid-oxide electrochemical cells are an exciting new technology. Development of solid-oxide cells (SOCs) has advanced considerable in recent years and continues to progress rapidly. This thesis studies several aspects of SOCs and contributes useful information to their continued development. This LDRD involved a collaboration between Sandia and the Colorado School of Mines (CSM) ins solid-oxide electrochemical reactors targeted at solid

  10. Using reactor operating experience to improve the design of a new Broad Application Test Reactor

    SciTech Connect

    Fletcher, C.D.; Ryskamp, J.M.; Drexler, R.L.; Leyse, C.F.

    1993-07-01

    Increasing regulatory demands and effects of plant aging are limiting the operation of existing test reactors. Additionally, these reactors have limited capacities and capabilities for supporting future testing missions. A multidisciplinary team of experts developed sets of preliminary safety requirements, facility user needs, and reactor design concepts for a new Broad Application Test Reactor (BATR). Anticipated missions for the new reactor include fuels and materials irradiation testing, isotope production, space testing, medical research, fusion testing, intense positron research, and transmutation doping. The early BATR design decisions have benefited from operating experiences with existing reactors. This paper discusses these experiences and highlights their significance for the design of a new BATR.

  11. Reactor tank UT acceptance criteria

    SciTech Connect

    Daugherty, W.L.

    1990-01-30

    The SRS reactor tanks are constructed of type 304 stainless steel, with 0.5 inch thick walls. An ultrasonic (UT) in-service inspection program has been developed for examination of these tanks, in accordance with the ISI Plan for the Savannah River Production Reactors Process Water System (DPSTM-88-100-1). Prior to initiation of these inspections, criteria for the disposition of any indications that might be found are required. A working group has been formed to review available information on the SRS reactor tanks and develop acceptance criteria. This working group includes nationally recognized experts in the nuclear industry. The working group has met three times and produced three documents describing the proposed acceptance criteria, the technical basis for the criteria and a proposed initial sampling plan. This report transmits these three documents, which were prepared in accordance with the technical task plan and quality assurance plan for this task, task 88-001-A- 1. In addition, this report summarizes the acceptance criteria and proposed sampling plan, and provides further interpretation of the intent of these three documents where necessary.

  12. The ARIES tokamak reactor study

    SciTech Connect

    Not Available

    1989-10-01

    The ARIES study is a community effort to develop several visions of tokamaks as fusion power reactors. The aims are to determine the potential economics, safety, and environmental features of a range of possible tokamak reactors, and to identify physics and technology areas with the highest leverage for achieving the best tokamak reactor. Three ARIES visions are planned, each having a different degree of extrapolation from the present data base in physics and technology. The ARIES-I design assumes a minimum extrapolation from current tokamak physics (e.g., 1st stability) and incorporates technological advances that can be available in the next 20 to 30 years. ARIES-II is a DT-burning tokamak which would operate at a higher beta in the 2nd MHD stability regime. It employs both potential advances in the physics and expected advances in technology and engineering. ARIES-II will examine the potential of the tokamak and the D{sup 3}He fuel cycle. This report is a collection of 14 papers on the results of the ARIES study which were presented at the IEEE 13th Symposium on Fusion Engineering (October 2-6, 1989, Knoxville, TN). This collection describes the ARIES research effort, with emphasis on the ARIES-I design, summarizing the major results, the key technical issues, and the central conclusions.

  13. Prospects for Tokamak Fusion Reactors

    SciTech Connect

    Sheffield, J.; Galambos, J.

    1995-04-01

    This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant.

  14. Reference worldwide model for antineutrinos from reactors

    NASA Astrophysics Data System (ADS)

    Baldoncini, Marica; Callegari, Ivan; Fiorentini, Giovanni; Mantovani, Fabio; Ricci, Barbara; Strati, Virginia; Xhixha, Gerti

    2015-03-01

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework, we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency. We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO +), and proposed (Juno, RENO-50, LENA, and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation, and detection processes are estimated using a Monte Carlo-based approach, which provides an overall site-dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes, and we estimate a 2.4% increase of the unoscillated event rate in the geoneutrino energy window due to the storage of spent nuclear fuels in the cooling pools. We predict that the research reactors contribute to less than 0.2% to the commercial reactor signal in the investigated 14 sites. We perform a multitemporal analysis of the expected reactor signal over a time lapse of ten years using reactor operational records collected in a comprehensive database published at www.fe.infn.it/antineutrino.

  15. Reference worldwide model for antineutrinos from reactors

    E-print Network

    Marica Baldoncini; Ivan Callegari; Giovanni Fiorentini; Fabio Mantovani; Barbara Ricci; Virginia Strati; Gerti Xhixha

    2015-02-16

    Antineutrinos produced at nuclear reactors constitute a severe source of background for the detection of geoneutrinos, which bring to the Earth's surface information about natural radioactivity in the whole planet. In this framework we provide a reference worldwide model for antineutrinos from reactors, in view of reactors operational records yearly published by the International Atomic Energy Agency (IAEA). We evaluate the expected signal from commercial reactors for ongoing (KamLAND and Borexino), planned (SNO+) and proposed (Juno, RENO-50, LENA and Hanohano) experimental sites. Uncertainties related to reactor antineutrino production, propagation and detection processes are estimated using a Monte Carlo based approach, which provides an overall site dependent uncertainty on the signal in the geoneutrino energy window on the order of 3%. We also implement the off-equilibrium correction to the reference reactor spectra associated with the long-lived isotopes and we estimate a 2.4% increase of the unoscillated event rate in the geoneutrino energy window due to the storage of spent nuclear fuels in the cooling pools. We predict that the research reactors contribute to less than 0.2% to the commercial reactor signal in the investigated 14 sites. We perform a multitemporal analysis of the expected reactor signal over a time lapse of 10 years using reactor operational records collected in a comprehensive database published at www.fe.infn.it/antineutrino.

  16. PID Control Effectiveness for Surface Reactor Concepts

    NASA Astrophysics Data System (ADS)

    Dixon, David D.; Marsh, Christopher L.; Poston, David I.

    2007-01-01

    Control of space and surface fission reactors should be kept as simple as possible, because of the need for high reliability and the difficulty to diagnose and adapt to control system failures. Fortunately, compact, fast-spectrum, externally controlled reactors are very simple in operation. In fact, for some applications it may be possible to design low-power surface reactors without the need for any reactor control after startup; however, a simple proportional, integral, derivative (PID) controller can allow a higher performance concept and add more flexibility to system operation. This paper investigates the effectiveness of a PID control scheme for several anticipated transients that a surface reactor might experience. To perform these analyses, the surface reactor transient code FRINK was modified to simulate control drum movements based on bulk coolant temperature.

  17. Molten-Salt Depleted-Uranium Reactor

    E-print Network

    Dong, Bao-Guo; Gu, Ji-Yuan

    2015-01-01

    The supercritical, reactor core melting and nuclear fuel leaking accidents have troubled fission reactors for decades, and greatly limit their extensive applications. Now these troubles are still open. Here we first show a possible perfect reactor, Molten-Salt Depleted-Uranium Reactor which is no above accident trouble. We found this reactor could be realized in practical applications in terms of all of the scientific principle, principle of operation, technology, and engineering. Our results demonstrate how these reactors can possess and realize extraordinary excellent characteristics, no prompt critical, long-term safe and stable operation with negative feedback, closed uranium-plutonium cycle chain within the vessel, normal operation only with depleted-uranium, and depleted-uranium high burnup in reality, to realize with fission nuclear energy sufficiently satisfying humanity long-term energy resource needs, as well as thoroughly solve the challenges of nuclear criticality safety, uranium resource insuffic...

  18. Antineutrino monitoring for heavy water reactors.

    PubMed

    Christensen, Eric; Huber, Patrick; Jaffke, Patrick; Shea, Thomas E

    2014-07-25

    In this Letter we discuss the potential application of antineutrino monitoring to the Iranian heavy water reactor at Arak, the IR-40, as a nonproliferation measure. An above ground detector positioned right outside the IR-40 reactor building could meet IAEA verification goals for reactor plutonium inventories. While detectors with the needed spectral sensitivity have been demonstrated below ground, additional research and development is needed to demonstrate an above-ground detector with this same level of sensitivity. In addition to monitoring the reactor during operation, observing antineutrino emissions from long-lived fission products could also allow monitoring the reactor when it is shut down, provided very low detector backgrounds can be achieved. Antineutrino monitoring could also be used to distinguish different levels of fuel enrichment. Most importantly, these capabilities would not require a complete reactor operational history and could provide a means to reestablish continuity of knowledge in safeguards conclusions should this become necessary. PMID:25105612

  19. Jules Horowitz Reactor: a high performance material testing reactor

    NASA Astrophysics Data System (ADS)

    Iracane, Daniel; Chaix, Pascal; Alamo, Ana

    2008-04-01

    The physical modelling of materials' behaviour under severe conditions is an indispensable element for developing future fission and fusion systems: screening, design, optimisation, processing, licensing, and lifetime assessment of a new generation of structure materials and fuels, which will withstand high fast neutron flux at high in-service temperatures with the production of elements like helium and hydrogen. JANNUS and other analytical experimental tools are developed for this objective. However, a purely analytical approach is not sufficient: there is a need for flexible experiments integrating higher scales and coupled phenomena and offering high quality measurements; these experiments are performed in material testing reactors (MTR). Moreover, complementary representative experiments are usually performed in prototypes or dedicated facilities such as IFMIF for fusion. Only such a consistent set of tools operating on a wide range of scales, can provide an actual prediction capability. A program such as the development of silicon carbide composites (600-1200 °C) illustrates this multiscale strategy. Facing the long term needs of experimental irradiations and the ageing of present MTRs, it was thought necessary to implement a new generation high performance MTR in Europe for supporting existing and future nuclear reactors. The Jules Horowitz Reactor (JHR) project copes with this context. It is funded by an international consortium and will start operation in 2014. JHR will provide improved performances such as high neutron flux ( 10 n/cm/s above 0.1 MeV) in representative environments (coolant, pressure, temperature) with online monitoring of experimental parameters (including stress and strain control). Experimental devices designing, such as high dpa and small thermal gradients experiments, is now a key objective requiring a broad collaboration to put together present scientific state of art, end-users requirements and advanced instrumentation. To cite this article: D. Iracane et al., C. R. Physique 9 (2008).

  20. In-reactor performance of pressure tubes in CANDU reactors

    NASA Astrophysics Data System (ADS)

    Rodgers, D. K.; Coleman, C. E.; Griffiths, M.; Bickel, G. A.; Theaker, J. R.; Muir, I.; Bahurmuz, A. A.; Lawrence, S. St.; Resta Levi, M.

    2008-12-01

    The pressure tubes in CANDU reactors have been operating for times up to about 25 years. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behaviour and discusses the factors controlling the behaviour of these components in currently operating CANDU reactors. The mechanical properties (such as ultimate tensile strength, UTS, and fracture toughness), and delayed-hydride-cracking properties (crack growth rate Vc, and threshold stress intensity factor, KIH) change with irradiation; the former reach a limiting value at a fluence of <1 × 10 25 n m -2, while Vc and KIH reach a steady-state condition after a fluence of about 3 × 10 25 n m -2 and 3 × 10 24 n m -2, respectively. At saturation the UTS is raised by about 200 MPa, toughness is reduced to about 40% of its initial value, Vc increases by about a factor of ten while KIH is only slightly reduced. The role of microstructure and trace elements in these behaviours is described. Pressure tubes exhibit elongation and diametral expansion. The deformation behaviour is a function of operating conditions and material properties that vary from tube-to-tube and as a function of axial location. Semi-empirical predictive models have been developed to describe the deformation response of average tubes as a function of operating conditions. For corrosion and, more importantly deuterium pickup, semi-empirical predictive models have also been developed to represent the behaviour of an average tube. The effect of material variability on corrosion behaviour is less well defined compared with other properties. Improvements in manufacturing have increased fracture resistance by minimising trace elements, especially H and Cl, and reduced variability by tightening controls on forming parameters, especially hot-working temperatures.

  1. LPT. EBOR (TAN646) reactor vault. Concrete blocks for reactor shielding ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    LPT. EBOR (TAN-646) reactor vault. Concrete blocks for reactor shielding going into shield test facility pool. Photographer: Comiskey. Date: February 19, 1965. INEEL negative no. 65-933 - Idaho National Engineering Laboratory, Test Area North, Scoville, Butte County, ID

  2. A review on the reactor antineutrino flux and reactor based neutrino experiments

    NASA Astrophysics Data System (ADS)

    Wang, Wei

    2013-10-01

    The pure antineutrino flux from nuclear reactors has unique contributions to neutrino physics. We will first briefly review the past and the current reactor based neutrino experiments with more emphases on the current generation short-baseline reactor experiments Daya Bay, RNEO and Double Chooz. In coming years, while the current generation of short-baseline reactor neutrino experiments provides more accurate measurement of theta13 and reactor antineutrino flux, very short-baseline and medium-baseline reactor neutrino experiments are being carried out and proposed. The VSBL experiments are aimming at shedding light at the so-called reactor anomaly problem and provide pure isotope antineutrino flux measurements. The medium-baseline effort is taking the challenge of determining the neutrino mass hierachy and measuring the solar sector oscillation parameters and atmospheric mass-squared split to <1% level. We will also briefly review the reactor antineutrino flux calculation and its key issues and impact to future experiments.

  3. Investigation of the basic reactor physics characteristics of the Dalat Nuclear Research Reactor

    SciTech Connect

    Huy, N.Q. [Center of Nuclear Techniques, Hochiminh City (Viet Nam); Thong, H.V.; Khang, N.P. [Nuclear Research Institute, Dalat (Viet Nam)

    1994-12-31

    The Dalat nuclear research reactor was reconstructed from the TRIGA Mark II reactor, built in 1963 with a nominal power of 250 kW, and reached its planned nominal power of 500 kW for the first time in February 1984. The Dalat reactor has some characteristics distinct from the former TRIGA reactor. Investigation of its characteristics is carried out by the determination of the reactor physics parameters. This paper represents the experimental results obtained for the effective fraction of the delayed photoneutrons, the extraneous neutron source left after the reactor is shut down, the lowest power levels of reactor critical states, the relative axial and radial distributions of thermal neutrons, the safe positive reactivity inserted into the reactor at a deep subcritical state, the reactivity temperature coefficient of water, the temperature on the surface of the fuel elements, etc.

  4. Core design of the upgraded TREAT reactor

    Microsoft Academic Search

    D. C. Wade; S. K. Bhattacharyya; W. C. Lipinski; C. C. Stone

    1982-01-01

    The upgrading of the TREAT reactor involves the replacement of the central 11 x 11 subzone of the 19 x 19 fuel assembly array by new, Inconel-clad, high-temperature fuel assemblies, and the additions of a new reactor control system, a safety-grade plant protection system, and an enhanced reactor filtration\\/coolant system. The final design of these modifications will be completed in

  5. Solid tags for identifying failed reactor components

    DOEpatents

    Bunch, Wilbur L. (Richland, WA); Schenter, Robert E. (Richland, WA)

    1987-01-01

    A solid tag material which generates stable detectable, identifiable, and measurable isotopic gases on exposure to a neutron flux to be placed in a nuclear reactor component, particularly a fuel element, in order to identify the reactor component in event of its failure. Several tag materials consisting of salts which generate a multiplicity of gaseous isotopes in predetermined ratios are used to identify different reactor components.

  6. Studies and thoughts on nuclear reactor systems

    Microsoft Academic Search

    M. Grenon

    1976-01-01

    Several types of nuclear power reactor systems are compared in terms of: costs, environmental impact, security against diversion of fissile materials, dependence on dwindling high-grade uranium reserves, fuel cycle, wastes disposal problems, and research\\/development problems and needs. The review is based on a Battelle team study of high temperature gas-cooled reactors (HTGR), molten salt breeders (MSBR), gas-cooled fast reactors (GCFR),

  7. Energy deposition in STARFIRE reactor components

    SciTech Connect

    Gohar, Y.; Brooks, J.N.

    1985-04-01

    The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry.

  8. Interactive simulators for AGN and TRIGA reactors

    SciTech Connect

    Crawford, K.C. (Idaho State Unvi., Pocatello (United States))

    1993-01-01

    Students and faculty at Idaho State University (ISU) have developed several interactive research reactor simulators that operate on personal computer compatible computers. The development of the simulators was undertaken primarily to provide a framework to test feedback control algorithms prior to implementation on real systems. As the project developed, it became evident that these simulators could also be used to familiarize students and reactor operator trainees with dynamic reactor behavior before allowing them to manipulate the controls of the facility.

  9. Nuclear reactor shield including magnesium oxide

    DOEpatents

    Rouse, Carl A. (Del Mar, CA); Simnad, Massoud T. (La Jolla, CA)

    1981-01-01

    An improvement in nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux, the reactor shielding including means providing structural support, neutron moderator material, neutron absorber material and other components as described below, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron.

  10. D-D tokamak reactor studies

    SciTech Connect

    Evans, K.E. Jr.; Baker, C.C.; Brooks, J.N.; Ehst, D.A.; Finn, P.A.; Jung, J.; Mattas, R.F.; Misra, B.; Smith, D.L.; Stevens, H.C.

    1980-11-01

    A tokamak D-D reactor design, utilizing the advantages of a deuterium-fueled reactor but with parameters not unnecessarily extended from existing D-T designs, is presented. Studies leading to the choice of a design and initial studies of the design are described. The studies are in the areas of plasma engineering, first-wall/blanket/shield design, magnet design, and tritium/fuel/vacuum requirements. Conclusions concerning D-D tokamak reactors are stated.

  11. Dynamics of heat-pipe reactors

    NASA Technical Reports Server (NTRS)

    Niederauer, G. F.

    1971-01-01

    A split-core heat pipe reactor, fueled with either U(233)C or U(235)C in a tungsten cermet and cooled by 7-Li-W heat pipes, was examined for the effects of the heat pipes on reactor while trying to safely absorb large reactivity inputs through inherent shutdown mechanisms. Limits on ramp reactivity inputs due to fuel melting temperature and heat pipe wall heat flux were mapped for the reactor in both startup and at-power operating modes.

  12. The fixed bed nuclear reactor concept

    Microsoft Academic Search

    Sümer ?ahin; Farhang Sefidvash

    2008-01-01

    In the present work, the basic features of a new reactor type, the so-called Fixed Bed Nuclear Reactor (FBNR) is presented. FBNR is a small reactor (40MWe) without the need of on-site refueling. It utilizes the PWR technology but uses the HTGR type fuel elements. It has the characteristics of being simple in design, modular, inherent safety, passive cooling, proliferation

  13. Identification and Control of Polymerization Reactors

    Microsoft Academic Search

    Eric J. Hukkanen; Jeremy G. VanAntwerp; Richard D. Braatz

    This chapter considers the identification and control of free-radical polymerization reactors. A discussion of the modeling\\u000a and simulation of such reactors is followed by an optimal control study that demonstrates the potential of optimal control\\u000a of the molecular-weight distribution based on mechanistic models. Achieving this potential in a batch reactor requires an\\u000a accurate estimation of the free-radical polymerization kinetic parameters.

  14. Development of the reactor safety film

    SciTech Connect

    Sheheen, N.N.

    1980-01-01

    This paper summarizes the text and describes the processes followed to develop the first computer-generated film of LASL's Reactor Safety efforts. The 11-1/2 min film with narrative and musical background gives a brief overview of reactor components, of how LASL's Reactor Safety groups develop and verify computer codes to anticipate accidents, and of how these codes were applied to the Three Mile Island accident.

  15. Fuel handling apparatus for a nuclear reactor

    DOEpatents

    Hawke, Basil C. (Solana Beach, CA)

    1987-01-01

    Fuel handling apparatus for transporting fuel elements into and out of a nuclear reactor and transporting them within the reactor vessel extends through a penetration in the side of the reactor vessel. A lateral transport device carries the fuel elements laterally within the vessel and through the opening in the side of the vessel, and a reversible lifting device raises and lowers the fuel elements. In the preferred embodiment, the lifting device is supported by a pair of pivot arms.

  16. Synergetic Potential of Light Water Reactors

    SciTech Connect

    Barchevtsev, V.; Artisyuk, V.; Ninokata, H. [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1 O-okayama, Meguro-ku, Tokyo 152-8550 (Japan)

    2002-07-01

    To satisfy growing electricity demands and recent trends in nuclear power, the paper proposes an innovative fuel cycle scenario based on well-established light water-cooled reactor technology. The originality of this study relies on the synergetic potential of Pressurized Water Reactor (PWR) and Pressure Tube Graphite Reactor (PTGR) in achieving proliferation resistant and high burnup fuel cycle. Suggested approach is suitable for countries or remote territories that are not provided with sufficient nuclear infrastructure. (authors)

  17. Nuclear weapons and power-reactor plutonium

    Microsoft Academic Search

    Amory B. Lovins

    1980-01-01

    1-10 that for making nuclear bombs, 'reactor-grade' plutonium produced by the normal operation of uranium-fuelled power reactors is necessarily much inferior to specially made 'weapons-grade' Pu: so infe- rior in explosive power or predictability that its potential use by amateurs is not a serious problem and that governments would instead make the higher-performance weapons-grade Pu in special production reactors. Although

  18. Heat pipe reactors for space power applications

    Microsoft Academic Search

    D. R. Koenig; W. A. Ranken; E. W. Salmi

    1977-01-01

    A family of heat pipe reactors design concepts has been developed to provide heat to a variety of electrical conversion systems. Three power plants are described that span the power range 1-500 kWe and operate in the temperature range 1200-1700 K. The reactors are fast, compact, heat-pipe cooled, high-temperature nuclear reactors fueled with fully enriched refractory fuels, UC-ZrC or UO2.

  19. Neutron shielding panels for reactor pressure vessels

    DOEpatents

    Singleton, Norman R. (Murrysville, PA)

    2011-11-22

    In a nuclear reactor neutron panels varying in thickness in the circumferential direction are disposed at spaced circumferential locations around the reactor core so that the greatest radial thickness is at the point of highest fluence with lesser thicknesses at adjacent locations where the fluence level is lower. The neutron panels are disposed between the core barrel and the interior of the reactor vessel to maintain radiation exposure to the vessel within acceptable limits.

  20. Desirability of small reactors, HTGR in particular

    Microsoft Academic Search

    Yasuo Tsuchie

    2000-01-01

    Small reactors of about 100–300 MWe, High Temperature Gas Cooled Reactors (HTGRs) in particular, are considered desirable in future, based on the following ways of thinking;Global scale enhancement of nuclear energy is considered necessary from reduction of environment impact point of view.Small reactors are desirable, due to (a) enhanced safety in terms of fuel inventory and inherent safety, then (b)

  1. Adaptive state estimation for the Clinch River Breeder Reactor Plant

    Microsoft Academic Search

    DeMore

    1979-01-01

    This dissertation presents a study on a state estimator for a liquid metal fast breeder reactor specifically for the reactor to be used in the Clinch River Breeder Reactor Plant. The state estimator, which is a steady-state Kalman Filter, adapts to the non-linear behavior of the reactor over the operating range of 40% to 100% reactor power. A low order

  2. Plutonium breeding in liquid-metal fast breeder reactors and light water reactors

    Microsoft Academic Search

    Vendryes

    1985-01-01

    The possibilities of breeding in liquid-metal fast breeder reactors (LMFBRs) and light water reactors (LWRs) are compared in two ways. The feasibility of breeding has been demonstrated in the Phenix reactor with a measured gain of 0.14. The gain in Superphenix will amount to about0.20. The studies show that while maintaining the performance of commercial reactors their breeding gain can

  3. Engineering activities at the MIT research reactor in support of power reactor technology

    Microsoft Academic Search

    O. K. Harling; J. A. Bernard; M. J. Driscoll; G. E. Kohse; R. G. Ballinger

    1989-01-01

    The Massachusetts Institute of Technology (MIT) research reactor (MITR-II) is a 5-MW(thermal) light-water-cooled and-moderated reactor (LWR) with in-core neutron and gamma dose rates that closely approximate those in current LWRs. Compact in-pile loops that simulate pressurized water reactor (PWR) and boiling water reactor (BWR) thermal hydraulics and coolant chemistry have been designed for installation in the MITR-II. A PWR loop

  4. From High-Temperature Gas-Cooled Reactors to Gas-Cooled Fast Breeder Reactors

    Microsoft Academic Search

    R. H. Simon; G. J. Schlueter

    1973-01-01

    The evolution of gas-cooled reactors is described and the advantages of helium as a reactor coolant stemming from its inertness and the fact that it is a single-phase coolant are discussed. The High-Temperature Gas-Cooled Reactor (HTGR) forms the basis for the current design of a Gas-Cooled Fast Breeder Reactor (GCFR). The extensive use of existing HTGR technology and the similarities

  5. Heat dissipating nuclear reactor with metal liner

    DOEpatents

    Gluekler, Emil L. (San Jose, CA); Hunsbedt, Anstein (Los Gatos, CA); Lazarus, Jonathan D. (Sunnyvale, CA)

    1987-01-01

    Disclosed is a nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  6. Tanden Mirror Reactor Systems Code (TMRSC)

    SciTech Connect

    Reid, R.L.; Rothe, K.E.; Barrett, R.J.

    1985-01-01

    This paper describes a computer code developed to model a tandem mirror reactor. This is the first tandem mirror reactor model to couple the highly linked physics, magnetics, and neutronic analysis into a single code. Results from this code for two sensitivity studies are included in this paper. These studies are designed (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power and (2) to determine the impact of reactor power level on cost.

  7. Non-equilibrium radiation nuclear reactor

    NASA Technical Reports Server (NTRS)

    Thom, K.; Schneider, R. T. (inventors)

    1978-01-01

    An externally moderated thermal nuclear reactor is disclosed which is designed to provide output power in the form of electromagnetic radiation. The reactor is a gaseous fueled nuclear cavity reactor device which can operate over wide ranges of temperature and pressure, and which includes the capability of processing and recycling waste products such as long-lived transuranium actinides. The primary output of the device may be in the form of coherent radiation, so that the reactor may be utilized as a self-critical nuclear pumped laser.

  8. TA-2 Water Boiler Reactor Decommissioning Project

    SciTech Connect

    Durbin, M.E. (ed.); Montoya, G.M.

    1991-06-01

    This final report addresses the Phase 2 decommissioning of the Water Boiler Reactor, biological shield, other components within the biological shield, and piping pits in the floor of the reactor building. External structures and underground piping associated with the gaseous effluent (stack) line from Technical Area 2 (TA-2) Water Boiler Reactor were removed in 1985--1986 as Phase 1 of reactor decommissioning. The cost of Phase 2 was approximately $623K. The decommissioning operation produced 173 m{sup 3} of low-level solid radioactive waste and 35 m{sup 3} of mixed waste. 15 refs., 25 figs., 3 tabs.

  9. Heat dissipating nuclear reactor with metal liner

    DOEpatents

    Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  10. Nuclear reactor multiphysics via bond graph formalism

    E-print Network

    Sosnovsky, Eugeny

    2014-01-01

    This work proposes a simple and effective approach to modeling nuclear reactor multiphysics problems using bond graphs. Conventional multiphysics simulation paradigms normally use operator splitting, which treats the ...

  11. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    Microsoft Academic Search

    Lee O. Nelson

    2011-01-01

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered

  12. REACTOR PRESSURE VESSEL TEMPERATURE ANALYSIS OF CANDIDATE VERY HIGH TEMPERATURE REACTOR DESIGNS

    Microsoft Academic Search

    Hans D. Gougar; Cliff B. Davis; George Hayner; Kevan Weaver

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated

  13. Theoretical and experimental investigations of reactor parameters in a U-233 fuelled research reactor

    Microsoft Academic Search

    D. K. Mohapatra; E. Radha; P. Mohanakrishnan

    2004-01-01

    The Kalpakkam Mini Reactor (KAMINI) has the unique distinction of being the only operating pool type research reactor in the world at present, that utilizes U-233 as the fissile material. It is a 30 kW pool type reactor, operated with an alloy of U-233 and aluminium as fuel, light water as moderator and beryllium oxide as reflector. It is designed

  14. Space power reactor ground test in the Experimental Gas Cooled Reactor (EGCR) at Oak Ridge

    Microsoft Academic Search

    M. H. Fontana; R. S. Holcomb; R. H. Cooper

    1992-01-01

    The Experimental Gas Cooled Reactor (EGCR) facility and the supporting technical infrastructure at the Oak Ridge National Laboratory have the capabilities of performing ground tests of space nuclear power reactor systems. A candidate test would be a 10 MWt lithium cooled reactor, generating potassium vapor that would drive a power turbine. The facility is a large containment vessel originally intended

  15. COMPARISON OF BOILING WATER REACTORS WITH OTHER TYPES OF POWER REACTORS

    Microsoft Academic Search

    H. Bergua; L. Palacios

    1958-01-01

    A comoparison is made of the boiling water reactors with the pressurized ; water reactor and the gas-cooled, graphite-moderated reactors of the Calder Hall ; type. The initial cost, operating characteristics, malntenance cost, and future ; development are considered. (J.S.R.);

  16. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Microsoft Academic Search

    F. Faghihi; S. M. Mirvakili

    2009-01-01

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (?ex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and

  17. High-rate treatment of molasses wastewater by combination of an acidification reactor and a USSB reactor

    Microsoft Academic Search

    Takashi Onodera; Shinya Sase; Pairaya Choeisai; Wilasinee Yoochatchaval; Haruhiko Sumino; Takashi Yamaguchi; Yoshitaka Ebie; Kaiqin Xu; Noriko Tomioka; Kazuaki Syutsubo

    2011-01-01

    A combination of an acidification reactor and an up-flow staged sludge bed (USSB) reactor was applied for treatment of molasses wastewater containing a large amount of organic compounds and sulfate. The USSB reactor had three gas-solid separators (GSS) along the height of the reactor. The combined system was continuously operated at mesophilic temperature over 400 days. In the acidification reactor,

  18. Effects of nuclear island connected buildings on seismic behaviour of reactor internals in a pool type fast breeder reactor

    Microsoft Academic Search

    P. Chellapandi; S. C. Chetal; Baldev Raj

    2007-01-01

    The seismic analysis of reactor assembly housing the primary circuit of a typical 500MWe capacity pool type fast breeder reactor (PFBR) is reported. The reactor assembly is supported on the reactor vault within the nuclear island connected buildings (NICB). The seismic responses, viz. critical displacements, sloshing heights, stresses and strain energy values in the vessels are determined for the reactor

  19. Neural antecedents of the endowment effect Supplementary Material

    E-print Network

    Knutson, Brian

    the Godiva chocolate instead of $4.00. You would choose "No" if you want $4.00 instead of the Godiva Chocolate. Choose Product? Chocolates> Godiva Chocolate $4.00 Press the Space Bar to continue of the Godiva chocolate. You would choose "No" if you want the Godiva Chocolate instead of $4.00. Choose Money

  20. Fast pulse nonthermal plasma reactor

    DOEpatents

    Rosocha, Louis A.

    2005-06-14

    A fast pulsed nonthermal plasma reactor includes a discharge cell and a charging assembly electrically connected thereto. The charging assembly provides plural high voltage pulses to the discharge cell. Each pulse has a rise time between one and ten nanoseconds and a duration of three to twenty nanoseconds. The pulses create nonthermal plasma discharge within the discharge cell. Accordingly, the nonthermal plasma discharge can be used to remove pollutants from gases or break the gases into smaller molecules so that they can be more efficiently combusted.

  1. Integral data for fast reactors

    SciTech Connect

    Collins, P.J.; Poenitz, W.P.; McFarlane, H.F.

    1988-01-01

    Requirements at Argonne National Laboratory to establish the best estimates and uncertainties for LMR design parameters have lead to an extensive evaluation of the available critical experiment database. Emphasis has been put upon selection of a wide range of cores, including both benchmark, assemblies covering a range of spectra and compositions and power reactor mock-up assemblies with diverse measured parameters. The integral measurements have been revised, where necessary, using the most recent reference data and a covariance matrix constructed. A sensitivity database has been calculated, embracing all parameters, which enables quantification of the relevance of the integral data to parameters calculated with ENDF/B-V.2 cross sections.

  2. Liquid metal reactor absorber technology

    SciTech Connect

    Pitner, A.L.

    1990-10-01

    The selection of boron carbide as the reference liquid metal reactor absorber material is supported by results presented for irradiation performance, reactivity worth compatibility, and benign failure consequences. Scram response requirements are met easily with current control rod configurations. The trend in absorber design development is toward larger sized pins with fewer pins per bundle, providing economic savings and improved hydraulic characteristics. Very long-life absorber designs appear to be attainable with the application of vented pin and sodium-bonded concepts. 3 refs., 3 figs.

  3. Operate Your Own Tokamak Reactor!

    NSDL National Science Digital Library

    Stotler, Daren.

    2001-01-01

    Princeton University's Plasma Physics Laboratory's Interactive Plasma Physics Education Experience Web site (last mentioned in the July 26, 2002 NSDL Physical Sciences ) has updated its interactive Virtual Tokamak. The Java applet is designed to illustrate the basic principles of magnetically confined fusion, and users can now type in the three parameters that include the heating power, magnetic field, and plasma density. Although the applet doesn't work on older PCs, older browsers, and on most Macs, it's worth finding a newer PC to interactively learn about these specific types of reactors.

  4. Autonomous Control of Space Nuclear Reactors

    NASA Technical Reports Server (NTRS)

    Merk, John

    2013-01-01

    Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the safety requirements of a nuclear reactor and provides high availability to the host system. The RICS is intended to interface with a host computer (the computer of the spacecraft where the reactor is mounted). The RICS leverages the safety features inherent in Earth-based reactors and also integrates the wide range neutron detector (WRND). A neutron detector provides the input that allows the RICS to do its job. The RICS is based on proven technology currently in use at a nuclear research facility. In its most basic form, the RICS is a ruggedized, compact data-acquisition and control system that could be adapted to support a wide variety of harsh environments. As such, the RICS could be a useful instrument outside the scope of a nuclear reactor, including military applications where failsafe data acquisition and control is required with stringent size, weight, and power constraints.

  5. Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

    NASA Astrophysics Data System (ADS)

    Wright, Steven A.; Houts, Michael

    2001-02-01

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

  6. Coupled Reactor Kinetics and Heat Transfer Model for Heat Pipe Cooled Reactors

    SciTech Connect

    WRIGHT,STEVEN A.; HOUTS,MICHAEL

    2000-11-22

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). The paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities.

  7. Looking Northeast in Oxide Building at Reactors on Second Floor ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    Looking Northeast in Oxide Building at Reactors on Second Floor Including Reactor One (Left) and Reactor Two (Right) - Hematite Fuel Fabrication Facility, Oxide Building & Oxide Loading Dock, 3300 State Road P, Festus, Jefferson County, MO

  8. 10 CFR 1.44 - Office of New Reactors.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ... false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR REGULATORY COMMISSION...1.44 Office of New Reactors. The Office of New...or the safeguarding of nuclear reactor facilities licensed...

  9. APPLICATION OF DATA ANALYSIS TECHNIQUES TO NUCLEAR REACTOR

    E-print Network

    Kunz, Robert Francis

    1 APPLICATION OF DATA ANALYSIS TECHNIQUES TO NUCLEAR REACTOR SYSTEMS CODE ACCURACY ASSESSMENT) has been developed by the authors to provide quantitative comparisons between nuclear reactor systems. 1. INTRODUCTION In recent years, the commercial nuclear reactor industry has focused significant

  10. 77 FR 38742 - Non-Power Reactor License Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-29

    ...3150-AI96 Non-Power Reactor License Renewal AGENCY: Nuclear Regulatory Commission...and Rulemaking, Office of Nuclear Reactor Regulation, Mail Stop...and Rulemaking, Office of Nuclear Reactor Regulation. [FR...

  11. 10 CFR 1.44 - Office of New Reactors.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ... false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR REGULATORY COMMISSION...1.44 Office of New Reactors. The Office of New...or the safeguarding of nuclear reactor facilities licensed...

  12. 77 FR 60039 - Non-Power Reactor License Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-10-02

    ...3150-AI96 Non-Power Reactor License Renewal AGENCY: Nuclear Regulatory Commission...and Rulemaking, Office of Nuclear Reactor Regulation, U.S. Nuclear...and Rulemaking, Office of Nuclear Reactor Regulation. [FR...

  13. 10 CFR 1.44 - Office of New Reactors.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ... false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR REGULATORY COMMISSION...1.44 Office of New Reactors. The Office of New...or the safeguarding of nuclear reactor facilities licensed...

  14. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ...2011-01-01 2011-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43... Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a) Develops,...

  15. 10 CFR 1.44 - Office of New Reactors.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ... false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR REGULATORY COMMISSION...1.44 Office of New Reactors. The Office of New...or the safeguarding of nuclear reactor facilities licensed...

  16. 10 CFR 1.44 - Office of New Reactors.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ... false Office of New Reactors. 1.44 Section 1.44 Energy NUCLEAR REGULATORY COMMISSION...1.44 Office of New Reactors. The Office of New...or the safeguarding of nuclear reactor facilities licensed...

  17. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ...2013-01-01 2013-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43... Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a) Develops,...

  18. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-09

    ...Licensing Examination Standards for Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft NUREG...New Reactors; or Timothy Kolb, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington,...

  19. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ...2010-01-01 2010-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43... Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a) Develops,...

  20. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ...2012-01-01 2012-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43... Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a) Develops,...

  1. 10 CFR 1.43 - Office of Nuclear Reactor Regulation.

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ...2014-01-01 2014-01-01 false Office of Nuclear Reactor Regulation. 1.43 Section 1.43... Program Offices § 1.43 Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation— (a) Develops,...

  2. REMOVAL OF CREOSOTE FROM SOIL BY BIOSLURRY REACTORS

    EPA Science Inventory

    Biological slurry reactors were tested for removal of polynuclear aromatic hydrocarbons (PAHs) from creosote contaminated soil. ive bioslurry reactors, operated in parallel, kept the soil aerated, partially suspended and well mixed. he reactors were inoculated with indigenous mic...

  3. 75 FR 36715 - Advisory Committee on Reactor Safeguards; Meeting

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-06-28

    ...Advisory Committee on Reactor Safeguards (ACRS...Margin Analysis for New Reactors Based on PRA...Margin Analysis for New Reactors Based on PRA...Respect to Emerging Technology in Nuclear Power Plants,'' and...

  4. 75 FR 51500 - Advisory Committee on Reactor Safeguards

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-08-20

    ...on Pressurized Water Reactor Sump Performance (Open...Advanced Boiling Water Reactor (ABWR) Design (Open...Simplified Boiling Water Reactor (ESBWR) Design (Open...Respect to Emerging Technology in Nuclear Power Plants,'' and...

  5. 75 FR 21046 - Advisory Committee on Reactor Safeguards

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-04-22

    ...COMMISSION Advisory Committee on Reactor Safeguards In accordance with...the Advisory Committee on Reactor Safeguards (ACRS) will hold...interest. 1 p.m.-4 p.m.: Boiling Water Reactor (BWR) Owners Group...

  6. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    SciTech Connect

    Chang, Y.I.

    1992-07-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  7. A next-generation reactor concept: The Integral Fast Reactor (IFR)

    SciTech Connect

    Chang, Y.I.

    1992-01-01

    The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

  8. Research on plasma core reactors

    NASA Technical Reports Server (NTRS)

    Jarvis, G. A.; Barton, D. M.; Helmick, H. H.; Bernard, W.; White, R. H.

    1976-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with one-meter diameter by one-meter long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17 cm thick by 89 cm diameter beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF6 container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000 cu cm aluminum canister in the central region was fueled with UF6 gas and fission density distributions determined. These results are to be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  9. Simulated nuclear reactor fuel assembly

    DOEpatents

    Berta, Victor T. (Idaho Falls, ID)

    1993-01-01

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

  10. Simulated nuclear reactor fuel assembly

    DOEpatents

    Berta, V.T.

    1993-04-06

    An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end.

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

    Microsoft Academic Search

    Rosenthal; Murray Wilford

    2009-01-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and

  12. URANIUM TRIOXIDE IN A FLUIDIZED BED REACTOR

    Microsoft Academic Search

    W. C. Philoon; E. F. Sanders; W. T. Trask

    1960-01-01

    A continuous method for the production of uranium trioxide from uranyl ; nitrate solution was developed, and its advantages over the batch process pointed ; out. Uranyl nitrate and fluidizing air are sprayed into the reactor, heat being ; supplied by molten heat transfer salt. The UOâ product overflows from the ; reactor into a packaging station. The best range

  13. New Control Techniques for Nerva Reactors

    Microsoft Academic Search

    H. S. Kirschbaum; W. R. Morris; G. L. Hohmann

    1966-01-01

    The development of the NERVA reactor has reached the point where it can be operated predictably for sustained periods of time. This success in mechanical and nuclear design has given to the control engineers the opportunity to design and experimentally verify improvements in reactor control. This paper presents a summary of the control systems used to test the early KIWI

  14. CONTROL ELEMENTS FOR SODIUM GRAPHITE REACTORS

    Microsoft Academic Search

    Shaw

    1961-01-01

    An investigation of three control element designs for sodium-graphite ; reactors is presented: the first design utilizes wire rope supporting a rod of ; neutron absorber material, permitting installation of the complete actuator in ; the upper end of a control rod thimble below the reactor loading face; the second ; concept uses overlapping fuel and absorber elements, enabling the

  15. Fusion reactor design studies. [ARIES Tokamak

    SciTech Connect

    Emmert, G.A.; Kulcinski, G.L.; Santarius, J.F.

    1990-10-12

    This report discusses the following topics on the ARIES tokamak: systems; plasma power balance; impurity control and fusion ash removal; fusion product ripple loss; energy conversion; reactor fueling; first wall design; shield design; reactor safety; and fuel cost and resources. (LSP)

  16. Reactor assessments of advanced bumpy torus configurations

    Microsoft Academic Search

    N. A. Uckan; L. W. Owen; D. A. Spong; R. L. Miller; W. B. Ard; J. F. Pipkins; R. J. Schmitt

    1984-01-01

    Recently, several innovative approaches were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator - snakey torus). Preliminary evaluations of reactor implications of each approach have been carried

  17. Reactor assessments of advanced bumpy torus configurations

    Microsoft Academic Search

    N. A. Uckan; L. W. Owen; D. A. Spong; R. L. Miller; W. B. Ard; J. F. Pipkins; R. J. Schmitt

    1983-01-01

    Recently, several configurational approaches and concept improvement schemes were introduced for enhancing the performance of the basic ELMO Bumpy Torus (EBT) concept and for improving its reactor potential. These configurations include planar racetrack and square geometries, Andreoletti coil systems, and bumpy torus-stellarator hybrids (which include twisted racetrack and helical axis stellarator-snakey torus). Preliminary evaluations of reactor implications of each of

  18. Agglomeration characteristics of fast reactor HCDA aerosols

    Microsoft Academic Search

    G. W. Parker; G. E. Creek; A. L. Jr. Sutton

    1978-01-01

    The behavior of vaporized mixed oxide fuel aerosols postulated to result from fast reactor core disruptive accidents is a subject which is intensely evaluated in fast reactor safety analysis, containment design, and site selection licensing procedure. In this program, surrogate uranium oxide aerosols produced by vapor condensation of super-heated liquid UOâ have been produced in a variety of ways and

  19. Nonlinear, inelastic fast reactor subassembly interaction analyses

    Microsoft Academic Search

    W. H. Sutherland; F. E. Bard

    1983-01-01

    Liquid Metal Fast Breeder Reactor (LMFBR) core structural design is complicated by the trade-offs associated with keeping the subassemblies closely packed for the neutronic considerations and accommodating the volumetric changes associated with irradiation swelling. The environmental variation across the reactor core results in temperature and neutron flux gradients across the subassemblies which in turn cause the subassemblies to bow as

  20. Review of reactor graphite distortion problems

    Microsoft Academic Search

    K. D. Coughren; F. J. Kempf; C. A. Munro

    1962-01-01

    The purpose of this document is primarily to discuss the effects of current retubing programs on the graphite moderator of the K Reactors. A secondary purpose of this report is to present general information related to graphite arrangement and keying patterns at the other operating reactors and to review, in a limited way, the types of graphite distortion observed at

  1. Heat pipe thermionic reactor shield optimization studies

    Microsoft Academic Search

    Vahé Keshishan; Terry E. Dix

    1992-01-01

    Shield optimization studies were conducted for a thermionic reactor, that uses heat pipes for both reactor heat removal and radiator. The radiator was placed on the opposite side of the payload to more efficiency reject the heat without affecting the LiH shadow shield. Neutron scattering off the radiator was an important consideration. The shield that was added to reduce the

  2. Reactor for production of U-233

    Microsoft Academic Search

    L. W. Lang; R. L. Stetson

    1983-01-01

    The production of a novel nuclear fuel utilizing clean uranium 233 in combination with other nuclear materials is made possible by utilization of an equally novel reactor configuration and method of operation. Clean uranium 233 is produced from thorium in a light water reactor while utilizing discrete separation of the thorium being irradiated from the fissile fuel. This clean uranium

  3. Nuclear Cross Sections For Fast Reactors

    Microsoft Academic Search

    S. Yiftah; M. Sieger

    1964-01-01

    The 16-group YOM cross section set for fast reactor analysis has been widely used by fast reactor centers during the last three years. Since its publication at the end of 1960, a considerable amount of experimental results as well as some theoretical investigations have become available. Also, various 'users' have kindly sent queries and remarks on the different cross sections

  4. A Pebble Bed Reactor cross section methodology

    Microsoft Academic Search

    Nathanael H. Hudson; Abderrafi. M. Ougouag; Farzad Rahnema; Hans Gougar

    2009-01-01

    A method is presented for the evaluation of microscopic cross sections for the Pebble Bed Reactor (PBR) neutron diffusion computational models during convergence to an equilibrium (asymptotic) fuel cycle. This method considers the isotopics within a core spectral zone and the leakages from such a zone as they arise during reactor operation. The randomness of the spatial distribution of fuel

  5. Liquid metal fast breeder reactor safety

    Microsoft Academic Search

    P. Murray; J. E. Quinn; W. B. Wolfe; H. J. Larson

    1979-01-01

    The many years of operating experience with sodium reactors and test loops gives confidence that sodium breeder reactors can be operated safely and that there is a high degree of safety in LMFBR plants. The lessons learned from Three Mile Island will also provide a valuable input to the breeder program. A review has also been made of foreign LMFBR

  6. Gas-cooled fast breeder reactor studies

    Microsoft Academic Search

    J. B. Dee; P. Fortescue; J. A. Larrimore

    1973-01-01

    Recent design and assessment work performed on gas-cooled fast breeder ; reactors by Gulf General Atomic is outlined. A description is glven of the 300 ; MW(e) gas-cooled fast breeder reactor demonstration plant design, and safety ; aspects of the plant are discussed, including potential flow blockage and coolant ; depressurization accidents. In addition preliminary studies of larger GCFR ;

  7. Steam Generators for Future Fast Breeder Reactors

    Microsoft Academic Search

    R. Nandakumar; S. Athmalingam; V. Balasubramaniyan; S. C. Chetal

    2011-01-01

    The operating experience of fast reactors worldwide has highlighted the need for reliable performance of Steam Generator (SG) as it is one of the most critical components deciding the plant availability. Based on design, development and manufacture of steam generator for 500 MWe Prototype Fast Breeder Reactor (PFBR) and also the reassuring operational feedback from experimental Steam Generator Test Facility

  8. Status of fast breeder reactor safety

    Microsoft Academic Search

    1982-01-01

    The current state of knowledge of fast breeder reactors is reviewed. The primary focus is on the analysis of postulated accident sequences and on the implications to fast reactor design. The accidents considered include loss of coolant flow and transient overpower, both with a postulated failure to scram. The associated accident phenomena is considered largely related to the potential for

  9. Safety problems in fast breeder reactors

    Microsoft Academic Search

    1973-01-01

    The approach to safety analysis of nuclear reactors has generally been ; to start with high conservatism owing to insufficient knowledge of the answers to ; the safety problems and to gradually progress to a more realistic evaluation ; fully supported by experimental safety studies. In particular, safety analyses ; of the first few fast reactors to be built were

  10. Towards intrinsically safe light-water reactors

    Microsoft Academic Search

    Hannerz

    1983-01-01

    The reactor-safety issue is one of the principal problems threatening the future of the nuclear option, at least in participatory democracies. It has contributed to widespread public distrust and is the direct cause of the escalation in design complexity and quality assurance requirements that are rapidly eroding the competitive advantage of nuclear power. Redesign of the light-water reactor can eliminate

  11. LIQUID METAL FUEL REACTOR WITH RECYCLED PLUTONIUM

    Microsoft Academic Search

    F. T. Miles; T. V. Sheehan; D. H. Gurinsky; H. J. C. Kouts

    1958-01-01

    A liquid metal reactor (LMFR) fueled with recycled plutonium dissolved ; in bismuth is described. The LMFR plutonium burner discussed was designed to use ; technology developed for a proposed U²³³ breeder. The design is ; conservative in that it attempts to avoid the problem associated with two fluids ; in the reactor core, e.g., the leakage of fluids into

  12. Technical issues in fusion reactors; A review

    Microsoft Academic Search

    V. K. Rohatgi; T. Vijayan

    1989-01-01

    In this paper the issues in fusion reactor technology are examined. Rapid progress in fusion technology research in recent years can be attributed to the advances in various technologies. The commercial generation of fusion power greatly depends on the evolution and improvements in these technologies. With better understanding of plasma physics, fusion reactor designs are becoming more and more realistic

  13. Selective purge for hydrogenation reactor recycle loop

    DOEpatents

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

    2001-01-01

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

  14. ECONOMICS OF HEAVY WATER POWER REACTORS

    Microsoft Academic Search

    1962-01-01

    Three types of the heavy-water-moderated power reactor that are part of ; the Canadian development program for nuclear power are discussed. Competitive ; sources of power and the place for nuclear power in Canadian utility systems are ; suggested. The interest of other countries in heavywater power reactors is ; described. The advantages and disadvantages of heavy water as a

  15. Aerosol reactor production of uniform submicron powders

    DOEpatents

    Flagan, Richard C. (Pasadena, CA); Wu, Jin J. (Pasadena, CA)

    1991-02-19

    A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

  16. Fuel element study for Reactor Overbore Program

    Microsoft Academic Search

    J. T. Stringer; W. A. Blanton

    1960-01-01

    Recent studies have confirmed that large incentives exist for overboring the reactor process channels approximately 500 mils in the C and five old reactors under the proposed Plant Improvement Program. Conservative estimates of the incentives for overboring indicate a payout period of about two years for the proposed work, an increase in plutonium production of 15--18%, derived from increased conversion

  17. Hydrogasification reactor and method of operating same

    DOEpatents

    Hobbs, Raymond; Karner, Donald; Sun, Xiaolei; Boyle, John; Noguchi, Fuyuki

    2013-09-10

    The present invention provides a system and method for evaluating effects of process parameters on hydrogasification processes. The system includes a hydrogasification reactor, a pressurized feed system, a hopper system, a hydrogen gas source, and a carrier gas source. Pressurized carbonaceous material, such as coal, is fed to the reactor using the carrier gas and reacted with hydrogen to produce natural gas.

  18. Mechanical cutting of irradiated reactor internal components

    SciTech Connect

    Anderson, Michael G. [MOTA Corporation: 3410 Sunset Boulevard, West Columbia, SC, 29169 (United States)

    2008-01-15

    Mechanical cutting methods to volume reduce and package reactor internal components are now a viable solution for stakeholders challenged with the retirement of first generation nuclear facilities. The recent completion of the removal of the Reactor Vessel Internals (RVI) from within the Sacramento Municipal Utility District's (SMUD) Rancho Seco Nuclear Power Plant demonstrates that unlike previous methods, inclusive of plasma arc and abrasive water-jet cutting, mechanical cutting minimizes exposure to workers, costly water cleanup, and excessive secondary waste generation. Reactor internal components were segmented, packaged, and removed from the reactor building for shipment or storage, allowing the reactor cavity to be drained and follow-on reactor segmentation activities to proceed in the dry state. Area exposure rates at the work positions during the segmentation process were generally 1 mR per hr. Radiological exposure documented for the underwater segmentation processes totaled 13 person rem. The reactor internals weighing 343,000 pounds were segmented into over 200 pieces for maximum shipping package efficiency and produced 5,600 lb of stainless steel chips and shavings which were packaged in void spaces of existing disposal containers, therefore creating no additional disposal volume. Because no secondary waste was driven into suspension in the reactor cavity water, the water was free released after one pass through a charcoal bed and ion exchange filter system. Mechanical cutting techniques are capable of underwater segmentation of highly radioactive components on a large scale. This method minimized radiological exposure and costly water cleanup while creating no secondary waste.

  19. Nuclear reactor safety: Physics and engineering aspects

    Microsoft Academic Search

    G. H. Kinchin

    1982-01-01

    In order to carry out the sort of probabilistic risk analysis referred to in Professor F. R. Farmer's paper (Farmer 1981), it is necessary to have a good understanding of the processes involved in both normal and accident conditions in a nuclear reactor. Some of these processes, for a variety of different reactor systems, are considered in the following sections

  20. The First Reactor, 40th Anniversary (rev.)

    SciTech Connect

    Allardice, Corbin; Trapnell, Edward R.; Fermi, Enrico; Fermi, Laura; Williams, Robert C.

    1982-12-01

    This booklet, an updated version of the original booklet describing the first nuclear reactor, was written in honor of the 40th anniversary of the first reactor or "pile". It is based on firsthand accounts told to Corbin Allardice and Edward R. Trapnell, and includes recollections of Enrico and Laura Fermi.

  1. Aerosol reactor production of uniform submicron powders

    NASA Technical Reports Server (NTRS)

    Flagan, Richard C. (Inventor); Wu, Jin J. (Inventor)

    1991-01-01

    A method of producing submicron nonagglomerated particles in a single stage reactor includes introducing a reactant or mixture of reactants at one end while varying the temperature along the reactor to initiate reactions at a low rate. As homogeneously small numbers of seed particles generated in the initial section of the reactor progress through the reactor, the reaction is gradually accelerated through programmed increases in temperature along the length of the reactor to promote particle growth by chemical vapor deposition while minimizing agglomerate formation by maintaining a sufficiently low number concentration of particles in the reactor such that coagulation is inhibited within the residence time of particles in the reactor. The maximum temperature and minimum residence time is defined by a combination of temperature and residence time that is necessary to bring the reaction to completion. In one embodiment, electronic grade silane and high purity nitrogen are introduced into the reactor and temperatures of approximately 770.degree. K. to 1550.degree. K. are employed. In another embodiment silane and ammonia are employed at temperatures from 750.degree. K. to 1800.degree. K.

  2. Low power reactor for remote applications

    Microsoft Academic Search

    K. L. Meier; R. G. Palmer; W. L. Kirchner

    1985-01-01

    A compact, low power reactor is being designed to provide electric power for remote, unattended applications. Because of the high fuel and maintenance costs for conventional power sources such as diesel generators, a reactor power supply appears especially attractive for remote and inaccessible locations. Operating at a thermal power level of 135 kWt, the power supply achieves a gross electrical

  3. Application of Reactor Antineutrinos: Neutrinos for Peace

    NASA Astrophysics Data System (ADS)

    Suekane, F.

    2013-02-01

    In nuclear reactors, 239Pu are produced along with burn-up of nuclear fuel. 239Pu is subject of safeguard controls since it is an explosive component of nuclear weapon. International Atomic Energy Agency (IAEA) is watching undeclared operation of reactors to prevent illegal production and removal of 239Pu. In operating reactors, a huge numbers of anti electron neutrinos (?) are produced. Neutrino flux is approximately proportional to the operating power of reactor in short term and long term decrease of the neutrino flux per thermal power is proportional to the amount of 239Pu produced. Thus rector ?'s carry direct and real time information useful for the safeguard purposes. Since ? can not be hidden, it could be an ideal medium to monitor the reactor operation. IAEA seeks for novel technologies which enhance their ability and reactor neutrino monitoring is listed as one of such candidates. Currently neutrino physicists are performing R&D of small reactor neutrino detectors to use specifically for the safeguard use in response to the IAEA interest. In this proceedings of the neutrino2012 conference, possibilities of such reactor neutrinos application and current world-wide R&D status are described.

  4. Scaling study for SP100 reactor technology

    Microsoft Academic Search

    A. C. Marshall; B. McKissock; OH Cleveland

    1989-01-01

    Several ways were explored of extending SP-100 reactor technology to higher power levels. One approach was to use the reference SP-100 pin design and increase the fuel pin length and the number of fuel pins as needed to provide higher capability. The impact on scaling of a modified and advanced SP-100 reactor technology was also explored. Finally, the effect of

  5. James P. Mosquera Director, Reactor Plant Components

    E-print Network

    of the application of nuclear reactor power to capital ships of the U.S. Navy, and other assigned projects. Mr for steam generator technology (within the Nuclear Components Division); and power plant systems engineerJames P. Mosquera Director, Reactor Plant Components and Auxiliary Equipment Naval Sea Systems

  6. Modularity Approach Modular Pebble Bed Reactor (MPBR)

    E-print Network

    · On--line Refueling #12;4/23/03 MIT NED MPBR Reference Plant Modular Pebble Bed Reactor Thermal Power4/23/03 MIT NED MPBR Modularity Approach of the Modular Pebble Bed Reactor (MPBR) Marc Berte Professor Andrew Kadak Massachusetts Institute of Technology Nuclear Engineering Department Nuclear Energy

  7. SOLASE: a conceptual laser fusion reactor design

    Microsoft Academic Search

    R. W. Conn; S. I. Abdel-Khalik; G. A. Moses

    1977-01-01

    The SOLASE conceptual laser fusion reactor has been designed to elucidate the technological problems posed by inertial confinement fusion reactors. This report contains a detailed description of all aspects of the study including the physics of pellet implosion and burn, optics and target illumination, last mirror design, laser system analysis, cavity design, pellet fabrication and delivery, vacuum system requirements, blanket

  8. Operational control of boiling water reactor stability

    Microsoft Academic Search

    C. M. Mowry; I. Nir; D. W. Newkirk

    1995-01-01

    Boiling water reactor cores are susceptible to instabilities, which generate power oscillations. Specific reactor operating practices can provide a mechanism for control of the instability phenomenon. An axial separation of the core into a single-phase region and a two-phase region resolves the influence of axial flux shapes on core stability. This separation provides the means to derive a core stability

  9. Stability monitoring for boiling water reactors

    Microsoft Academic Search

    Miguel Cecenas-Falcon

    1999-01-01

    A methodology is presented to evaluate the stability properties of Boiling Water Reactors based on a reduced order model, power measurements, and a non-linear estimation technique. For a Boiling Water Reactor, the feedback reactivity imposed by the thermal-hydraulics has an important effect in the system stability, where the dominant contribution to this feedback reactivity is provided by the void reactivity.

  10. Dynamic behavior of boiling water reactors

    Microsoft Academic Search

    March-Leuba

    1984-01-01

    A study of the basic processes involved in boiling water nuclear reactor dynamics is presented. The main emphasis of this research was placed on the physical interpretation of these processes. It is shown that this type of reactor has two regimes of operation: linear, during normal operation, and nonlinear, if they become unstable due to the thermohydraulic feedback. Both of

  11. Coatings for fast breeder reactor components

    Microsoft Academic Search

    R. Johnson

    1984-01-01

    Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on

  12. The Economics of Fast Breeder Reactors

    Microsoft Academic Search

    M. Rapin; F. J. Barclay; R. H. Allardice

    1990-01-01

    The overall status of the fast breeder reactor (FBR) system is periodically reviewed in France. In 1983, a report was prepared on the status and prospects of the FBR system at the request of the then Minister of Industry. Five years later, Electricite de France (EdF) and the French Atomic Energy Commission (CEA) jointly updated this report. The FBR reactor

  13. BURNABLE POISONS IN SMALL POWER REACTORS

    Microsoft Academic Search

    A. M. Smith; J. Jeffrey

    1959-01-01

    In the design of small reactors suitable for propulsion purposes the ; need for a large amount of excess reactivity to compensate for the effects of ; depletion and fission product poisoning leads to a difficult engineering problem ; in the accommodation of an adequate amount of control rods to shut the reactor ; down. Moreover, unless a complex control

  14. Soluble poison flux quenching in nuclear reactors

    Microsoft Academic Search

    P. Ravetto; B. D. Ganapol

    1990-01-01

    Recently, there have been significant developments in the area of inherently safe nuclear reactor conceptual designs. Among these there is much interest in the so-called PIUS reactor, where safety is to be guaranteed by the timely introduction of borated water into the core. In the event of an accidental reactivity insertion followed by a power excursion, the poison introduction would

  15. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2012 CFR

    2012-01-01

    ...associated bases for other types of nuclear power reactors. II. Explanation of Terms...Containment Structures for Nuclear Reactors,” March 16, 1972. In addition...to the Director, Office of Nuclear Reactor Regulation or Director,...

  16. 75 FR 79423 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-20

    ...attention of the Director, Office of Nuclear Reactor Regulation under 10 CFR 50...21. The Director, Office of Nuclear Reactor Regulation, may, in writing...writing to the Director, Office of Nuclear Reactor Regulation, U.S....

  17. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2013 CFR

    2013-01-01

    ...associated bases for other types of nuclear power reactors. II. Explanation of Terms...Containment Structures for Nuclear Reactors,” March 16, 1972. In addition...to the Director, Office of Nuclear Reactor Regulation or Director,...

  18. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2014 CFR

    2014-01-01

    ...associated bases for other types of nuclear power reactors. II. Explanation of Terms...Containment Structures for Nuclear Reactors,” March 16, 1972. In addition...to the Director, Office of Nuclear Reactor Regulation or Director,...

  19. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2011 CFR

    2011-01-01

    ...associated bases for other types of nuclear power reactors. II. Explanation of Terms...Containment Structures for Nuclear Reactors,” March 16, 1972. In addition...to the Director, Office of Nuclear Reactor Regulation or Director,...

  20. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Code of Federal Regulations, 2010 CFR

    2010-01-01

    ...associated bases for other types of nuclear power reactors. II. Explanation of Terms...Containment Structures for Nuclear Reactors,” March 16, 1972. In addition...to the Director, Office of Nuclear Reactor Regulation or Director,...

  1. 75 FR 70042 - In the Matter of All Power Reactor Licensees and Research Reactor Licensees Who Transport Spent...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-11-16

    ...attention of the Director, Office of Nuclear Reactor Regulation under 10 CFR 50...21. The Director, Office of Nuclear Reactor Regulation, may, in writing...writing to the Director, Office of Nuclear Reactor Regulation, U.S....

  2. Burning of Reactor Grade Plutonium Mixed with Thorium in a Hybrid Reactor

    NASA Astrophysics Data System (ADS)

    Ac?r, Adem; Übeyli, Mustafa

    2007-09-01

    Significant amount of plutonium have been discharged and accumulated from the conventional LWRs and CANDU reactors. Reducing this reactor grade (RG) plutonium is very important because it may be misused and/or released accidentally into the environment. Fusion-fission (hybrid) reactors have strong potential on burning plutonium effectively. This study presents the burning of RG plutonium mixed with thorium in a hybrid reactor for an operation period of 24 months. The effect of various fuel mixtures (98% ThO2 + 2% RG-PuO2, 94% ThO2 + 6% RG-PuO2 and 90% ThO2 + 10% RG-PuO2) and coolants (Flinabe, natural lithium and Li20Sn80) on the reactor's performance was investigated. Numerical results showed that utilization of RG plutonium in the mixed fuel in such a hybrid reactor not only enhanced the reactor's performance but also reduced its 239Pu content significantly.

  3. Reactivity control assembly for nuclear reactor. [LMFBR

    DOEpatents

    Bollinger, L.R.

    1982-03-17

    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.

  4. Fast-acting nuclear reactor control device

    SciTech Connect

    Kotlyar, O.M.; West, P.B.

    1992-12-31

    This invention consists of a fast-acting nuclear reactor control device for moving and positioning a safety control rod to desired elevations within the core of the reactor between a run position in which the safety control rod is outside the reactor core, and a shutdown position in which the rod is fully inserted in the reactor core. The device employs a hydraulic pump motor, an electric gear motor, and a solenoid valve to drive the safety control rod into the reactor core through the entire stroke of the safety control rod. An overrunning clutch, allows the safety control rod to freely travel toward a safe position in the event of a partial drive system failure.

  5. Status of the US stellarator reactor study

    SciTech Connect

    Lyon, J.F. [Oak Ridge National Lab., TN (United States); Gulec, K. [Univ. of Tennessee, Knoxville (United States); Miller, R.L. [Los Alamos National Lab., NM (United States); El-Guebaly, L. [Univ. of Wisconsin, Madison (United States)

    1994-09-01

    Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the ARIES costing algorithms is 62.5 mill/kW(e)h in constant 1992 dollars for a 1-GW(e) Compact Torsatron reactor reference case. The COE is relatively insensitive (< 10% variation) over a wide range of assumptions including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio.

  6. Transient analyses of multimegawatt space reactors

    NASA Astrophysics Data System (ADS)

    Hanan, Nelson A.; Bhattacharyya, Samit K.

    Several systems have been proposed for the generation of multimegawatt power for space applications. In some concepts, a nuclear reactor is proposed to meet all the power requirements; for other concepts, a nuclear reactor is proposed to deliver the housekeeping and alert power, and other systems are proposed to meet the burst power requirements. A common requirement for all the nuclear concepts is a fast response for change of power level from the housekeeping mode to either the alert or burst mode of operation. First, the requirements imposed on the control system of two reactor concepts, a Liquid Metal Reactor and a gas cooled reactor, having as its main objective the determination of the maximum input reactivity rates and control drum speed to rotation necessary to meet the power requirements, are presented. Next, preliminary analysis to validate the point kinetics against detailed space-time kinetics calculations is discussed.

  7. Advanced PPA Reactor and Process Development

    NASA Technical Reports Server (NTRS)

    Wheeler, Raymond; Aske, James; Abney, Morgan B.; Miller, Lee A.; Greenwood, Zachary

    2012-01-01

    Design and development of a second generation Plasma Pyrolysis Assembly (PPA) reactor is currently underway as part of NASA s Atmosphere Revitalization Resource Recovery effort. By recovering up to 75% of the hydrogen currently lost as methane in the Sabatier reactor effluent, the PPA helps to minimize life support resupply costs for extended duration missions. To date, second generation PPA development has demonstrated significant technology advancements over the first generation device by doubling the methane processing rate while, at the same time, more than halving the required power. One development area of particular interest to NASA system engineers is fouling of the PPA reactor with carbonaceous products. As a mitigation plan, NASA MSFC has explored the feasibility of using an oxidative plasma based upon metabolic CO2 to regenerate the reactor window and gas inlet ports. The results and implications of this testing are addressed along with the advanced PPA reactor development work.

  8. [Fission Working Group -- Molten salt reactors

    SciTech Connect

    Gat, U.; Engel, J.R.

    1992-01-01

    This report provides an assessment of molten salt reactors (MSRs) which are fluid fuel reactors and, as such, have several unique features, some which are important to the burning of fissile material from dismantled weapons. This material can be added on-line during operation in either continuous or batch form. The added fuel need only be in an acceptable chemical form, but no fuel manufacturing or minimum discrete amounts for a fuel element are required. Fluid fuel reactors can have partial or full on-line fuel processing. When online fuel processing is utilized, a particular fuel component, for example the plutonium, can be burned completely, or in some sense can be converted to other kinds of fuel, for example into [sup 233]U. There is no equivalent of fuel burnup in continuous processing reactors, and no need for reprocessing in external plants and manufacturing of fuel elements, transportation, and reinsertion in the reactor.

  9. [Fission Working Group -- Molten salt reactors

    SciTech Connect

    Gat, U.; Engel, J.R.

    1992-12-31

    This report provides an assessment of molten salt reactors (MSRs) which are fluid fuel reactors and, as such, have several unique features, some which are important to the burning of fissile material from dismantled weapons. This material can be added on-line during operation in either continuous or batch form. The added fuel need only be in an acceptable chemical form, but no fuel manufacturing or minimum discrete amounts for a fuel element are required. Fluid fuel reactors can have partial or full on-line fuel processing. When online fuel processing is utilized, a particular fuel component, for example the plutonium, can be burned completely, or in some sense can be converted to other kinds of fuel, for example into {sup 233}U. There is no equivalent of fuel burnup in continuous processing reactors, and no need for reprocessing in external plants and manufacturing of fuel elements, transportation, and reinsertion in the reactor.

  10. Self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M. (San Jose, CA); Brummond, Willian A (Livermore, CA); Peterson, Leslie F. (Danville, CA)

    1988-01-01

    A control system for the automatic or self-actuated shutdown or "scram" of a nuclear reactor. The system is capable of initiating scram insertion by a signal from the plant protection system or by independent action directly sensing reactor conditions of low-flow or over-power. Self-actuation due to a loss of reactor coolant flow results from a decrease of pressure differential between the upper and lower ends of an absorber element. When the force due to this differential falls below the weight of the element, the element will fall by gravitational force to scram the reactor. Self-actuation due to high neutron flux is accomplished via a valve controlled by an electromagnet and a thermionic diode. In a reactor over-power, the diode will be heated to a change of state causing the electromagnet to be shorted thereby actuating the valve which provides the changed flow and pressure conditions required for scramming the absorber element.

  11. Nuclear reactor internals alignment configuration

    DOEpatents

    Gilmore, Charles B. (Greensburg, PA); Singleton, Norman R. (Murrysville, PA)

    2009-11-10

    An alignment system that employs jacking block assemblies and alignment posts around the periphery of the top plate of a nuclear reactor lower internals core shroud to align an upper core plate with the lower internals and the core shroud with the core barrel. The distal ends of the alignment posts are chamfered and are closely received within notches machined in the upper core plate at spaced locations around the outer circumference of the upper core plate. The jacking block assemblies are used to center the core shroud in the core barrel and the alignment posts assure the proper orientation of the upper core plate. The alignment posts may alternately be formed in the upper core plate and the notches may be formed in top plate.

  12. Power reactor events and issues

    SciTech Connect

    Not Available

    1992-10-01

    This publication reviews selected operating events that have occurred at nuclear power plants and presents the results of NRC-sponsored analysis of pertinent operating issues. This study was initiated following several instances of water hammer involving the service water system at Arkansas Nuclear One. The task was to evaluate the need to reissue previous NRC guidance about water hammer or to suggest additional measures to prevent or mitigate their occurrence. Twelve from PWR and BWR type reactors were studied. This study concluded that the frequency of reported water hammer occurrences continues to drop and no new phenomena were identified as causes of water hammer. In addition, this study supports prior NRC conclusions regarding water hammer; however, some aspects that could impact safety and were identified in this study had not been previously emphasized.

  13. Liquid-metal-cooled reactor

    DOEpatents

    Hutter, Ernest (Wilmette, IL)

    1982-01-01

    A perforated depressor plate extending across the bottom of the instrument ree of a fast breeder reactor cooperates with a circular cylindrical metal bellows forming a part of the upper adapter of each core assembly and bearing on the bottom of the depressor plate to restrict flow of coolant between core assemblies, thereby reducing significantly the pressure differential between the coolant inside the core assemblies and the coolant outside of the core assemblies. Openings in the depressor plate are slightly smaller than the top of the upper adapter so the depressor plate will serve as a backup mechanical holddown for the core. In addition coolant mixing devices and locating devices are provided attached to the depressor plate.

  14. Nuclear reactor composite fuel assembly

    DOEpatents

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

    1980-01-01

    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.

  15. Reactor pressure vessel vented head

    DOEpatents

    Sawabe, J.K.

    1994-01-11

    A head for closing a nuclear reactor pressure vessel shell includes an arcuate dome having an integral head flange which includes a mating surface for sealingly mating with the shell upon assembly therewith. The head flange includes an internal passage extending therethrough with a first port being disposed on the head mating surface. A vent line includes a proximal end disposed in flow communication with the head internal passage, and a distal end disposed in flow communication with the inside of the dome for channeling a fluid therethrough. The vent line is fixedly joined to the dome and is carried therewith when the head is assembled to and disassembled from the shell. 6 figures.

  16. Reactor operation environmental information document

    SciTech Connect

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

    1989-12-01

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

  17. Antineutrinos from nuclear reactors: recent oscillation measurements

    NASA Astrophysics Data System (ADS)

    Dwyer, D. A.

    2015-02-01

    Nuclear reactors are the most intense man-made source of antineutrinos, providing a useful tool for the study of these particles. Oscillation due to the neutrino mixing angle {{? }13} is revealed by the disappearance of reactor {{\\bar{? }}e} over ˜km distances. Use of additional identical detectors located near nuclear reactors reduce systematic uncertainties related to reactor {{\\bar{? }}e} emission and detector efficiency, significantly improving the sensitivity of oscillation measurements. The Double Chooz, RENO, and Daya Bay experiments set out in search of {{? }13} using these techniques. All three experiments have recently observed reactor {{\\bar{? }}e} disappearance, and have estimated values for {{? }13} of 9.3? ± 2.1°, 9.2? ± 0.9°, and 8.7? ± 0.4° respectively. The energy-dependence of {{\\bar{? }}e} disappearance has also allowed measurement of the effective neutrino mass difference, \\mid ? mee2\\mid ? \\mid ? m312\\mid . Comparison with \\mid ? m? ? 2\\mid ? \\mid ? m322\\mid from accelerator {{? }? } measurements supports the three-flavor model of neutrino oscillation. The current generation of reactor {{\\bar{? }}e} experiments are expected to reach ˜3% precision in both {{? }13} and \\mid ? mee2\\mid . Precise knowledge of these parameters aids interpretation of planned {{? }? } measurements, and allows future experiments to probe the neutrino mass hierarchy and possible CP-violation in neutrino oscillation. Absolute measurements of the energy spectra of {{\\bar{? }}e} deviate from existing models of reactor emission, particularly in the range of 5–7 MeV.

  18. Flexible Conversion Ratio Fast Reactor Systems Evaluation

    SciTech Connect

    Neil Todreas; Pavel Hejzlar

    2008-06-30

    Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

  19. Transients in reactors for power systems compensation

    NASA Astrophysics Data System (ADS)

    Abdul Hamid, Haziah

    This thesis describes new models and investigations into switching transient phenomena related to the shunt reactors and the Mechanically Switched Capacitor with Damping Network (MSCDN) operations used for reactive power control in the transmission system. Shunt reactors and MSCDN are similar in that they have reactors. A shunt reactor is connected parallel to the compensated lines to absorb the leading current, whereas the MSCDN is a version of a capacitor bank designed as a C-type filter for use in the harmonic-rich environment. In this work, models have been developed and transient overvoltages due to shunt reactor deenergisation were estimated analytically using MathCad, a mathematical program. Computer simulations used the ATP/EMTP program to reproduce both single-phase and three-phase shunt reactor switching at 275 kV operational substations. The effect of the reactor switching on the circuit breaker grading capacitor was also examined by considering various switching conditions.. The main original achievement of this thesis is the clarification of failure mechanisms occurring in the air-core filter reactor due to MSCDN switching operations. The simulation of the MSCDN energisation was conducted using the ATP/EMTP program in the presence of surge arresters. The outcome of this simulation shows that extremely fast transients were established across the air-core filter reactor. This identified transient event has led to the development of a detailed air-core reactor model, which accounts for the inter-turn RLC parameters as well as the stray capacitances-to-ground. These parameters are incorporated into the transient simulation circuit, from which the current and voltage distribution across the winding were derived using electric field and equivalent circuit modelling. Analysis of the results has revealed that there are substantial dielectric stresses imposed on the winding insulation that can be attributed to a combination of three factors. (i) First, the surge arrester operation during the MSCDN energisation, which causes steep voltage change at the reactor terminal. (ii) Second, the nonuniform voltage distribution, resulting in high stresses across the top inter-turn windings. (iii) Third, the rapid rate-of-change of voltage in the assumed worst-case reactor winding location. This is accompanied by a high dielectric current through the inter-turn winding insulation..

  20. Advances in ICF power reactor design

    SciTech Connect

    Hogan, W.J.; Kulcinski, G.L.

    1985-04-17

    Fifteen ICF power reactor design studies published since 1980 are reviewed to illuminate the design trends they represent. There is a clear, continuing trend toward making ICF reactors inherently safer and environmentally benign. Since this trend accentuates inherent advantages of ICF reactors, we expect it to be further emphasized in the future. An emphasis on economic competitiveness appears to be a somewhat newer trend. Lower cost of electricity, smaller initial size (and capital cost), and more affordable development paths are three of the issues being addressed with new studies.

  1. Reactor power system/spacecraft integration

    NASA Technical Reports Server (NTRS)

    Elms, R. V.

    1985-01-01

    The new national initiative in space reactor technology evaluation and development is strongly tied to mission applications and to spacecraft and space transportation system (STS) compatibility. This paper discusses the power system integration interfaces with potential using spacecraft and the STS, and the impact of these requirements on the design. The integration areas of interest are mechanical, thermal, electrical, attitude control, and mission environments. The mission environments include space vacuum, solar input, heat sink, space radiation, weapons effects, and reactor power system radiation environments. The natural, reactor, and weapons effects radiation must be evaluated and combined to define the design requirements for spacecraft electronic equipment.

  2. Safety design of prototype fast breeder reactor

    SciTech Connect

    Bhoje, S.B.; Chetal, S.C.; Singh, Om Pal [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    2004-07-01

    The basic design and safety design of Prototype Fast Breeder Reactor (PFBR) is presented. Design aspects covered include safety classification, seismic categorization, design basis conditions, design safety limits, core physics, core monitoring, shutdown system, decay heat removal system, protection against sodium leaks and tube leaks in steam generator, plant layout, radiation protection, event analysis, beyond design basis accidents, integrity of primary containment, reactor containment building and design pressure resulting from core disruptive accident. The measures provided in the design represent a robust case of the safety of the reactor. (authors)

  3. Integral Fast Reactor concept inherent safety features

    SciTech Connect

    Marchaterre, J.F.; Sevy, R.H.; Cahalan, J.E.

    1986-01-01

    The Integral Fast Reactor (IFR) is an innovative 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. The design features that together fulfill these goals are: (1) a liquid metal (sodium) coolant, (2) a pool-type reactor primary system configuration, (3) an advanced ternary alloy metallic fuel, and (4) an integral fuel cycle. This paper reviews the design features that contribute to the safety margins inherent to the IFR concept. Special emphasis is placed on the ability of the IFR design to accommodate anticipated transients without scram (ATWS).

  4. The Integral Fast Reactor (IFR) concept

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1989-01-01

    In addition to maintaining the viability of its present commercial nuclear technology, a principal challenge in the US in the 1990s and beyond will be to regain and maintain a position among the world leadership in advanced reactor research and development. In this paper we'll discuss factors which we believe should today provide the rationale and focus for advanced reactor R and D, and we will then review the status of the major US effort, the Integral Fast Reactor (IFR) program.

  5. Liquid metal-cooled nuclear reactor

    SciTech Connect

    Costes, D.

    1985-04-02

    The invention relates to a fast neutron nuclear reactor in which the primary pumps and exchangers are suspended on the rigid slab sealing the vessel containing the reactor core. The slab has in its thickness housings of reduced dimensions in which are confined the heads of the exchangers and the primary pumps. The flywheels of the pumps and part of the pipes of the secondary circuits are also contained in the housings. Other housings can be provided in the slab, particularly for the handling of fuels. Application to the improvement of the safety and reliability of fast neutron reactors is taught.

  6. Nuclear reactor fissile isotopes antineutrino spectra

    E-print Network

    V. Sinev

    2012-07-30

    Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using specific method. They could be used for neutrino experiments data analysis made at any fuel composition of reactor core.

  7. Optimizing N reactor fuel spacer disposal

    SciTech Connect

    Duncan, G.M.; Day, R.S. [Bechtel Hanford, Inc., Richland, WA (United States); Trumble, H.R. [Dept. of Energy, Richland, WA (United States)

    1996-12-31

    Deactivation of the Hanford Site`s N Reactor included the disposal of a large number of radioactive reactor fuel spacers buried in silos outside the reactor building. Optimizing the disposal effort involved simultaneously solving objectives of low cost, schedule, dose minimization, and contamination control while satisfying a variety of regulatory requirements. The selected approach included removing the radioactive fuel spacers by crane-mounted electromagnet, spray painting for airborne control, using grouted large-diameter pipe sections for the burial container, and shipping by truck. Cost and schedule goals were achieved, doses were higher than predicted mainly due to delays in operations, surface contamination was minimized. and airborne contamination was precluded.

  8. Optimized Conditioning of Activated Reactor Graphite

    SciTech Connect

    Tress, G.; Doehring, L.; Pauli, H.; Beer, H.-F.

    2002-02-25

    The research reactor DIORIT at the Paul Scherrer Institute was decommissioned in 1993 and is now being dismantled. One of the materials to be conditioned is activated reactor graphite, approximately 45 tons. A cost effective conditioning method has been developed. The graphite is crushed to less than 6 mm and added to concrete and grout. This graphite concrete is used as matrix for embedding dismantling waste in containers. The waste containers that would have been needed for separate conditioning and disposal of activated reactor graphite are thus saved. Applying the new method, the cost can be reduced from about 55 SFr/kg to about 17 SFr/kg graphite.

  9. Joint analysis of spectral reactor neutrino experiments

    E-print Network

    V. V. Sinev

    2013-04-18

    The analysis of experiments at nuclear reactors where inverse beta decay reaction positron spectrum was measured at different distances from reactor core is presented here. It was found that there appear three enclosed zones of neutrino oscillation parameters when joint analysis is applied on the ${\\Delta}m^{2}-\\sin^{2}2{\\theta}$ plane. The parameters that found are partially crossed with similar regions originating from other non reactor experiments where they observed neutrino oscillations having unusual mass parameter about 1 eV^2 and amplitude about 0.04. Confidence level for observed regions achieves the value of 99.9%.

  10. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    SciTech Connect

    Harto, Andang Widi [Engineering Physics Department, Faculty of Engineering, Gadjah Mada University (Indonesia)

    2012-06-06

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  11. TREATMENT OF METHANOLIC WASTEWATER BY ANAEROBIC DOWN-FLOW HANGING SPONGE (ANDHS) REACTOR AND UASB REACTOR

    NASA Astrophysics Data System (ADS)

    Sumino, Haruhiko; Wada, Keiji; Syutsubo, Kazuaki; Yamaguchi, Takashi; Harada, Hideki; Ohashi, Akiyoshi

    Anaerobic down-flow hanging sponge (AnDHS) reactor and UASB reactor were operated at 30℃ for over 400 days in order to investigate the process performance and the sludge characteristics of treating methanolic wastewater (2 gCOD/L). The settings OLR of AnDHS reactor and of UASB reactor were 5.0 -10.0 kgCOD/m3/d and 5.0 kgCOD/m3/d. The average of the COD removal demonstrated by both reactors were over 90% throughout the experiment. From the results of methane producing activities and the PCR-DGGE method, most methanol was directly converted to methane in both reactors. The conversion was carried out by different methanogens: one closely related to Methanomethylovorans hollandica in the AnDHS retainted sludge and the other closely related to Methanosarcinaceae and Metanosarciales in the UASB retainted sludge.

  12. Passive compact molten salt reactor (PCMSR), modular thermal breeder reactor with totally passive safety system

    NASA Astrophysics Data System (ADS)

    Harto, Andang Widi

    2012-06-01

    Design Study Passive Compact Molten Salt Reactor (PCMSR) with totally passive safety system has been performed. The term of Compact in the PCMSR name means that the reactor system is designed to have relatively small volume per unit power output by using modular and integral concept. In term of modular, the reactor system consists of three modules, i.e. reactor module, turbine module and fuel management module. The reactor module is an integral design that consists of reactor, primary and intermediate heat exchangers and passive post shutdown cooling system. The turbine module is an integral design of a multi heating, multi cooling, regenerative gas turbine. The fuel management module consists of all equipments related to fuel preparation, fuel reprocessing and radioactive handling. The preliminary calculations show that the PCMSR has negative temperature and void reactivity coefficient, passive shutdown characteristic related to fuel pump failure and possibility of using natural circulation for post shutdown cooling system.

  13. Recent measurements supporting N reactor graphite stack stored energy analysis

    Microsoft Academic Search

    D. A. Tollefson; H. Toffer; C. L. Mohr

    1987-01-01

    During the accident at the Chernobyl reactor in the Soviet Union last April, many speculated that Wigner stored energy in the graphite stack contributed to the burning of the moderator. This speculation prompted a reevaluation of the stored-energy question in a graphite-moderated reactor such as the Hanford N Reactor. This reactor is operated by UNC Nuclear Industries, Inc., the for

  14. CRITICALITY CALCULATIONS OF THE FIXED BED NUCLEAR REACTOR

    Microsoft Academic Search

    Farhang Sefidvash; Tomas Matela; Robson Silva da Silva

    Small nuclear reactors without the need for on-site refuelling have greater simplicity, better compliance with passive safety systems, and are more adequate for countries with small electric grids and limited investment capabilities. The Fixed Bed Nuclear Reactor (FBNR) is a small 40 MWe reactor based on the Pressurized Water Reactor (PWR) technology. FBNR is an integrated primary circuit and simple

  15. Heat Exchangers for the Next Generation of Nuclear Reactors

    Microsoft Academic Search

    Li Xiuqing; Renaud Le Pierres; Stephen John Dewson

    2006-01-01

    The realisation that fossil fuel resources are finite, the associated rising price and a growing concern about greenhouse gas emissions, has resulted in renewed interest in nuclear energy. Generation IV and other programmes are looking at a variety of new reactors. These reactors vary in type from Very High Temperature Gas Cooled Reactors (VHTR) to Liquid Metal Fast Reactors (LFR

  16. Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency

    Microsoft Academic Search

    R. Wigeland; K. Hamman

    2009-01-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future

  17. A Compact Nuclear Fusion Reactor for Space Flights

    Microsoft Academic Search

    Anatoly F. Nastoyashchiy; Anatoly F

    2006-01-01

    A small-scale nuclear fusion reactor is suggested based on the concepts of plasma confinement (with a high pressure gas) which have been patented by the author. The reactor considered can be used as a power setup in space flights. Among the advantages of this reactor is the use of a D3He fuel mixture which at burning gives main reactor products

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

    E-print Network

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

    2008-08-06

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

  19. LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS

    E-print Network

    Bazhenov, Maxim

    LIMITED POWER BURSTS IN DISTRIBUTED MODELS OF NUCLEAR REACTORS M. V. Bazhenov and E. F. Sabaev UDC employed for analyzing reactor dynamics. Equations of this type are used for analyzing the stability of the reactor power, etc. Among these problems the question of the boundedness of reactor power bursts

  20. ME 361E Nuclear Reactor Engineering ABET EC2000 syllabus

    E-print Network

    Ben-Yakar, Adela

    ME 361E ­ Nuclear Reactor Engineering Page 1 ABET EC2000 syllabus ME 361E ­ Nuclear Reactor students should be able to: · Compare and contrast numerous nuclear reactor designs · Calculate the effects of nuclear fuel burnup · Summarize the mechanism that affect the control of a nuclear reactor Topics Covered

  1. Novel reactor configuration for synthesis gas conversion to alcohols

    SciTech Connect

    Akgerman, A.; Anthony, R.G.

    1991-01-01

    Our objectives during this quarter were to complete construction and shake down runs on our new reactor assemblies, obtain activity data on the catalyst employing the slurry reactor, incorporate pressure drop and change in number of moles options into our reactor simulation code and start experiments in the trickle bed reactor.

  2. Utilization of Egyptian Research Reactor and Modes of Collaboration

    Microsoft Academic Search

    M. K. Shaat; M. A. Gaheen

    The new Egyptian Research Reactor (ETRR-2) is a Material Testing Reactor (MTR) and it was commissioned in 1997. It is open pool Research Reactor (RR) using low enriched MTR fuel elements (less than 20% enrichment), cooled and moderated with light water and reflected by beryllium. The reactor power is 22 MW with high neutron flux irradiation positions (flux > 10

  3. Figure of merit for a magnetic fusion reactor

    Microsoft Academic Search

    Kazawa

    1986-01-01

    This paper deals with the economics of magnetic fusion reactors. The author proposes and defines figure of merit for a fusion reactor as a method of evaluating the economics of fusion reactors. The figures of merit are useful to compare the economics of fusion and fission reactors of various types. Using the above-mentioned formula, trial calculations were made as to

  4. Development of 1000 MWe Advanced Boiling Water Reactor

    Microsoft Academic Search

    Kazuo Hisajima; Ken Uchida; Keiji Matsumoto; Koichi Kondo; Shigeki Yokoyama; Takuya Miyagawa

    2006-01-01

    1000 MWe Advanced Boiling Water Reactor has only two main steam lines and six reactor internal pumps, whereas 1350 MWe ABWR has four main steam lines and ten reactor internal pumps. In order to confirm how the differences affect hydrodynamic conditions in the dome and lower plenum of the reactor pressure vessel, fluid analyses have been performed. The results indicate

  5. Biological Treatment of Dairy Wastewater by Sequencing Batch Reactor

    Microsoft Academic Search

    A Mohseni-Bandpi; H Bazari

    2004-01-01

    A bench scale aerobic Sequencing Batch Reactor (SBR) was investigated to treat the wastewater from an industrial milk factory. The reactor was constructed from plexi glass material and its volume was 22.5 L. The reactor was supplied with oxygen by fine bubble air diffuser. The reactor was fed with milk factory and synthetic wastewater under different operational conditions. The COD

  6. Improved reactor trip breaker reliability for pressurized water reactors

    SciTech Connect

    Redmon, J.R.; Borchart, D.J.

    1985-11-01

    Reactor trip breaker (RTB) reliability has been a major concern of the industry and the US Nuclear Regulatory Commission because of the incident with Westinghouse RTBs at Salem Unit I. A similar problem occurred at San Onofre Units 2 and 3 with General Electric (GE) RTBs. An investigation concluded that the reliability of GE AK-2-25 low voltage breakers used as RTBs can be substantially improved by replacing the undervoltage trip device with an alternating current (ac) shunt trip device. This paper proposes a design change that has been experimentally verified as sufficient torque force to reliably open the breaker and is considered failsafe. Problems with GE AK-2-25 breakers used as RTBs have been almost exclusively related to the undervoltage trip (UV) device's ability to open the breaker. RTBs typically employ two means of opening on a trip signal: (a) an undervoltage trip attachment operating in a loss-of-voltage (failsafe) mode and (b) a shunt trip attachment requiring electric power to function (diverse method). Investigation determined breaker bearing lubrication and UV device electrical/mechanical adjustments are critical to breaker openings as the UV device has marginal tripping force.

  7. Safety Features Comparative Study between Advanced Boiling Water Reactor and Supercritical Water-Cooling Reactor

    Microsoft Academic Search

    Hangding Wang; Tao Zhou; Zhousen Hou

    2011-01-01

    Supercritical water-cooling reactor is one of the advanced nuclear power plants as the fourth generation reactor of internationally recognized. The plant system design is once-through cycle, feed-water flow rate are provided from pressure head by pump through the reactor core became steam, and directly into the turbine. There is not a recirculation loop in this cycle, this is the the

  8. PBF Reactor Building (PER620). After lowering reactor vessel onto blocks, ...

    Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

    PBF Reactor Building (PER-620). After lowering reactor vessel onto blocks, it is rolled on logs into PBF. Metal framework under vessel is handling device. Various penetrations in reactor bottom were for instrumentation, poison injection, drains. Large one, below center "manhole" was for primary coolant. Photographer: Larry Page. Date: February 13, 1970. INEEL negative no. 70-736 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

  9. Using fast reactor component evaluation for pressurized water reactor life extension

    Microsoft Academic Search

    T. R. Allen; J. I. Cole; E. A. Kenik; H. Tsai; S. Ukai; S. Mizuta; T. Yoshitake

    1999-01-01

    An understanding of the effects of long-term, low-dose-rate radiation on core components is critical to light-water reactor\\u000a plant life extension. Following reactor shutdown, materials that had experienced long exposures to low-dose-rate irradiation\\u000a were retrieved from the EBR-II research reactor for analysis. These components are being analyzed to provide insight into\\u000a pressurized water reactor life extension. In this work, three examples

  10. Evolution of the liquid metal reactor: The Integral Fast Reactor (IFR) concept

    SciTech Connect

    Till, C.E.; Chang, Y.I.

    1989-01-01

    The Integral Fast Reactor (IFR) concept has been under development at Argonne National Laboratory since 1984. A key feature of the IFR concept is the metallic fuel. Metallic fuel was the original choice in early liquid metal reactor development. Solid technical accomplishments have been accumulating year after year in all aspects of the IFR development program. But as we make technical progress, the ultimate potential offered by the IFR concept as a next generation advanced reactor becomes clearer and clearer. The IFR concept can meet all three fundamental requirements needed in a next generation reactor. This document discusses these requirements: breeding, safety, and waste management. 5 refs., 4 figs.

  11. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    NASA Astrophysics Data System (ADS)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  12. Reactor physics project progress report no. 2

    E-print Network

    Driscoll, Michael J.

    1969-01-01

    This is the second annual report in an experimental and theoretical program to develop and apply single and few element heterogeneous methods for the determination of reactor lattice parameters. During the period covered ...

  13. Water-Gas Shift Membrane Reactor Studies

    E-print Network

    Coal, Petroleum coke, Biomass, Waste, etc. Gasifier Particulate Removal Air Separator Oxygen Air Steam syngas components (CO2, H2O, CO) - conduct forward WGS using a membrane reactor at favorable conditions

  14. EMERGING TECHNOLOGY BULLETIN: SPOUTED BED REACTOR

    EPA Science Inventory

    The Spouted Bed Reactor (SBR) technology utilizes the unique attributes of the "spouting " fluidization regime, which can provide heat transfer rates comparable to traditional fluid beds, while providing robust circulation of highly heterogeneous solids, concurrent with very agg...

  15. Small Scale Reactor for Ultraviolet Photochemistry

    NASA Astrophysics Data System (ADS)

    Gano, James E.; Gano, Andrew J.; Garry, Patricia; Sekher, Padmanabhan

    2002-11-01

    A simple, versatile, inexpensive, synthetic photochemical reactor is described. A variety of wavelengths, such as 214 (Zn), 229 (Cd), and 254 nm (Hg), can be applied to gram quantities of material. It can be constructed using modest glassblowing skills.

  16. Small Scale Reactor for Ultraviolet Photochemistry

    Microsoft Academic Search

    James E. Gano; Andrew J. Gano; Patricia Garry; Padmanabhan Sekher

    2002-01-01

    A simple, versatile, inexpensive, synthetic photochemical reactor is described. A variety of wavelengths, such as 214 (Zn), 229 (Cd), and 254 nm (Hg), can be applied to gram quantities of material. It can be constructed using modest glassblowing skills.

  17. Corrosion Minimization for Research Reactor Fuel

    SciTech Connect

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  18. Actinide minimization using pressurized water reactors

    E-print Network

    Visosky, Mark Michael

    2006-01-01

    Transuranic actinides dominate the long-term radiotoxity in spent LWR fuel. In an open fuel cycle, they impose a long-term burden on geologic repositories. Transmuting these materials in reactor systems is one way to ease ...

  19. Rethinking the light water reactor fuel cycle

    E-print Network

    Shwageraus, Evgeni, 1973-

    2004-01-01

    The once through nuclear fuel cycle adopted by the majority of countries with operating commercial power reactors imposes a number of concerns. The radioactive waste created in the once through nuclear fuel cycle has to ...

  20. Self-Sustaining Thorium Boiling Water Reactors

    E-print Network

    Ganda, Francesco

    A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar ...

  1. Physics design of the upgraded TREAT reactor

    SciTech Connect

    Bhattacharyya, S.K.; Lell, R.M.; Liaw, J.R.; Ulrich, A.J.; Wade, D.C.; Yang, S.T.

    1980-01-01

    With the deferral of the Safety Test Facility (STF), the TREAT Upgrade (TU) reactor has assumed a lead role in the US LMFBR safety test program for the foreseeable future. The functional requirements on TU require a significant enhancement of the capability of the current TREAT reactor. A design of the TU reactor has been developed that modifies the central 11 x 11 fuel assembly array of the TREAT reactor such as to provide the increased source of hard spectrum neutrons necessary to meet the functional requirements. A safety consequence of the increased demands on TU is that the self limiting operation capability of TREAT has proved unattainable, and reliance on a safety grade Plant Protection System is necessary to ensure that no clad damage occurs under postulated low-probability reactivity accidents. With that constraint, the physics design of TU provides a means of meeting the functional requirements with a high degree of confidence.

  2. Control system for a small fission reactor

    NASA Astrophysics Data System (ADS)

    Burelbach, J. P.; Kann, W. J.; Saiveau, J. G.

    1985-02-01

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

  3. Critical assessment of thorium reactor technology

    E-print Network

    Drenkhahn, Robert (Robert A.)

    2012-01-01

    Thorium-based fuels for nuclear reactors are being considered for use with current and future designs in both large and small-scale energy production. Thorium-232 is as abundant on Earth as lead, far more common than all ...

  4. CANDU reactor system: An appropriate technology

    Microsoft Academic Search

    J. A. L. Robertson; J. A. L

    1978-01-01

    Information on CANDU-type reactors is presented concerning design characteristics, development history, operating experience, future developments, fuel supply, nuclear waste management, and internatonal aspects of fuel cycle safeguards.

  5. N Reactor Deactivation Program Plan. Revision 4

    SciTech Connect

    Walsh, J.L.

    1993-12-01

    This N Reactor Deactivation Program Plan is structured to provide the basic methodology required to place N Reactor and supporting facilities {center_dot} in a radiologically and environmentally safe condition such that they can be decommissioned at a later date. Deactivation will be in accordance with facility transfer criteria specified in Department of Energy (DOE) and Westinghouse Hanford Company (WHC) guidance. Transition activities primarily involve shutdown and isolation of operational systems and buildings, radiological/hazardous waste cleanup, N Fuel Basin stabilization and environmental stabilization of the facilities. The N Reactor Deactivation Program covers the period FY 1992 through FY 1997. The directive to cease N Reactor preservation and prepare for decommissioning was issued by DOE to WHC on September 20, 1991. The work year and budget data supporting the Work Breakdown Structure in this document are found in the Activity Data Sheets (ADS) and the Environmental Restoration Program Baseline, that are prepared annually.

  6. Reactor antineutrino anomaly with known ?13

    NASA Astrophysics Data System (ADS)

    Zhang, C.; Qian, X.; Vogel, P.

    2013-04-01

    We revisit the reactor antineutrino anomaly using the recent reactor flux independent determination of sizable ?13 by considering the full set of the absolute reactor ?¯e flux measurements. When normalized to the predicted flux of Mueller [Phys. Rev. C 83, 054615 (2011)]PRVCAN0556-2813, the new world average, after including results from Palo Verde, Chooz, and Double Chooz, is 0.959±0.009 (experiment uncertainty) ±0.027 (flux systematics). Including the data with kilometer baseline, the new world average is only about 1.4? lower than the unity, weakening the significance of the reactor antineutrino anomaly. The upcoming results from Daya Bay, RENO, and the Double Chooz will provide further information about this issue.

  7. Reactor Antineutrino Anomaly with known ?_{13}

    E-print Network

    C. Zhang; X. Qian; P. Vogel

    2013-04-18

    We revisit the reactor antineutrino anomaly using the recent reactor flux independent determination of sizable theta?13 by considering the full set of the absolute reactor electron antineutrino flux measurements. When normalized to the predicted flux of Mueller et al. [1], the new world average, after including results from Palo Verde, Chooz, and Double Chooz, is 0.959 +- 0.009 (experiment uncertainty) +- 0.027 (flux systematics). Including the data with kilometer baseline, the new world average is only about 1.4 sigma lower than the unity, weakening the significance of the reactor antineutrino anomaly. The upcoming results from Daya Bay, RENO, and the Double Chooz will provide further information about this issue.

  8. Reactor Antineutrino Anomaly with known ?13

    NASA Astrophysics Data System (ADS)

    Zhang, Chao

    2013-10-01

    We revisit the reactor antineutrino anomaly using the recent reactor flux independent determination of sizable ?13 by considering the full set of the absolute reactor ?e flux measurements. When normalized to the predicted flux of Mueller et al. [Phys. Rev. C 83, 054615 (2011)], the new world average, after including results from Palo Verde, Chooz, and Double Chooz, is 0.959 +/- 0.009 (experiment uncertainty) +/- 0.027 (flux systematics). Including the data with kilometer baseline, the new world average is only about 1.4 ? lower than the unity, weakening the significance of the reactor antineutrino anomaly. The upcoming results from Daya Bay, RENO, and the Double Chooz will provide further information about this issue.

  9. Prismatic modular reactor analysis with melcor 

    E-print Network

    Zhen, Ni

    2009-05-15

    Hydrogen, a more sustainable source of energy, is a potential substitute for hydrocarbon fuel for power generation. The Very High Temperature gas-cooled Reactor (VHTR) concept can produce hydrogen with high efficiency and ...

  10. Control system for a small fission reactor

    DOEpatents

    Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

    1985-02-08

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

  11. Removal of hydrogen bubbles from nuclear reactors

    NASA Technical Reports Server (NTRS)

    Jenkins, R. V.

    1980-01-01

    Method proposed for removing large hydrogen bubbles from nuclear environment uses, in its simplest form, hollow spheres of palladium or platinum. Methods would result in hydrogen bubble being reduced in size without letting more radioactivity outside reactor.

  12. Heat pipe nuclear reactor for space power

    NASA Technical Reports Server (NTRS)

    Koening, D. R.

    1976-01-01

    A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

  13. Passive heat transfer means for nuclear reactors

    DOEpatents

    Burelbach, James P. (Glen Ellyn, IL)

    1984-01-01

    An improved passive cooling arrangement is disclosed for maintaining adjacent or related components of a nuclear reactor within specified temperature differences. Specifically, heat pipes are operatively interposed between the components, with the vaporizing section of the heat pipe proximate the hot component operable to cool it and the primary condensing section of the heat pipe proximate the other and cooler component operable to heat it. Each heat pipe further has a secondary condensing section that is located outwardly beyond the reactor confinement and in a secondary heat sink, such as air ambient the containment, that is cooler than the other reactor component. Means such as shrouding normally isolated the secondary condensing section from effective heat transfer with the heat sink, but a sensor responds to overheat conditions of the reactor to open the shrouding, which thereby increases the cooling capacity of the heat pipe. By having many such heat pipes, an emergency passive cooling system is defined that is operative without electrical power.

  14. THE FAST NEUTRON BREEDER REACTOR: THE \\

    Microsoft Academic Search

    Van Dievoet

    1958-01-01

    The heat exchange systemo in the Fermi Fast Breeder Reactor is described. ; The primoary circuit, the intermoediate circuit, the water-steam circuit, and the ; natural circulation circuit for accidental loss of pumoping power are discussed. ; (J.S.R.);

  15. Environmental Management Brookhaven Graphite Research Reactor

    E-print Network

    Homes, Christopher C.

    Environmental Management Brookhaven Graphite Research Reactor (BGRR) Project Completion John transfer to Office of Science (October 2013) Sign on exterior door, Building 701 Environmental Management Sattler Federal Project Director Office of Environmental Management U.S. Department of Energy BNL

  16. Advanced Test Reactor National Scientific User Facility

    SciTech Connect

    Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

    2011-08-01

    The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

  17. Reactor monitoring with Neutrinos Michel Cribier

    E-print Network

    Paris-Sud XI, Université de

    production. Nevertheless the net balance in plutonium is positive and a standard pressurized water power reactor produces around 200 kg of plutonium per year. Every fission of a fissile isotope produce two

  18. Molten salt reactors - safety options galore

    SciTech Connect

    Gat, U. [Oak Ridge National Lab., TN (United States); Dodds, H.L. [Univ. of Tennessee, Knoxville, TN (United States)

    1997-03-01

    Safety features and attributes of molten salt reactors (MSR) are described. The unique features of fluid fuel reactors of on-line continuous processing and the ability for so-called external cooling result in simple and safe designs with low excess reactivity, low fission product inventory, and small source term. These, in turn, make a criticality accident unlikely and reduce the severity of a loss of coolant to where they are no longer severe accidents. A melt down is not an accident for a reactor that uses molten fuel. The molten salts are stable, non-reactive and efficient heat transfer media that operate at high temperatures at low pressures and are highly compatible with selected structural materials. All these features reduce the accident plethora. Freeze valves can be used for added safety. An ultimate safe reactor (U.S.R) is described with safety features that are passive, inherent and non-tamperable (PINT).

  19. Space Propulsion via Spherical Torus Fusion Reactor

    SciTech Connect

    Williams, Craig H.; Juhasz, Albert J.; Borowski, Stanley K.; Dudzinski, Leonard A. [NASA Glenn Research Center (United States)

    2003-01-15

    A conceptual vehicle design enabling fast outer solar system travel was produced predicated on a small aspect ratio spherical torus nuclear fusion reactor. Analysis revealed that the vehicle could deliver a 108 mt crew habitat payload to Saturn rendezvous in 204 days, with an initial mass in low Earth orbit of 1630 mt. Engineering conceptual design, analysis, and assessment were performed on all major systems including nuclear fusion reactor, magnetic nozzle, power conversion, fast wave plasma heating, fuel pellet injector, startup/re-start fission reactor and battery, and other systems. Detailed fusion reactor design included analysis of plasma characteristics, power balance and utilization, first wall, toroidal field coils, heat transfer, and neutron/X-ray radiation.

  20. Producing tritium in a homogenous reactor

    DOEpatents

    Cawley, William E. (Richland, WA)

    1985-01-01

    A method and apparatus are described for the joint production and separation of tritium. Tritium is produced in an aqueous homogenous reactor and heat from the nuclear reaction is used to distill tritium from the lower isotopes of hydrogen.