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Sample records for reactor plant wwer-1200

  1. Modeling of corrosion product migration in the secondary circuit of nuclear power plants with WWER-1200

    NASA Astrophysics Data System (ADS)

    Kritskii, V. G.; Berezina, I. G.; Gavrilov, A. V.; Motkova, E. A.; Zelenina, E. V.; Prokhorov, N. A.; Gorbatenko, S. P.; Tsitser, A. A.

    2016-04-01

    Models of corrosion and mass transfer of corrosion products in the pipes of the condensate-feeding and steam paths of the secondary circuit of NPPs with WWER-1200 are presented. The mass transfer and distribution of corrosion products over the currents of the working medium of the secondary circuit were calculated using the physicochemical model of mass transfer of corrosion products in which the secondary circuit is regarded as a cyclic system consisting of a number of interrelated elements. The circuit was divided into calculated regions in which the change in the parameters (flow rate, temperature, and pressure) was traced and the rates of corrosion and corrosion products entrainment, high-temperature pH, and iron concentration were calculated. The models were verified according to the results of chemical analyses at Kalinin NPP and iron corrosion product concentrations in the feed water at different NPPs depending on pH at 25°C (pH25) for service times τ ≥ 5000 h. The calculated pH values at a coolant temperature t (pH t ) in the secondary circuit of NPPs with WWER-1200 were presented. The calculation of the distribution of pH t and ethanolamine and ammonia concentrations over the condensate feed (CFC) and steam circuits is given. The models are designed for developing the calculation codes. The project solutions of ATOMPROEKT satisfy the safety and reliability requirements for power plants with WWER-1200. The calculated corrosion and corrosion product mass transfer parameters showed that the model allows the designer to choose between the increase of the correcting reagent concentration, the use of steel with higher chromium contents, and intermittent washing of the steam generator from sediments as the best solution for definite regions of the circuit.

  2. NEUTRONIC REACTOR POWER PLANT

    DOEpatents

    Metcalf, H.E.

    1962-12-25

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

  3. Plant maintenance and advanced reactors, 2005

    SciTech Connect

    Agnihotri, Newal

    2005-09-15

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

  4. Plant maintenance and advanced reactors, 2007

    SciTech Connect

    Agnihotri, Newal

    2007-09-15

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

  5. Generic small modular reactor plant design.

    SciTech Connect

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

    2012-12-01

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

  6. Modular stellarator reactor: a fusion power plant

    SciTech Connect

    Miller, R.L.; Bathke, C.G.; Krakowski, R.A.; Heck, F.M.; Green, L.; Karbowski, J.S.; Murphy, J.H.; Tupper, R.B.; DeLuca, R.A.; Moazed, A.

    1983-07-01

    A comparative analysis of the modular stellarator and the torsatron concepts is made based upon a steady-state ignited, DT-fueled, reactor embodiment of each concept for use as a central electric-power station. Parametric tradeoff calculations lead to the selection of four design points for an approx. 4-GWt plant based upon Alcator transport scaling in l = 2 systems of moderate aspect ratio. The four design points represent high-aspect ratio. The four design points represent high-(0.08) and low-(0.04) beta versions of the modular stellarator and torsatron concepts. The physics basis of each design point is described together with supporting engineering and economic analyses. The primary intent of this study is the elucidation of key physics and engineering tradeoffs, constraints, and uncertainties with respect to the ultimate power reactor embodiment.

  7. Analysis of UF6 breeder reactor power plants

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1976-01-01

    Gaseous UF6 fueled breeder reactor design and technical applications of such concepts are summarized. Special attention was given to application in nuclear power plants and to reactor efficiency and safety factors.

  8. Plant maintenance and advanced reactors issue, 2004

    SciTech Connect

    Agnihotri, Newal

    2004-09-15

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

  9. Plant maintenance and advanced reactors issue, 2008

    SciTech Connect

    Agnihotri, Newal

    2009-09-15

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

  10. Gas Reactor Plant Analyzer and Simulator for Hydrogen Production

    2004-01-01

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

  11. Liquid metal cooled nuclear reactor plant system

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

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

  12. European simplified boiling water reactor (ESBWR) plant

    SciTech Connect

    Posta, B.A.; Goldenberg, E.A.; Sawhney, P.S.; Rao, A.S.

    1996-07-01

    This paper covers innovative ideas which made possible the redesign of the US 660-MW Simplified Boiling Water Reactor (SBWR) Reactor Island for a 1,200-MW size reactor while actually reducing the building cost. This was achieved by breaking down the Reactor Island into multiple buildings separating seismic-1 from non-seismic-1 areas, providing for better space utilization, shorter construction schedule, easier maintainability and better postaccident accessibility.

  13. Design considerations for an interial confinement fusion reactor power plant

    NASA Astrophysics Data System (ADS)

    Massey, J. V.; Simpson, J. E.

    1981-08-01

    A conceptual design study to further define the engineering and economic concerns for inertial confinement fusion reactors is presented. Alternatives to the Livermore HYLIFE concept were examined and information from liquid metal cooled fast breeder reactor power plant studies was incorporated into the design. Laser and target physics models were employed in a reactor design with a low coolant flowrate and a high driver repetition rate. An example of such a design is the JADE concept. In addition to a power plant design developed using the JADE example, the applicability of the energy absorbing gas lithium ejector concept was investigated.

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

    SciTech Connect

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

    1990-09-01

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

  15. Design considerations for an inertial confinement fusion reactor power plant

    SciTech Connect

    Massey, J.V.; Simpson, J.E.

    1981-08-10

    To further define the engineering and economic concerns for inertial confinement fusion reactors (ICR's), a conceptual design study was performed by Bechtel Group Incorporated under the direction of Lawrence Livermore National Laboratory (LLNL). The study examined alternatives to the LLNL HYLIFE concept and expanded the previous balance of plant design to incorporate information from recent liquid metal cooled fast breeder reactor (LMFBR) power plant studies. The majority of the effort was to incorporate present laser and target physics models into a reactor design with a low coolant flowrate and a high driver repetition rate. An example of such a design is the LLNL JADE concept. In addition to producing a power plant design for LLNL using the JADE example, Bechtel has also examined the applicability of the EAGLE (Energy Absorbing Gas Lithium Ejector) concept.

  16. a Decade of Dosimetry for Magnox Reactor Plants

    NASA Astrophysics Data System (ADS)

    Lewis, T. A.; Thornton, D. A.

    2003-06-01

    This paper reviews the reactor dosimetry program that has supported steel pressure vessel integrity assessments for magnox power plants over the last ten years. The dosimetry program has aimed to achieve consistent:. • calculated and measured fast and thermal neutron doses. • data for surveillance specimens and reactor pressure vessels. Throughout the program, the flux measurements on the plants have been judged essential for any doses where a high degree of confidence is required. The work to support operation is now largely complete and the dosimetry is being extended to assess radioactive inventories as part of the decommissioning process.

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

    SciTech Connect

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

    2006-12-11

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

  18. Nuclear plant-aging research on reactor protection systems

    SciTech Connect

    Meyer, L.C.

    1988-01-01

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

  19. UF6 breeder reactor power plants for electric power generation

    NASA Technical Reports Server (NTRS)

    Rust, J. H.; Clement, J. D.; Hohl, F.

    1976-01-01

    The reactor concept analyzed is a U-233F6 core surrounded by a molten salt (Li(7)F, BeF2, ThF4) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. Advantages of the gas core breeder reactor (GCBR) are as follows: (1) high efficiency; (2) simplified on-line reprocessing; (3) inherent safety considerations; (4) high breeding ratio; (5) possibility of burning all or most of the long-lived nuclear waste actinides; and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion.

  20. Annealing the reactor vessel at the Palisades Plant

    SciTech Connect

    Fenech, R.A.

    1996-03-01

    In the way of background, Palisades was licensed in 1967 and went commercial in 1971. Jumping to two years ago, we faced at that time three issues that challenged our ability to operate to end-of-license, which would be 2007 without any extensions. The three items were regulatory performance, economic performance, and reactor vessel embrittlement. We had not been operating the plant with the kind of conservative decisions and with the kind of safety margins that one is expected to operate a plant in the United States at this time. Our economic performance was not satisfactory in that our capacity factor was low and our costs high. In the area of reactor vessel embrittlement, our analysis showed that we would reach the NRC screening criteria for embrittlement in the year 2004. Over the last two years, we have made significant improvements in the first two areas. Our decision-making has changed. Our performance, especially over the last year and a half, has been excellent. In addition, we have gotten our capacity factors up and our costs under control. Clearly, sustained performance is what is going to carry the day but from what we can see and from where we are, we are in more of a maintenance-of-performance than in a turn-around situation. On the other hand, in the area of reactor vessel embrittlement, about a year and a half ago we had a bit of a setback. We had taken material from retired steam generators that had welds identical to the welds in our reactor vessel. When we analyzed the welds from our steam generators, we were given some surprises about the chemistry makeup. When we applied the new information to our analysis, we changed the date on which we would reach our screening criteria from 2004 to late 1999.

  1. Membrane bio-reactor for textile wastewater treatment plant upgrading.

    PubMed

    Lubello, C; Gori, R

    2005-01-01

    Textile industries carry out several fiber treatments using variable quantities of water, from five to forty times the fiber weight, and consequently generate large volumes of wastewater to be disposed of. Membrane Bio-reactors (MBRs) combine membrane technology with biological reactors for the treatment of wastewater: micro or ultrafiltration membranes are used for solid-liquid separation replacing the secondary settling of the traditional activated sludge system. This paper deals with the possibility of realizing a new section of one existing WWTP (activated sludge + clariflocculation + ozonation) for the treatment of treating textile wastewater to be recycled, equipped with an MBR (76 l/s as design capacity) and running in parallel with the existing one. During a 4-month experimental period, a pilot-scale MBR proved to be very effective for wastewater reclamation. On average, removal efficiency of the pilot plant (93% for COD, and over 99% for total suspended solids) was higher than the WWTP ones. Color was removed as in the WWTP. Anionic surfactants removal of pilot plant was lower than that of the WWTP (90.5 and 93.2% respectively), while the BiAS removal was higher in the pilot plant (98.2 vs. 97.1). At the end cost analysis of the proposed upgrade is reported.

  2. Dual-phase reactor plant with partitioned isolation condenser

    DOEpatents

    Hui, Marvin M.

    1992-01-01

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

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

    DOEpatents

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

    1984-08-30

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

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

    DOEpatents

    Huebotter, Paul R.; McLennan, George A.

    1985-01-01

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

  5. Initiating Events for Multi-Reactor Plant Sites

    SciTech Connect

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

    2014-09-01

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

  6. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-01-20

    ... COMMISSION Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors..., ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors.'' DATES... developed using this Catalog along with the Operator Licensing Examination Standards for Power...

  7. REACTOR-FLASH BOILER-FLYWHEEL POWER PLANT

    DOEpatents

    Loeb, E.

    1961-01-17

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

  8. Biodenitrification of gaseous diffusion plant aqueous wastes: fluid bed reactor

    SciTech Connect

    Kowalchuk, M.

    1982-01-04

    Decontamination and uranium recovery operations at Portsmouth generate raffinates which contain nitrate. Nitrate discharges are now within EPA limits. However, more stringent limits go into effect on October 2, 1982. These limits cannot be met by present operating methods without seriously restricting decontamination and recovery operations. A biodentrification process will therefore be used at Portsmouth to reduce the nitrate concentration to acceptable levels. Pilot plant studies using a fluid bed reactor were carried out at ORNL. Process operating parameters were characterized and design criteria for the full-scale facility which is to be built at Portsmouth were devised. When operations were completed, the pilot plant, equipped with a 20-inch bioreactor, was shipped to Portsmouth. It will be installed during FY-1982, and will be operated until the full-scale facility is built. It will allow GAT to meet EPA limts and will accommodate 9000 liters of raffinate per month. The projected post CIP/CUP raffinate generation rate is 15,000 liters per month. Recovery operations will be limited to some extent until the fullscale biodenitrification facility is built.

  9. Modular high temperature gas-cooled reactor plant design duty cycle. Revision 3

    SciTech Connect

    Chan, T.

    1989-12-31

    This document defines the Plant Design Duty Cycle (PCDC) for the Modular High Temperature Gas-cooled Reactor (MHTGR). The duty cycle is a set of events and their design number of occurrences over the life of the plant for which the MHTGR plant shall be designed to ensure that the plant meets all the top-level requirements. The duty cycle is representative of the types of events to be expected in multiple reactor module-turbine plant configurations of the MHTGR. A synopsis of each PDDC event is presented to provide an overview of the plant response and consequence. 8 refs., 1 fig., 4 tabs.

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

    SciTech Connect

    1995-08-01

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

  11. Two-reactor, high-recovery sulfur plant and process

    SciTech Connect

    Reed, R.L.; Palm, J.W.

    1989-04-18

    This patent describes a process for the recovery of sulfur wherein an acid gas feedstream comprising hydrogen sulfide is processed for the recovery of sulfur in a Claus process sulfur recovery plant. The process consists of: (a) passing the acid gas feedstream successively through the thermal reaction zone, the first position Claus catalytic reaction zone, and the second position Claus catalytic reaction zone for the recovery of sulfur; (b) preconditioning the first position Claus catalytic reaction zone by introducing thereinto a cold stream having an inlet temperature effective for condensing sulfur on at least a portion of the catalyst and passing the resulting stream through a remaining substantial portion of the catalyst, the cold stream thus used for preconditioning being produced by cooling acid gas feedstream effluent from the thermal reaction zone to the first position catalytic reaction zone to the temperature; and (c) switching the thus preconditioned Claus catalytic reaction zone in the first position into the second position and continuing cooling the thus preconditioned freshly regenerated reactor in the second position concurrently with forming and depositing sulfur on catalyst therein, and switching the Claus catalytic reaction zone in the second position into the first position and continuing the process according to (a), (b), and (c).

  12. Experiments on rehabilitation of radioactive metallic waste (RMW) of reactor stainless steels of Siberian chemical plant

    NASA Astrophysics Data System (ADS)

    Kolpakov, G. N.; Zakusilov, V. V.; Demyanenko, N. V.; Mishin, A. S.

    2016-06-01

    Stainless steel pipes, used to cool a reactor plant, have a high cost, and after taking a reactor out of service they must be buried together with other radioactive waste. Therefore, the relevant problem is the rinse of pipes from contamination, followed by returning to operation.

  13. Clinch River Breeder Reactor Plant Project. Summary edition. 1980 technical progress report, October 1979-September 1980

    SciTech Connect

    Not Available

    1980-01-01

    This technical progress report on the CRBRP Project describes the objectives, design decisions, and major accomplishments achieved in the planning, organizing, design, and execution of the Project during the period October 1, 1979, through September 30, 1980. It is a summary of the 1980 CRBRP Technical Progress Report, which was prepared by the Advanced Reactors Division of Westinghouse Electric Corporation, the Lead Reactor Manufacturer for the Clinch River Breeder Reactor Plant Project, in fulfillment of contract requirements with the United States Department of Energy. It includes inputs from the CRBRP Architect-Engineer (Burns and Roe, Inc.), from the Constructor (Stone and Webster Engineering Corporation), and from the supporting Reactor Manufacturers (Atomics International Division of the Energy Systems Group of Rockwell International Corporation, the Advanced Reactor Systems Department of General Electric Company, and the Advanced Reactors Division of Westinghouse Electric Corporation).

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

    SciTech Connect

    Not Available

    1993-05-13

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

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

    SciTech Connect

    M. J. Russell

    2006-06-01

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

  16. 76 FR 4738 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-26

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant License... Canyon Power Plant, Units 1 and 2 and the associated Safety Evaluation Report (SER) with Open Items....

  17. Study of hydrogen generation plant coupled to high temperature gas cooled reactor

    NASA Astrophysics Data System (ADS)

    Brown, Nicholas Robert

    Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Several thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. Eight unique case studies are performed based on a thorough literature review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without SCRAM, (6) total failure of the chemical plant, (7) parametric study of the temperature in an individual reaction chamber, and (8) control rod insertion in the nuclear reactor. Various parametric

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

    SciTech Connect

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

    1982-05-01

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

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

    SciTech Connect

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

    2002-07-01

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

  20. High Efficiency Nuclear Power Plants using Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITS of 950 K and 1200 K are presented. Power plant performance data were obtained for TITS ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo -generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  1. High Efficiency Nuclear Power Plants Using Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.; Rarick, Richard A.; Rangarajan, Rajmohan

    2009-01-01

    An overall system analysis approach is used to propose potential conceptual designs of advanced terrestrial nuclear power plants based on Oak Ridge National Laboratory (ORNL) Molten Salt Reactor (MSR) experience and utilizing Closed Cycle Gas Turbine (CCGT) thermal-to-electric energy conversion technology. In particular conceptual designs for an advanced 1 GWe power plant with turbine reheat and compressor intercooling at a 950 K turbine inlet temperature (TIT), as well as near term 100 MWe demonstration plants with TITs of 950 and 1200 K are presented. Power plant performance data were obtained for TITs ranging from 650 to 1300 K by use of a Closed Brayton Cycle (CBC) systems code which considered the interaction between major sub-systems, including the Liquid Fluoride Thorium Reactor (LFTR), heat source and heat sink heat exchangers, turbo-generator machinery, and an electric power generation and transmission system. Optional off-shore submarine installation of the power plant is a major consideration.

  2. EPRI`s nuclear power plant instrumentation and control program and its applicability to advanced reactors

    SciTech Connect

    Naser, J.; Torok, R.; Wilkinson, D.

    1997-12-01

    I&C systems in nuclear power plants need to be upgraded over the lifetime of the plant in a reliable and cost-effective manner to replace obsolete equipment, to reduce O&M costs, to improve plant performance, and to maintain safety. This applies to operating plants now and will apply to advanced reactors in the future. The major drivers for the replacement of the safety, control, and information systems in nuclear power plants are the obsolescence of the existing hardware and the need for more cost-effective power production. Competition between power producers is dictating more cost-effective power production. The increasing O&M costs to maintain systems experiencing obsolescence problems is counter to the needs for more cost-effective power production and improved competitiveness. This need for increased productivity applies to government facilities as well as commercial plants. Increasing competition will continue to be a major factor in the operation of both operating plants and advanced reactors. It will continue to dictate the need for improved productivity and cost-effectiveness. EPRI and its member nuclear utilities are working together on an industry wide I&C Program to address I&C issues and to develop cost-effective solutions. A majority of the I&C products and demonstrations being developed under this program will benefit advanced reactors in both the design and operational phases of their life cycle as well as it will benefit existing plants. 20 refs.

  3. Knowledges and abilities catalog for nuclear power plant operators: Savannah River Site (SRS) production reactors

    SciTech Connect

    Not Available

    1990-06-20

    The Knowledges and Abilities Catalog for Nuclear Power Plant Operations: Savannah River Site (SRS) Production Reactors, provides the basis for the development of content-valid certification examinations for Senior Reactor Operators (SROs) and Central Control Room Supervisors (SUP). The position of Shift Technical Engineer (STE) has been included in the catalog for completeness. This new SRS reactor operating shift crew position is held by an individual holding a CCR Supervisor Certification who has received special engineering and technical training. Also, the STE has a Bachelor of Science degree in engineering or a related technical field. The SRS catalog contains approximately 2500 knowledge and ability (K/A) statements for SROs and SUPs at heavy water moderated production reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring the health and safety of the public. The SRS K/A catalog is presently organized into five major sections: Plant Systems grouped by Safety Function, Plant Wide Generic K/As, Emergency Plant Evolutions, Theory and Components (to be developed).

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

    SciTech Connect

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

    1990-01-01

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

  5. 76 FR 5220 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-01-28

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant Designs The ACRS Subcommittee on Future Plant Designs will hold a meeting on February 9, 2011, at 11545... small modular reactor applications. The Subcommittee will hear presentations by and hold...

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

    SciTech Connect

    Wayne Moe

    2013-05-01

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

  7. Overall plant design specification Modular High Temperature Gas-cooled Reactor. Revision 9

    SciTech Connect

    1990-05-01

    Revision 9 of the ``Overall Plant Design Specification Modular High Temperature Gas-Cooled Reactor,`` DOE-HTGR-86004 (OPDS) has been completed and is hereby distributed for use by the HTGR Program team members. This document, Revision 9 of the ``Overall Plant Design Specification`` (OPDS) reflects those changes in the MHTGR design requirements and configuration resulting form approved Design Change Proposals DCP BNI-003 and DCP BNI-004, involving the Nuclear Island Cooling and Spent Fuel Cooling Systems respectively.

  8. 75 FR 62610 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-12

    ... Guide DG-1237, ``Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors,'' Interim Staff Guidance (ISG) NSIR/DPR-ISG-01, ``Emergency Planning for Nuclear Power Plants,'' and NUREG/CR 7002... From the Federal Register Online via the Government Publishing Office NUCLEAR...

  9. Nuclear Technology Series. Nuclear Reactor (Plant) Operator Trainee. A Suggested Program Planning Guide. Revised June 80.

    ERIC Educational Resources Information Center

    Center for Occupational Research and Development, Inc., Waco, TX.

    This program planning guide for a two-year postsecondary nuclear reactor (plant) operator trainee program is designed for use with courses 1-16 of thirty-five in the Nuclear Technology Series. The purpose of the guide is to describe the nuclear power field and its job categories for specialists, technicians and operators; and to assist planners,…

  10. 76 FR 40406 - Advisory Committee on Reactor Safeguards; Meeting of the ACRS Subcommittee on Plant Operations...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-07-08

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards; Meeting of the ACRS Subcommittee on Plant Operations and Fire Protection; Revision to an ACRS Subcommittee Meeting Federal Register Notice The Federal Register Notice for the ACRS Subcommittee Meeting...

  11. 77 FR 56239 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-09-12

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant License...-Besse Nuclear Power station and the associated draft Safety Evaluation Report (SER) with open items....

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

    SciTech Connect

    1995-08-01

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

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

    SciTech Connect

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

    2010-11-01

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

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

    SciTech Connect

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

    1981-09-15

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

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

    SciTech Connect

    Chang H. Oh; Eung Soo Kim; Steven Sherman

    2008-04-01

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

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

    SciTech Connect

    L.E. Demick

    2010-09-01

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

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

    SciTech Connect

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

    2007-03-21

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

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

    SciTech Connect

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

    2006-08-01

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

  19. Overview of thermal-buoyancy-induced phenomena in reactor-plant components. [LMFBR

    SciTech Connect

    Kasza, K.E.; Kuzay, T.M.; Oras, J.J.

    1984-01-01

    Studies related to delineating the influence of thermal-buoyancy forces on the thermal-hydraulics of Liquid Metal Fast Breeder Reactor plant components under low-flow thermal transient and steady state conditions have generated unique information which will aid design of these components. Various buoyancy force induced phenomena such as thermal stratification, flow recirculation, stagnation, and channeling are described and the importance to component performance are discussed. The water based studies have been conducted in the Mixing Components Test Facility, a large multi program facility capable of performing generic studies of fluid flow and heat transfer in reactor components under programmed transient and steady state conditions.

  20. REACTOR

    DOEpatents

    Christy, R.F.

    1961-07-25

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

  1. REACTOR

    DOEpatents

    Szilard, L.

    1963-09-10

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

  2. REACTOR

    DOEpatents

    Roman, W.G.

    1961-06-27

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

  3. Thermal analysis of heat and power plant with high temperature reactor and intermediate steam cycle

    NASA Astrophysics Data System (ADS)

    Fic, Adam; Składzień, Jan; Gabriel, Michał

    2015-03-01

    Thermal analysis of a heat and power plant with a high temperature gas cooled nuclear reactor is presented. The main aim of the considered system is to supply a technological process with the heat at suitably high temperature level. The considered unit is also used to produce electricity. The high temperature helium cooled nuclear reactor is the primary heat source in the system, which consists of: the reactor cooling cycle, the steam cycle and the gas heat pump cycle. Helium used as a carrier in the first cycle (classic Brayton cycle), which includes the reactor, delivers heat in a steam generator to produce superheated steam with required parameters of the intermediate cycle. The intermediate cycle is provided to transport energy from the reactor installation to the process installation requiring a high temperature heat. The distance between reactor and the process installation is assumed short and negligable, or alternatively equal to 1 km in the analysis. The system is also equipped with a high temperature argon heat pump to obtain the temperature level of a heat carrier required by a high temperature process. Thus, the steam of the intermediate cycle supplies a lower heat exchanger of the heat pump, a process heat exchanger at the medium temperature level and a classical steam turbine system (Rankine cycle). The main purpose of the research was to evaluate the effectiveness of the system considered and to assess whether such a three cycle cogeneration system is reasonable. Multivariant calculations have been carried out employing the developed mathematical model. The results have been presented in a form of the energy efficiency and exergy efficiency of the system as a function of the temperature drop in the high temperature process heat exchanger and the reactor pressure.

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

    SciTech Connect

    Wood, RT

    2004-09-27

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

  5. Manufacturing demonstration of microbially mediated zinc sulfide nanoparticles in pilot-plant scale reactors.

    PubMed

    Moon, Ji-Won; Phelps, Tommy J; Fitzgerald, Curtis L; Lind, Randall F; Elkins, James G; Jang, Gyoung Gug; Joshi, Pooran C; Kidder, Michelle; Armstrong, Beth L; Watkins, Thomas R; Ivanov, Ilia N; Graham, David E

    2016-09-01

    The thermophilic anaerobic metal-reducing bacterium Thermoanaerobacter sp. X513 efficiently produces zinc sulfide (ZnS) nanoparticles (NPs) in laboratory-scale (≤ 24-L) reactors. To determine whether this process can be up-scaled and adapted for pilot-plant production while maintaining NP yield and quality, a series of pilot-plant scale experiments were performed using 100-L and 900-L reactors. Pasteurization and N2-sparging replaced autoclaving and boiling for deoxygenating media in the transition from small-scale to pilot plant reactors. Consecutive 100-L batches using new or recycled media produced ZnS NPs with highly reproducible ~2-nm average crystallite size (ACS) and yields of ~0.5 g L(-1), similar to the small-scale batches. The 900-L pilot plant reactor produced ~320 g ZnS without process optimization or replacement of used medium; this quantity would be sufficient to form a ZnS thin film with ~120 nm thickness over 0.5 m width × 13 km length. At all scales, the bacteria produced significant amounts of acetic, lactic, and formic acids, which could be neutralized by the controlled addition of sodium hydroxide without the use of an organic pH buffer, eliminating 98 % of the buffer chemical costs. The final NP products were characterized using XRD, ICP-OES, TEM, FTIR, PL, DLS, HPLC, and C/N analyses, which confirmed that the growth medium without organic buffer enhanced the ZnS NP properties by reducing carbon and nitrogen surface coatings and supporting better dispersivity with similar ACS. PMID:27118014

  6. Manufacturing demonstration of microbially mediated zinc sulfide nanoparticles in pilot-plant scale reactors.

    PubMed

    Moon, Ji-Won; Phelps, Tommy J; Fitzgerald, Curtis L; Lind, Randall F; Elkins, James G; Jang, Gyoung Gug; Joshi, Pooran C; Kidder, Michelle; Armstrong, Beth L; Watkins, Thomas R; Ivanov, Ilia N; Graham, David E

    2016-09-01

    The thermophilic anaerobic metal-reducing bacterium Thermoanaerobacter sp. X513 efficiently produces zinc sulfide (ZnS) nanoparticles (NPs) in laboratory-scale (≤ 24-L) reactors. To determine whether this process can be up-scaled and adapted for pilot-plant production while maintaining NP yield and quality, a series of pilot-plant scale experiments were performed using 100-L and 900-L reactors. Pasteurization and N2-sparging replaced autoclaving and boiling for deoxygenating media in the transition from small-scale to pilot plant reactors. Consecutive 100-L batches using new or recycled media produced ZnS NPs with highly reproducible ~2-nm average crystallite size (ACS) and yields of ~0.5 g L(-1), similar to the small-scale batches. The 900-L pilot plant reactor produced ~320 g ZnS without process optimization or replacement of used medium; this quantity would be sufficient to form a ZnS thin film with ~120 nm thickness over 0.5 m width × 13 km length. At all scales, the bacteria produced significant amounts of acetic, lactic, and formic acids, which could be neutralized by the controlled addition of sodium hydroxide without the use of an organic pH buffer, eliminating 98 % of the buffer chemical costs. The final NP products were characterized using XRD, ICP-OES, TEM, FTIR, PL, DLS, HPLC, and C/N analyses, which confirmed that the growth medium without organic buffer enhanced the ZnS NP properties by reducing carbon and nitrogen surface coatings and supporting better dispersivity with similar ACS.

  7. Dynamic analysis of the condensate feedwater system in boiling water reactor plants

    SciTech Connect

    Tanji, J.; Omori, T.

    1982-05-01

    The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents.

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

    SciTech Connect

    Blaine Grover

    2010-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data, including irradiation creep, at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant (NGNP) Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain six stacks of graphite specimens, with half of the graphite specimens in each stack under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will have differing compressive loads applied to the top half of each pair of specimen stacks, while a seventh stack will not have a compressive load. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any oxidation or off-gassing of the specimens occurs during initial start-up of

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

    SciTech Connect

    J. K. Wright; R. N. Wright

    2008-04-01

    The U.S. Department of Energy has selected the High Temperature Gas-cooled Reactor design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic, or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development Program is responsible for performing research and development on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. Studies of potential Reactor Pressure Vessel (RPV) steels have been carried out as part of the pre-conceptual design studies. These design studies generally focus on American Society of Mechanical Engineers (ASME) Code status of the steels, temperature limits, and allowable stresses. Three realistic candidate materials have been identified by this process: conventional light water reactor RPV steels A508/533, 2¼Cr-1Mo in the annealed condition, and modified 9Cr 1Mo ferritic martenistic steel. Based on superior strength and higher temperature limits, the modified 9Cr-1Mo steel has been identified by the majority of design engineers as the preferred choice for the RPV. All of the vendors have

  10. Manufacturing demonstration of microbially mediated zinc sulfide nanoparticles in pilot-plant scale reactors

    DOE PAGES

    Moon, Ji-Won; Phelps, Tommy J.; Fitzgerald Jr, Curtis L.; Lind, Randall F.; Elkins, James G.; Jang, Gyoung Gug; Joshi, Pooran C.; Kidder, Michelle; Armstrong, Beth L.; Watkins, Thomas R.; et al

    2016-04-27

    The thermophilic anaerobic metal-reducing bacterium Thermoanaerobacter sp. X513 efficiently produces zinc sulfide (ZnS) nanoparticles (NPs) in laboratory-scale ( ≤24-L) reactors. To determine whether this process can be up-scaled and adapted for pilot-plant production while maintaining NP yield and quality, a series of meso-scale experiments were performed using 100-l and 900-l reactors. Pasteurization and N2-sparging replaced autoclaving and boiling for deoxygenating media in the transition from small-scale to pilot-plant reactors. Consecutive 100-L batches using new or recycled media produced ZnS NPs with highly reproducible ~2 nm average crystallite size (ACS) and yields of ~0.5g L-1, similar to small-scale batches. The 900-Lmore » pilot plant reactor produced ~ 320 g ZnS without process optimization or replacement of used medium; this quantity would be sufficient to form a ZnS thin film with ~120 nm thickness over 0.5 m width 13 km length. At all scales, the bacteria produced significant amounts of acetic, lactic and formic acids, which could be neutralized by the controlled addition of sodium hydroxide without the use of an organic pH buffer, eliminating 98% of the buffer chemical costs. In conclusion, the final NP products were characterized using XRD, ICP-OES, FTIR, DLS, and C/N analyses, which confirmed the growth medium without organic buffer enhanced the ZnS NP properties by reducing carbon and nitrogen surface coatings and supporting better dispersivity with similar ACS.« less

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

    SciTech Connect

    J. K. Wright; R. N. Wright

    2010-07-01

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

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

    SciTech Connect

    1997-06-01

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

  13. Maintenance Cycle Extension in the IRIS Advanced Light Water Reactor Plant Design

    SciTech Connect

    Galvin, Mark R.; Todreas, Neil E.; Conway, Larry E.

    2003-09-15

    New nuclear power generation in the United States will be realized only if the economic performance can be made competitive with other methods of electrical power generation. The economic performance of a nuclear power plant can be significantly improved by increasing the time spent on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described that can be used to resolve, in the design phase, maintenance-related operating cycle length barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the International Reactor, Innovative and Secure (IRIS) design. IRIS is an advanced light water nuclear power plant that is being designed to maximize this on-line generating time by increasing the operating cycle length. This is consequently a maintenance strategy paper using the IRIS plant as the example.Potential IRIS operating cycle length maintenance-related barriers, determined by modification of an earlier operating pressurized water reactor (PWR) plant cycle length analysis to account for differences between the design of IRIS and this operating PWR, are presented. The proposed methodology to resolve these maintenance-related barriers by the design process is described. The results of applying the methodology to two potential IRIS cycle length barriers, relief valve testing and emergency heat removal system testing, are presented.

  14. Replacement of outboard main steam isolation valves in a boiling water reactor plant

    SciTech Connect

    Schlereth, J.R.; Pennington, D.

    1996-12-01

    Most Boiling Water Reactor plants utilize wye pattern globe valves for main steam isolation valves for both inboard and outboard isolation. These valves have required a high degree of maintenance attention in order to pass the plant local leakage rate testing (LLRT) requirements at each outage. Northern States Power made a decision in 1993 to replace the outboard valves at it`s Monticello plant with double disc gate valves. The replacement of the outboard valves was completed during the fall outage in 1994. During the spring outage in April of 1996 the first LLRT testing was performed with excellent results. This presentation will address the decision process, time requirements and planning necessary to accomplish the task as well as the performance results and cost effectiveness of replacing these components.

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

    SciTech Connect

    Phillip Mills

    2012-02-01

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

  16. Research on pressure control of pressurizer in pressurized water reactor nuclear power plant

    NASA Astrophysics Data System (ADS)

    Dai, Ling; Yang, Xuhong; Liu, Gang; Ye, Jianhua; Qian, Hong; Xue, Yang

    2010-07-01

    Pressurizer is one of the most important components in the nuclear reactor system. Its function is to keep the pressure of the primary circuit. It can prevent shutdown of the system from the reactor accident under the normal transient state while keeping the setting value in the normal run-time. This paper is mainly research on the pressure system which is running in the Daya Bay Nuclear Power Plant. A conventional PID controller and a fuzzy controller are designed through analyzing the dynamic characteristics and calculating the transfer function. Then a fuzzy PID controller is designed by analyzing the results of two controllers. The fuzzy PID controller achieves the optimal control system finally.

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

    SciTech Connect

    1993-09-15

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

  18. Interim results of the study of control room crew staffing for advanced passive reactor plants

    SciTech Connect

    Hallbert, B.P.; Sebok, A.; Haugset, K.

    1996-03-01

    Differences in the ways in which vendors expect the operations staff to interact with advanced passive plants by vendors have led to a need for reconsideration of the minimum shift staffing requirements of licensed Reactor Operators and Senior Reactor Operators contained in current federal regulations (i.e., 10 CFR 50.54(m)). A research project is being carried out to evaluate the impact(s) of advanced passive plant design and staffing of control room crews on operator and team performance. The purpose of the project is to contribute to the understanding of potential safety issues and provide data to support the development of design review guidance. Two factors are being evaluated across a range of plant operating conditions: control room crew staffing; and characteristics of the operating facility itself, whether it employs conventional or advanced, passive features. This paper presents the results of the first phase of the study conducted at the Loviisa nuclear power station earlier this year. Loviisa served as the conventional plant in this study. Data collection from four crews were collected from a series of design basis scenarios, each crew serving in either a normal or minimum staffing configuration. Results of data analyses show that crews participating in the minimum shift staffing configuration experienced significantly higher workload, had lower situation awareness, demonstrated significantly less effective team performance, and performed more poorly as a crew than the crews participating in the normal shift staffing configuration. The baseline data on crew configurations from the conventional plant setting will be compared with similar data to be collected from the advanced plant setting, and a report prepared providing the results of the entire study.

  19. 76 FR 17160 - Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-28

    ... COMMISSION Office of New Reactors; Final Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby Emergency Alternating Current Power System AGENCY: Nuclear Regulatory... Guidance (ISG) DC/COL-ISG-021 titled ``Interim Staff Guidance on the Review of Nuclear Power Plant...

  20. 75 FR 58448 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee On Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-24

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee On Future Plant Designs The ACRS Subcommittee on Future Plant Designs will hold a meeting on October 21, 2010, at 11545... Subcommittee will review current Design Acceptance Criteria associated with Digital Instrumentation and...

  1. 77 FR 74698 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-12-17

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant Designs; Notice of Meeting The ACRS Subcommittee on Future Plant Designs will hold a meeting on January 17... were published in the Federal Register on October 18, 2012, (77 FR 64146-64147). Detailed...

  2. 76 FR 64123 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-17

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant Designs; Notice of Meeting The ACRS Subcommittee on Future Plant Designs will hold a meeting on November 2..., (75 FR 65038-65039). Detailed meeting agendas and meeting transcripts are available on the NRC...

  3. 78 FR 17945 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-03-25

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant Designs; Notice of Meeting The ACRS Subcommittee on Future Plant Designs will hold a meeting on April 9... were published in the Federal Register on October 18, 2012, (77 FR 64146-64147). Detailed...

  4. 76 FR 16016 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-03-22

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Future Plant Designs The ACRS Subcommittee on Future Plant Designs will hold a meeting on April 5, 2011, at 11545..., 10 a.m. until 5 p.m. The Subcommittee will review the NRC Staff's High Temperature Gas Cooled...

  5. Reactor

    DOEpatents

    Evans, Robert M.

    1976-10-05

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

  6. 14C release from a Soviet-designed pressurized water reactor nuclear power plant.

    PubMed

    Uchrin, G; Csaba, E; Hertelendi, E; Ormai, P; Barnabas, I

    1992-12-01

    The Paks Nuclear Power Plant in Hungary runs with four pressurized water reactors, each of 440-MWe capacity. Sampling systems have been developed and used to determine the 14C of various chemical forms (14CO2, 14CO, 14CnHm) in the airborne releases. The average normalized yearly discharge rates for the time period 1988-1991 are equal to 0.77 TBq GWe-1 y-1 for hydrocarbons and 0.05 TBq GWe-1 y-1 for CO2. The contribution of 14CO was less than 0.5% of the total emission. The 14C discharge rate is estimated to be four times higher than the corresponding mean data of Western European pressurized water reactors. The calculated effective dose equivalent to individuals living in the vicinity of the power plant, due to 14C release, was 0.64 microSv in 1991 while the effective dose equivalent due to the natural 14C level was 15 microSv y-1. The long-term global impact of the 14C release in the operational period of the plant (1982-1991) was 1,270 man-Sv. The 14C excess in the environmental air has been measured since 1989 by taking biweekly samples at a distance of 1.7 km from the nuclear power plant. The long-term average of radiocarbon excess coming from the power plant was 2 mBq m-3. The local 14C deposition was followed by tree ring analysis, too. No 14C increase higher than the uncertainty of the measurement (four per thousand = 0.17 mBq m-3) was observed.

  7. Heat exchanger design considerations for high temperature gas-cooled reactor (HTGR) plants

    SciTech Connect

    McDonald, C.F.; Vrable, D.L.; Van Hagan, T.H.; King, J.H.; Spring, A.H.

    1980-02-01

    Various aspects of the high-temperature heat exchanger conceptual designs for the gas turbine (HTGR-GT) and process heat (HTGR-PH) plants are discussed. Topics include technology background, heat exchanger types, surface geometry, thermal sizing, performance, material selection, mechanical design, fabrication, and the systems-related impact of installation and integration of the units in the prestressed concrete reactor vessel. The impact of future technology developments, such as the utilization of nonmetallic materials and advanced heat exchanger surface geometries and methods of construction, is also discussed.

  8. Assessments of Longevity of Equipment Metal of Nuclear Power Plants equipped with Reactors VVER-1000

    SciTech Connect

    Gorbatykh, V.P.; Al Kassem, S.N.

    2004-07-01

    Characteristics of damage processes of metal of coffer-dams of steam generators collectors at nuclear power plants (NPPs) equipped with reactors VVER-1000 have been mentioned; principles of construction of longevity function has been cited and new approach has been shown while solving the problem of the longevity of the metal resource by substantiating the technological actions with new mode characteristics, performed with the help of specially developed equations and formulae, where practically all damage processes and all influencing factors can be accounted. (authors)

  9. Individual plant examination program: Perspectives on reactor safety and plant performance. Part 1: Final summary report; Volume 1

    SciTech Connect

    1997-12-01

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

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

    SciTech Connect

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

    2012-07-01

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

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

    SciTech Connect

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

    2007-09-01

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

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

    NASA Astrophysics Data System (ADS)

    Sayre, Edwin D.; Ring, Peter J.; Brown, Neil; Elsner, Norbert B.; Bass, John C.

    2008-01-01

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

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

    SciTech Connect

    Sayre, Edwin D.; Ring, Peter J.; Brown, Neil; Elsner, Norbert B.; Bass, John C.

    2008-01-21

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

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

    SciTech Connect

    Bylkin, Boris K.; Davydova, Galina B.; Zverkov, Yuri A.; Krayushkin, Alexander V.; Neretin, Yuri A.; Nosovsky, Anatoly V.; Seyda, Valery A.; Short, Steven M.

    2001-10-15

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

  15. Kalkar nuclear power plant (SNR-300) - A sodium-cooled fast breeder reactor prototype

    SciTech Connect

    Morgenstern, F.H.

    1987-09-01

    The status of the Kalkar nuclear power plant in early summer 1986 is that, apart from later alterations to the workshop building, the assembly and non-nuclear commissioning work has practically been completed. From a technical point of view, nuclear commissioning of the plant can begin, but vital factors for this are the necessary nuclear licenses. The most important licensing prerequisites have been fulfilled;all essential appraisals have been available since January/February 1986. At the beginning of April 1986, the Reactor Safety Commission and the Radiation Protection Commission cast a positive vote for initial fuel loading. Before the accident in Chernobyl, but particularly since then, the issuing of the licenses has come under the political pressure of the commencing election campaign phase for the federal elections in January 1987. The initial project definition phase, the organizational boundary conditions, and the major requirements for the construction of the plant are summarized in chronological form. To provide the total picture, references dealing with general and technical aspects of the project are listed.

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

    SciTech Connect

    J. M. Beck; L. F. Pincock

    2011-04-01

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

  17. Natural and man-made radioactivity in soils and plants around the research reactor of Inshass.

    PubMed

    Higgy, R H; Pimpl, M

    1998-12-01

    The specific radioactivities of the U-series, 232Th, 137Cs and 40K were measured in soil samples around the Inshass reactor in Cairo, using a gamma-ray spectrometer with a HpGe detector. The alpha activity of 238U, 234U and 235U was measured in the same soil samples by surface barrier detectors after radiochemical separation and the obtained results were compared with the specific activities determined by gamma-measurements. The alpha-activity of 238Pu, 239+240Pu, 241Am, 242Cm and 244Cm was measured after radiochemical separation by surface barrier detectors for both soil and plant samples. Then beta-activity of 241Pu was measured using liquid scintillation spectrometry. PMID:9776618

  18. Clinch River Breeder Reactor Plant Steam Generator Few Tube Test model post-test examination

    SciTech Connect

    Impellezzeri, J.R.; Camaret, T.L.; Friske, W.H.

    1981-03-11

    The Steam Generator Few Tube Test (FTT) was part of an extensive testing program carried out in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design. The testing of full-length seven-tube evaporator and three-tube superheater models of the CRBRP design was conducted to provide steady-state thermal/hydraulic performance data to full power per tube and to verify the absence of multi-year endurance problems. This paper describes the problems encountered with the mechanical features of the FTT model design which led to premature test termination, and the results of the post-test examination. Conditions of tube bowing and significant tube and tube support gouging was observed. An interpretation of the visual and metallurgical observations is also presented. The CRBRP steam generator has undergone design evaluations to resolve observed deficiences found in the FFTM.

  19. LWR (Light Water Reactor) power plant simulations using the AD10 and AD100 systems

    SciTech Connect

    Wulff, W.; Cheng, H.S.; Chien, C.J.; Jang, J.Y.; Lin, H.C.; Mallen, A.N.; Wang, S.J.; Institute of Nuclear Energy Research, Lung-Tan; Tawian Power Co., Taipei; Brookhaven National Lab., Upton, NY; Institute of Nuclear Energy Research, Lung-Tan )

    1989-01-01

    Boiling (BWR) and Pressurized (PWR) Water Reactor Power Plants are being simulated at BNL with the AD10 and AD100 Peripheral Processor Systems. The AD10 system has been used for BWR simulations since 1984 for safety analyses, emergency training and optimization studies. BWR simulation capabilities have been implemented recently on the AD100 system and PWR simulation capabilities are currently being developed under the auspices of international cooperation. Modeling and simulation methods are presented with emphasis on the simulation of the Nuclear Steam Supply System. Results are presented for BWR simulation and performance characteristics are compared of the AD10 and AD100 systems. It will be shown that the AD100 simulates two times faster than two AD10 processors operating in parallel and that the computing capacity of one AD100 (with FMU processor) is twice as large as that of two AD10 processors. 9 refs., 5 figs., 1 tab.

  20. 75 FR 5632 - Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-03

    ... emergency alternating current power system. This ISG document provides guidance on the implementation of... EGTG systems that are air cooled and diesel oil fueled are considered in this interim guidance. DATES... COMMISSION Office of New Reactors; Interim Staff Guidance on the Review of Nuclear Power Plant Designs...

  1. 75 FR 61227 - Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Future Plant Designs...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-10-04

    ..., General Electric--Hitachi Nuclear Energy (GEH), and their contractors, pursuant to 5 U.S.C. 552b(c)(4... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Future Plant...

  2. Sedimentation and erosion trends of the Savannah River Plant reactor-discharge creeks. Final report

    SciTech Connect

    Ruby, C.H.; Reinhart, P.J.; Reel, C.L.

    1981-07-01

    The principal objective of this project was to quantify the sedimentological changes that have taken place as a result of the Savannah River Plant's thermal discharges into Steel Creek, Pen Branch, and Four Mile Creek. These thermal discharges have resulted in modifications to creek drainage patterns, channel depths, and surrounding vegetation densities and species. Additionally, they have deposited small deltas at their creek mouths where the creeks enter the broader Savannah River floodplain. The first section of the report, Historical Analysis, delineates sequential changes for each of the creeks from predischarge to 1979. Discharge data for each of the creeks are tabulated. The second section, Profile Analysis, discusses changes in cross-sectional areas of each of the creeks resulting from the various reactor discharges. All profiles are displayed in the appendix. The third section, Stratigraphic Cross-Section Analysis, identifies the depositional sedimentary wedge deposited on each of the creek deltas. Within each cross-section, the predischarge surface has been identified. Depositional facies are described, and the genesis of each is discussed. The appendix contains the profiles and a table that shows the results of the grain-size analysis. Each of the grain-size samples corresponds to a Savannah River Plant sample split.

  3. Neutronic design studies of a conceptual DCLL fusion reactor for a DEMO and a commercial power plant

    NASA Astrophysics Data System (ADS)

    Palermo, I.; Veredas, G.; Gómez-Ros, J. M.; Sanz, J.; Ibarra, A.

    2016-01-01

    Neutronic analyses or, more widely, nuclear analyses have been performed for the development of a dual-coolant He/LiPb (DCLL) conceptual design reactor. A detailed three-dimensional (3D) model has been examined and optimized. The design is based on the plasma parameters and functional materials of the power plant conceptual studies (PPCS) model C. The initial radial-build for the detailed model has been determined according to the dimensions established in a previous work on an equivalent simplified homogenized reactor model. For optimization purposes, the initial specifications established over the simplified model have been refined on the detailed 3D design, modifying material and dimension of breeding blanket, shield and vacuum vessel in order to fulfil the priority requirements of a fusion reactor in terms of the fundamental neutronic responses. Tritium breeding ratio, energy multiplication factor, radiation limits in the TF coils, helium production and displacements per atom (dpa) have been calculated in order to demonstrate the functionality and viability of the reactor design in guaranteeing tritium self-sufficiency, power efficiency, plasma confinement, and re-weldability and structural integrity of the components. The paper describes the neutronic design improvements of the DCLL reactor, obtaining results for both DEMO and power plant operational scenarios.

  4. Impact of mechanical- and maintenance-induced failures of main reactor coolant pump seals on plant safety

    SciTech Connect

    Azarm, M A; Boccio, J L; Mitra, S

    1985-12-01

    This document presents an investigation of the safety impact resulting from mechanical- and maintenance-induced reactor coolant pump (RCP) seal failures in nuclear power plants. A data survey of the pump seal failures for existing nuclear power plants in the US from several available sources was performed. The annual frequency of pump seal failures in a nuclear power plant was estimated based on the concept of hazard rate and dependency evaluation. The conditional probability of various sizes of leak rates given seal failures was then evaluated. The safety impact of RCP seal failures, in terms of contribution to plant core-melt frequency, was also evaluated for three nuclear power plants. For leak rates below the normal makeup capacity and the impact of plant safety were discussed qualitatively, whereas for leak rates beyond the normal make up capacity, formal PRA methodologies were applied. 22 refs., 17 figs., 19 tabs.

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

    SciTech Connect

    Curtis Smith; Scott Beck; Bill Galyean

    2005-09-01

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

  6. An investigation of temperature measurement methods in nuclear power plant reactor pressure vessel annealing

    SciTech Connect

    Acton, R.U.; Gill, W.; Sais, D.J.; Schulze, D.H.; Nakos, J.T.

    1996-05-01

    The objective of this project was to provide an assessment of several methods by which the temperature of a commercial nuclear power plant reactor pressure vessel (RPV) could be measured during an annealing process. This project was a coordinated effort between DOE`s Office of Nuclear Energy, Science and Technology; DOE`s Light Water Reactor Technology Center at Sandia National Laboratories; and the Electric Power Research Institute`s Non- Destructive Evaluation Center. Ball- thermocouple probes similar to those described in NUREG/CR-5760, spring-loaded, metal- sheathed thermocouple probes, and 1778 air- suspended thermocouples were investigated in experiments that heated a section of an RPV wall to simulate a thermal annealing treatment. A parametric study of ball material, emissivity, thermal conductivity, and thermocouple function locations was conducted. Also investigated was a sheathed thermocouple failure mode known as shunting (electrical breakdown of insulation separating the thermocouple wires). Large errors were found between the temperature as measured by the probes and the true RPV wall temperature during heat-up and cool-down. At the annealing soak temperature, in this case 454{degrees}C [850`F], all sensors measured the same temperature within about {plus_minus}5% (23.6{degrees}C [42.5{degrees}F]). Because of these errors, actual RPV wall heating and cooling rates differed from those prescribed (by up to 29%). Shunting does not appear to be a problem under these conditions. The large temperature measurement errors led to the development of a thermal model that predicts the RPV wall temperature from the temperature of a ball- probe. Comparisons between the model and the experimental data for ball-probes indicate that the model could be a useful tool in predicting the actual RPV temperature based on the indicated ball- probe temperature. The model does not predict the temperature as well for the spring-loaded and air suspended probes.

  7. Aging assessment and license renewals: Plant life management for the first stage boiling water reactor

    SciTech Connect

    Tezuka, Kenichi; Kawamura, Shinichi; Aoki, Masataka; Mori, Tsuguo

    1996-09-01

    The first stage Boiling Water Reactors (BWRs) have been operating more than 25 years. Some components have potential of failure by aging. So, evaluations have been done for the main components such as Reactor Pressure Vessel (RPV), Reactor Internals, Primary Coolant Piping, Reactor Recirculation Pump, Cable (Inside PCV), Primary Containment Vessel (PCV) and Concrete Structure. This evaluation has been done by joint study between electric utilities and manufacturers to confirm integrity and identify necessary development.

  8. Demonstration of a 100-kWth high-temperature solar thermochemical reactor pilot plant for ZnO dissociation

    NASA Astrophysics Data System (ADS)

    Koepf, E.; Villasmil, W.; Meier, A.

    2016-05-01

    Solar thermochemical H2O and CO2 splitting is a viable pathway towards sustainable and large-scale production of synthetic fuels. A reactor pilot plant for the solar-driven thermal dissociation of ZnO into metallic Zn has been successfully developed at the Paul Scherrer Institute (PSI). Promising experimental results from the 100-kWth ZnO pilot plant were obtained in 2014 during two prolonged experimental campaigns in a high flux solar simulator at PSI and a 1-MW solar furnace in Odeillo, France. Between March and June the pilot plant was mounted in the solar simulator and in-situ flow-visualization experiments were conducted in order to prevent particle-laden fluid flows near the window from attenuating transparency by blocking incoming radiation. Window flow patterns were successfully characterized, and it was demonstrated that particle transport could be controlled and suppressed completely. These results enabled the successful operation of the reactor between August and October when on-sun experiments were conducted in the solar furnace in order to demonstrate the pilot plant technology and characterize its performance. The reactor was operated for over 97 hours at temperatures as high as 2064 K; over 28 kg of ZnO was dissociated at reaction rates as high as 28 g/min.

  9. Enrichment of AOB and NOB Population by Applying a BABE Reactor in an Activated Sludge Pilot Plant.

    PubMed

    Gatti, Marcela N; Giménez, Juan B; Carretero, Laura; Ruano, María V; Borrás, Luis; Serralta, Joaquín; Seco, Aurora

    2015-04-01

    This paper deals with the effect of a bioaugmentation batch enhanced (BABE) reactor implementation in a biological nutrient removal pilot plant on the populations of ammonia-oxidizing bacteria (AOB) and nitrite-oxidizing bacteria (NOB). The results of fluorescence in situ hybridization (FISH) technique showed that AOB and NOB populations were significantly enhanced, from 4 to 8% and from 2 to 9%, respectively, as a result of the BABE reactor implementation. Regarding AOB, the percentage of Nitrosomonas oligotropha was mainly increased (3 to 6%). Regarding NOB, Nitrospirae spp was greatly enhanced (1 to 7%). Both species are considered K-strategist (high affinity to the substrate, low maximum growth rates) and they usually predominate in reactors with low ammonium and nitrite concentrations, respectively. PMID:26462081

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

    SciTech Connect

    Not Available

    1994-01-15

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

  11. Immobilised native plant cysteine proteases: packed-bed reactor for white wine protein stabilisation.

    PubMed

    Benucci, Ilaria; Lombardelli, Claudio; Liburdi, Katia; Acciaro, Giuseppe; Zappino, Matteo; Esti, Marco

    2016-02-01

    This research presents a feasibility study of using a continuous packed-bed reactor (PBR), containing immobilised native plant cysteine proteases, as a specific and mild alternative technique relative to the usual bentonite fining for white wine protein stabilisation. The operational parameters for a PBR containing immobilised bromelain (PBR-br) or immobilised papain (PBR-pa) were optimised using model wine fortified with synthetic substrate (Bz-Phe-Val-Arg-pNA). The effectiveness of PBR-br, both in terms of hazing potential and total protein decrease, was significantly higher than PBR-pa, in all the seven unfined, white wines used. Among the wines tested, Sauvignon Blanc, given its total protein content as well as its very high intrinsic instability, was selected as a control wine to evaluate the effect of the treatment on wine as to its soluble protein profile, phenolic composition, mineral component, and sensory properties. The treatment in a PBR containing immobilised bromelain appeared effective in decreasing both wine hazing potential and total protein amount, while it did not significantly affect the phenol compounds, the mineral component nor the sensory quality of wine. The enzymatic treatment in PBR was shown to be a specific and mild technique for use as an alternative to bentonite fining for white wine protein stabilisation. PMID:27162393

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

    SciTech Connect

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

    1997-04-01

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

  13. Immobilised native plant cysteine proteases: packed-bed reactor for white wine protein stabilisation.

    PubMed

    Benucci, Ilaria; Lombardelli, Claudio; Liburdi, Katia; Acciaro, Giuseppe; Zappino, Matteo; Esti, Marco

    2016-02-01

    This research presents a feasibility study of using a continuous packed-bed reactor (PBR), containing immobilised native plant cysteine proteases, as a specific and mild alternative technique relative to the usual bentonite fining for white wine protein stabilisation. The operational parameters for a PBR containing immobilised bromelain (PBR-br) or immobilised papain (PBR-pa) were optimised using model wine fortified with synthetic substrate (Bz-Phe-Val-Arg-pNA). The effectiveness of PBR-br, both in terms of hazing potential and total protein decrease, was significantly higher than PBR-pa, in all the seven unfined, white wines used. Among the wines tested, Sauvignon Blanc, given its total protein content as well as its very high intrinsic instability, was selected as a control wine to evaluate the effect of the treatment on wine as to its soluble protein profile, phenolic composition, mineral component, and sensory properties. The treatment in a PBR containing immobilised bromelain appeared effective in decreasing both wine hazing potential and total protein amount, while it did not significantly affect the phenol compounds, the mineral component nor the sensory quality of wine. The enzymatic treatment in PBR was shown to be a specific and mild technique for use as an alternative to bentonite fining for white wine protein stabilisation.

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

    SciTech Connect

    Heckle, Wm. Lloyd; Bolian, Tricia W.

    2006-07-01

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

  15. Application of Computational Fluid Dynamics Model to Disinfection Reactors in Water Reclamation Plants

    NASA Astrophysics Data System (ADS)

    Helmns, Andrea; Texeira, Pablo; Issakhanian, Emin; Saez, Jose

    2014-11-01

    California's current drought has renewed public interest in recycled water from Water Reclamation Plants (WRPs). It is critical that the recycled water meets public health standards. This project consists of simulating the transport of an instantaneous conservative tracer through the chlorine contact tanks at two WRPs in California, where recycled water regulations stipulate a minimum 90-minute modal contact time during disinfection at peak dry weather design flow. Computational Fluid Dynamics (CFD) is used to model the turbulent flow, transport, and contact time of a conservative solute for several real operating scenarios. Given as-built drawings and operation parameters, the chlorine contact tanks are modeled to match actual geometries and flow conditions. The turbulent flow solutions are used as the basis to model the transport of a turbulently diffusing conservative tracer added instantaneously to the inlet of the reactors. This tracer simulates the transport through advection and dispersion of chlorine in the WRPs. Breakthrough curves of the tracer at the outlet are used to determine the modal contact times.

  16. Space and Terrestrial Power System Integration Optimization Code BRMAPS for Gas Turbine Space Power Plants With Nuclear Reactor Heat Sources

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.

    2007-01-01

    In view of the difficult times the US and global economies are experiencing today, funds for the development of advanced fission reactors nuclear power systems for space propulsion and planetary surface applications are currently not available. However, according to the Energy Policy Act of 2005 the U.S. needs to invest in developing fission reactor technology for ground based terrestrial power plants. Such plants would make a significant contribution toward drastic reduction of worldwide greenhouse gas emissions and associated global warming. To accomplish this goal the Next Generation Nuclear Plant Project (NGNP) has been established by DOE under the Generation IV Nuclear Systems Initiative. Idaho National Laboratory (INL) was designated as the lead in the development of VHTR (Very High Temperature Reactor) and HTGR (High Temperature Gas Reactor) technology to be integrated with MMW (multi-megawatt) helium gas turbine driven electric power AC generators. However, the advantages of transmitting power in high voltage DC form over large distances are also explored in the seminar lecture series. As an attractive alternate heat source the Liquid Fluoride Reactor (LFR), pioneered at ORNL (Oak Ridge National Laboratory) in the mid 1960's, would offer much higher energy yields than current nuclear plants by using an inherently safe energy conversion scheme based on the Thorium --> U233 fuel cycle and a fission process with a negative temperature coefficient of reactivity. The power plants are to be sized to meet electric power demand during peak periods and also for providing thermal energy for hydrogen (H2) production during "off peak" periods. This approach will both supply electric power by using environmentally clean nuclear heat which does not generate green house gases, and also provide a clean fuel H2 for the future, when, due to increased global demand and the decline in discovering new deposits, our supply of liquid fossil fuels will have been used up. This is

  17. Review of the cost estimate and schedule for the 2240-MWt high-temperature gas-cooled reactor steam-cycle/cogeneration lead plant

    SciTech Connect

    Not Available

    1983-09-01

    This report documents Bechtel's review of the cost estimate and schedule for the 2240 MWt High Temperature Gas-Cooled Reactor Steam Cycle/Cogeneration (HTGR-SC/C) Lead Plant. The overall objective of the review is to verify that the 1982 update of the cost estimate and schedule for the Lead Plant are reasonable and consistent with current power plant experience.

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

    SciTech Connect

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

    2010-06-01

    A reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production was developed to provide a basis for comparing the HTE concept with other hydrogen production concepts. The reference plant design is driven by a high-temperature helium-cooled nuclear reactor coupled to a direct Brayton power cycle. The reference design reactor power is 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 540°C and 900°C, respectively. The electrolysis unit used to produce hydrogen includes 4,009,177 cells with a per-cell active area of 225 cm2. The optimized design for the reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes an air-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The inlet air for the air-sweep system is compressed to the system operating pressure of 5.0 MPa in a four-stage compressor with intercooling. The alternating current (AC) to direct current (DC) conversion efficiency is 96%. The overall system thermal-to-hydrogen production efficiency (based on the lower heating value of the produced hydrogen) is 47.1% at a hydrogen production rate of 2.356 kg/s. An economic analysis of this plant was performed using the standardized H2A Analysis Methodology developed by the Department of Energy (DOE) Hydrogen Program, and using realistic financial and cost estimating assumptions. The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.23/kg of hydrogen was calculated assuming an internal rate of return of 10%.

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

    SciTech Connect

    Zdarek, J.; Pecinka, L.

    1997-04-01

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

  20. Modeling of Coolant Flow in the Fuel Assembly of the Reactor of a Floating Nuclear Power Plant Using the Logos CFD Program

    NASA Astrophysics Data System (ADS)

    Dmitriev, S. M.; Dobrov, A. A.; Legchanov, M. A.; Khrobostov, A. E.

    2015-09-01

    Results of computer modeling of coolant flow in the fuel assembly of the reactor of a floating nuclear power plant using the LOGOS CFD programs have been given. The possibility of using the obtained results to improve models built into the engineering programs of thermohydraulic calculation of nuclear-reactor cores has been considered.

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

    SciTech Connect

    Lata

    1996-09-26

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

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

    SciTech Connect

    S. Blaine Grover

    2009-05-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Program will be irradiating six gas reactor graphite creep experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The ATR has a long history of irradiation testing in support of reactor development and the INL has been designated as the new United States Department of Energy’s lead laboratory for nuclear energy development. The ATR is one of the world’s premiere test reactors for performing long term, high flux, and/or large volume irradiation test programs. These graphite irradiations are being accomplished to support development of the next generation reactors in the United States. The graphite experiments will be irradiated over the next six to eight years to support development of a graphite irradiation performance data base on the new nuclear grade graphites now available for use in high temperature gas reactors. The goals of the irradiation experiments are to obtain irradiation performance data at different temperatures and loading conditions to support design of the Next Generation Nuclear Plant Very High Temperature Gas Reactor, as well as other future gas reactors. The experiments will each consist of a single capsule that will contain seven separate stacks of graphite specimens. Six of the specimen stacks will have half of their graphite specimens under a compressive load, while the other half of the specimens will not be subjected to a compressive load during irradiation. The six stacks will be organized into pairs with a different compressive load being applied to the top half of each pair of specimen stacks. The seventh stack will not have a compressive load on the graphite specimens during irradiation. The specimens will be irradiated in an inert sweep gas atmosphere with on-line temperature and compressive load monitoring and control. There will also be the capability of sampling the sweep gas effluent to determine if any

  3. The use of supercritical parameters of a coolant—A promising path to development of nuclear power plant water-cooled reactors in the 21st century

    NASA Astrophysics Data System (ADS)

    Gabaraev, B. A.; Smolin, V. N.; Solov'ev, S. L.

    2006-09-01

    A modern concept of the development of water-cooled reactors of nuclear power plants (NPPs) is considered. Data on the design of NPPs with supercritical-parameters coolant and the results of experimental studies are presented.

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

    SciTech Connect

    Moisseytsev, A.; Sienicki, J.

    2011-04-12

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

  5. Aging of the containment pressure boundary in light-water reactor plants

    SciTech Connect

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

    1997-01-01

    Research is being conducted by the Oak Ridge National Laboratory to address aging of the containment pressure boundary in light-water reactor plants. The objectives of this work are to (1) identify the significant factors related to occurrence of corrosion, efficacy of inspection, and structural capacity reduction of steel containments and liners of concrete containments, and to make recommendations on use of risk models in regulatory decisions; (2) provide NRC reviewers a means of establishing current structural capacity margins for steel containments, and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by NRC reviewers in assessing the seriousness of reported incidences of containment degradation. In meeting these objectives research is being conducted in two primary task areas - pressure boundary condition assessment and root-cause resolution practices, and reliability-based condition assessments. Under the first task area a degradation assessment methodology was developed for use in characterizing the in-service condition of metal and concrete containment pressure boundary components and quantifying the amount of damage that is present. An assessment of available destructive and nondestructive techniques for examining steel containments and liners is ongoing. Under the second task area quantitative structural reliability analysis methods are being developed for application to degraded metallic pressure boundaries to provide assurances that they will be able to withstand future extreme loads during the desired service period with a level of reliability that is sufficient for public safety. To date, mathematical models that describe time-dependent changes in steel due to aggressive environmental factors have been identified, and statistical data supporting their use in time-dependent reliability analysis have been summarized.

  6. Coagulant recovery from water treatment plant sludge and reuse in post-treatment of UASB reactor effluent treating municipal wastewater.

    PubMed

    Nair, Abhilash T; Ahammed, M Mansoor

    2014-09-01

    In the present study, feasibility of recovering the coagulant from water treatment plant sludge with sulphuric acid and reusing it in post-treatment of upflow anaerobic sludge blanket (UASB) reactor effluent treating municipal wastewater were studied. The optimum conditions for coagulant recovery from water treatment plant sludge were investigated using response surface methodology (RSM). Sludge obtained from plants that use polyaluminium chloride (PACl) and alum coagulant was utilised for the study. Effect of three variables, pH, solid content and mixing time was studied using a Box-Behnken statistical experimental design. RSM model was developed based on the experimental aluminium recovery, and the response plots were developed. Results of the study showed significant effects of all the three variables and their interactions in the recovery process. The optimum aluminium recovery of 73.26 and 62.73 % from PACl sludge and alum sludge, respectively, was obtained at pH of 2.0, solid content of 0.5 % and mixing time of 30 min. The recovered coagulant solution had elevated concentrations of certain metals and chemical oxygen demand (COD) which raised concern about its reuse potential in water treatment. Hence, the coagulant recovered from PACl sludge was reused as coagulant for post-treatment of UASB reactor effluent treating municipal wastewater. The recovered coagulant gave 71 % COD, 80 % turbidity, 89 % phosphate, 77 % suspended solids and 99.5 % total coliform removal at 25 mg Al/L. Fresh PACl also gave similar performance but at higher dose of 40 mg Al/L. The results suggest that coagulant can be recovered from water treatment plant sludge and can be used to treat UASB reactor effluent treating municipal wastewater which can reduce the consumption of fresh coagulant in wastewater treatment.

  7. Modified ADM1 for modelling an UASB reactor laboratory plant treating starch wastewater and synthetic substrate load tests.

    PubMed

    Hinken, L; Huber, M; Weichgrebe, D; Rosenwinkel, K-H

    2014-11-01

    A laboratory plant consisting of two UASB reactors was used for the treatment of industrial wastewater from the wheat starch industry. Several load tests were carried out with starch wastewater and the synthetic substrates glucose, acetate, cellulose, butyrate and propionate to observe the impact of changing loads on gas yield and effluent quality. The measurement data sets were used for calibration and validation of the Anaerobic Digestion Model No. 1 (ADM1). For a precise simulation of the detected glucose degradation during load tests with starch wastewater and glucose, it was necessary to incorporate the complete lactic acid fermentation into the ADM1, which contains the formation and degradation of lactate and a non-competitive inhibition function. The modelling results of both reactors based on the modified ADM1 confirm an accurate calculation of the produced gas and the effluent concentrations. Especially, the modelled lactate effluent concentrations for the load cases are similar to the measurements and justified by literature.

  8. Performance evaluation of planted and unplanted subsurface-flow constructed wetlands for the post-treatment of UASB reactor effluents.

    PubMed

    Dornelas, Filipe Lima; Machado, Matheus Boechat; von Sperling, Marcos

    2009-01-01

    A system comprised by a UASB (Upflow Anaerobic Sludge Blanket) reactor followed by two horizontal subsurface-flow constructed wetlands in parallel was evaluated for the treatment of the wastewater generated in the city of Belo Horizonte, Brazil (50 inhabitants each unit). One unit was planted (Typha latifolia) and the other was unplanted. Influent and effluent samples were collected for a period of seven months. The systems were able to produce final effluents with low concentrations of organic matter and suspended solids, but showed not to be efficient in the removal of nutrients. Mean effluent concentrations for the planted and unplanted units were, respectively: BOD: 15 and 19 mg/L; COD: 42 and 64 mg/L; TSS: 3 and 5 mg/L; TN: 27 and 33 mg/L; N-NH(3): 25 and 29 mg/L; P Total: 1.2 and 1.5 mg/L. The planted wetland presented effluent concentrations and removal efficiencies significantly (Wilcoxon matched-pairs test, 5% significance level) better than the unplanted unit for most constituents. The study shows that horizontal subsurface-flow constructed wetlands can be effectively used as a post-treatment option for the effluent from UASB reactors.

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

    SciTech Connect

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

    1984-07-01

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

  10. Analysis and selection of high pressure heaters design for a new generation of NPP with BN-1200 reactor plant

    NASA Astrophysics Data System (ADS)

    Yurchenko, A. Yu.; Sukhorukov, Yu. G.; Trifonov, N. N.; Grigor'eva, E. B.; Esin, S. B.; Svyatkin, F. A.; Nikolaenkova, E. K.; Prikhod'ko, P. Yu.; Nazarov, V. V.

    2016-09-01

    In the development of advanced high-power steam-turbine plants (STP), special attention is placed on the design of reliable and economical high-pressure heater (HPH) capable to maintain the specified thermal hydraulic performance during the entire service life. Comparative analysis of the known designs of HPH, such as the spiral-collector HPH, the collector-coiled HPH, the collector-platen HPH, modular HPH, and the chamber HPH, was carried out. The advantages and disadvantages of each design were pointed. For better comparison, the heaters are separated into two groups—horizontal and vertical ones. The weight and dimension characteristics, the materials and features of the basic elements, and operating features of variety HPH are presented. At operating the spiral-collector HPH used in the majority of regenerative schemes of high-pressure STP of thermal and nuclear power plants, the disadvantages reducing the economy and reliability of their operation were revealed. The recommendations directed to the reliability growth of HPH, the decrease of subcooling the feed water, the increase of compactness are stated. Some of these were developed by the specialists of OAO NPO TsKTI and are successfully implemented on the thermal power plants and nuclear power plants. Technical solutions to reduce the cost of regeneration system and the weight of chamber HPH, reduce the thickness of the tube plate of HPH, and reliability assurance of the cooler of steam and condensate built in the HPH casing under all operating conditions were proposed. Three types of feed water chambers for vertical and horizontal chamber HPH are considered in detail, the constructive solutions that have been implemented in HPH of the regeneration system of turbines of 1000 and 1200 MW capacity with water-moderated water-cooled power reactor (WMWCPR) are described. The optimal design of HPH for the regeneration system of high-pressure turbine plant with BN-1200 reactor was selected.

  11. Dynamic Comparison of Three- and Four-Equation Reactor Core Models in a Full-Scope Power Plant Training Simulator

    SciTech Connect

    Espinosa-Paredes, Gilberto; Alvarez-Ramirez, Jose; Nunez-Carrera, Alejandro; Garcia-Gutierrez, Alfonso; Martinez-Mendez, Elizabeth Jeannette

    2004-02-15

    A comparative analysis of the dynamic behavior of a boiling water reactor in a full-scope power plant simulator for operator training is presented. Three- and four-equation reactor core models were used to examine three transients following tests described in acceptance test procedures: scram, loss of feedwater flow, and closure of main isolation valves. The three-equation model consists of water and steam mixture momentum, including mass and energy balances. The four-equation model is based on liquid and gas phase mass balances, together with a drift-flux approach for the analysis of phase separation. Analysis of the models showed that the scram transient was slightly different for three- and four-equation models. The drift-flux effects can explain such differences. Regarding the loss-of-feedwater transient, the predicted steam flow after scram is larger for the three-equation model. Finally, for the transient related to the closure of main steam isolation valves, the three-equation model provides slightly different results for the pressure change, which affects reactor level behavior.

  12. Utilization of the Philippine Research Reactor as a training facility for nuclear power plant operators

    SciTech Connect

    Palabrica, R.J.

    1981-01-01

    The Philippines has a 1-MW swimming-pool reactor facility operated by the Philippine Atomic Energy Commission (PAEC). The reactor is light-water moderated and cooled, graphite reflected, and fueled with 90% enriched uranium. Since it became critical in 1963 it has been utilized for research, radioisotope production, and training. It was used initially in the training of PAEC personnel and other research institutions and universities. During the last few years, however, it has played a key role in training personnel for the Philippine Nuclear Power Project (PNPP).

  13. Reactor Physics Parametric and Depletion Studies in Support of TRISO Particle Fuel Specification for the Next Generation Nuclear Plant

    SciTech Connect

    James W. Sterbentz; Bren Phillips; Robert L. Sant; Gray S. Chang; Paul D. Bayless

    2003-09-01

    Reactor physics calculations were initiated to answer several major questions related to the proposed TRISO-coated particle fuel that is to be used in the prismatic Very High Temperature Reactor (VHTR) or the Next Generation Nuclear Plant (NGNP). These preliminary design evaluation calculations help ensure that the upcoming fuel irradiation tests will test appropriate size and type of fuel particles for a future NGNP reactor design. Conclusions from these calculations are expected to confirm and suggest possible modifications to the current particle fuel parameters specified in the evolving Fuel Specification. Calculated results dispel the need for a binary fuel particle system, which is proposed in the General Atomics GT-MHR concept. The GT-MHR binary system is composed of both a fissile and fertile particle with 350- and 500- micron kernel diameters, respectively. For the NGNP reactor, a single fissile particle system (single UCO kernel size) can meet the reactivity and power cycle length requirements demanded of the NGNP. At the same time, it will provide substantial programmatic cost savings by eliminating the need for dual particle fabrication process lines and dual fuel particle irradiation tests required of a binary system. Use of a larger 425-micron kernel diameter single fissile particle (proposed here), as opposed to the 350-micron GT-MHR fissile particle size, helps alleviate current compact particle packing fractions fabrication limitations (<35%), improves fuel block loading for higher n-batch reload options, and tracks the historical correlation between particle size and enrichment (10 and 14 wt% U-235 particle enrichments are proposed for the NGNP). Overall, the use of the slightly larger kernel significantly broadens the NGNP reactor core design envelope and provides increased design margin to accommodate the (as yet) unknown final NGNP reactor design. Maximum power-peaking factors are calculated for both the initial and equilibrium NGNP cores

  14. The SAM software system for modeling severe accidents at nuclear power plants equipped with VVER reactors on full-scale and analytic training simulators

    NASA Astrophysics Data System (ADS)

    Osadchaya, D. Yu.; Fuks, R. L.

    2014-04-01

    The architecture of the SAM software package intended for modeling beyond-design-basis accidents at nuclear power plants equipped with VVER reactors evolving into a severe stage with core melting and failure of the reactor pressure vessel is presented. By using the SAM software package it is possible to perform comprehensive modeling of the entire emergency process from the failure initiating event to the stage of severe accident involving meltdown of nuclear fuel, failure of the reactor pressure vessel, and escape of corium onto the concrete basement or into the corium catcher with retention of molten products in it.

  15. Reference modular High Temperature Gas-Cooled Reactor Plant: Concept description report

    SciTech Connect

    Not Available

    1986-10-01

    This report provides a summary description of the Modular High Temperature Gas-Cooled Reactor (MHTGR) concept and interim results of assessments of costs, safety, constructibility, operability, maintainability, and availability. Conceptual design of this concept was initiated in October 1985 and is scheduled for completion in 1987. Participating industrial contractors are Bechtel National, Inc. (BNI), Stone and Webster Engineering Corporation (SWEC), GA Technologies, Inc. (GA), General Electric Co. (GE), and Combustion Engineering, Inc. (C-E).

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

    SciTech Connect

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

    1984-07-01

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

  17. On the fusion triple product and fusion power gain of tokamak pilot plants and reactors

    NASA Astrophysics Data System (ADS)

    Costley, A. E.

    2016-06-01

    The energy confinement time of tokamak plasmas scales positively with plasma size and so it is generally expected that the fusion triple product, nTτ E, will also increase with size, and this has been part of the motivation for building devices of increasing size including ITER. Here n, T, and τ E are the ion density, ion temperature and energy confinement time respectively. However, tokamak plasmas are subject to operational limits and two important limits are a density limit and a beta limit. We show that when these limits are taken into account, nTτ E becomes almost independent of size; rather it depends mainly on the fusion power, P fus. In consequence, the fusion power gain, Q fus, a parameter closely linked to nTτ E is also independent of size. Hence, P fus and Q fus, two parameters of critical importance in reactor design, are actually tightly coupled. Further, we find that nTτ E is inversely dependent on the normalised beta, β N; an unexpected result that tends to favour lower power reactors. Our findings imply that the minimum power to achieve fusion reactor conditions is driven mainly by physics considerations, especially energy confinement, while the minimum device size is driven by technology and engineering considerations. Through dedicated R&D and parallel developments in other fields, the technology and engineering aspects are evolving in a direction to make smaller devices feasible.

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

    SciTech Connect

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

    2013-05-01

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

  19. Purge gas recovery of ammonia synthesis plant by integrated configuration of catalytic hydrogen-permselective membrane reactor and solid oxide fuel cell as a novel technology

    NASA Astrophysics Data System (ADS)

    Siavashi, Fakhteh; Saidi, Majid; Rahimpour, Mohammad Reza

    2014-12-01

    The purge gas emission of ammonia synthesis plant which contains hazardous components is one of the major sources of environmental pollution. Using integrated configuration of catalytic hydrogen-permselective membrane reactor and solid oxide fuel cell (SOFC) system is a new approach which has a great impact to reduce the pollutant emission. By application of this method, not only emission of ammonia and methane in the atmosphere is prevented, hydrogen is produced through the methane steam reforming and ammonia decomposition reactions that take place simultaneously in a catalytic membrane reactor. The pure generated hydrogen by recovery of the purge gas in the Pd-Ag membrane reactor is used as a feed of SOFC. Since water is the only byproduct of the electrochemical reaction in the SOFC, it is recycled to the reactor for providing the required water of the reforming reaction. Performance investigation of the reactor represents that the rate of hydrogen permeation increases with enhancing the reactor temperature and pressure. Also modeling results indicate that the SOFC performance improves with increasing the temperature and fuel utilization ratio. The generated power by recovery of the purging gas stream of ammonia synthesis plant in the Razi petrochemical complex is about 8 MW.

  20. dsrAB-based analysis of sulphate-reducing bacteria in moving bed biofilm reactor (MBBR) wastewater treatment plants.

    PubMed

    Biswas, Kristi; Taylor, Michael W; Turner, Susan J

    2014-08-01

    Sulphate-reducing bacteria (SRB) are important members of the sulphur cycle in wastewater treatment plants (WWTPs). In this study, we investigate the diversity and activity of SRB within the developing and established biofilm of two moving bed biofilm reactor (MBBR) systems treating municipal wastewater in New Zealand. The larger of the two WWTPs (Moa Point) generates high levels of sulphide relative to the smaller Karori plant. Clone libraries of the dissimilatory (bi)sulphite reductase (dsrAB) genes and quantitative real-time PCR targeting dsrA transcripts were used to compare SRB communities between the two WWTPs. Desulfobulbus (35-53 % of total SRB sequences) and genera belonging to the family Desulfobacteraceae (27-41 %) dominated the SRB fraction of the developing biofilm on deployed plastic carriers at both sites, whereas Desulfovibrio and Desulfomicrobium were exclusively found at Moa Point. In contrast, the established biofilms from resident MBBR carriers were largely dominated by Desulfomonile tiedjei-like organisms (58-100 % of SRB sequences). The relative transcript abundance of dsrA genes (signifying active SRBs) increased with biofilm weight yet remained low overall, even in the mature biofilm stage. Our results indicate that although SRB are both present and active in the microbial community at both MBBR study sites, differences in the availability of sulphate may be contributing to the observed differences in sulphide production at these two plants.

  1. Assessment of impacts at the advanced test reactor as a result of chemical releases at the Idaho Chemical Processing Plant

    SciTech Connect

    Rood, A.S.

    1991-02-01

    This report provides an assessment of potential impacts at the Advanced Test Reactor Facility (ATR) resulting from accidental chemical spill at the Idaho Chemical Processing Plant (ICPP). Spills postulated to occur at the Lincoln Blvd turnoff to ICPP were also evaluated. Peak and time weighted average concentrations were calculated for receptors at the ATR facility and the Test Reactor Area guard station at a height above ground level of 1.0 m. Calculated concentrations were then compared to the 15 minute averaged Threshold Limit Value - Short Term Exposure Limit (TLV-STEL) and the 30 minute averaged Immediately Dangerous to Life and Health (IDLH) limit. Several different methodologies were used to estimate source strength and dispersion. Fifteen minute time weighted averaged concentrations of hydrofluoric acid and anhydrous ammonia exceeded TLV-STEL values for the cases considered. The IDLH value for these chemicals was not exceeded. Calculated concentrations of ammonium hydroxide, hexone, nitric acid, propane, gasoline, chlorine and liquid nitrogen were all below the TLV-STEL value.

  2. Removal of COD and nitrogen from animal food plant wastewater in an intermittently-aerated structured-bed reactor.

    PubMed

    Wosiack, Priscila Arcoverde; Lopes, Deize Dias; Rissato Zamariolli Damianovic, Márcia Helena; Foresti, Eugenio; Granato, Daniel; Barana, Ana Cláudia

    2015-05-01

    This study evaluated the performance of a continuous flow structured-bed reactor in the simultaneous removal of total nitrogen (TN) and chemical oxygen demand (COD) in the effluent from an animal food plant. The reactor had an intermittent aeration system; hydraulic retention time (HRT) of one day; temperature of 30 °C; and recirculation ratio of five times the flow. An experimental central composite rotational delineation (CCRD) type design was used to define the aeration conditions and nitrogen load (factors) to be studied. Response surface methodology was used to analyse the influence of the factors above the results, the removal of TN and COD. It was observed that the aeration factor showed the greatest significance for the results and that the affluent TKN concentration did not have a significant effect, at a 95% level of confidence, on COD removal. Throughout the experiment, the COD/N ratio remained between 3.2 and 3.8. The best results for COD and TN removal, 80% and 88%, respectively, were obtained with 158 min of aeration on a cycle of 180 min and 255 mg L(-1) of Total Kjeldahl Nitrogen (TKN) in the substrate.

  3. Helium heater design for the helium direct cycle component test facility. [for gas-cooled nuclear reactor power plant

    NASA Technical Reports Server (NTRS)

    Larson, V. R.; Gunn, S. V.; Lee, J. C.

    1975-01-01

    The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.

  4. The French nuclear power plant reactor building containment contributions of prestressing and concrete performances in reliability improvements and cost savings

    SciTech Connect

    Rouelle, P.; Roy, F.

    1998-12-31

    The Electricite de France`s N4 CHOOZ B nuclear power plant, two units of the world`s largest PWR model (1450 Mwe each), has earned the Electric Power International`s 1997 Powerplant Award. This lead NPP for EDF`s N4 series has been improved notably in terms of civil works. The presentation will focus on the Reactor Building`s inner containment wall which is one of the main civil structures on a technical and safety point of view. In order to take into account the necessary evolution of the concrete technical specification such as compressive strength low creep and shrinkage, the HSC/HPC has been used on the last N4 Civaux 2 NPP. As a result of the use of this type of professional concrete, the containment withstands an higher internal pressure related to severe accident and ensures higher level of leak-tightness, thus improving the overall safety of the NPP. On that occasion, a new type of prestressing has been tested locally through 55 C 15 S tendons using a new C 1500 FE Jack. These updated civil works techniques shall allow EDF to ensure a Reactor Containment lifespan for more than 50 years. The gains in terms of reliability and cost saving of these improved techniques will be developed hereafter.

  5. 75 FR 38564 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Subcommittee on Plant Operations...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-07-02

    ... published in the Federal Register on October 14, 2009 (74 FR 58268-58269). Detailed meeting agendas and... and Fire Protection The ACRS Subcommittee on Plant Operations and Fire Protection will hold a...

  6. Energy potential and alternative usages of biogas and sludge from UASB reactors: case study of the Laboreaux wastewater treatment plant.

    PubMed

    Rosa, A P; Conesa, J A; Fullana, A; Melo, G C B; Borges, J M; Chernicharo, C A L

    2016-01-01

    This work assessed the energy potential and alternative usages of biogas and sludge generated in upflow anaerobic sludge blanket reactors at the Laboreaux sewage treatment plant (STP), Brazil. Two scenarios were considered: (i) priority use of biogas for the thermal drying of dehydrated sludge and the use of the excess biogas for electricity generation in an ICE (internal combustion engine); and (ii) priority use of biogas for electricity generation and the use of the heat of the engine exhaust gases for the thermal drying of the sludge. Scenario 1 showed that the electricity generated is able to supply 22.2% of the STP power demand, but the thermal drying process enables a greater reduction or even elimination of the final volume of sludge to be disposed. In Scenario 2, the electricity generated is able to supply 57.6% of the STP power demand; however, the heat in the exhaust gases is not enough to dry the total amount of dehydrated sludge.

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

    SciTech Connect

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

    2011-10-01

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

  8. Gas Turbine Energy Conversion Systems for Nuclear Power Plants Applicable to LiFTR Liquid Fluoride Thorium Reactor Technology

    NASA Technical Reports Server (NTRS)

    Juhasz, Albert J.

    2014-01-01

    This panel plans to cover thermal energy and electric power production issues facing our nation and the world over the next decades, with relevant technologies ranging from near term to mid-and far term.Although the main focus will be on ground based plants to provide baseload electric power, energy conversion systems (ECS) for space are also included, with solar- or nuclear energy sources for output power levels ranging tens of Watts to kilo-Watts for unmanned spacecraft, and eventual mega-Watts for lunar outposts and planetary surface colonies. Implications of these technologies on future terrestrial energy systems, combined with advanced fracking, are touched upon.Thorium based reactors, and nuclear fusion along with suitable gas turbine energy conversion systems (ECS) will also be considered by the panelists. The characteristics of the above mentioned ECS will be described, both in terms of their overall energy utilization effectiveness and also with regard to climactic effects due to exhaust emissions.

  9. Acetylcholinesterase immobilized capillary reactors coupled to protein coated magnetic beads: A new tool for plant extract ligand screening

    PubMed Central

    Vanzolini, Kenia Lourenço; Jiang, Zhengjin; Zhang, Xiaoqi; Vieira, Lucas Campos Curcino; Corrêa, Arlene Gonçalvez; Cardoso, Carmen Lucia; Cass, Quezia Bezerra; Moaddel, Ruin

    2013-01-01

    The use of immobilized capillary enzyme reactors (ICERs) and enzymes coated to magnetic beads ((NT or CT)-MB) for ligand screening has been adopted as a new technique of high throughput screening (HTS). In this work the selected target was the enzyme acetylcholinesterase (AChE), which acts on the central nervous system and is a validated target for the treatment of Alzheimer’s disease, as well as for new insecticides. A new approach for the screening of plant extracts was developed based on the ligand fishing experiments and zonal chromatography. For that, the magnetic beads were used for the ligand fishing experiments and capillary bioreactors for the activity assays. The latter was employed also under non-linear conditions to determine the affinity constants of known ligands, for the first time, as well as for the active fished ligand. PMID:24148457

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

    SciTech Connect

    Simon, A J

    2009-08-21

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

  11. Light water reactor program

    SciTech Connect

    Franks, S.M.

    1994-12-31

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

  12. Analysis of operation of filters for post-accident decontamination of pressurized rooms of a nuclear power plants with a type VVER-440 reactor

    NASA Astrophysics Data System (ADS)

    Zaichik, L. I.; Zeigarnik, Yu. A.; Rotinov, A. G.; Sidorov, A. S.; Silina, N. N.; Chalyi, R. F.

    2007-05-01

    Operation of filters of postaccident decontamination of pressurized rooms of a nuclear power plant with a type-VVER-440 reactor is analyzed. The distribution of radioactive nuclides over filter stages, the time variation of the thermal state of filter, and the characteristic features of the processes of sorption in the section of fine cleaning are considered.

  13. Light Water Reactor Sustainability Program: Computer-Based Procedures for Field Activities: Results from Three Evaluations at Nuclear Power Plants

    SciTech Connect

    Oxstrand, Johanna; Le Blanc, Katya; Bly, Aaron

    2014-09-01

    The Computer-Based Procedure (CBP) research effort is a part of the Light-Water Reactor Sustainability (LWRS) Program, which is a research and development (R&D) program sponsored by Department of Energy (DOE) and performed in close collaboration with industry R&D programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants. One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. One area that could yield tremendous savings in increased efficiency and safety is in improving procedure use. Nearly all activities in the nuclear power industry are guided by procedures, which today are printed and executed on paper. This paper-based procedure process has proven to ensure safety; however, there are improvements to be gained. Due to its inherent dynamic nature, a CBP provides the opportunity to incorporate context driven job aids, such as drawings, photos, and just-in-time training. Compared to the static state of paper-based procedures (PBPs), the presentation of information in CBPs can be much more flexible and tailored to the task, actual plant condition, and operation mode. The dynamic presentation of the procedure will guide the user down the path of relevant steps, thus minimizing time spent by the field worker to evaluate plant conditions and decisions related to the applicability of each step. This dynamic presentation of the procedure also minimizes the risk of conducting steps out of order and/or incorrectly assessed applicability of steps.

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

    SciTech Connect

    Not Available

    1991-07-01

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

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

    SciTech Connect

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

    2013-03-01

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

  16. International safeguards for a light-water reactor fuels reprocessing plant: containment and surveillance concepts

    SciTech Connect

    Cameron, C.P.; Bleck, M.E.

    1980-12-01

    Concepts for containment/surveillance for reprocessing plants are described, conceptual designs are developed, and their effectiveness is evaluated. A technical approach to design of containment/surveillance systems is presented, and design considerations are discussed. This is the second in a series of reports. The first described the basis for the study of international safeguards for reprocessing plants. In this second report, only containment/surveillance is discussed. The third report will discuss the integration of concepts for containment/surveillance and material accountancy.

  17. 75 FR 55365 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-10

    ... licensing activities associated with the staff's review of the Watts Bar Nuclear Plant Unit 2 Operating..., (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available on the NRC Web... the Chairman as necessary to facilitate the conduct of the meeting, persons planning to attend...

  18. 76 FR 38211 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-29

    ... Report, Supplement 23 (SSER23) associated with the staff's review of the Watts Bar Nuclear Plant, Unit 2... were published in the Federal Register on October 21, 2010, (75 FR 65038- 65039). Detailed meeting... conduct of the meeting, persons planning to attend should check with these references if such...

  19. Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

    NASA Astrophysics Data System (ADS)

    Zvonaryov, Yu. A.; Budaev, M. A.; Volchek, A. M.; Gorbaev, V. A.; Zagryazkin, V. N.; Kiselyov, N. P.; Kobzar', V. L.; Konobeev, A. V.; Tsurikov, D. F.

    2013-12-01

    The present paper reports the results of the preliminary design estimate of the behavior of the core melt in vessels of reactors of the VVER-600 and VVER-1300 types (a standard optimized and informative nuclear power unit based on VVER technology—VVER TOI) in the case of beyond-design-basis severe accidents. The basic processes determining the state of the core melt in the reactor vessel are analyzed. The concept of molten core confinement within the vessel based on the idea of outside cooling is discussed. Basic assumptions and models, as well as the results of calculation of the interaction between molten materials of the core and the wall of the reactor vessel performed by means of the SOCRAT severe accident code, are presented and discussed. On the basis of the data obtained, the requirements on the operation of the safety systems are determined, upon the fulfillment of which there will appear potential prerequisites for implementing the concept of the confinement of the core melt within the reactor in cases of severe accidents at nuclear power plants equipped with VVER reactors.

  20. Advances in Multi-Sensor Scanning and Visualization of Complex Plants: the Utmost Case of a Reactor Building

    NASA Astrophysics Data System (ADS)

    Hullo, J.-F.; Thibault, G.; Boucheny, C.

    2015-02-01

    In a context of increased maintenance operations and workers generational renewal, a nuclear owner and operator like Electricité de France (EDF) is interested in the scaling up of tools and methods of "as-built virtual reality" for larger buildings and wider audiences. However, acquisition and sharing of as-built data on a large scale (large and complex multi-floored buildings) challenge current scientific and technical capacities. In this paper, we first present a state of the art of scanning tools and methods for industrial plants with very complex architecture. Then, we introduce the inner characteristics of the multi-sensor scanning and visualization of the interior of the most complex building of a power plant: a nuclear reactor building. We introduce several developments that made possible a first complete survey of such a large building, from acquisition, processing and fusion of multiple data sources (3D laser scans, total-station survey, RGB panoramic, 2D floor plans, 3D CAD as-built models). In addition, we present the concepts of a smart application developed for the painless exploration of the whole dataset. The goal of this application is to help professionals, unfamiliar with the manipulation of such datasets, to take into account spatial constraints induced by the building complexity while preparing maintenance operations. Finally, we discuss the main feedbacks of this large experiment, the remaining issues for the generalization of such large scale surveys and the future technical and scientific challenges in the field of industrial "virtual reality".

  1. Rheology Measurements for Online Monitoring of Solids in Activated Sludge Reactors of Municipal Wastewater Treatment Plant

    PubMed Central

    Papo, Adriano; Goi, Daniele

    2014-01-01

    Rheological behaviour of recycled sludge from a secondary clarifier of a municipal wastewater treatment plant was studied by using the rate controlled coaxial cylinder viscometer Rotovisko-Haake 20, system M5-osc., measuring device NV. The tests (hysteresis cycles) were performed under continuous flow conditions and following an ad hoc measurement protocol. Sludge shear stress versus shear rate curves were fitted very satisfactorily by rheological models. An experimental equation correlating the solid concentration of sludge to relative viscosity and fitting satisfactorily flow curves at different Total Suspended Solids (TTS%) was obtained. Application of the empirical correlation should allow the monitoring of the proper functioning of a wastewater treatment plant measuring viscosity of sludge. PMID:24550715

  2. Evaluation of cracking in feedwater piping adjacent to the steam generators in Nine Pressurized Water Reactor Plants

    SciTech Connect

    Goldberg, A.; Streit, R.D.; Scott, R.G.

    1980-06-25

    Cracking in ASTM A106-B and A106-C feedwater piping was detected near the inlet to the steam generators in a number of pressurized water reactor plants. We received sections with cracks from nine of the plants with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Variations were observed in piping surface irregularities, corrosion-product, pit, and crack morphology, surface elmental and crystal structure analyses, and steel microstructures and mechanical properties. However, with but two exceptions, namely, arrest bands and major surface irregularities, we were unable to relate the extent of cracking to any of these factors. Tensile and fracture toughness (J/sub Ic/ and tearing modulus) properties were measured over a range of temperatures and strain rates. No unusual properties or microstructures were observed that could be related to the cracking problem. All crack surfaces contained thick oxide deposits and showed evidence of cyclic events in the form of arrest bands. Transmission electron microscopy revealed fatigue striations on replicas of cleaned crack surfaces from one plant and possibly from three others. Calculations based on the observed striation spacings gave a value of ..delta..sigma = 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses. Although surface irregularities and corrosion pits were sources for crack initiation and corrosion may have contributed to crack propagation, it is proposed that the overriding factor in the cracking problem is the presence of unforeseen cyclic loads.

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

    SciTech Connect

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

    2005-06-01

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

  4. Reducing NO(x) emissions from a nitric acid plant of domestic petrochemical complex: enhanced conversion in conventional radial-flow reactor of selective catalytic reduction process.

    PubMed

    Abbasfard, Hamed; Hashemi, Seyed Hamid; Rahimpour, Mohammad Reza; Jokar, Seyyed Mohammad; Ghader, Sattar

    2013-01-01

    The nitric acid plant of a domestic petrochemical complex is designed to annually produce 56,400 metric tons (based on 100% nitric acid). In the present work, radial-flow spherical bed reactor (RFSBR) for selective catalytic reduction of nitric oxides (NO(x)) from the stack of this plant was modelled and compared with the conventional radial-flow reactor (CRFR). Moreover, the proficiency of a radial-flow (water or nitrogen) membrane reactor was also compared with the CRFR which was found to be inefficient at identical process conditions. In the RFSBR, the space between the two concentric spheres is filled by a catalyst. A mathematical model, including conservation of mass has been developed to investigate the performance of the configurations. The model was checked against the CRFR in a nitric acid plant located at the domestic petrochemical complex. A good agreement was observed between the modelling results and the plant data. The effects of some important parameters such as pressure and temperature on NO(x) conversion were analysed. Results show 14% decrease in NO(x) emission annually in RFSBR compared with the CRFR, which is beneficial for the prevention of NO(x) emission, global warming and acid rain.

  5. Reducing NO(x) emissions from a nitric acid plant of domestic petrochemical complex: enhanced conversion in conventional radial-flow reactor of selective catalytic reduction process.

    PubMed

    Abbasfard, Hamed; Hashemi, Seyed Hamid; Rahimpour, Mohammad Reza; Jokar, Seyyed Mohammad; Ghader, Sattar

    2013-01-01

    The nitric acid plant of a domestic petrochemical complex is designed to annually produce 56,400 metric tons (based on 100% nitric acid). In the present work, radial-flow spherical bed reactor (RFSBR) for selective catalytic reduction of nitric oxides (NO(x)) from the stack of this plant was modelled and compared with the conventional radial-flow reactor (CRFR). Moreover, the proficiency of a radial-flow (water or nitrogen) membrane reactor was also compared with the CRFR which was found to be inefficient at identical process conditions. In the RFSBR, the space between the two concentric spheres is filled by a catalyst. A mathematical model, including conservation of mass has been developed to investigate the performance of the configurations. The model was checked against the CRFR in a nitric acid plant located at the domestic petrochemical complex. A good agreement was observed between the modelling results and the plant data. The effects of some important parameters such as pressure and temperature on NO(x) conversion were analysed. Results show 14% decrease in NO(x) emission annually in RFSBR compared with the CRFR, which is beneficial for the prevention of NO(x) emission, global warming and acid rain. PMID:24527652

  6. Development Program of IS Process Pilot Test Plant for Hydrogen Production With High-Temperature Gas-Cooled Reactor

    SciTech Connect

    Jin Iwatsuki; Atsuhiko Terada; Hiroyuki Noguchi; Yoshiyuki Imai; Masanori Ijichi; Akihiro Kanagawa; Hiroyuki Ota; Shinji Kubo; Kaoru Onuki; Ryutaro Hino

    2006-07-01

    At the present time, we are alarmed by depletion of fossil energy and effects on global environment such as acid rain and global warming, because our lives depend still heavily on fossil energy. So, it is universally recognized that hydrogen is one of the best energy media and its demand will be increased greatly in the near future. In Japan, the Basic Plan for Energy Supply and Demand based on the Basic Law on Energy Policy Making was decided upon by the Cabinet on 6 October, 2003. In the plan, efforts for hydrogen energy utilization were expressed as follows; hydrogen is a clean energy carrier without carbon dioxide (CO{sub 2}) emission, and commercialization of hydrogen production system using nuclear, solar and biomass, not fossil fuels, is desired. However, it is necessary to develop suitable technology to produce hydrogen without CO{sub 2} emission from a view point of global environmental protection, since little hydrogen exists naturally. Hydrogen production from water using nuclear energy, especially the high-temperature gas-cooled reactor (HTGR), is one of the most attractive solutions for the environmental issue, because HTGR hydrogen production by water splitting methods such as a thermochemical iodine-sulfur (IS) process has a high possibility to produce hydrogen effectively and economically. The Japan Atomic Energy Agency (JAEA) has been conducting the HTTR (High-Temperature Engineering Test Reactor) project from the view to establishing technology base on HTGR and also on the IS process. In the IS process, raw material, water, is to be reacted with iodine (I{sub 2}) and sulfur dioxide (SO{sub 2}) to produce hydrogen iodide (HI) and sulfuric acid (H{sub 2}SO{sub 4}), the so-called Bunsen reaction, which are then decomposed endo-thermically to produce hydrogen (H{sub 2}) and oxygen (O{sub 2}), respectively. Iodine and sulfur dioxide produced in the decomposition reactions can be used again as the reactants in the Bunsen reaction. In JAEA, continuous

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

    SciTech Connect

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

    2004-08-30

    The Modular Pebble Bed Reactor system (MPBR) requires a gas turbine cycle (Brayton cycle) as the power conversion system for it to achieve economic competitiveness as a Generation IV nuclear system. The availability of controllable helium turbomachinery and compact heat exchangers are thus the critical enabling technology for the gas turbine cycle. The development of an initial reference design for an indirect helium cycle has been accomplished with the overriding constraint that this design could be built with existing technology and complies with all current codes and standards. Using the initial reference design, limiting features were identified. Finally, an optimized reference design was developed by identifying key advances in the technology that could reasonably be expected to be achieved with limited R&D. This final reference design is an indirect, intercooled and recuperated cycle consisting of a three-shaft arrangement for the turbomachinery system. A critical part of the design process involved the interaction between individual component design and overall plant performance. The helium cycle overall efficiency is significantly influenced by performance of individual components. Changes in the design of one component, a turbine for example, often required changes in other components. To allow for the optimization of the overall design with these interdependencies, a detailed steady state and transient control model was developed. The use of the steady state and transient models as a part of an iterative design process represents a key contribution of this work. A dynamic model, MPBRSim, has been developed. The model integrates the reactor core and the power conversion system simultaneously. Physical parameters such as the heat exchangers; weights and practical performance maps such as the turbine characteristics and compressor characteristics are incorporated into the model. The individual component models as well as the fully integrated model of the

  8. Energy potential and alternative usages of biogas and sludge from UASB reactors: case study of the Laboreaux wastewater treatment plant.

    PubMed

    Rosa, A P; Conesa, J A; Fullana, A; Melo, G C B; Borges, J M; Chernicharo, C A L

    2016-01-01

    This work assessed the energy potential and alternative usages of biogas and sludge generated in upflow anaerobic sludge blanket reactors at the Laboreaux sewage treatment plant (STP), Brazil. Two scenarios were considered: (i) priority use of biogas for the thermal drying of dehydrated sludge and the use of the excess biogas for electricity generation in an ICE (internal combustion engine); and (ii) priority use of biogas for electricity generation and the use of the heat of the engine exhaust gases for the thermal drying of the sludge. Scenario 1 showed that the electricity generated is able to supply 22.2% of the STP power demand, but the thermal drying process enables a greater reduction or even elimination of the final volume of sludge to be disposed. In Scenario 2, the electricity generated is able to supply 57.6% of the STP power demand; however, the heat in the exhaust gases is not enough to dry the total amount of dehydrated sludge. PMID:27054741

  9. Almost twenty years' search of transuranium isotopes in effluents discharged to air from nuclear power plants with VVER reactors.

    PubMed

    Hölgye, Z; Filgas, R

    2006-04-01

    Airborne effluents of 5 stacks (stacks 1-5) of three nuclear power plants, with 9 pressurized water reactors VVER of 4,520 MWe total power, were searched for transuranium isotopes in different time periods. The search started in 1985. The subject of this work is a presentation of discharge data for the period of 1998-2003 and a final evaluation. It was found that 238Pu, 239,240Pu, 241Am, 242Cm, and 244Cm can be present in airborne effluents. Transuranium isotope contents in most of the quarterly effluent samples from stacks 2, 4 and 5 were not measurable. Transuranium isotopes were present in the effluents from stack l during all 9 years of the study and from stack 3 since the 3rd quarter of 1996 as a result of a defect in the fuel cladding. A relatively high increase of transuranium isotopes in effluents from stack 3 occurred in the 3rd quarter of 1999, and a smaller increase occurred in the 3rd quarter of 2003. In each instance 242Cm prevailed in the transuranium isotope mixtures. 238Pu/239,240Pu, 241Am/239,240Pu, 242Cm/239,240Pu, and 244Cm/239,240Pu ratios in fuel for different burn-up were calculated, and comparison of these ratios in fuel and effluents was performed.

  10. A UASB reactor coupled to a hybrid aerobic MBR as innovative plant configuration to enhance the removal of organic micropollutants.

    PubMed

    Alvarino, T; Suárez, S; Garrido, M; Lema, J M; Omil, F

    2016-02-01

    An innovative plant configuration based in an UASB reactor coupled to a hybrid aerobic membrane bioreactor designed for sustainable treatment of municipal wastewater at ambient temperatures and low hydraulic retention time was studied in terms of organic micropollutants (OMPs) removal. OMPs removal mechanisms, as well as the potential influence of biomass activity and physical conformation were assessed. Throughout all periods of operation (150 days) high organic matter removals were maintained (>95%) and, regarding OMPs removal, this innovative system has shown to be more efficient than conventional technologies for those OMPs which are prone to be biotransformed under anaerobic conditions. For instance, sulfamethoxazole and trimethoprim have both shown to be biodegradable under anaerobic conditions with similar efficiencies (removal efficiencies above 84%). OMPs main removal mechanism was found to be biotransformation, except in the case of musk fragrances which showed medium sorption onto sludge. OMPs removal was strongly dependent on the efficiency of the primary metabolism (organic matter degradation and nitrification) and the type of biomass. PMID:26386770

  11. Contamination level of four priority phthalates in North Indian wastewater treatment plants and their fate in sequencing batch reactor systems.

    PubMed

    Gani, Khalid Muzamil; Rajpal, Ankur; Kazmi, Absar Ahmad

    2016-03-01

    The contamination level of four phthalates in untreated and treated wastewater of fifteen wastewater treatment plants (WWTPs) and their fate in a full scale sequencing batch reactor (SBR) based WWTP was evaluated in this study. The four phthalates were diethyl phthalate (DEP), dibutyl phthalate (DBP), benzylbutyl phthalate (BBP) and diethylhexyl phthalate (DEHP). All compounds were present in untreated wastewater with DEHP being present in the highest mean concentration of 28.4 ± 5.3 μg L(-1). The concentration was in the range of 7.3 μg L(-1) (BBP) to 28.4 μg L(-1) (DEHP) in untreated wastewater and 1.3 μg L(-1) (DBP) to 2.6 μg L(-1) (DEHP) in treated wastewater. The nutrient removal process and advance tertiary treatment based WWTPs showed the highest phthalate removal efficiencies of 87% and 93%, respectively. The correlation between phthalate removal and conventional performance of WWTPs was positive. Fate analysis of these phthalates in a SBR based WWTP showed that total removal of the sum of phthalates in a primary settling tank and SBR was 84% out of which 55% is removed by biodegradation and 29% was removed by sorption to primary and secondary sludge. The percentage removal of four phthalates in primary settling tanks was 18%. Comparison of the diluted effluent DEHP concentration with its environmental quality standards showed that the dilution in an effluent receiving water body can reduce the DEHP emissions to acceptable values.

  12. Alternate analysis criteria for the seismic qualification of the supplementary safety system in Reactor Building 105-L, Savannah River Plant

    SciTech Connect

    Quan, C.N.; Wong, P.W.

    1982-09-01

    This system consists of a series of stainless-steel, safety-related pipelines, 1/2 to 2 in. in diameter, which run from the control room to the reactor tank. The alternate analysis criteria were developed for the seismic qualification of the piping system, according to the requirements of the 1967 Housner criteria. The application of alternate analysis criteria is a widely employed and accepted procedure for the seismic qualification of large lengths of nuclear power plant small-diameter piping (generally 2 in. in diameter and smaller). Objective of this procedure is to eliminate the need for extensive and costly individual mathematical modeling and dynamic analysis of a great number of small-bore piping systems. The procedure used to develop the alternate analysis criteria consisted of applying equivalent static seismic loading to the various pipe sizes to determine maximum support spacings for each pipe size based on the allowable stress or deflection limits. Guidelines were also developed to provide for sufficient thermal growth capacity, which is typically in direct conflict with seismic requirements.

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

    SciTech Connect

    Not Available

    1994-08-01

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

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

    SciTech Connect

    Not Available

    1994-08-01

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

  15. Effect of hydraulic retention time on inorganic nutrient recovery and biodegradable organics removal in a biofilm reactor treating plant biomass leachate

    NASA Technical Reports Server (NTRS)

    Krumins, Valdis; Hummerick, Mary; Levine, Lanfang; Strayer, Richard; Adams, Jennifer L.; Bauer, Jan

    2002-01-01

    A fixed-film (biofilm) reactor was designed and its performance was determined at various retention times. The goal was to find the optimal retention time for recycling plant nutrients in an advanced life support system, to minimize the size, mass, and volume (hold-up) of a production model. The prototype reactor was tested with aqueous leachate from wheat crop residue at 24, 12, 6, and 3 h hydraulic retention times (HRTs). Biochemical oxygen demand (BOD), nitrates and other plant nutrients, carbohydrates, total phenolics, and microbial counts were monitored to characterize reactor performance. BOD removal decreased significantly from 92% at the 24 h HRT to 73% at 3 h. Removal of phenolics was 62% at the 24 h retention time, but 37% at 3 h. Dissolved oxygen concentrations, nitric acid consumption, and calcium and magnesium removals were also affected by HRT. Carbohydrate removals, carbon dioxide (CO2) productions, denitrification, potassium concentrations, and microbial counts were not affected by different retention times. A 6 h HRT will be used in future studies to determine the suitability of the bioreactor effluent for hydroponic plant production.

  16. Thermal and neutron-physical features of the nuclear reactor for a power pulsation plant for space applications

    NASA Astrophysics Data System (ADS)

    Gordeev, É. G.; Kaminskii, A. S.; Konyukhov, G. V.; Pavshuk, V. A.; Turbina, T. A.

    2012-05-01

    We have explored the possibility of creating small-size reactors with a high power output with the provision of thermal stability and nuclear safety under standard operating conditions and in emergency situations. The neutron-physical features of such a reactor have been considered and variants of its designs preserving the main principles and approaches of nuclear rocket engine technology are presented.

  17. Treatment of Common Effluent Treatment Plant Wastewater in a Sequential Anoxic-Oxic Batch Reactor by Developed Bacterial Consortium VN11.

    PubMed

    Chattaraj, Sananda; Purohit, Hemant J; Sharma, Abhinav; Jadeja, Niti B; Madamwar, Datta

    2016-06-01

    A laboratory-scale anoxic-oxic sequential reactor system was seeded with acclimatized mixed microbial consortium for the treatment of common effluent treatment plant (CETP) wastewater having 7000-7400 mg L(-1) of COD and 3000-3400 mg L(-1) of BOD. Initially, CETP wastewater was treated under anoxic reactor at 5000 mg L(-1) of MLSS concentrations, 5.26 ± 0.27 kg COD m(-3) day(-1) of organic loading rate (OLR) and 36 h of hydraulic retention time (HRT). Further, the effluent of anoxic reactor was treated in oxic reactor with an OLR of 6.6 ± 0.31 kg COD m(-3) day(-1) and 18 h HRT. Maximum color and COD removal were found to be 72 and 85 % at total HRT of 2.25 days under anoxic-oxic sequential reactor at 37 °C and pH 7.0. The UV-VIS, FTIR, NMR and GCMS studies showed that majority of peaks observed in untreated wastewater were either shifted or disappeared after sequential treatment. Phytotoxicity study with the seeds of Vigna radiata and Triticum aestivum showed more sensitivity toward the CETP wastewater, while the products obtained after sequential treatment does not have any inhibitory effects. The results demonstrated that the anoxic-oxic reactor fed with bacterial consortium VN11 could bring about efficient bioremediation of industrial wastewaters.

  18. Degradation of polycyclic aromatic hydrocarbons in a coking wastewater treatment plant residual by an O3/ultraviolet fluidized bed reactor.

    PubMed

    Lin, Chong; Zhang, Wanhui; Yuan, Mengyang; Feng, Chunhua; Ren, Yuan; Wei, Chaohai

    2014-09-01

    Coking wastewater treatment plant (CWWTP) represents a typical point source of polycyclic aromatic hydrocarbons (PAHs) to the water environment and threatens the safety of drinking water in downstream regions. To enhance the removal of residual PAHs from bio-treated coking wastewater, a pilot-scale O3/ultraviolet (UV) fluidized bed reactor (O3/UV FBR) was designed and different operating factors including UV irradiation intensity, pH, initial concentration, contact time, and hydraulic retention time (HRT) were investigated at an ozone level of 240 g h(-1) and 25 ± 3 °C. A health risk evaluation and cost analysis were also carried out under the continuous-flow mode. As far as we know, this is the first time an O3/UV FBR has been explored for PAHs treatment. The results indicated that between 41 and 75 % of 18 target PAHs were removed in O3/UV FBR due to synergistic effects of UV irradiation. Both increased reaction time and increased pH were beneficial for the removal of PAHs. The degradation of the target PAHs within 8 h can be well fitted by the pseudo-first-order kinetics (R (2) > 0.920). The reaction rate was also positively correlated with the initial concentrations of PAHs. The health risk assessment showed that the total amount of carcinogenic substance exposure to surface water was reduced by 0.432 g day(-1). The economic analysis showed that the O3/UV FBR was able to remove 18 target PAHs at a cost of US$0.34 m(-3). These results suggest that O3/UV FBR is efficient in removing residuals from CWWTP, thus reducing the accumulation of persistent pollutant released to surface water.

  19. Contamination level of four priority phthalates in North Indian wastewater treatment plants and their fate in sequencing batch reactor systems.

    PubMed

    Gani, Khalid Muzamil; Rajpal, Ankur; Kazmi, Absar Ahmad

    2016-03-01

    The contamination level of four phthalates in untreated and treated wastewater of fifteen wastewater treatment plants (WWTPs) and their fate in a full scale sequencing batch reactor (SBR) based WWTP was evaluated in this study. The four phthalates were diethyl phthalate (DEP), dibutyl phthalate (DBP), benzylbutyl phthalate (BBP) and diethylhexyl phthalate (DEHP). All compounds were present in untreated wastewater with DEHP being present in the highest mean concentration of 28.4 ± 5.3 μg L(-1). The concentration was in the range of 7.3 μg L(-1) (BBP) to 28.4 μg L(-1) (DEHP) in untreated wastewater and 1.3 μg L(-1) (DBP) to 2.6 μg L(-1) (DEHP) in treated wastewater. The nutrient removal process and advance tertiary treatment based WWTPs showed the highest phthalate removal efficiencies of 87% and 93%, respectively. The correlation between phthalate removal and conventional performance of WWTPs was positive. Fate analysis of these phthalates in a SBR based WWTP showed that total removal of the sum of phthalates in a primary settling tank and SBR was 84% out of which 55% is removed by biodegradation and 29% was removed by sorption to primary and secondary sludge. The percentage removal of four phthalates in primary settling tanks was 18%. Comparison of the diluted effluent DEHP concentration with its environmental quality standards showed that the dilution in an effluent receiving water body can reduce the DEHP emissions to acceptable values. PMID:26923228

  20. Problems and prospects connected with development of high-temperature filtration technology at nuclear power plants equipped with VVER-1000 reactors

    NASA Astrophysics Data System (ADS)

    Shchelik, S. V.; Pavlov, A. S.

    2013-07-01

    Results of work on restoring the service properties of filtering material used in the high-temperature reactor coolant purification system of a VVER-1000 reactor are presented. A quantitative assessment is given to the effect from subjecting a high-temperature sorbent to backwashing operations carried out with the use of regular capacities available in the design process circuit in the first years of operation of Unit 3 at the Kalinin nuclear power plant. Approaches to optimizing this process are suggested. A conceptual idea about comprehensively solving the problem of achieving more efficient and safe operation of the high-temperature active water treatment system (AWT-1) on a nuclear power industry-wide scale is outlined.

  1. Experimental investigations of thermal-hydraulic processes arising during operation of the passive safety systems used in new projects of nuclear power plants equipped with VVER reactors

    NASA Astrophysics Data System (ADS)

    Morozov, A. V.; Remizov, O. V.; Kalyakin, D. S.

    2014-05-01

    The results obtained from experimental investigations into thermal-hydraulic processes that take place during operation of the passive safety systems used in new-generation reactor plants constructed on the basis of VVER technology are presented. The experiments were carried out on the model rigs available at the Leipunskii Institute for Physics and Power Engineering. The processes through which interaction occurs between the opposite flows of saturated steam and cold water moving in the vertical steam line of the additional system for passively flooding the core from the second-stage hydro accumulators are studied. The specific features pertinent to undeveloped boiling of liquid on a single horizontal tube heated by steam and steam-gas mixture that is typical for of the condensing operating mode of a VVER reactor steam generator are investigated.

  2. Quantification method of N2O emission from full-scale biological nutrient removal wastewater treatment plant by laboratory batch reactor analysis.

    PubMed

    Lim, Yesul; Kim, Dong-Jin

    2014-08-01

    This study proposes a simplified method for the quantification of N2O emission from a biological nutrient removal wastewater treatment plant (WWTP). The method incorporates a laboratory-scale batch reactor which had almost the same operational (wastewater and sludge flow rates) condition of a unit operation/process of the WWTP. Cumulative N2O emissions from the batch reactor at the corresponding hydraulic retention times of the full-scale units (primary and secondary clarifiers, pre-anoxic, anaerobic, anoxic and aerobic basins) were used for the quantification of N2O emission. The analysis showed that the aerobic basin emitted 95% of the total emission and the emission factor (yield) reached 0.8% based on the influent nitrogen load. The method successfully estimated N2O emission from the WWTP and it has shown advantages in measurement time and cost over the direct field measurement (floating chamber) method. PMID:24690468

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

    SciTech Connect

    Myers, Carl W; Elkins, Ned Z

    2008-01-01

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

  4. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, J.

    1996-03-19

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

  5. Pressurized fluidized bed reactor

    DOEpatents

    Isaksson, Juhani

    1996-01-01

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

  6. Start-up of an anaerobic hybrid (UASB/filter) reactor treating wastewater from a coffee processing plant.

    PubMed

    Bello-Mendoza, R; Castillo-Rivera, M F

    1998-10-01

    The ability of an anaerobic hybrid reactor, treating coffee wastewater, to achieve a quick start-up was tested at pilot scale. The unacclimatized seed sludge used showed a low specific methanogenic activity of 26.47 g CH4 as chemical oxygen demand (COD)/kg volatile suspended solids (VSS) x day. This strongly limited the reactor performance. After a few days of operation, a COD removal of 77.2% was obtained at an organic loading rate (OLR) of 1.89 kg COD/m3 x day and a hydraulic retention time (HRT) of 22 h. However, suddenly increasing OLR above 2.4 kg COD/m3 x day resulted in a deterioration in treatment efficiency. The reactor recovered from shock loads after shutdowns of 1 week. The hybrid design of the anaerobic reactor prevented the biomass from washing-out but gas clogging in the packing material was also observed. Wide variations in wastewater strength and flow rates prevented stable reactor operation in the short period of the study.

  7. Preparation of LWBR (Light Water Breeder Reactor) spent fuel for shipment to ICPP (Idaho Chemical Processing Plant) for long term storage (LWBR Development Program). [Shippingport Atomic Power Station

    SciTech Connect

    Hodges, B.W.

    1987-10-01

    After successfully operating for 29,047 effective full power hours, the Light Water Breeder Reactor (LWBR) core was defueled prior to total decommissioning of the Shippingport facility. All nuclear fuel and much of the reactor internal hardware was removed from the reactor vessel. Non-fuel components were prepared for shipment to disposal sites, and the fuel assemblies were partially disassembled and shipped to the Expended Core Facility (ECF) in Idaho. At ECF, the fuel modules underwent further disassembly to provide fuel rods for nondestructive testing to establish the core's breeding efficiency and to provide core components for examinations to assess their performance characteristics. This report presents a basic description of the processes and equipment used to prepare and to ship all LWBR nuclear fuel to the Idaho Chemical Processing Plant (ICPP) for long-term storage. Preparation processes included the underwater loading of LWBR fuel into storage liners, the sealing, dewatering and drying of the storage liners, and the final pressurization of the storage liners with inert neon gas. Shipping operations included the underwater installation of the fuel loaded storage liner into the Peach Bottom shipping cask, cask removal from the waterpit, cask preparations for shipping, and cask shipment by tractor trailer to the ICPP facility for long-term storage. The ICPP facility preparations for LWBR fuel storage and the ICPP process for discharge of the fuel into underground silos are presented. 10 refs., 42 figs.

  8. The effectiveness of using the combined-cycle technology in a nuclear power plant unit equipped with an SVBR-100 reactor

    NASA Astrophysics Data System (ADS)

    Kasilov, V. F.; Dudolin, A. A.; Gospodchenkov, I. V.

    2015-05-01

    The design of a modular SVBR-100 reactor with a lead-bismuth alloy liquid-metal coolant is described. The basic thermal circuit of a power unit built around the SVBR-100 reactor is presented together with the results of its calculation. The gross electrical efficiency of the turbine unit driven by saturated steam at a pressure of 6.7 MPa is estimated at η{el/gr} = 35.5%. Ways for improving the efficiency of this power unit and increasing its power output by applying gas-turbine and combined-cycle technologies are considered. With implementing a combined-cycle power-generating system comprising two GE-6101FA gas-turbine units with a total capacity of 140 MW, it becomes possible to obtain the efficiency of the combined-cycle plant equipped with the SVBR-100 reactor η{el/gr} = 45.39% and its electrical power output equal to 328 MW. The heat-recovery boiler used as part of this power installation generates superheated steam with a temperature of 560°C, due to which there is no need to use a moisture separator/steam reheater in the turbine unit thermal circuit.

  9. Final Safety Analysis Addenda to Hazards Summary Report, Experimental Breeder Reactor II (EBR-II): upgrading of plant protection system. Volume II

    SciTech Connect

    Allen, N. L.; Keeton, J. M.; Sackett, J. I.

    1980-06-01

    This report is the second in a series of compilations of the formal Final Safety Analysis Addenda (FSAA`s) to the EBR-II Hazard Summary Report and Addendum. Sections 2 and 3 are edited versions of the original FSAA`s prepared in support of certain modifications to the reactor-shutdown-system portion of the EBR-II plant-protection system. Section 4 is an edited version of the original FSAA prepared in support of certain modifications to a system classified as an engineered safety feature. These sections describe the pre- and postmodification system, the rationale for the modification, and required supporting safety analysis. Section 5 provides an updated description and analysis of the EBR-II emergency power system. Section 6 summarizes all significant modifications to the EBR-II plant-protection system to date.

  10. Performance and behaviour of planted and unplanted units of a horizontal subsurface flow constructed wetland system treating municipal effluent from a UASB reactor.

    PubMed

    da Costa, Jocilene Ferreira; de Paoli, André Cordeiro; Seidl, Martin; von Sperling, Marcos

    2013-01-01

    A system composed of two horizontal subsurface flow constructed wetlands operating in parallel was evaluated for the post-treatment of UASB (upflow anaerobic sludge blanket) reactor effluent, for a population equivalent of 50 inhabitants per unit. One unit was planted with cattail (Typha latifolia) and the other was unplanted. The study was undertaken over a period of 4 years, comprising monitoring of influent and effluent constituents together with a full characterization of the behaviour of the units (tracer studies, mathematical modelling of chemical oxygen demand (COD) decay, characterization of solids in the filter medium). The mean value of the surface hydraulic load was 0.11 m(3)m(-2)d(-1), and the theoretical hydraulic retention time was 1.1 d in each unit. Using tracer tests with (82)Br, dispersion number (d) values of 0.084 and 0.079 for the planted and unplanted units were obtained, indicating low to moderate dispersion. The final effluent had excellent quality in terms of organic matter and suspended solids, but the system showed low capacity for nitrogen removal. Four-year mean effluent concentration values from the planted and unplanted units were, respectively: biochemical oxygen demand (BOD(5)): 25 and 23 mg L(-1); COD: 50 and 55 mg L(-1); total suspended solids (TSS): 9 and 9 mg L(-1); N-ammonia: 27 and 28 mg L(-1). The COD decay coefficient K for the traditional plug-flow model was 0.81 and 0.84 d(-1) for the planted and unplanted units. Around 80% of the total solids present in the filter medium were inorganic, and most of them were present in the interstices rather than attached to the support medium. As an overall conclusion, horizontal subsurface flow wetlands can be a very suitable post-treatment method for municipal effluents from anaerobic reactors.

  11. Advanced light water reactor requirements document: Chapter 4, Reactor systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the following for Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR): reactor pressure vessel, nozzles and safe-ends, reactor internals, in-vessel portions of fluid systems (including reactor internal pumps (Emergency Core Cooling System (ECCS) piping and spargers), nuclear fuel, and the control rods and control rod drive system (including hydraulic supply and accumulators). Special tools required for reactor system maintenance, inspection and testing are also covered.

  12. Reactor performance of a 750 m(3) anaerobic digestion plant: varied substrate input conditions impacting methanogenic community.

    PubMed

    Wagner, Andreas Otto; Malin, Cornelia; Lins, Philipp; Gstraunthaler, Gudrun; Illmer, Paul

    2014-10-01

    A 750 m(3) anaerobic digester was studied over a half year period including a shift from good reactor performance to a reduced one. Various abiotic parameters like volatile fatty acids (VFA) (formic-, acetic-, propionic-, (iso-)butyric-, (iso-)valeric-, lactic acid), total C, total N, NH4 -N, and total proteins, as well as the organic matter content and dry mass were determined. In addition several process parameters such as temperature, pH, retention time and input of substrate and the concentrations of CH4, H2, CO2 and H2S within the reactor were monitored continuously. The present study aimed at the investigation of the abundance of acetogens and total cell numbers and the microbial methanogenic community as derived from PCR-dHPLC analysis in order to put it into context with the determined abiotic parameters. An influence of substrate quantity on the efficiency of the anaerobic digestion process was found as well as a shift from a hydrogenotrophic in times of good reactor performance towards an acetoclastic dominated methanogenic community in times of reduced reactor performance. After the change in substrate conditions it took the methano-archaeal community about 5-6 weeks to be affected but then changes occurred quickly.

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

    NASA Astrophysics Data System (ADS)

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

    2012-02-01

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

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

    SciTech Connect

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

    2012-02-15

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

  15. Field application of a planted fixed bed reactor (PFR) for support media and rhizosphere investigation using undisturbed samples from full-scale constructed wetlands.

    PubMed

    Barreto, A B; Vasconcellos, G R; von Sperling, M; Kuschk, P; Kappelmeyer, U; Vasel, J L

    2015-01-01

    This study presents a novel method for investigations on undisturbed samples from full-scale horizontal subsurface-flow constructed wetlands (HSSFCW). The planted fixed bed reactor (PFR), developed at the Helmholtz Center for Environmental Research (UFZ), is a universal test unit for planted soil filters that reproduces the operational conditions of a constructed wetland (CW) system in laboratory scale. The present research proposes modifications on the PFR original configuration in order to allow its operation in field conditions. A mobile device to obtain undisturbed samples from real-scale HSSFCW was also developed. The experimental setting is presented with two possible operational configurations. The first allows the removal and replacement of undisturbed samples in the CW bed for laboratory investigations, guaranteeing sample integrity with a mobile device. The second allows the continuous operation of the PFR and undisturbed samples as a fraction of the support media, reproducing the same environmental conditions outside the real-scale system. Investigations on the hydrodynamics of the adapted PFR were carried out with saline tracer tests, validating the proposed adaptation. Six adapted PFR units were installed next to full-scale HSSFCW beds and fed with interstitial liquid pumped from two regions of planted and unplanted support media. Fourteen points were monitored along the system, covering carbon fractions, nitrogen and sulfate. The results indicate the method as a promising tool for investigations on CW support media, rhizosphere and open space for studies on CW modeling, respirometry, kinetic parameters, microbial communities, redox potential and plant influence on HSSFCW. PMID:26247753

  16. 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.

  17. PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

    NASA Astrophysics Data System (ADS)

    Leclercq, Sylvain; Lidbury, David; Van Dyck, Steven; Moinereau, Dominique; Alamo, Ana; Mazouzi, Abdou Al

    2010-11-01

    In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels). PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users' Group and Training sub-project shall allow representatives of constructors, utilities, research organizations… from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach young

  18. Slurry reactor design studies

    SciTech Connect

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

    1990-06-01

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

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

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2012-05-10

    Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO{sub 2}) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO{sub 2} cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO{sub 2}. It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO{sub 2} heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO{sub 2}-to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO{sub 2} turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the

  20. Co3O4-based honeycombs as compact redox reactors/heat exchangers for thermochemical storage in the next generation CSP plants

    NASA Astrophysics Data System (ADS)

    Pagkoura, Chrysoula; Karagiannakis, George; Halevas, Eleftherios; Konstandopoulos, Athanasios G.

    2016-05-01

    Over the last years, several research groups have focused on developing efficient thermochemical heat storage (THS) systems, in-principle capable of being coupled with next generation high temperature Concentrated Solar Power plants. Among systems studied, the Co3O4/CoO redox system is a promising candidate. Currently, research efforts extend beyond basic level identification of promising materials to more application-oriented approaches aiming at validation of THS performance at pilot scale reactors. The present work focuses on the investigation of cobalt oxide based honeycomb structures as candidate reactors/heat exchangers to be employed for such purposes. In the evaluation conducted and presented here, cobalt oxide-based structures with different composition and geometrical characteristics were subjected to redox cycles in the temperature window between 800 and 1000°C under air flow. Basic aspects related to redox performance of each system are briefly discussed but the main focus lies on the evaluation of the segments structural stability after multi-cyclic operation. The latter is based on macroscopic visual observation and also supplemented by pre- (i.e. fresh samples) and post-characterization (i.e. after long term exposure) of extruded honeycombs via combined mercury porosimetry and SEM analysis.

  1. Appling hydrolysis acidification-anoxic-oxic process in the treatment of petrochemical wastewater: From bench scale reactor to full scale wastewater treatment plant.

    PubMed

    Wu, Changyong; Zhou, Yuexi; Sun, Qingliang; Fu, Liya; Xi, Hongbo; Yu, Yin; Yu, Ruozhen

    2016-05-15

    A hydrolysis acidification (HA)-anoxic-oxic (A/O) process was adopted to treat a petrochemical wastewater. The operation optimization was carried out firstly by a bench scale experimental reactor. Then a full scale petrochemical wastewater treatment plant (PCWWTP, 6500 m(3) h(-1)) was operated with the same parameters. The results showed that the BOD5/COD of the wastewater increased from 0.30 to 0.43 by HA. The effluent COD was 54.4 mg L(-1) for bench scale reactor and 60.9 mg L(-1) for PCWWTP when the influent COD was about 480 mg L(-1) on optimized conditions. The organics measured by gas chromatography-mass spectrometry (GC-MS) reduced obviously and the total concentration of the 5 organics (1,3-dioxolane, 2-pentanone, ethylbenzene, 2-chloromethyl-1,3-dioxolane and indene) detected in the effluent was only 0.24 mg L(-1). There was no obvious toxicity of the effluent. However, low acute toxicity of the effluent could be detected by the luminescent bacteria assay, indicating the advanced treatment is needed. The clone library profiling analysis showed that the dominant bacteria in the system were Acidobacteria, Proteobacteria and Bacteriodetes. HA-A/O process is suitable for the petrochemical wastewater treatment.

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

    SciTech Connect

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

    2013-07-01

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

  3. Design of pilot-scale solar photocatalytic reactor for the generation of hydrogen from alkaline sulfide wastewater of sewage treatment plant.

    PubMed

    Priya, R; Kanmani, S

    2013-01-01

    Experiments were conducted for photocatalytic generation of renewable fuel hydrogen from sulphide wastewater from the sewage treatment plant. In this study, pilot-scale solar photocatalytic reactor was designed for treating 1 m3 of sulphide wastewater and also for the simultaneous generation of hydrogen. Bench-scale studies were conducted both in the batch recycle and continuous modes under solar irradiation at similar experimental conditions. The maximum of 89.7% conversion was achieved in the continuous mode. The length of the pilot-scale solar photocatalytic reactor was arrived using the design parameters such as volumetric flow rate (Q) (11 x 10(-2) m3/s), inlet concentration of sulphide ion (C(in)) (28 mol/m3), conversion (89.7%) and average mass flow destruction rate (3.488 x 10(-6) mol/m2 s). The treatment cost of the process was estimated to be 6 US$/m3. This process would be suitable for India like sub-tropical country where sunlight is abundantly available throughout the year.

  4. 1-Dimensional simulation of thermal annealing in a commercial nuclear power plant reactor pressure vessel wall section

    SciTech Connect

    Nakos, J.T.; Rosinski, S.T.; Acton, R.U.

    1994-11-01

    The objective of this work was to provide experimental heat transfer boundary condition and reactor pressure vessel (RPV) section thermal response data that can be used to benchmark computer codes that simulate thermal annealing of RPVS. This specific protect was designed to provide the Electric Power Research Institute (EPRI) with experimental data that could be used to support the development of a thermal annealing model. A secondary benefit is to provide additional experimental data (e.g., thermal response of concrete reactor cavity wall) that could be of use in an annealing demonstration project. The setup comprised a heater assembly, a 1.2 in {times} 1.2 m {times} 17.1 cm thick [4 ft {times} 4 ft {times} 6.75 in] section of an RPV (A533B ferritic steel with stainless steel cladding), a mockup of the {open_quotes}mirror{close_quotes} insulation between the RPV and the concrete reactor cavity wall, and a 25.4 cm [10 in] thick concrete wall, 2.1 in {times} 2.1 in [10 ft {times} 10 ft] square. Experiments were performed at temperature heat-up/cooldown rates of 7, 14, and 28{degrees}C/hr [12.5, 25, and 50{degrees}F/hr] as measured on the heated face. A peak temperature of 454{degrees}C [850{degrees}F] was maintained on the heated face until the concrete wall temperature reached equilibrium. Results are most representative of those RPV locations where the heat transfer would be 1-dimensional. Temperature was measured at multiple locations on the heated and unheated faces of the RPV section and the concrete wall. Incident heat flux was measured on the heated face, and absorbed heat flux estimates were generated from temperature measurements and an inverse heat conduction code. Through-wall temperature differences, concrete wall temperature response, heat flux absorbed into the RPV surface and incident on the surface are presented. All of these data are useful to modelers developing codes to simulate RPV annealing.

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

    SciTech Connect

    Smith, C

    2010-02-22

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

  6. NEUTRONIC REACTOR

    DOEpatents

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

    1958-09-16

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

  7. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1962-12-18

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

  8. Eliminating non-renewable CO2 emissions from sewage treatment: an anaerobic migrating bed reactor pilot plant study.

    PubMed

    Hartley, Ken; Lant, Paul

    2006-10-20

    The aim of this work was to demonstrate at pilot scale a high level of energy recovery from sewage utilising a primary Anaerobic Migrating Bed Reactor (AMBR) operating at ambient temperature to convert COD to methane. The focus is the reduction in non-renewable CO(2) emissions resulting from reduced energy requirements for sewage treatment. A pilot AMBR was operated on screened sewage over the period June 2003 to September 2004. The study was divided into two experimental phases. In Phase 1 the process operated at a feed rate of 10 L/h (HRT 50 h), SRT 63 days, average temperature 28 degrees C and mixing time fraction 0.05. In Phase 2 the operating parameters were 20 L/h, 26 days, 16 degrees C and 0.025. Methane production was 66% of total sewage COD in Phase 1 and 23% in Phase 2. Gas mixing of the reactor provided micro-aeration which suppressed sulphide production. Intermittent gas mixing at a useful power input of 6 W/m(3) provided satisfactory process performance in both phases. Energy consumption for mixing was about 1.5% of the energy conversion to methane in both operating phases. Comparative analysis with previously published data confirmed that methane supersaturation resulted in significant losses of methane in the effluent of anaerobic treatment systems. No cases have been reported where methane was considered to be supersaturated in the effluent. We have shown that methane supersaturation is likely to be significant and that methane losses in the effluent are likely to have been greater than previously predicted. Dissolved methane concentrations were measured at up to 2.2 times the saturation concentration relative to the mixing gas composition. However, this study has also demonstrated that despite methane supersaturation occurring, micro-aeration can result in significantly lower losses of methane in the effluent (<11% in this study), and has demonstrated that anaerobic sewage treatment can genuinely provide energy recovery. The goal of demonstrating a

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

    SciTech Connect

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

    1998-12-01

    The objective of this project is to develop and demonstrate a continuous, low energy process for the conversion of cellulosics to ethanol. This process involves a pretreatment step followed by enzymatic release of sugars and the consecutive simultaneous saccharification/fermentation (SSF) of cellulose (glucans) followed by hemi-cellulose (pentosans) in a multi-stage continuous stirred reactor separator (CSRS). During quarters 3 and 4, we have completed a literature survey on cellulase production, activated one strain of Trichoderma reesei. We continued developing our proprietary Steep Delignification (SD) process for biomass pretreatment. Some problems with fermentations were traces to bad cellulase enzyme. Using commercial cellulase enzymes from Solvay & Genecor, SSF experiments with wheat straw showed 41 g/L ethanol and free xylose of 20 g/L after completion of the fermentation. From corn stover, we noted 36 g/L ethanol production from the cellulose fraction of the biomass, and 4 g/L free xylose at the completion of the SSF. We also began some work with paper mill sludge as a cellulose source, and in some preliminary experiments obtained 23 g/L ethanol during SSF of the sludge. During year 2, a 130 L process scale unit will be operated to demonstrate the process using straw or cornstalks. Co-sponsors of this project include the Indiana Biomass Grants Program, Bio-Process Innovation.

  10. Effects of selected pharmaceutically active compounds on treatment performance in sequencing batch reactors mimicking wastewater treatment plants operations.

    PubMed

    Wang, Shuyi; Gunsch, Claudia K

    2011-05-01

    The impact of four pharmaceutically active compounds (PhACs) introduced both individually and in mixtures was ascertained on the performance of laboratory-scale wastewater treatment sequencing batch reactors (SBRs). When introduced individually at concentrations of 0.1, 1 and 10 μM, no significant differences were observed with respect to chemical oxygen demand (COD) and ammonia removal. Microbial community analyses reveal that although similarity index values generally decreased over time with an increase in PhAC concentrations as compared to the controls, no major microbial community shifts were observed for total bacteria and ammonia-oxidizing bacteria (AOB) communities. However, when some PhACs were introduced in mixtures, they were found to both inhibit nitrification and alter AOB community structure. Ammonia removal decreased by up to 45% in the presence of 0.25 μM gemfibrozil and 0.75 μM naproxen. PhAC mixtures did not however affect COD removal performance suggesting that heterotrophic bacteria are more robust to PhACs than AOB. These results highlight that the joint action of PhACs in mixtures may have significantly different effects on nitrification than the individual PhACs. This phenomenon should be further investigated with a wider range of PhACs so that toxicity effects can more accurately be predicted.

  11. Conceptual design of a laser-fusion power plant. Part II. Two technical options: 1. JADE reactor; 2. Heat transfer by heat pipes

    SciTech Connect

    Not Available

    1981-07-01

    A laser fusion reactor concept is described that employs liquid metal walls. The concept envisions a porous medium, called the JADE, of specific geometry lining the reactor cavity. Some advantages and disadvantages of the concept are pointed out. The possibility of using heat pipes for passive cooling in ICF reactors is discussed. Some of the problems are outlined. (MOW)

  12. Fossil fuel furnace reactor

    DOEpatents

    Parkinson, William J.

    1987-01-01

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

  13. Molecular characterization of denitrifying bacteria isolated from the anoxic reactor of a modified DEPHANOX plant performing enhanced biological phosphorus removal.

    PubMed

    Zafiriadis, Ilias; Ntougias, Spyridon; Mirelis, Paraskevi; Kapagiannidis, Anastasios G; Aivasidis, Alexander

    2012-06-01

    Enhanced Biological Phosphorus Removal (EBPR) under anoxic conditions was achieved using a Biological Nutrient Removal (BNR) system based on a modification of the DEPHANOX configuration. Double-probe Fluorescence in Situ Hybridization (FISH) revealed that Polyphosphate Accumulating Organisms (PAOs) comprised 12.3 +/- 3.2% of the total bacterial population in the modified DEPHANOX plant. The growing bacterial population on blood agar and Casitone Glycerol Yeast Autolysate agar (CGYA) medium was 16.7 +/- 0.9 x 10(5) and 3.0 +/- 0.6 x 10(5) colony forming units (cfu) mL(-1) activated sludge, respectively. A total of 121 bacterial isolates were characterized according to their denitrification ability, with 26 bacterial strains being capable of reducing nitrate to gas. All denitrifying isolates were placed within the alpha-, beta-, and gamma-subdivisions of Proteobacteria and the family Flavobacteriaceae. Furthermore, a novel denitrifying bacterium within the genus Pseudomonas was identified. This is the first report on the isolation and molecular characterization of denitrifying bacteria from EBPR sludge using a DEPHANOX-type plant.

  14. Assessment of the radionuclide composition of "hot particles" sampled in the Chernobyl nuclear power plant fourth reactor unit.

    PubMed

    Bondarkov, Mikhail D; Zheltonozhsky, Viktor A; Zheltonozhskaya, Maryna V; Kulich, Nadezhda V; Maksimenko, Andrey M; Farfán, Eduardo B; Jannik, G Timothy; Marra, James C

    2011-10-01

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

  15. Reactor operation environmental information document

    SciTech Connect

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

    1989-12-01

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

  16. A pilot plant scale Continuous Stirred Reactor/Separator for the production of ethanol from corn grits/starch and biomass streams. [Quarterly progress report, May 1, 1994--August 1, 1994

    SciTech Connect

    Dale, M.C.

    1994-10-01

    This project seeks to develop and demonstrate an improved reactor for the production of ethanol from starch and ligno-cellulosic streams. Bio-Process Innovations holds a patent on this reactor technology, and is directing the project. A Continuous Stirred Reactor Separator (CSRS) is being built on a pilot plant scale for testing at a small Iowa ethanol plant (Permeate Refining) while bench scale tests on the reactor system are being performed at Purdue University. The CSRS unit combines several operations within the confines of the reactor vessel- (1) complex carbohydrates are reduced to simple sugars by enzymatic breakdown, (2) sugars are converted to ethanol by yeast or bacteria, and (3) the ethanol is separated by a stripping gas stream. The ethanol is removed from the stripping gas in an absorber, and then taken to an extractive distillation column. This unit should allow concentrated feeds to be converted to ethanol, and the use of bottoms recycle will be extensively tested to establish the limits of minimizing net bottoms water production leaving the plant. Lab scale tests raw starch conversion in a 4-liter CSRS unit have been completed, a flocculating yeast strain has been selected with some ongoing characterizations studies being performed, a xylose fermenting yeast strain has been selected from among five or six best performing strains. Ongoing studies yet to be completed include testing of the effects of bottoms water recirculation on fermentation performance, testing of raw starch degrading enzymes as produced by various strains of Aspergillus, lignocellulosic conversion to ethanol, and molasses conversion to ethanol.

  17. Estimation of Te-132 distribution in Fukushima Prefecture at the early stage of the Fukushima Daiichi Nuclear Power Plant reactor failures.

    PubMed

    Tagami, Keiko; Uchida, Shigeo; Ishii, Nobuyoshi; Zheng, Jian

    2013-05-21

    Tellurium-132 ((132)Te, half-life: 3.2 d) has been assessed as the radionuclide with the third largest release from the Fukushima Daiichi Nuclear Power Plant (FDNPP) in March 2011; thus it would have made some dose contribution during the early stage of the reactor failures. The available data for (132)Te are, however, limited. In this study, available reported values of other isotopes of Te were compiled to estimate (132)Te concentration (in MBq m(-2)). It was found that (132)Te and (129m)Te (half-life: 33.6 d) concentrations were well correlated (R = 0.99, p < 0.001) by t test. Thus, (132)Te concentrations on March 11, 2011 were estimated from (129m)Te using the concentration conversion factor ((132)Te /(129m)Te) of 14.5. It was also found that since deposited (129m)Te was well retained in the soil, the data collected in March-May of 2011 were applicable to (132)Te estimation. It was possible to obtain the first (132)Te concentration contour map for the eastern part of Fukushima Prefecture, including data from within the 20-km exclusion zone around the FDNPP, using these newly available estimated (132)Te data sets.

  18. Considerations about the enantioselective transformation of polycyclic musks in wastewater, treated wastewater and sewage sludge and analysis of their fate in a sequencing batch reactor plant.

    PubMed

    Berset, J D; Kupper, T; Etter, R; Tarradellas, J

    2004-11-01

    The present work consists of two distinct parts: in the first part enantioselective GC was used to separate the different enantiomeric/diastereomeric polycyclic musks, PCMs (HHCB, AHTN, AHDI, ATII and DPMI) including the main transformation product of HHCB, HHCB-lactone, in wastewater and sewage sludge. After optimization all PCMs were resolved on a cyclodextrin containing Rt-BDEXcst capillary GC column. Enantiomeric ratios of PCMs in a technical mixture were determined and compared to those obtained from enantioselective separation of wastewater and sewage sludge samples. In general, enantiomeric ratios were similar for most materials in influent, effluent and stabilized sewage sludge. However, the ratios for HHCB, AHDI and particularly ATII suggest some stereospecific removal of these compounds. In the second part, a field study was conducted on a wastewater treatment plant comprising a sequencing batch reactor. Concentrations of HHCB, AHTN, ADBI, AHDI, ATII, DPMI and HHCB-lactone were determined by non-enantioselective GC in daily samples of influent, effluent and activated sludge during one week. Mean concentrations in influent were 6900 and 1520 ng/l for HHCB and AHTN, respectively. The other PCMs exhibited contents 200 ng/l. Mean percent removal was between 61% (AHDI) and 87% (HHCB) resulting in mean effluent concentrations below 860 ng/l. HHCB-lactone concentration increased during wastewater treatment with a mean in the influent of 430 ng/l and in the effluent of 900 ng/l, respectively, indicating a degradation of HHCB.

  19. Assessment of potential impact of the Clinch River Breeder Reactor Plant thermal effluent on the Watts Bar Reservoir striped bass population

    SciTech Connect

    Heuer, J H; McIntosh, D; Ostrowski, P; Tomljanovich, D A

    1983-11-01

    This report is an assessment of potential adverse impact to striped bass (Morone saxatilis) in Watts Bar Reservoir caused by thermal effluent from operation of the Clinch River Breeder Reactor Plant (CRBRP). The Clinch River arm of Watts Bar Reservoir is occupied by adult striped bass during the warmest months of the year. Concern was raised that operation of the CRBRP, specifically thermal discharges, could conflict with management of striped bass. In all cases examined the thermal plume becomes nearly imperceptible within a short distance from the discharge pipe (about 30 ft (10 m)) compared to river width (about 630 ft (190 m)). Under worst case conditions any presence of the plume in the main channel (opposite side of the river from the discharge) will be confined to the surface layer of the water. An ample portion of river cross sections containing ambient temperature water for passage or residence of adult striped bass will always be available in the vicinity of this thermal effluent. Although a small portion of river cross section would exceed the thermal tolerance of striped bass, the fish would naturally avoid this area and seek out adjacent cooler water. Therefore, it is concluded the CRBRP thermal effluent will not significantly affect the integrity of the striped bass thermal refuge in the Clinch River arm of Watts Bar Reservoir. At this time there is no need to consider alternative diffuser designs and thermal modeling. 8 references, 3 figures, 2 tables.

  20. Survey for the presence of Naegleria fowleri amebae in lake water used to cool reactors at a nuclear power generating plant.

    PubMed

    Jamerson, Melissa; Remmers, Kenneth; Cabral, Guy; Marciano-Cabral, Francine

    2009-04-01

    Water from Lake Anna in Virginia, a lake that is used to cool reactors at a nuclear power plant and for recreational activities, was assessed for the presence of Naegleria fowleri, an ameba that causes primary amebic meningoencephalitis (PAM). This survey was undertaken because it has been reported that thermally enriched water fosters the propagation of N. fowleri and, hence, increases the risk of infection to humans. Of 16 sites sampled during the summer of 2007, nine were found to be positive for N. fowleri by a nested polymerase chain reaction assay. However, total ameba counts, inclusive of N. fowleri, never exceeded 12/50 mL of lake water at any site. No correlation was obtained between the conductivity, dissolved oxygen, temperature, and pH of water and presence of N. fowleri. To date, cases of PAM have not been reported from this thermally enriched lake. It is postulated that predation by other protozoa and invertebrates, disturbance of the water surface from recreational boating activities, or the presence of bacterial or fungal toxins, maintain the number N. fowleri at a low level in Lake Anna.

  1. Pyrolysis of oil-plant wastes in a TGA and a fixed-bed reactor: Thermochemical behaviors, kinetics, and products characterization.

    PubMed

    Chen, Jianbiao; Fan, Xiaotian; Jiang, Bo; Mu, Lin; Yao, Pikai; Yin, Hongchao; Song, Xigeng

    2015-09-01

    Pyrolysis characteristics of four distinct oil-plant wastes were investigated using TGA and fixed-bed reactor coupled with GC. TGA experiments showed that the pyrolysis behaviors were related to biomass species and heating rates. As the heating rate increased, TG and DTG curves shifted to the higher temperatures, and the comprehensive devolatilization index obviously increased. The remaining chars from TGA experiments were higher than those obtained from the fixed-bed experiments. The crack of tars at high temperatures enhanced the formation of non-condensable gases. During the pyrolysis, C-O and CO2 were the major gases. Chars FTIR showed that the functional groups of O-H, C-H(n), C=O, C-O, and C-C gradually disappeared from 400 °C on. The kinetic parameters were calculated by Coats-Redfern approach. The results manifested that the most appropriate pyrolysis mechanisms were the order reaction models. The existence of kinetic compensation effect was evident. PMID:26093253

  2. Anammox moving bed biofilm reactor pilot at the 26th Ward wastewater treatment plants in Brooklyn, New York: start-up, biofilm population diversity and performance optimization.

    PubMed

    Mehrdad, M; Park, H; Ramalingam, K; Fillos, J; Beckmann, K; Deur, A; Chandran, K

    2014-01-01

    New York City Environmental Protection in conjunction with City College of New York assessed the application of the anammox process in the reject water treatment using a moving bed biofilm reactor (MBBR) located at the 26th Ward wastewater treatment plant, in Brooklyn, NY. The single-stage nitritation/anammox MBBR was seeded with activated sludge and consequently was enriched with its own 'homegrown' anammox bacteria (AMX). Objectives of this study included collection of additional process kinetic and operating data and assessment of the effect of nitrogen loading rates on process performance. The initial target total inorganic nitrogen removal of 70% was limited by the low alkalinity concentration available in the influent reject water. Higher removals were achieved after supplementing the alkalinity by adding sodium hydroxide. Throughout startup and process optimization, quantitative real-time polymerase chain reaction (qPCR) analyses were used for monitoring the relevant species enriched in the biofilm and in the suspension. Maximum nitrogen removal rate was achieved by stimulating the growth of a thick biofilm on the carriers, and controlling the concentration of dissolved oxygen in the bulk flow and the nitrogen loading rates per surface area; all three appear to have contributed in suppressing nitrite-oxidizing bacteria activity while enriching AMX density within the biofilm.

  3. Anammox moving bed biofilm reactor pilot at the 26th Ward wastewater treatment plants in Brooklyn, New York: start-up, biofilm population diversity and performance optimization.

    PubMed

    Mehrdad, M; Park, H; Ramalingam, K; Fillos, J; Beckmann, K; Deur, A; Chandran, K

    2014-01-01

    New York City Environmental Protection in conjunction with City College of New York assessed the application of the anammox process in the reject water treatment using a moving bed biofilm reactor (MBBR) located at the 26th Ward wastewater treatment plant, in Brooklyn, NY. The single-stage nitritation/anammox MBBR was seeded with activated sludge and consequently was enriched with its own 'homegrown' anammox bacteria (AMX). Objectives of this study included collection of additional process kinetic and operating data and assessment of the effect of nitrogen loading rates on process performance. The initial target total inorganic nitrogen removal of 70% was limited by the low alkalinity concentration available in the influent reject water. Higher removals were achieved after supplementing the alkalinity by adding sodium hydroxide. Throughout startup and process optimization, quantitative real-time polymerase chain reaction (qPCR) analyses were used for monitoring the relevant species enriched in the biofilm and in the suspension. Maximum nitrogen removal rate was achieved by stimulating the growth of a thick biofilm on the carriers, and controlling the concentration of dissolved oxygen in the bulk flow and the nitrogen loading rates per surface area; all three appear to have contributed in suppressing nitrite-oxidizing bacteria activity while enriching AMX density within the biofilm. PMID:25401307

  4. Pyrolysis of oil-plant wastes in a TGA and a fixed-bed reactor: Thermochemical behaviors, kinetics, and products characterization.

    PubMed

    Chen, Jianbiao; Fan, Xiaotian; Jiang, Bo; Mu, Lin; Yao, Pikai; Yin, Hongchao; Song, Xigeng

    2015-09-01

    Pyrolysis characteristics of four distinct oil-plant wastes were investigated using TGA and fixed-bed reactor coupled with GC. TGA experiments showed that the pyrolysis behaviors were related to biomass species and heating rates. As the heating rate increased, TG and DTG curves shifted to the higher temperatures, and the comprehensive devolatilization index obviously increased. The remaining chars from TGA experiments were higher than those obtained from the fixed-bed experiments. The crack of tars at high temperatures enhanced the formation of non-condensable gases. During the pyrolysis, C-O and CO2 were the major gases. Chars FTIR showed that the functional groups of O-H, C-H(n), C=O, C-O, and C-C gradually disappeared from 400 °C on. The kinetic parameters were calculated by Coats-Redfern approach. The results manifested that the most appropriate pyrolysis mechanisms were the order reaction models. The existence of kinetic compensation effect was evident.

  5. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1961-09-01

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

  6. NEUTRONIC REACTOR

    DOEpatents

    Daniels, F.

    1959-10-27

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

  7. Research reactors

    SciTech Connect

    Tonneson, L.C.; Fox, G.J.

    1996-04-01

    There are currently 284 research reactors in operation, and 12 under construction around the world. Of the operating reactors, nearly two-thirds are used exclusively for research, and the rest for a variety of purposes, including training, testing, and critical assembly. For more than 50 years, research reactor programs have contributed greatly to the scientific and educational communities. Today, six of the world`s research reactors are being shut down, three of which are in the USA. With government budget constraints and the growing proliferation concerns surrounding the use of highly enriched uranium in some of these reactors, the future of nuclear research could be impacted.

  8. CONVECTION REACTOR

    DOEpatents

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

    1960-03-22

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

  9. Technical-and-economic analysis and optimization of the full flow charts of processing of radioactive wastes on a polyfunctional plant of pyrochemical processing of the spent nuclear fuel of fast reactors

    NASA Astrophysics Data System (ADS)

    Gupalo, V. S.; Chistyakov, V. N.; Kormilitsyn, M. V.; Kormilitsyna, L. A.; Osipenko, A. G.

    2015-12-01

    When considering the full flow charts of processing of radioactive wastes (RAW) on a polyfunctional plant of pyrochemical processing of the spent nuclear fuel of NIIAR fast reactors, we corroborate optimum technical solutions for the preparation of RAW for burial from a standpoint of heat release, dose formation, and technological storage time with allowance for technical-and-economic and ecological indices during the implementation of the analyzed technologies and equipment for processing of all RAW fluxes.

  10. Estimate of radiation-induced steel embrittlement in the BWR core shroud and vessel wall from reactor-grade MOX/UOX fuel for the nuclear power plant at Laguna Verde, Veracruz, Mexico

    NASA Astrophysics Data System (ADS)

    Vickers, Lisa Rene

    The government of Mexico has expressed interest to utilize the Laguna Verde boiling water reactor (BWR) nuclear power plant for the disposition of reprocessed spent uranium oxide (UOX) fuel in the form of reactor-grade mixed-oxide (MOX) fuel. MOX fuel would replace spent UOX fuel as a fraction in the core from 18--30% depending on the fuel loading cycle. MOX fuel is expected to increase the neutron fluence, flux, fuel centerline temperature, reactor core pressure, and yield higher energy neutrons. There is concern that a core with a fraction of MOX fuel (i.e., increased 239Pu wt%) would increase the radiation-induced steel embrittlement within the core shroud and vessel wall as compared to only conventional, enriched UOX fuel in the core. The evaluation of radiation-induced steel embrittlement within the core shroud and vessel wall is a concern because of the potentially adverse affect to plant and public safety, environment, and operating life of the reactor. This dissertation provides computational results of the neutron fluence, flux, energy spectrum, and radiation damage displacements per atom per second (dpa-s-1) in steel within the core shroud and vessel wall of the Laguna Verde Unit 1 BWR. The results were computed using the nuclear data processing code NJOY99 and the continuous energy Monte Carlo Neutral Particle transport code MCNP4B. The MCNP4B model of the reactor core was for maximum core loading fractions of ⅓ MOX and ⅔ UOX reactor-grade fuel in an equilibrium core. The primary conclusion of this dissertation was that the addition of the maximum fraction of ⅓ MOX fuel to the LV1 BWR core did significantly accelerate the radiation-induced steel embrittlement such that without mitigation of steel embrittlement by periodic thermal annealing or reduction in operating parameters such as, neutron fluence, core temperature and pressure, it posed a potentially adverse affect to the plant and public safety, environment, and operating life of the reactor.

  11. Unique features of space reactors

    SciTech Connect

    Buden, D.

    1990-01-01

    Space reactors are designed to meet a unique set of requirements; they must be sufficiently compact to be launched in a rocket to their operational location, operate for many years without maintenance and servicing, operate in extreme environments, and reject heat by radiation to space. To meet these restrictions, operating temperatures are much greater than in terrestrial power plants, and the reactors tend to have a fast neutron spectrum. Currently, a new generation of space reactor power plants is being developed. The major effort is in the SP-100 program, where the power plant is being designed for seven years of full power, and no maintenance operation at a reactor outlet operating temperature of 1350 K. 8 refs., 3 figs., 1 tab.

  12. NEUTRONIC REACTOR

    DOEpatents

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

    1961-11-21

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

  13. REACTOR COOLING

    DOEpatents

    Quackenbush, C.F.

    1959-09-29

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

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

    SciTech Connect

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

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

  15. Advanced light water reactor requirements document: Chapter 3, Reactor coolant system and reactor non-safety auxiliary systems

    SciTech Connect

    Not Available

    1987-06-01

    The purpose of this chapter of the Advanced Light Water Reactor (ALWR) Plant Requirements Document is to establish utility requirements for the design of the Reactor Coolant System and the Reactor Non-safety Auxiliary Systems of Advanced LWR plants consistent with the objectives and principles of the ALWR program. The scope of this chapter covers the reactor coolant system and reactor non-safety auxiliary systems for Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). Non-safety auxiliaries include systems which are required for normal operation of the plant but are not required to operate for accident mitigation or to bring the plant to a safe shutdown condition. For PWRs, the reactor coolant system, steam generator system, chemical and volume control system and boron recycle system are included. For BWRs, the reactor coolant system and reactor water cleanup system are included. The chapter also includes requirements for the above systems which are common to BWRs and PWRs and requirements for process sampling for BWRs and PWRs.

  16. 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.

  17. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1958-04-22

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

  18. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.; Johnson, H.W.

    1961-04-01

    BS>A nuclear reactor incorporating fuel rods passing through a moderator and including tubes of a material of higher Thermal conductivity than the fuel in contact with the fuel is described. The tubes extend beyond the active portion of the reactor into contant with a fiuld coolant.

  19. Reactor building

    SciTech Connect

    Hista, J. C.

    1984-09-18

    Reactor building comprising a vessel shaft anchored in a slab which is peripherally locked. This reactor building comprises a confinement enclosure within which are positioned internal structures constituted by an internal structure floor, a vessel shaft, a slab being positioned between the general floor and the internal structure floor, the vesse

  20. Using fast reactor component evaluation to pressurized water reactor life extension.

    SciTech Connect

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

    1999-10-01

    An understanding of the effects of long-term, low-dose-rate radiation on core components is critical to light-water reactor plant life extension. Following reactor shutdown, materials that had experienced long exposures to low-dose-rate irradiation were retrieved from the EBR-II research reactor for analysis. These components are being analyzed to provide insight into pressurized water reactor life extension. In this work, three examples of EBR-II materials analyses are highlighted: radiation-induced segregation in 304 stainless steel, stress relaxation in Inconel X750, and swelling in 316 stainless steel.;EVALUATION;FAST REACTORS;PWR TYPE REACTORS;

  1. 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.

  2. Temperature effects on chemical reactor

    NASA Astrophysics Data System (ADS)

    Azzouzi, M.

    2008-06-01

    In this paper we had to study some characteristics of the chemical reactors, from which we can understand the reactor operation in different circumstances; from these and the most important factor that has a great effect on the reactor operation is the temperature, it is a mathematical processing of a chemical problem that was already studied, but it may be developed by introducing new strategies of control; in our case we deal with the analysis of a liquid-gas reactor which can make the flotation of the benzene to produce the ethylene; this type of reactors can be used in vast domains of the chemical industry, especially in refinery plants where we find the oil separation and its extractions whether they are gases or liquids which become necessary for industrial technology, especially in our century.

  3. Compact Reactor

    SciTech Connect

    Williams, Pharis E.

    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.

  4. NUCLEAR REACTOR

    DOEpatents

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

    1958-12-01

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

  5. Power Plant Systems Analysis

    NASA Technical Reports Server (NTRS)

    Williams, J. R.; Yang, Y. Y.

    1973-01-01

    Three basic thermodynamic cycles of advanced nuclear MHD power plant systems are studied. The effect of reactor exit temperature and space radiator temperature on the overall thermal efficiency of a regenerative turbine compressor power plant system is shown. The effect of MHD pressure ratio on plant efficiency is also described, along with the dependence of MHD power output, compressor power requirement, turbine power output, mass flow rate of H2, and overall plant efficiency on the reactor exit temperature for a specific configuration.

  6. Proceedings: 2001 PWR/BWR Plant Chemistry Meeting

    SciTech Connect

    2001-05-01

    This report presents proceedings of EPRI's 2001 Plant Chemistry Conference, which brought together approximately 100 industry representatives to discuss experiences and issues regarding nuclear plant chemistry at both pressurized water reactor (PWR) and boiling water reactor (BWR) plants.

  7. Evaluation of the resilience of a full-scale down-flow hanging sponge reactor to long-term outages at a sewage treatment plant in India.

    PubMed

    Onodera, Takashi; Takayama, Daisuke; Ohashi, Akiyoshi; Yamaguchi, Takashi; Uemura, Shigeki; Harada, Hideki

    2016-10-01

    Resilience to process outages is an essential requirement for sustainable wastewater treatment systems in developing countries. In this study, we evaluated the ability of a full-scale down-flow hanging sponge (DHS) reactor to recover after a 10-day outage. The DHS tested in this study uses polyurethane sponge as packing material. This full-scale DHS reactor has been tested over a period of about 4 years in India with a flow rate of 500 m(3)/day. Water was not supplied to the DHS reactor that was subjected to the 10-day outage; however, the biomass did not dry out because the sponge was able to retain enough water. Soon after the reactor was restarted, a small quantity of biomass, amounting to only 0.1% of the total retained biomass, was eluted. The DHS effluent achieved satisfactory removal of suspended solids, chemical oxygen demand, and ammonium nitrogen within 90, 45, and 90 min, respectively. Conversely, fecal coliforms in the DHS effluent did not reach satisfactory levels within 540 min; instead, the normal levels of fecal coliforms were achieved within 3 days. Overall, the tests demonstrated that the DHS reactor was sufficiently robust to withstand long-term outages and achieved steady state soon after restart. This reinforces the suitability of this technology for developing countries. PMID:27450993

  8. Evaluation of the resilience of a full-scale down-flow hanging sponge reactor to long-term outages at a sewage treatment plant in India.

    PubMed

    Onodera, Takashi; Takayama, Daisuke; Ohashi, Akiyoshi; Yamaguchi, Takashi; Uemura, Shigeki; Harada, Hideki

    2016-10-01

    Resilience to process outages is an essential requirement for sustainable wastewater treatment systems in developing countries. In this study, we evaluated the ability of a full-scale down-flow hanging sponge (DHS) reactor to recover after a 10-day outage. The DHS tested in this study uses polyurethane sponge as packing material. This full-scale DHS reactor has been tested over a period of about 4 years in India with a flow rate of 500 m(3)/day. Water was not supplied to the DHS reactor that was subjected to the 10-day outage; however, the biomass did not dry out because the sponge was able to retain enough water. Soon after the reactor was restarted, a small quantity of biomass, amounting to only 0.1% of the total retained biomass, was eluted. The DHS effluent achieved satisfactory removal of suspended solids, chemical oxygen demand, and ammonium nitrogen within 90, 45, and 90 min, respectively. Conversely, fecal coliforms in the DHS effluent did not reach satisfactory levels within 540 min; instead, the normal levels of fecal coliforms were achieved within 3 days. Overall, the tests demonstrated that the DHS reactor was sufficiently robust to withstand long-term outages and achieved steady state soon after restart. This reinforces the suitability of this technology for developing countries.

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

    SciTech Connect

    Dale, M.C.; Lei, S.; Zhou, C.

    1995-11-01

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

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

    SciTech Connect

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

    1995-10-01

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

  11. 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.

  12. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.

    1960-04-01

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

  13. NEUTRONIC REACTOR

    DOEpatents

    Anderson, H.L.

    1960-09-20

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

  14. Chemical Reactors.

    ERIC Educational Resources Information Center

    Kenney, C. N.

    1980-01-01

    Describes a course, including content, reading list, and presentation on chemical reactors at Cambridge University, England. A brief comparison of chemical engineering education between the United States and England is also given. (JN)

  15. NEUTRONIC REACTOR

    DOEpatents

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

    1959-03-24

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

  16. NUCLEAR REACTOR

    DOEpatents

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

    1958-01-21

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

  17. REACTOR SHIELD

    DOEpatents

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

    1959-02-17

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

  18. NUCLEAR REACTOR

    DOEpatents

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

    1962-10-23

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

  19. NEUTRONIC REACTORS

    DOEpatents

    Vernon, H.C.

    1959-01-13

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

  20. NUCLEAR REACTOR

    DOEpatents

    Anderson, C.R.

    1962-07-24

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

  1. NUCLEAR REACTOR

    DOEpatents

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

    1961-06-20

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

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

    SciTech Connect

    Cale, M. Clark; Moelhman, Mark

    1999-09-30

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

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

    SciTech Connect

    Dale, M. Clark; Moelhman, Mark

    1999-09-30

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

  4. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-07-14

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

  5. Reactor antineutrino monitoring with a plastic scintillator array as a new safeguards method

    NASA Astrophysics Data System (ADS)

    Oguri, S.; Kuroda, Y.; Kato, Y.; Nakata, R.; Inoue, Y.; Ito, C.; Minowa, M.

    2014-09-01

    We developed a segmented reactor-antineutrino detector made of plastic scintillators for application as a tool in nuclear safeguards inspection and performed mostly unmanned field operations at a commercial power plant reactor. At a position outside the reactor building, we measured the difference in reactor antineutrino flux above the ground when the reactor was active and inactive.

  6. Fire detection and alarm subsystem design description: 4 x 350 MW(t) Modular HTGR [High-Temperature Gas-Cooled Reactor] Plant

    SciTech Connect

    1986-06-01

    Fire Detection and Alarm is an early warning system used to detect and report the presence of a fire within the plant. It detects, annunciates, and records plant-wide fire alarms, subsystem trouble, and fire console operator actions.

  7. The integral fast reactor and its role in a new generation of nuclear power plants, Tokai, Japan, November 19-21, 1986

    SciTech Connect

    Smith, R.R.

    1986-01-01

    This report presents information on the Integral Fast Reactor and its role in the future. Information is presented in the areas of: inherent safety; other virtues of sodium-cooled breeder; and solving LWR fuel cycle problems with IFR technologies. (JDB)

  8. Research reactors - an overview

    SciTech Connect

    West, C.D.

    1997-03-01

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

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

    SciTech Connect

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

    2014-07-10

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

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

    NASA Astrophysics Data System (ADS)

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

    2014-07-01

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

  11. NEUTRONIC REACTOR

    DOEpatents

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

    1958-09-01

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

  12. Design analysis of the upgraded TREAT reactor

    SciTech Connect

    Bhattacharyya, S.K.

    1982-01-01

    The TREAT reactor, fueled by a dilute dispersion of fully enriched UO/sub 2/ in graphite, has been a premier transient testing facility since 1959. A major Upgrade of the reactor is in progress to enhance its transient testing capability in support of the LMFBR safety program. The TREAT Upgrade (TU) reactor features a modified central zone of the core with higher fissile loadings of the same fuel, clad in Inconel to allow operation at higher temperatures. The demanding functional requirements on the reactor necessitated the use of unique features in the core design which, in turn, presented major calculational complexities in the analysis. Special design methods had to be used in many cases to treat these complexities. The addition of an improved Reactor Control System, a safety grade Plant Protection System and an enhanced Coolant/Filtration System produces a reactor that can meet the functional requirements on the reactor in a safe manner.

  13. POWER REACTOR

    DOEpatents

    Zinn, W.H.

    1958-07-01

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

  14. NEUTRONIC REACTOR

    DOEpatents

    Metcalf, H.E.

    1957-10-01

    A reactor of the type which preferably uses plutonium as the fuel and a liquid moderator, preferably ordinary water, and which produces steam within the reactor core due to the heat of the chain reaction is described. In the reactor shown the fuel elements are essentially in the form of trays and are ventically stacked in spaced relationship. The water moderator is continuously supplied to the trays to maintain a constant level on the upper surfaces of the fuel element as it is continually evaporated by the heat. The steam passes out through the spaces between the fuel elements and is drawn off at the top of the core. The fuel elements are clad in aluminum to prevent deterioration thereof with consequent contamimation of the water.

  15. NUCLEAR REACTOR

    DOEpatents

    Christy, R.F.

    1958-07-15

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

  16. Bioconversion reactor

    DOEpatents

    McCarty, Perry L.; Bachmann, Andre

    1992-01-01

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

  17. REACTOR CONTROL

    DOEpatents

    Fortescue, P.; Nicoll, D.

    1962-04-24

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

  18. Catalytic reactor

    SciTech Connect

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

    2009-03-10

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

  19. NEUTRONIC REACTORS

    DOEpatents

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

    1958-10-14

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

  20. On reactor type comparisons for the next generation of reactors

    SciTech Connect

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

    1991-08-22

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

  1. 75 FR 78777 - Advisory Committee On Reactor Safeguards; Renewal

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-16

    ...: The Advisory Committee on Reactor Safeguards was established by Section 29 of the Atomic Energy Act... accident phenomena; design of nuclear power plant structures, systems and components; materials...

  2. NUCLEAR REACTOR

    DOEpatents

    Young, G.

    1963-01-01

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

  3. NEUTRONIC REACTOR

    DOEpatents

    Fermi, E.; Szilard, L.

    1957-09-24

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

  4. NEUTRONIC REACTOR

    DOEpatents

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

    1958-04-15

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

  5. Selective purge for hydrogenation reactor recycle loop

    DOEpatents

    Baker, Richard W.; Lokhandwala, Kaaeid A.

    2001-01-01

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

  6. Design concepts for the reactor protection and control process instrumentation digital upgrade project at the Donald C. Cook Nuclear Plant units 1 and 2

    SciTech Connect

    Carruth, R.C.; Sotos, W.G.

    1996-06-01

    As the nation`s nuclear power plants age, the need to consider upgrading of their electronic protection and control systems becomes more urgent. Hardware obsolescence and mechanical wear out resulting from frequent calibration and surveillance play a major role in defining their useful life. At Cook Nuclear Plant, a decision was made to replace a major portion of the plant`s protection and control systems with newer technology. This paper describes the engineering processes involved in this successful upgrade and explains the basis for many decisions made while performing the digital upgrade.

  7. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward for 750–800°C Reactor Outlet Temperature

    SciTech Connect

    John Collins

    2009-08-01

    This document presents the NGNP Critical PASSCs and defines their technical maturation path through Technology Development Roadmaps (TDRMs) and their associated Technology Readiness Levels (TRLs). As the critical PASSCs advance through increasing levels of technical maturity, project risk is reduced and the likelihood of within-budget and on-schedule completion is enhanced. The current supplier-generated TRLs and TDRMs for a 750–800°C reactor outlet temperature (ROT) specific to each supplier are collected in Appendix A.

  8. Upscaling of an electronic nose for completely stirred tank reactor stability monitoring from pilot-scale to real-scale agricultural co-digestion biogas plant.

    PubMed

    Adam, Gilles; Lemaigre, Sébastien; Goux, Xavier; Delfosse, Philippe; Romain, Anne-Claude

    2015-02-01

    This study investigated the use of an electronic nose for on-line anaerobic reactor state monitoring at the pilot-scale level and then upscaling to the full-scale level. E-nose indicator was compared to classical state indicators such as pH, alkalinity, volatile fatty acids concentration and to other gas phase compounds. Multivariate statistical process control method, based on principal component analysis and the Hotelling's T(2) statistics was used to derive an indicator representative of the reactor state. At the pilot-scale level, the e-nose indicator was relevant and could distinguish 3 process states: steady-state, transient and collapsing process. At the full-scale level, the e-nose indicator could provide the warning of the major disturbance whereas two slight disturbances were not detected and it gave one major false alarm. This work showed that gas phase relation with anaerobic process should be deeper investigated, as an e-nose could indicate the reactor state, focusing on the gas phase.

  9. Neutronic reactor

    DOEpatents

    Wende, Charles W. J.

    1976-08-17

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

  10. Neutronic reactor

    DOEpatents

    Carleton, John T.

    1977-01-25

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

  11. NEUTRONIC REACTOR

    DOEpatents

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

    1959-10-27

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

  12. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation.

  13. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1962-12-25

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

  14. NEUTRONIC REACTORS

    DOEpatents

    Anderson, H.L.

    1958-10-01

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

  15. Sonochemical Reactors.

    PubMed

    Gogate, Parag R; Patil, Pankaj N

    2016-10-01

    Sonochemical reactors are based on the generation of cavitational events using ultrasound and offer immense potential for the intensification of physical and chemical processing applications. The present work presents a critical analysis of the underlying mechanisms for intensification, available reactor configurations and overview of the different applications exploited successfully, though mostly at laboratory scales. Guidelines have also been presented for optimum selection of the important operating parameters (frequency and intensity of irradiation, temperature and liquid physicochemical properties) as well as the geometric parameters (type of reactor configuration and the number/position of the transducers) so as to maximize the process intensification benefits. The key areas for future work so as to transform the successful technique at laboratory/pilot scale into commercial technology have also been discussed. Overall, it has been established that there is immense potential for sonochemical reactors for process intensification leading to greener processing and economic benefits. Combined efforts from a wide range of disciplines such as material science, physics, chemistry and chemical engineers are required to harness the benefits at commercial scale operation. PMID:27573503

  16. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1960-09-27

    A unit assembly is described for a neutronic reactor comprising a tube and plurality of spaced parallel sandwiches in the tube extending lengthwise thereof, each sandwich including a middle plate having a central opening for plutonium and other openings for fertile material at opposite ends of the plate.

  17. Development of tokamak reactor system analysis code NEW-TORSAC

    NASA Astrophysics Data System (ADS)

    Kasai, Masao; Ida, Toshio; Nishikawa, Masana; Kameari, Akihisa; Nishio, Satoshi; Tone, Tatsuzo

    1987-07-01

    A systems analysis code named NEW-TORSAC (TOkamak Reactor Systems Analysis Code) has been developed by modifying the TORSAC which had been already developed by us. The NEW-TORSAC is available for tokamak reactor designs and evaluations from experimental machines to commercial reactor plants. It has functions to design tokamaks automatically from plasma parameter setting to determining configurations of reactor equipments and calculating main characteristics parameters of auxiliary systems and the capital costs. In the case of analyzing tokamak reactor plants, the code can calculate busbar energy costs. In addition to numerical output, some output of this code such as a reactor configuration, plasma equilibrium, electro-magnetic forces, etc., are graphically displayed. The code has been successfully applied to the scoping studies of the next generation machines and commercial reactor plants.

  18. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1959-12-15

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

  19. NUCLEAR REACTOR

    DOEpatents

    Treshow, M.

    1958-08-19

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

  20. REACTOR MONITORING

    DOEpatents

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

    1959-02-01

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

  1. NEUTRONIC REACTOR

    DOEpatents

    Wade, E.J.

    1958-09-16

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

  2. Nuclear reactor

    DOEpatents

    Wade, Elman E.

    1979-01-01

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

  3. REACTOR UNLOADING

    DOEpatents

    Leverett, M.C.

    1958-02-18

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

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

    SciTech Connect

    Seniuk, P.J.

    1996-12-01

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

  5. Savannah River Site production reactor technical specifications. K Production Reactor

    SciTech Connect

    1996-02-01

    These technical specifications are explicit restrictions on the operation of the Savannah River Site K Production Reactor. They are designed to preserve the validity of the plant safety analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and with the operable equipment that is assumed to mitigate the consequences of an accident. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. This report describes requirements on thermal-hydraulic limits; limiting conditions for operation and surveillance for the reactor, power distribution control, instrumentation, process water system, emergency cooling and emergency shutdown systems, confinement systems, plant systems, electrical systems, components handling, and special test exceptions; design features; and administrative controls.

  6. Neutronic reactor

    DOEpatents

    Lewis, Warren R.

    1978-05-30

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

  7. NUCLEAR REACTORS

    DOEpatents

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

    1961-12-01

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

  8. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashby, J.W.

    1958-09-16

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

  9. NEUTRONIC REACTORS

    DOEpatents

    Anderson, J.B.

    1960-01-01

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

  10. NUCLEAR REACTOR

    DOEpatents

    Grebe, J.J.

    1961-01-24

    A core structure for neutronic reactors adapted for the propulsion of aircraft and rockets is offered. The core is designed for cooling by gaseous media, and comprises a plurality of hollow tapered tubular segments of a porous moderating material impregniated with fissionable fuel nested about a common axis. Alternate ends of the segments are joined. In operation a coolant gas passes through the porous structure and is heated.

  11. REACTOR CONTROL

    DOEpatents

    Ruano, W.J.

    1957-12-10

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

  12. Equilibrium and kinetic studies of in situ generation of ammonia from urea in a batch reactor for flue gas conditioning of thermal power plants

    SciTech Connect

    Sahu, J.N.; Patwardhan, A.V.; Meikap, B.C.

    2009-03-15

    Ammonia has long been known to be useful in the treatment of flue/tail/stack gases from industrial furnaces, incinerators, and electric power generation industries. In this study, urea hydrolysis for production of ammonia, in different application areas that require safe use of ammonia at in situ condition, was investigated in a batch reactor. The equilibrium and kinetic study of urea hydrolysis was done in a batch reactor at reaction pressure to investigate the effect of reaction temperature, initial feed concentration, and time on ammonia production. This study reveals that conversion increases exponentially with an increase in temperature but with increases in initial feed concentration of urea the conversion decreases marginally. Further, the effect of time on conversion has also been studied; it was found that conversion increases with increase in time. Using collision theory, the temperature dependency of forward rate constant developed from which activation energy of the reaction and the frequency factor has been calculated. The activation energy and frequency factor of urea hydrolysis reaction at atmospheric pressure was found to be 73.6 kJ/mol and 2.89 x 10{sup 7} min{sup -1}, respectively.

  13. Rapid starting methanol reactor system

    DOEpatents

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

    1984-01-01

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

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

    SciTech Connect

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

    2008-08-06

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

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

    SciTech Connect

    Dale, M.C.

    1998-06-01

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

  16. 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…

  17. Nuclear reactor

    DOEpatents

    Pennell, William E.; Rowan, William J.

    1977-01-01

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

  18. REACTOR COMPONETN

    DOEpatents

    Creutz, E.C.

    1959-10-27

    A reactor fuel element comprised of a slug of fissionable material disposed in a sheath of corrosion resistantmaterial is described. The sheath is in the form of a tubular container closed at one end and is in tight-fitting engagement with the peripheral sunface of the slug. An inner cap is insented into the open end of the sheath against the slug, which end is then bent around the inner cap and welded thereto. An outer cap is then welded around its peripheny to the bent portion of the container.

  19. NUCLEAR REACTORS

    DOEpatents

    Long, E.; Ashley, J.W.

    1958-12-16

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

  20. A novel sorbent for transport reactors and fluidized bed reactors

    SciTech Connect

    Copeland, R.; Cesario, M.; Gershanovich, Y.; Sibold, J.; Windecker, B.

    1998-12-31

    Coal Fired Gasifier Combined Cycles (GCC) have both high efficiency and very low emissions. GCCs critically need a method of removing the H{sub 2}S produced from the sulfur in the coal from the hot gases. There has been extensive research on hot gas cleanup systems, focused on the use of a zinc oxide based sorbent (e.g., zinc titanate). TDA Research, Inc. (TDA) is developing a novel sorbent with improved attrition resistance for transport reactors and fluidized bed reactors. The authors are testing sorbents at conditions simulating the operating conditions of the Pinon Pine clean coal technology plant. TDA sulfided several different formulations at 538 C and found several that have high sulfur capacity when tested in a fluidized bed reactor. TDA initiated sorbent regeneration at 538 C. The sorbents retained chemical activity with multiple cycles. Additional tests will be conducted to evaluate the best sorbent formulation.

  1. Nuclear Power from Fission Reactors. An Introduction.

    ERIC Educational Resources Information Center

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  2. The Cascade inertial confinement fusion reactor concept

    NASA Astrophysics Data System (ADS)

    Pitts, J. H.; Hogan, W. J.; Tobin, M. T.; Bourque, R. F.; Meier, W. R.

    1990-12-01

    The Cascade reactor concept has the potential of converting inertial confinement fusion (ICF) energy into electrical power safely, efficiently, and with low activation. Its flexibility permits a number of options for materials, blankets, fuel-pellet designs and drivers. This report documents a theoretical and experimental study culminating in a reference Cascade conceptual design that produces 890 MW of electrical power with a net plant efficiency of 47 percent if a heavy-ion driver if used. The reactor is double-cone shaped and rotates at 50 rpm. A ceramic-granule blanket flows through the reactor held against the reactor wall by the rotation. The blanket serves several functions: it absorbs energy from fusion reactions that occur at 5 Hz in the center of the reactor, thereby protecting the reactor wall and extending its lifetime to that of power plant; it acts as a heat-exchange medium to transfer fusion energy to high-pressure helium gas used in power conversion; and it produces tritium to replace that burned in the fusion process. Cascade's illumination geometry is restricted, so that good energy coupling to presently-envisioned fuel pellets is practical only with heavy-ion drivers. Laser drivers would require the use of fuel pellets with advanced design features. We discuss the reactor concept, heat exchanger, balance of plant, other systems that would be necessary for a full-scale production of electrical power, and experiments that prove the feasibility of a flowing granular blanket. A cost study predicts that Cascade, using a heavy-ion driver, could produce electricity for between 5.5 and 6.8 cents/kWh-comparable to the cost of power using modular high-temperature gas-cooled reactors, pressurized-water reactors, or coal-fired power plants. Finally, we include an annotated bibliography of the over 50 reports which have been written about Cascade.

  3. Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 1, Final report

    SciTech Connect

    Waganer, L.M.; Driemeyer, D.E.; Lee, V.D.

    1992-03-01

    This report contains a review of design studies for inertial confinement reactors. The first of three volumes briefly discusses the following: Introduction; Key objectives, requirements, and assumptions; Systems modeling and trade studies; Prometheus-L reactor plant design overview; Prometheus-H reactor plant design overview; Key technical issues and R&D requirements; Comparison of IFE designs; and study conclusions.

  4. 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.

  5. Hybrid adsorptive membrane reactor

    DOEpatents

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

    2011-03-01

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

  6. ESBWR... An Evolutionary Reactor Design

    SciTech Connect

    Gamble, Robert E.; Hinds, David H.; Hucik, Steven A.; Maslak, Chris E.

    2006-07-01

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

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

    NASA Astrophysics Data System (ADS)

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

    1999-10-01

    An understanding of the effects of long-term, low-dose-rate radiation on core components is critical to light-water reactor plant life extension. Following reactor shutdown, materials that had experienced long exposures to low-dose-rate irradiation were retrieved from the EBR-II research reactor for analysis. These components are being analyzed to provide insight into pressurized water reactor life extension. In this work, three examples of EBR-II materials analyses are highlighted: radiation-induced segregation in 304 stainless steel, stress relaxation in Inconel X750, and swelling in 316 stainless steel.

  8. Comparison between polishing (maturation) ponds and subsurface flow constructed wetlands (planted and unplanted) for the post-treatment of the effluent from UASB reactors.

    PubMed

    von Sperling, M; Dornelas, F L; Assunção, F A L; de Paoli, A C; Mabub, M O A

    2010-01-01

    This paper presents the results of a comparison of the performance of two treatment systems operating in parallel, with the same influent wastewater. The investigated systems are (i) UASB + three polishing ponds in series + coarse filter (200 population equivalents) and (ii) UASB + subsurface flow constructed wetlands (50 population equivalents). Two wetland units, operating in parallel, were analysed, being one planted (Typha latifolia) and the other unplanted. The systems were located in Belo Horizonte, Brazil. The wetland systems showed to be more efficient in the removal of organic matter and suspended solids, leading to good effluent BOD and COD concentrations and excellent SS concentrations. The planted wetland performed better than the unplanted unit, but the latter was also able to provide a good effluent quality. The polishing pond system was more efficient in the removal of nitrogen (ammonia) and coliforms (E. coli). Land requirements and cost considerations are presented.

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

    SciTech Connect

    Jeffrey D Bryan

    2009-09-01

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

  10. Development of high-strength concrete mix designs in support of the prestressed concrete reactor vessel design for a HTGR steam cycle/cogeneration plant

    SciTech Connect

    Naus, D.J.; Oland, C.B.

    1985-01-01

    Design optimization studies indicate that a significant reduction in the size of the PCRV for a 2240 MW(t) HTGR plant can be effected through utilization of high-strength concrete in conjunction with large capacity prestressing systems. A three-phase test program to develop and evaluate high-strength concretes (>63.4 MPa) is described. Results obtained under Phase I of the investigation related to materials selection-evaluation and mix design development are presented. 3 refs., 4 figs.

  11. Plant heat cycles, vessel internal arrangement, and auxiliary systems. Volume five

    SciTech Connect

    Not Available

    1986-01-01

    This volume covers nuclear power plant heat cycles (type of nuclear power cycles, power cycle refinements, BWR/PWR power cycle, BWR/PWR reactor coolant system), reactor vessel internal arrangement (reactor vessel features, BWR/PWR reactor vessel and internals, BWR/PWR reactor core), reactor auxiliary systems (purpose of reactor auxiliary systems, PWR and BWR reactor auxiliary systems, PWR and BWR control rod drive mechanisms).

  12. Control Means for Reactor

    DOEpatents

    Manley, J. H.

    1961-06-27

    An apparatus for controlling a nuclear reactor includes a tank just below the reactor, tubes extending from the tank into the reactor, and a thermally expansible liquid neutron absorbent material in the tank. The liquid in the tank is exposed to a beam of neutrons from the reactor which heats the liquid causing it to expand into the reactor when the neutron flux in the reactor rises above a predetermincd danger point. Boron triamine may be used for this purpose.

  13. Performance of a combined system of biotrickling filter and photocatalytic reactor in treating waste gases from a paint-manufacturing plant.

    PubMed

    Zeng, Peiyuan; Li, Jianjun; Liao, Dongqi; Tu, Xiang; Xu, Meiying; Sun, Guoping

    2016-01-01

    A pilot-scale biotrickling filter (BTF) was established in treating the waste gases that are intermittently produced from an automobile paint-manufacturing workshop. Results showed that the BTF required longer time to adapt to the aromatic compounds. The removal efficiencies (REs) for all aliphatic compounds reached more than 95% on day 80. Aromatic compounds were not easily removed by the BTF. The REs obtained by the BTF for toluene, ethylbenzene, m-xylene, o-xylene and p-xylene on day 80 were 72.7%, 77.2%, 71.9%, 74.8% and 60.0%, respectively. A maximum elimination capacity (EC) of 13.8 g-C m(-3) h(-1) of the BTF was achieved at an inlet loading rate of 19.4 g-C m(-3) h(-1) with an RE of 72%. Glucose addition promoted the biomass accumulation despite the fact that temporal decrease of REs for aromatic compounds occurred. When the inlet loading rates exceed 11.1 g-C m(-3) h(-1), the REs of the aromatic compounds decreased by 10% to 15%. This negative effect of shock loads on the performance of the BTF can be attenuated by the pre-treatment of the photocatalytic reactor. Nearly all components were removed by the combined system with REs of 99%.

  14. NEUTRONIC REACTOR

    DOEpatents

    Stewart, H.B.

    1958-12-23

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

  15. NEUTRONIC REACTOR

    DOEpatents

    Wigner, E.P.

    1957-09-17

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

  16. Nuclear reactor

    DOEpatents

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

    1977-01-01

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

  17. Upgrading of an activated sludge wastewater treatment plant by adding a moving bed biofilm reactor as pre-treatment and ozonation followed by biofiltration for enhanced COD reduction: design and operation experience.

    PubMed

    Kaindl, Nikolaus

    2010-01-01

    A paper mill producing 500,000 ton of graphic paper annually has an on-site wastewater treatment plant that treats 7,240,000 m³ of wastewater per year, mechanically first, then biologically and at last by ozonation. Increased paper production capacity led to higher COD load in the mill effluent while production of higher proportions of brighter products gave worse biodegradability. Therefore the biological capacity of the WWTP needed to be increased and extra measures were necessary to enhance the efficiency of COD reduction. The full scale implementation of one MBBR with a volume of 1,230 m³ was accomplished in 2000 followed by another MBBR of 2,475 m³ in 2002. An ozonation step with a capacity of 75 kg O₃/h was added in 2004 to meet higher COD reduction demands during the production of brighter products and thus keeping the given outflow limits. Adding a moving bed biofilm reactor prior to the existing activated sludge step gives: (i) cost advantages when increasing biological capacity as higher COD volume loads of MBBRs allow smaller reactors than usual for activated sludge plants; (ii) a relief of strain from the activated sludge step by biological degradation in the MBBR; (iii) equalizing of peaks in the COD load and toxic effects before affecting the activated sludge step; (iv) a stable volume sludge index below 100 ml/g in combination with an optimization of the activated sludge step allows good sludge separation--an important condition for further treatment with ozone. Ozonation and subsequent bio-filtration pre-treated waste water provide: (i) reduction of hard COD unobtainable by conventional treatment; (ii) controllable COD reduction in a very wide range and therefore elimination of COD-peaks; (iii) reduction of treatment costs by combination of ozonation and subsequent bio-filtration; (iv) decrease of the color in the ozonated wastewater. The MBBR step proved very simple to operate as part of the biological treatment. Excellent control of the COD

  18. [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.

  19. [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.

  20. Self-actuating reactor shutdown system

    DOEpatents

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

    1988-01-01

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

  1. Environmental Information Document: L-reactor reactivation

    SciTech Connect

    Mackey, H.E. Jr.

    1982-04-01

    Purpose of this Environmental Information Document is to provide background for assessing environmental impacts associated with the renovation, restartup, and operation of L Reactor at the Savannah River Plant (SRP). SRP is a major US Department of Energy installation for the production of nuclear materials for national defense. The purpose of the restart of L Reactor is to increase the production of nuclear weapons materials, such as plutonium and tritium, to meet projected needs in the nuclear weapons program.

  2. Hot Gas Desulfurization Using Transport Reactors

    SciTech Connect

    Moorehead, E.L.

    1996-12-31

    Sierra Pacific Power Company is building a 100 MW, IGCC power plant based on KRW fluid bed gasifier technology that utilizes transport reactors for hot gas desulfurization and sorbent regeneration. Use of a transport absorber avoids the need for pre-filtration of dust-laden gasifier effluent, while a transport regenerator allows for the use of 100% air without the need for heat exchange equipment. Selection of transport reactors for hot gas desulfurization using a proprietary sorbent, based on testing performed in a transport reactor test unit (TRTU) at the M. W. Kellogg Technology Development Center and in a fixed bed reactor at Morgantown Energy Technology Center (METC), is outlined. The results obtained in these two test facilities and reasons for selecting transport reactors for the IGCC power plant in preference to either fixed bed or fluidized bed reactors are discussed. This paper reviews the evolution of the hot gas desulfurization system designs and includes selected results on H{sub 2}S absorption and regeneration of sulfided sorbent over several absorption/regeneration cycles conducted in the TRTU and the METC fixed bed reactor. The original design for the Sierra Pacific Project was based on fixed bed reactors with zinc ferrite as the sorbent. Owing to the high steam requirements of this sorbent, zinc titanate was selected and tested in a fixed bed reactor and was found unacceptable due to loss of strength on cyclic absorption/regeneration operation. Another sorbent evaluated was Z-Sorb{reg_sign}, a proprietary sorbent developed by Phillips Petroleum Company, was found to have excellent sulfur capacity, structural strength and regenerability. Steam was found unsuitable as fixed bed regenerator diluent, this results in a requirement for a large amount of inert gas, whereas a transport regenerator requires no diluent. The final Sierra design features transport reactors for both desulfurization and regeneration steps using neat air. 3 refs., 3 figs., 2 tabs.

  3. 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.

  4. Expeditious method to determine uranium in the process control samples of chemical plant separating (233)U from thoria irradiated in power reactors.

    PubMed

    Kedari, C S; Kharwandikar, B K; Banerjee, K

    2016-11-01

    Analysis of U in the samples containing a significant proportion of (232)U and high concentration of Th is of great concern. Transmutation of Th in the nuclear power reactor produces a notable quantity of (232)U (half life 68.9 years) along with fissile isotope (233)U. The decay series of (232)U is initiated with (228)Th (half life 1.9 year) and it is followed by several short lived α emitting progenies, (224)Ra, (220)Rn, (216)Po, (212)Bi and (212)Po. Even at the smallest contamination of (228)Th in the sample, a very high pulse rate of α emission is obtained, which is to be counted for the radiometric determination of [U]. A commercially available anionic type of extractant Alamine®336 is used to obtain the selective extraction of U from other alpha active elements and fission products present in the sample. Experimental conditions of liquid-liquid extraction (LLE) are optimized for obtaining maximum decontamination and recovery of U in the organic phase. The effect of some interfering ionic impurities in the sample on the process of separation is investigated. Depending on the level of the concentration of U in the samples, spectrophotometry or radiometry methods are adopted for its determination after separation by LLE. Under optimized experimental conditions, i.e. 5.5M HCl in the aqueous phase and 0.27M Alamin®336 in the organic phase, the recovery of U is about 100%, the decontamination factor with respect to Th is >2000 and the extraction of fission products like (90)Sr, (144)Ce and (134,137)Cs is negligible. The detection limit for [U] using α radiometry is 10mg/L, even in presence of >100g/L of Th in the sample. Accuracy and precision for the determination of U is also assessed. Reproducibility of results is within 5%. This method shows very good agreement with the results obtained by mass spectrometry.

  5. Expeditious method to determine uranium in the process control samples of chemical plant separating (233)U from thoria irradiated in power reactors.

    PubMed

    Kedari, C S; Kharwandikar, B K; Banerjee, K

    2016-11-01

    Analysis of U in the samples containing a significant proportion of (232)U and high concentration of Th is of great concern. Transmutation of Th in the nuclear power reactor produces a notable quantity of (232)U (half life 68.9 years) along with fissile isotope (233)U. The decay series of (232)U is initiated with (228)Th (half life 1.9 year) and it is followed by several short lived α emitting progenies, (224)Ra, (220)Rn, (216)Po, (212)Bi and (212)Po. Even at the smallest contamination of (228)Th in the sample, a very high pulse rate of α emission is obtained, which is to be counted for the radiometric determination of [U]. A commercially available anionic type of extractant Alamine®336 is used to obtain the selective extraction of U from other alpha active elements and fission products present in the sample. Experimental conditions of liquid-liquid extraction (LLE) are optimized for obtaining maximum decontamination and recovery of U in the organic phase. The effect of some interfering ionic impurities in the sample on the process of separation is investigated. Depending on the level of the concentration of U in the samples, spectrophotometry or radiometry methods are adopted for its determination after separation by LLE. Under optimized experimental conditions, i.e. 5.5M HCl in the aqueous phase and 0.27M Alamin®336 in the organic phase, the recovery of U is about 100%, the decontamination factor with respect to Th is >2000 and the extraction of fission products like (90)Sr, (144)Ce and (134,137)Cs is negligible. The detection limit for [U] using α radiometry is 10mg/L, even in presence of >100g/L of Th in the sample. Accuracy and precision for the determination of U is also assessed. Reproducibility of results is within 5%. This method shows very good agreement with the results obtained by mass spectrometry. PMID:27591623

  6. Reactor safety method

    DOEpatents

    Vachon, Lawrence J.

    1980-03-11

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

  7. NEUTRONIC REACTOR MANIPULATING DEVICE

    DOEpatents

    Ohlinger, L.A.

    1962-08-01

    A cable connecting a control rod in a reactor with a motor outside the reactor for moving the rod, and a helical conduit in the reactor wall, through which the cable passes are described. The helical shape of the conduit prevents the escape of certain harmful radiations from the reactor. (AEC)

  8. Thermophilic co-digestion of organic fraction of municipal solid wastes with FOG wastes from a sewage treatment plant: reactor performance and microbial community monitoring.

    PubMed

    Martín-González, Lucia; Castro, Rita; Pereira, M Alcina; Alves, M Madalena; Font, Xavier; Vicent, Teresa

    2011-04-01

    Working at thermophilic conditions instead of mesophilic, and also the addition of a co-substrate, are both the ways to intend to improve the anaerobic digestion of the source-collected organic fraction of municipal solid wastes (SC-OFMSW). Addition of sewage treatment plant fat, oil and grease wastes (STP-FOGW), that are nowadays sent to landfill, would represent an opportunity to recover a wasted methane potential and, moreover, improve the whole process. In this study, after a first period feeding only SC-OFMSW, a co-digestion step was performed maintaining thermophilic conditions. During the co-digestion period enhancements in biogas production (52%) and methane yield (36%) were achieved. In addition, monitoring of microbial structure by using PCR-DGGE and cloning techniques showed that bacterial community profiles clustered in two distinct groups, before and after the extended contact with STP-FOGW, being more affected by the STP-FOGW addition than the archaeal one.

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

    SciTech Connect

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

    2012-09-14

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

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

    SciTech Connect

    Tadashi Narabayashi; Yoichiro Shimadu; Toshiiro Murase; Masatoshi Nagai; Michitsugu Mori; Shuichi Ohmori

    2006-07-01

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

  11. Inspecting the reactor vessel penetrations

    SciTech Connect

    Bodson, F.; Fleming, K.W.

    1995-08-01

    The susceptibility of Alloy 600 to Primary Water Stress Corrosion Cracking (PWSCC) continues to plague nuclear power plants. Recently, the problem of PWSCC cracking has manifested itself in Control Rod Drive Mechanism (CRDM) head penetrations in nuclear plants in Europe. Framatome has been extensively involved in the performance of both inspections and repairs of CRDM head penetrations at Electricite de France (EdF) plants. B and W Nuclear Technologies (BWNT), building on Framatome technology, has developed a fully integrated service package and robotic manipulator to inspect and repair CRDM head penetrations for US utilities. Reactor vessel bottom penetration are also made of Alloy 600 and to tackle this potential PWSCC problem at EdF plants, Framatome has been performing specific inspections in order to detect the appearance of the phenomenon. This paper describes the overall range of inspection techniques and toolings developed to address these issues.

  12. Nuclear reactor

    DOEpatents

    Thomson, Wallace B.

    2004-03-16

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

  13. Tritium issues in commercial pressurized water reactors

    SciTech Connect

    Jones, G.

    2008-07-15

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  14. Mechanical cutting of irradiated reactor internal components

    SciTech Connect

    Anderson, Michael G.

    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.

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

    SciTech Connect

    Ruger, C.J.; Higgins, J.C.

    1993-11-01

    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 the 1970`s and early 1980`s raised concerns from the US Nuclear Regulatory Commission (NRC) that gross failures may lead to reactor core uncovery and subsequent core damage. Some seal failure events resulted in a loss of primary coolant to the containment at flow rates greater than the normal makeup capacity of Pressurized Water Reactor (PWR) plants. This is an example of RCP seal failures resulting in a small Loss of Coolant Accident (LOCA). This paper discusses observed and potential causes of RCP seal failure and the recommendations for limiting the likelihood of a seal induced small LOCA. Issues arising out of the research supporting these recommendations and subsequent public comments by the utility industry on them, serve as lessons learned, which are applicable to the design of new reactor plants.

  16. Nuclear Power Plants. Revised.

    ERIC Educational Resources Information Center

    Lyerly, Ray L.; Mitchell, Walter, III

    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: Why Use Nuclear Power?; From Atoms to Electricity; Reactor Types; Typical Plant Design Features; The Cost of Nuclear Power; Plants in the United States; Developments in Foreign…

  17. NUCLEAR POWER PLANT

    DOEpatents

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

    1963-05-14

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

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

    SciTech Connect

    Rosenthal, Murray Wilford

    2009-08-01

    The Oak Ridge National Laboratory has built and operated 13 nuclear reactors in its 66-year history. The first was the graphite reactor, the world's first operational nuclear reactor, which served as a plutonium production pilot plant during World War II. It was followed by two aqueous-homogeneous reactors and two red-hot molten-salt reactors that were parts of power-reactor development programs and by eight others designed for research and radioisotope production. One of the eight was an all-metal fast burst reactor used for health physics studies. All of the others were light-water cooled and moderated, including the famous swimming-pool reactor that was copied dozens of times around the world. Two of the reactors were hoisted 200 feet into the air to study the shielding needs of proposed nuclear-powered aircraft. The final reactor, and the only one still operating today, is the High Flux Isotope Reactor (HFIR) that was built particularly for the production of californium and other heavy elements. With the world's highest flux and recent upgrades that include the addition of a cold neutron source, the 44-year-old HFIR continues to be a valuable tool for research and isotope production, attracting some 500 scientific visitors and guests to Oak Ridge each year. This report describes all of the reactors and their histories.

  19. Uncertainty quantification approaches for advanced reactor analyses.

    SciTech Connect

    Briggs, L. L.; Nuclear Engineering Division

    2009-03-24

    The original approach to nuclear reactor design or safety analyses was to make very conservative modeling assumptions so as to ensure meeting the required safety margins. Traditional regulation, as established by the U. S. Nuclear Regulatory Commission required conservatisms which have subsequently been shown to be excessive. The commission has therefore moved away from excessively conservative evaluations and has determined best-estimate calculations to be an acceptable alternative to conservative models, provided the best-estimate results are accompanied by an uncertainty evaluation which can demonstrate that, when a set of analysis cases which statistically account for uncertainties of all types are generated, there is a 95% probability that at least 95% of the cases meet the safety margins. To date, nearly all published work addressing uncertainty evaluations of nuclear power plant calculations has focused on light water reactors and on large-break loss-of-coolant accident (LBLOCA) analyses. However, there is nothing in the uncertainty evaluation methodologies that is limited to a specific type of reactor or to specific types of plant scenarios. These same methodologies can be equally well applied to analyses for high-temperature gas-cooled reactors and to liquid metal reactors, and they can be applied to steady-state calculations, operational transients, or severe accident scenarios. This report reviews and compares both statistical and deterministic uncertainty evaluation approaches. Recommendations are given for selection of an uncertainty methodology and for considerations to be factored into the process of evaluating uncertainties for advanced reactor best-estimate analyses.

  20. Automatic reactor model synthesis with genetic programming.

    PubMed

    Dürrenmatt, David J; Gujer, Willi

    2012-01-01

    Successful modeling of wastewater treatment plant (WWTP) processes requires an accurate description of the plant hydraulics. Common methods such as tracer experiments are difficult and costly and thus have limited applicability in practice; engineers are often forced to rely on their experience only. An implementation of grammar-based genetic programming with an encoding to represent hydraulic reactor models as program trees should fill this gap: The encoding enables the algorithm to construct arbitrary reactor models compatible with common software used for WWTP modeling by linking building blocks, such as continuous stirred-tank reactors. Discharge measurements and influent and effluent concentrations are the only required inputs. As shown in a synthetic example, the technique can be used to identify a set of reactor models that perform equally well. Instead of being guided by experience, the most suitable model can now be chosen by the engineer from the set. In a second example, temperature measurements at the influent and effluent of a primary clarifier are used to generate a reactor model. A virtual tracer experiment performed on the reactor model has good agreement with a tracer experiment performed on-site.

  1. 850/sup 0/C VHTR plant technical description

    SciTech Connect

    Not Available

    1980-06-01

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

  2. 78 FR 73898 - Operator Licensing Examination Standards for Power Reactors

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-12-09

    ... COMMISSION Operator Licensing Examination Standards for Power Reactors AGENCY: Nuclear Regulatory Commission... Standards for Power Reactors.'' DATES: Submit comments by February 7, 2014. Comments received after this..., and grading of examinations used for licensing operators at nuclear power plants pursuant to...

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

    SciTech Connect

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

    2015-12-15

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

  4. Small Modular Reactors (468th Brookhaven Lecture)

    SciTech Connect

    Bari, Robert

    2011-04-20

    With good reason, much more media attention has focused on nuclear power plants than solar farms, wind farms, or hydroelectric plants during the past month and a half. But as nations around the world demand more energy to power everything from cell phone batteries to drinking water pumps to foundries, nuclear plants are the only non-greenhouse-gas producing option that can be built to operate almost anywhere, and can continue to generate power during droughts, after the sun sets, and when winds die down. To supply this demand for power, designers around the world are competing to develop more affordable nuclear reactors of the future: small modular reactors. Brookhaven Lab is working with DOE to ensure that these reactors are designed to be safe for workers, members of surrounding communities, and the environment and to ensure that the radioactive materials and technology will only be used for peaceful purposes, not weapons. In his talk, Bari will discuss the advantages and challenges of small modular reactors and what drives both international and domestic interest in them. He will also explain how Brookhaven Lab and DOE are working to address the challenges and provide a framework for small modular reactors to be commercialized.

  5. Hybrid plasmachemical reactor

    SciTech Connect

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

    2015-04-15

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

  6. Attrition reactor system

    SciTech Connect

    Scott, Charles D.; Davison, Brian H.

    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.

  7. 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.

  8. The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants.

    PubMed

    Park, S D; Kim, J S; Han, S H; Ha, Y K; Song, K S; Jee, K Y

    2009-09-01

    In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

  9. Requirements for Reactor Physics Design

    SciTech Connect

    Diamond,D.J.

    2008-04-11

    It has been recognized that there is a need for requirements and guidance for design and operation of nuclear power plants. This is becoming more important as more reactors are being proposed to be built. In parallel with activities in individual countries are norms established by international organizations. This paper discusses requirements/guidance for neutronic design and operation as promulgated by the U.S. Nuclear Regulatory Commission (NRC). As an example, details are given for one reactor physics parameter, namely, the moderator temperature reactivity coefficient. The requirements/guidance from the NRC are discussed in the context of those generated for the International Atomic Energy Agency. The requirements/guidance are not identical from the two sources although they are compatible.

  10. Generic Stellarator-like Magnetic Fusion Reactor

    NASA Astrophysics Data System (ADS)

    Sheffield, John; Spong, Donald

    2015-11-01

    The Generic Magnetic Fusion Reactor paper, published in 1985, has been updated, reflecting the improved science and technology base in the magnetic fusion program. Key changes beyond inflation are driven by important benchmark numbers for technologies and costs from ITER construction, and the use of a more conservative neutron wall flux and fluence in modern fusion reactor designs. In this paper the generic approach is applied to a catalyzed D-D stellarator-like reactor. It is shown that an interesting power plant might be possible if the following parameters could be achieved for a reference reactor: R/ < a > ~ 4 , confinement factor, fren = 0.9-1.15, < β > ~ 8 . 0 -11.5 %, Zeff ~ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ~ 0.07, Bm ~ 14-16 T, and R ~ 18-24 m. J. Sheffield was supported under ORNL subcontract 4000088999 with the University of Tennessee.

  11. Heat pipe reactors for space power applications

    NASA Technical Reports Server (NTRS)

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

    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. Each fuel element is cooled by an axially located molybdenum heat pipe containing either sodium or lithium vapor. Virtues of the reactor designs are the avoidance of single-point failure mechanisms, the relatively high operating temperature, and the expected long lifetimes of the fuel element components.

  12. Study Gives Good Odds on Nuclear Reactor Safety

    ERIC Educational Resources Information Center

    Russell, Cristine

    1974-01-01

    Summarized is data from a recent study on nuclear reactor safety completed by Norman C. Rasmussen and others. Non-nuclear events are about 10,000 times more likely to produce large accidents than nuclear plants. (RH)

  13. Period meter for reactors

    DOEpatents

    Rusch, Gordon K.

    1976-01-06

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

  14. Reactor System Transient Code.

    1999-07-14

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

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

    SciTech Connect

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

    2009-03-01

    The goals of the Global Nuclear Energy Partnership (GNEP) are to expand the use of nuclear energy to meet increasing global energy demand in an environmentally sustainable manner, to address nuclear waste management issues without making separated plutonium, and to address nonproliferation concerns. The advanced burner reactor (ABR) is a fast reactor concept which supports the GNEP fuel cycle system. Since the integral fast reactor (IFR) and advanced liquid-metal reactor (ALMR) projects were terminated in 1994, there has been no major development on sodium-cooled fast reactors in the United States. Therefore, in support of the GNEP fast reactor program, the history of sodium-cooled reactor development was reviewed to support the initiation of this technology within the United States and to gain an understanding of the technology gaps that may still remain for sodium fast reactor technology. The fuel-handling system is a key element of any fast reactor design. The major functions of this system are to receive, test, store, and then load fresh fuel into the core; unload from the core; then clean, test, store, and ship spent fuel. Major requirements are that the system must be reliable and relatively easy to maintain. In addition, the system should be designed so that it does not adversely impact plant economics from the viewpoints of capital investment or plant operations. In this gap analysis, information on fuel-handling operating experiences in the following reactor plants was carefully reviewed: EBR-I, SRE, HNPF, Fermi, SEFOR, FFTF, CRBR, EBR-II, DFR, PFR, Rapsodie, Phenix, Superphenix, KNK, SNR-300, Joyo, and Monju. The results of this evaluation indicate that a standardized fuel-handling system for a commercial fast reactor is yet to be established. However, in the past sodium-cooled reactor plants, most major fuel-handling components-such as the rotatable plug, in-vessel fuel-handling machine, ex-vessel fuel transportation cask, ex-vessel sodium-cooled storage

  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. High solids fermentation reactor

    DOEpatents

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

    1993-01-01

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

  18. Efficient Silicon Reactor

    NASA Technical Reports Server (NTRS)

    Bates, H. E.; Hill, D. M.; Jewett, D. N.

    1983-01-01

    High-purity silicon efficiently produced and transferred by continuous two-cycle reactor. New reactor operates in relatively-narrow temperature rate and uses large surfaces area to minimize heat expenditure and processing time in producing silicon by hydrogen reduction of trichlorosilane. Two cycles of reactor consists of silicon production and removal.

  19. NUCLEAR REACTOR CONTROL SYSTEM

    DOEpatents

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

    1959-11-01

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

  20. 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.

  1. NEUTRONIC REACTOR SHIELDING

    DOEpatents

    Borst, L.B.

    1961-07-11

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

  2. Improved vortex reactor system

    DOEpatents

    Diebold, James P.; Scahill, John W.

    1995-01-01

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

  3. Advanced Test Reactor Tour

    ScienceCinema

    Miley, Don

    2016-07-12

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

  4. Probabilistic Safety Assessment of Tehran Research Reactor

    SciTech Connect

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

    2004-07-01

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

  5. Economic characteristics of a smaller, simpler reactor

    SciTech Connect

    LaBar, M.; Bowers, H.

    1988-01-01

    Reduced load growth and heightened concern with economic risk has led to an expressed utility preference for smaller capacity additions. The Modular High Temperature Reactor (MHTGR) plant has been developed as a small, simple plant that has limited financial risk and is economically competitive with comparatively sized coal plants. Competitive economics is achieved by the simplifications made possible in a small MHTGR, reduction in the quantity of nuclear grade construction and design standardization and certification. Assessments show the MHTGR plant to have an economic advantage over coal plants for plant sizes from 270 MWe to 1080 MWe. Financial risk is limited by small unit sizes and short lead times that allow incremental deployment. Evaluations show the MHTGR incremental deployment capability to reduce negative cash flows by almost a factor of 2 relative to that required by a single large nuclear plant.

  6. The polonium-210 problem in thermonuclear reactor

    SciTech Connect

    Shchipakhin, O.L.; Borisov, N.B.; Churkin, S.L.

    1993-12-31

    Polonium 210 forms in the lithium-lead eutectic blanket of a thermonuclear reactor. On the basis of obtained experimental data some estimates have been calculated on the ITER blanket accident consequences. The LOCA type accident represents the failure of eutectic circuit in the process of transfusion of liquid eutectic from blanket to the tritium reprocessing plant.

  7. NGNP Reactor Coolant Chemistry Control Study

    SciTech Connect

    Brian Castle

    2010-11-01

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

  8. Autonomous Control of Space Reactor Systems

    SciTech Connect

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

    2007-11-30

    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

  9. US graphite reactor D&D experience

    SciTech Connect

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

    1997-02-01

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

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

    SciTech Connect

    Not Available

    1986-01-01

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

  11. Nuclear reactor overflow line

    DOEpatents

    Severson, Wayne J.

    1976-01-01

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

  12. Reactor vessel support system

    DOEpatents

    Golden, Martin P.; Holley, John C.

    1982-01-01

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

  13. 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.

  14. Reactor water cleanup system

    DOEpatents

    Gluntz, Douglas M.; Taft, William E.

    1994-01-01

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

  15. 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.

  16. 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.

  17. Modular stellarator fusion reactor concept

    NASA Astrophysics Data System (ADS)

    Miller, R. L.; Krakowski, R. A.

    1981-08-01

    A steady-state ignited, DT-fueled, magnetic fusion reactor is proposed for use as a central electric-power station. The MSR concept combines the physics of the classic stellarator confinement topology with an innovative, modular-coil design. Parametric tradeoff calculations are described, leading to the selection of an interim design point for a 4-GWt plant based on Alcator transport scaling and an average beta value of 0.04 in an 1 = 2 system with a plasma aspect ratio of 11. The physical basis of the design point is described together with supporting magnetics, coil-force, and stress computations.

  18. Advanced reactor physics methods for heterogeneous reactor cores

    NASA Astrophysics Data System (ADS)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  19. Instrumentation and control improvements at Experimental Breeder Reactor II

    SciTech Connect

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I C systems of the next generation of liquid metal reactor (LMR) plants.

  20. Instrumentation and control improvements at Experimental Breeder Reactor II

    SciTech Connect

    Christensen, L.J.; Planchon, H.P.

    1993-03-01

    The purpose of this paper is to describe instrumentation and control (I&C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I&C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I&C systems of the next generation of liquid metal reactor (LMR) plants.

  1. Artificial intelligence program in a computer application supporting reactor operations

    SciTech Connect

    Stratton, R.C.; Town, G.G.

    1985-01-01

    Improving nuclear reactor power plant operability is an ever-present concern for the nuclear industry. The definition of plant operability involves a complex interaction of the ideas of reliability, safety, and efficiency. This paper presents observations concerning the issues involved and the benefits derived from the implementation of a computer application which combines traditional computer applications with artificial intelligence (AI) methodologies. A system, the Component Configuration Control System (CCCS), is being installed to support nuclear reactor operations at the Experimental Breeder Reactor II.

  2. The Angra Project: Monitoring Nuclear Reactors with Antineutrino Detectors

    SciTech Connect

    Anjos, J. C.; Barbosa, A. F.; Lima, H. P. Jr.; Bezerra, T. J. C.; Gonzalez, L. F. G.; Kemp, E.; Chimenti, P.; Leigui de Oliveira, M. A.; Lima, R. M.; Nunokawa, H.

    2010-03-30

    We present the status of the Angra Neutrino project, describing the development of an antineutrino detector aimed at monitoring nuclear reactor activity. The experiment will take place at the Brazilian nuclear power plant located in Angra dos Reis. The Angra II reactor, with 4 GW of thermal power, will be used as a source of antineutrinos. A water Cherenkov detector will be placed above ground in a commercial container outside the reactor containment, about 30 m from the reactor core. With a detector of one ton scale a few thousand antineutrino interactions per day are expected. We intend, in a first step, to use the measured neutrino event rate to monitor the on--off status and the thermal power delivered by the reactor. In addition to the safeguards issues the project will provide an alternative tool to have an independent measurement of the reactor power.

  3. A Precise Measurement of Reactor Antineutrino at RENO

    SciTech Connect

    Jang, J.S.

    2014-06-15

    RENO is the reactor experiment to measure the neutrino mixing angle θ{sub 1}3 by observing the disappearance of the reactor antineutrino. Antineutrinos from six reactors at Yonggwang Nuclear Power Plant in Korea, are detected and compared by two identical detectors located at 294 m and 1383 m, respectively, from the center of the reactor array. The far (near) detector observes 73 (780) electron antineutrino candidate events per day after background subtraction with the precise measurement of reactor antineutrino flux. In this paper, an updated result is presented about the energy spectra of antineutrino signals in RENO detectors. A precise measurement of reactor antineutrino flux is also presented in comparison with expectations.

  4. Slurry F-T reactor hydrodynamics and scale-up

    SciTech Connect

    Smith, D.N.; O'Dowd, W.; Ruether, J.A.; Stiegel, G.J.; Shah, Y.T.

    1984-01-01

    The Fischer-Tropsch (F-T) synthesis of hydrocarbons by the hydrogenation of carbon monoxide over a catalyst via the indirect liquefaction route has only been demonstrated on a commercial scale in fixed-bed and entrained-bed reactors. The slurry reactor has been found to offer important advantages, such as low H/sub 2//CO feed ratios and isothermal stability on the laboratory and pilot-plant scale. Recent findings concerning slurry bubble column hydrodynamics are presented, and their relevance to slurry F-T reactor design and scale-up are discussed. The need for optimization of a slurry F-T reactor is apparent owing in large part to the nonselective formation of products on the catalyst, and the influence of hydrodynamics on the overall conversion of reactants. In this regard, the important transport resistances in the formation of products, and the influence of reactor dimensions on reactor performance are discussed. 9 refs., 14 figs., 1 tab.

  5. Nuclear Reactor Physics

    NASA Astrophysics Data System (ADS)

    Stacey, Weston M.

    2001-02-01

    An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

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

    SciTech Connect

    Not Available

    1992-04-01

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

  7. Reliability assurance for regulation of advanced reactors

    SciTech Connect

    Fullwood, R.; Lofaro, R.; Samanta, P.

    1991-01-01

    The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. This paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics.

  8. Reliability assurance for regulation of advanced reactors

    SciTech Connect

    Fullwood, R.; Lofaro, R.; Samanta, P.

    1991-12-31

    The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. This paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics.

  9. Life extension approach to the reactor vessel of a nuclear production reactor: Revision 1

    SciTech Connect

    Sindelar, R.L.; Awadalla, N.G.; Baumann, N.P.; Mehta, H.S.

    1989-01-01

    The nuclear materials production reactors at the Savannah River Plant have been in service for over 35 years. All the primary components of the reactor system are readily accessible for repair or replacement as needed except for the reactor vessel and thermal shields. The reactor vessel of a Savannah River Plant reactor is a cylindrical tank approximately 16 feet in diameter and 14 feet high and is not pressurized except for a 5 psig helium blanket gas in addition to the hydrostatic head of the heavy water (D/sub 2/O) moderator. The vessels are made of American Iron and Steel Institute Type 304 stainless steel fabricated into cylindrical shells with four to six wrought plates per vessel, 1.27 cm (0.5-inches) thick. The shells were made up in flat in two half-sections for later rolling and welding. The vessel bottom section containing the moderator effluent nozzles was welded to the shell in a T-joint configuration. All joining was performed with multipass Metal Inert Gas (MIG) welding. The service life assessment of the reactor vessel addresses the corrosive effects of the D/sub 2/O moderator and the degradation of the vessel material properties through exposure of the vessel to neutron irradiation. Potential degradation mechanisms include radiation embrittlement, Intergranular Stress Corrosion Cracking; and Irradiation-Assisted Stress Corrosion Cracking. 15 refs., 11 figs.

  10. Completely automated nuclear reactors for long-term operation

    SciTech Connect

    Teller, E.; Ishikawa, M.; Wood, L.

    1996-01-01

    The authors discuss new types of nuclear fission reactors optimized for the generation of high-temperature heat for exceedingly safe, economic, and long-duration electricity production in large, long-lived central power stations. These reactors are quite different in design, implementation and operation from conventional light-water-cooled and -moderated reactors (LWRs) currently in widespread use, which were scaled-up from submarine nuclear propulsion reactors. They feature an inexpensive initial fuel loading which lasts the entire 30-year design life of the power-plant. The reactor contains a core comprised of a nuclear ignitor and a nuclear burn-wave propagating region comprised of natural thorium or uranium, a pressure shell for coolant transport purposes, and automatic emergency heat-dumping means to obviate concerns regarding loss-of-coolant accidents during the plant`s operational and post-operational life. These reactors are proposed to be situated in suitable environments at {approximately}100 meter depths underground, and their operation is completely automatic, with no moving parts and no human access during or after its operational lifetime, in order to avoid both error and misuse. The power plant`s heat engine and electrical generator subsystems are located above-ground.

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

    DOEpatents

    Hunsbedt, Anstein; Boardman, Charles E.

    1993-01-01

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

  12. HORIZONTAL BOILING REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1958-11-18

    Reactors of the boiling water type are described wherein water serves both as the moderator and coolant. The reactor system consists essentially of a horizontal pressure vessel divided into two compartments by a weir, a thermal neutronic reactor core having vertical coolant passages and designed to use water as a moderator-coolant posltioned in one compartment, means for removing live steam from the other compartment and means for conveying feed-water and water from the steam compartment to the reactor compartment. The system further includes auxiliary apparatus to utilize the steam for driving a turbine and returning the condensate to the feed-water inlet of the reactor. The entire system is designed so that the reactor is self-regulating and has self-limiting power and self-limiting pressure features.

  13. Modelling of biofilm reactors

    SciTech Connect

    Rodrigues, A.; Grasmick, A.; Elmaleh, S.

    1982-10-01

    Comprehensive models of biofilm reactors are developed. Model I assumes a zero-order reaction of a limiting substrate and a diffusional mass transport through the biofilm; in the diffusion-controlled regime the model is fully characterized by one parameter alpha. From this model the conversion of substrate or reactor efficiency can be calculated, for continuously stirred tank reactors (CSTRs) and plug flow reactors respectively, as follows: EA = )alpha(alpha + 2)) 1/2 - alpha; and Ep = (2 alpha) 1/2 - alpha/2: Validation of the model is tested for different experimental systems. Model II includes liquid film mass transfer resistance. The conversion gap between plug flow reactors and CSTRs is always lower than 25% and, as a first approximation, the biofilm reactor design does not then require accurate residence time distribution measurements. (Refs. 23).

  14. Hybrid reactors. [Fuel cycle

    SciTech Connect

    Moir, R.W.

    1980-09-09

    The rationale for hybrid fusion-fission reactors is the production of fissile fuel for fission reactors. A new class of reactor, the fission-suppressed hybrid promises unusually good safety features as well as the ability to support 25 light-water reactors of the same nuclear power rating, or even more high-conversion-ratio reactors such as the heavy-water type. One 4000-MW nuclear hybrid can produce 7200 kg of /sup 233/U per year. To obtain good economics, injector efficiency times plasma gain (eta/sub i/Q) should be greater than 2, the wall load should be greater than 1 MW.m/sup -2/, and the hybrid should cost less than 6 times the cost of a light-water reactor. Introduction rates for the fission-suppressed hybrid are usually rapid.

  15. Thermionic space reactors overview

    SciTech Connect

    Wetch, J.R.; Britt, E.J.; Fitzpatrick, G.O.; Rasor, N.S.

    1983-08-01

    The multi-national development of thermionic reactor systems is summarized in the context of the past general space nuclear reactor program and the recent renewed interest in space nuclear power. Comparison of various alternate reactor space power systems indicates that only the in-core thermionic reactor approach has the performance and growth potential required to provide the power levels potentially needed for shuttle-launchable systems by the year 2000 at reactor coolant and system temperatures that are near the current state-of-the-art. It is concluded that all shuttle- launchable high power space reactor systems require high-temperature, long-endurance nuclear fuels, and that high priority characterization and development of such fuels is essential to successfully realize power systems that can enable the space missions presently being considered.

  16. 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.

  17. The IRIS General Plant Arrangement

    SciTech Connect

    Robertson, J.; Love, J.; Morgan, R.; Conway, L.E.

    2002-07-01

    IRIS (International Reactor Innovative and Secure) is a light water cooled, 335 MWe power reactor which is being designed by an international consortium as part of the US DOE NERI Program. IRIS features an integral reactor vessel that contains all the major reactor coolant system components including the reactor core, the coolant pumps, the steam generators and the pressurizer. This integral design approach eliminates the large coolant loop piping, and thus eliminates large loss-of-coolant accidents (LOCAs) as well as the individual component pressure vessels and supports. In addition, IRIS is being designed with a long life core and enhanced safety to address the requirements defined by the US DOE for Generation IV reactors. Bechtel, with Westinghouse consultation, has performed a layout study of the IRIS plant and this paper will discuss the results of this design effort. (authors)

  18. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Dreffin, R.S.

    1959-12-15

    A control means for a nuclear reactor is described. Particularly a device extending into the active portion of the reactor consisting of two hollow elements coaxially disposed and forming a channel therebetween, the cross sectional area of the channel increasing from each extremity of the device towards the center thereof. An element of neutron absorbing material is slidably positionable within the inner hollow element and a fluid reactor poison is introduced into the channel defined by the two hollow elements.

  19. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Goett, J.J.

    1961-01-24

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

  20. Improved vortex reactor system

    DOEpatents

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

    1995-05-09

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

  1. THERMAL NEUTRONIC REACTOR

    DOEpatents

    Spinrad, B.I.

    1960-01-12

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

  2. FLOW SYSTEM FOR REACTOR

    DOEpatents

    Zinn, W.H.

    1963-06-11

    A reactor is designed with means for terminating the reaction when returning coolant is below a predetermined temperature. Coolant flowing from the reactor passes through a heat exchanger to a lower reservoir, and then circulates between the lower reservoir and an upper reservoir before being returned to the reactor. Means responsive to the temperature of the coolant in the return conduit terminate the chain reaction when the temperature reaches a predetermined minimum value. (AEC)

  3. Benchmarking NSP Reactors with CORETRAN-01

    SciTech Connect

    Hines, Donald D.; Grow, Rodney L.; Agee, Lance J

    2004-10-15

    As part of an overall verification and validation effort, the Electric Power Research Institute's (EPRIs) CORETRAN-01 has been benchmarked against Northern States Power's Prairie Island and Monticello reactors through 12 cycles of operation. The two Prairie Island reactors are Westinghouse 2-loop units with 121 asymmetric 14 x 14 lattice assemblies utilizing up to 8 wt% gadolinium while Monticello is a General Electric 484 bundle boiling water reactor. All reactor cases were executed in full core utilizing 24 axial nodes per assembly in the fuel with 1 additional reflector node above, below, and around the perimeter of the core. Cross-section sets used in this benchmark effort were generated by EPRI's CPM-3 as well as Studsvik's CASMO-3 and CASMO-4 to allow for separation of the lattice calculation effect from the nodal simulation method. These cases exercised the depletion-shuffle-depletion sequence through four cycles for each unit using plant data to follow actual operations. Flux map calculations were performed for comparison to corresponding measurement statepoints. Additionally, start-up physics testing cases were used to predict cycle physics parameters for comparison to existing plant methods and measurements.These benchmark results agreed well with both current analysis methods and plant measurements, indicating that CORETRAN-01 may be appropriate for steady-state physics calculations of both the Prairie Island and Monticello reactors. However, only the Prairie Island results are discussed in this paper since Monticello results were of similar quality and agreement. No attempt was made in this work to investigate CORETRAN-01 kinetics capability by analyzing plant transients, but these steady-state results form a good foundation for moving in that direction.

  4. HOMOGENEOUS NUCLEAR POWER REACTOR

    DOEpatents

    King, L.D.P.

    1959-09-01

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

  5. Nuclear plants - military hostages

    SciTech Connect

    Ramberg, B.

    1986-03-01

    Recent events suggest that nuclear reactors could make tempting military or terrorist targets. Despite the care with which most reactors are built, studies document their vulnerability to willful destruction through disruption of coolant mechanisms both inside and outside the containment building. In addition to reactors, such nuclear support facilities as fuel fabrication, reprocessing, and waste storage installations may be attractive military targets. A nuclear bomb which exploded in the vicinity of a reactor could increase its lethal effects by one-third. The implications of this is vulnerability for Middle East stability as well as to other volatile regions. The author suggests several avenues for controlling the dangers: international law, military and civil defense, facility siting, increasing plant safety, and the international management of nuclear energy. 21 references.

  6. 32. Photograph of a line drawing. 'REACTOR ARRANGEMENT, BUILDING 'D', ...

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

    32. Photograph of a line drawing. 'REACTOR ARRANGEMENT, BUILDING 'D', PLANT 'B'. Holston Ordnance Works, Tennessee Eastman Corporation. October 21, 1942. Delineator: Unknown. Drawing # 7651-1004.228. - Holston Army Ammunition Plant, RDX-and-Composition-B Manufacturing Line 9, Kingsport, Sullivan County, TN

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

    SciTech Connect

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

    2009-10-01

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

  8. Tritium leak triggers reactor shutdown in the US

    NASA Astrophysics Data System (ADS)

    Gwynne, Peter

    2010-04-01

    A US state has voted against renewing the operating licence for its only working nuclear reactor after a leak of tritium was found in the 38-year-old power plant. The decision in late February by Vermont's senate to close the 650 MW Vermont Yankee reactor has cast a shadow over the Obama administration's plans to encourage the construction of more nuclear power plants to meet the country's increasing electricity demands. The plant currently provides one-third of the state's electricity demands.

  9. Towards intrinsically safe light-water reactors

    SciTech Connect

    Hannerz, K

    1983-07-01

    Most of the present impediments to the rational use of the nuclear option have their roots in the reactor safety issue. The approach taken to satisfy the escalating safety concerns has resulted in excessively complex and expensive plant designs but has failed to create public confidence. This paper describes a new approach based on the principle of Process Inherent Ultimate Safety (PIUS). With the PIUS principle, ultimate safety is obtained by guaranteeing core integrity under all credible conditions. This is accomplished on the basis of the laws of gravity and thermohydraulics alone, interacting with the heat extraction process in an intact or damaged primary circuit, without recourse to engineered safety systems that may fail or dependence on error-prone human intervention. Application of the PIUS principle to the pressurized water reactor involves a substantial redesign of the reactor and primary system but builds on established PWR technology where long-term operation is needed for verification.

  10. Gas core reactors for coal gasification

    NASA Technical Reports Server (NTRS)

    Weinstein, H.

    1976-01-01

    The concept of using a gas core reactor to produce hydrogen directly from coal and water is presented. It is shown that the chemical equilibrium of the process is strongly in favor of the production of H2 and CO in the reactor cavity, indicating a 98% conversion of water and coal at only 1500 K. At lower temperatures in the moderator-reflector cooling channels the equilibrium strongly favors the conversion of CO and additional H2O to CO2 and H2. Furthermore, it is shown the H2 obtained per pound of carbon has 23% greater heating value than the carbon so that some nuclear energy is also fixed. Finally, a gas core reactor plant floating in the ocean is conceptualized which produces H2, fresh water and sea salts from coal.

  11. HYLIFE-II reactor chamber design refinements

    SciTech Connect

    House, P.A.

    1994-06-01

    Mechanical design features of the reactor chamber for the HYLIFE-II inertial confinement fusion power plant are presented. A combination of oscillating and steady, molten salt streams (Li{sub 2}BeF{sub 4}) are used for shielding and blast protection of the chamber walls. The system is designed for a 6 Hz repetition rate. Beam path clearing, between shots, is accomplished with the oscillating flow. The mechanism for generating the oscillating streams is described. A design configuration of the vessel wall allows adequate cooling and provides extra shielding to reduce thermal stresses to tolerable levels. The bottom portion of the reactor chamber is designed to minimize splash back of the high velocity (>12 m/s) salt streams and also recover up to half of the dynamic head. Cost estimates for a 1 GWe and 2 GWe reactor chamber are presented.

  12. Innovative energy production in fusion reactors

    NASA Astrophysics Data System (ADS)

    Iiyoshi, A.; Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Ohnishi, M.; Onozuka, M.; Uenosono, C.

    1993-10-01

    Concepts of innovative energy production in neutron-lean fusion reactors without having the conventional turbine-type generator are proposed for improving the plant efficiency. These concepts are: (1) traveling wave direct energy conversion of 14.7 MeV protons; (2) cusp type direct energy conversion of charged particles; (3) efficient use of radiation with semiconductor and supplying clean fuel in a form of hydrogen gas; and (4) direct energy conversion from deposited heat to electric power with semiconductor utilizing Nernst effect. The candidates of reactors such as a toroidal system and an open system are also studied for application of the new concepts. The study shows the above concepts for a commercial reactor are promising.

  13. NEUTRONIC REACTOR SYSTEM

    DOEpatents

    Treshow, M.

    1959-02-10

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

  14. Polymerization Reactor Engineering.

    ERIC Educational Resources Information Center

    Skaates, J. Michael

    1987-01-01

    Describes a polymerization reactor engineering course offered at Michigan Technological University which focuses on the design and operation of industrial polymerization reactors to achieve a desired degree of polymerization and molecular weight distribution. Provides a list of the course topics and assigned readings. (TW)

  15. The Integral Fast Reactor

    SciTech Connect

    Till, C.E.; Chang, Y.I. ); Lineberry, M.J. )

    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.

  16. Operating US power reactors

    SciTech Connect

    Silver, E.G.

    1982-07-01

    The operation of US power reactors during March and April 1982 is summarized. Events of special note are discussed in the text, and the operational performance of all licensed power reactors is presented. These data are taken from the monthly Operating Units Status Report prepared by the Nuclear Regulatory Commission (NRC).

  17. Compact torsatron reactors

    SciTech Connect

    Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

    1988-05-01

    Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R/sub 0/ = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R/sub 0/ approx. = 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs.

  18. REACTOR FUEL SCAVENGING MEANS

    DOEpatents

    Coffinberry, A.S.

    1962-04-10

    A process for removing fission products from reactor liquid fuel without interfering with the reactor's normal operation or causing a significant change in its fuel composition is described. The process consists of mixing a liquid scavenger alloy composed of about 44 at.% plutoniunm, 33 at.% lanthanum, and 23 at.% nickel or cobalt with a plutonium alloy reactor fuel containing about 3 at.% lanthanum; removing a portion of the fuel and scavenger alloy from the reactor core and replacing it with an equal amount of the fresh scavenger alloy; transferring the portion to a quiescent zone where the scavenger and the plutonium fuel form two distinct liquid layers with the fission products being dissolved in the lanthanum-rich scavenger layer; and the clean plutonium-rich fuel layer being returned to the reactor core. (AEC)

  19. Status of French reactors

    SciTech Connect

    Ballagny, A.

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  20. Reactor neutrino monitoring

    NASA Astrophysics Data System (ADS)

    Lhuillier, D.

    2009-03-01

    Nuclear reactors are the most intense man-controlled sources of antineutrinos and as such have hosted number of key physics experiments, from the antineutrino discovery to modern oscillation measurements. At the present time, both detection technology and understanding of fundamental physics are mature enough to think about antineutrinos as a new tool for reactor monitoring. We describe below how antineutrinos can provide online information on reactor operation and amount of plutonium accumulated in the core. Reactors are the only sources of plutonium on earth and this element can be chemically separated from the rest of the nuclear fuel and diverted into nuclear weapons. We present in the next sections the unique features antineutrino detectors could provide to safeguards agencies such as IAEA. We review the worldwide efforts to develop small ( 1m scale) antineutrino detectors dedicated to automated and non-intrusive reactor monitoring.

  1. Thermionic switched self-actuating reactor shutdown system

    DOEpatents

    Barrus, Donald M.; Shires, Charles D.; Brummond, William A.

    1989-01-01

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

  2. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    NASA Technical Reports Server (NTRS)

    Clement, J. D.; Rust, J. H.

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  3. Human Factors Aspects of Operating Small Reactors

    SciTech Connect

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

    2010-11-07

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

  4. Risk Management for Sodium Fast Reactors.

    SciTech Connect

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

    2015-01-01

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

  5. Simulation of a marine nuclear reactor

    SciTech Connect

    Kusunoki, Tsuyoshi; Kyouya, Masahiko; Kobayashi, Hideo; Ochiai, Masaaki

    1995-02-01

    A Nuclear-powered ship Engineering Simulation SYstem (NESSY) has been developed by the Japan Atomic Energy Research Institute as an advanced design tool for research and development of future marine reactors. A marine reactor must respond to changing loads and to the ship`s motions because of the ship`s maneuvering and its presence in a marine environment. The NESSY has combined programs for the reactor plant behavior calculations and the ship`s motion calculations. Thus, it can simulate reactor power fluctuations caused by changing loads and the ship`s motions. It can also simulate the behavior of water in the pressurizer and steam generators. This water sloshes in response to the ship`s motions. The performance of NESSY has been verified by comparing the simulation calculations with the measured data obtained by experiments performed using the nuclear ship Mutsu. The effects of changing loads and the ship`s motions on the reactor behavior can be accurately simulated by NESSY.

  6. Filamentous bacteria existence in aerobic granular reactors.

    PubMed

    Figueroa, M; Val del Río, A; Campos, J L; Méndez, R; Mosquera-Corral, A

    2015-05-01

    Filamentous bacteria are associated to biomass settling problems in wastewater treatment plants. In systems based on aerobic granular biomass they have been proposed to contribute to the initial biomass aggregation process. However, their development on mature aerobic granular systems has not been sufficiently studied. In the present research work, filamentous bacteria were studied for the first time after long-term operation (up to 300 days) of aerobic granular systems. Chloroflexi and Sphaerotilus natans have been observed in a reactor fed with synthetic wastewater. These filamentous bacteria could only come from the inoculated sludge. Thiothrix and Chloroflexi bacteria were observed in aerobic granular biomass treating wastewater from a fish canning industry. Meganema perideroedes was detected in a reactor treating wastewater from a plant processing marine products. As a conclusion, the source of filamentous bacteria in these mature aerobic granular systems fed with industrial effluents was the incoming wastewater.

  7. Structural materials issues for the next generation fission reactors

    NASA Astrophysics Data System (ADS)

    Chant, I.; Murty, K. L.

    2010-09-01

    Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

  8. Nuclear reactor control column

    DOEpatents

    Bachovchin, Dennis M.

    1982-01-01

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

  9. 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.

  10. 76 FR 77025 - Office of New Reactors; Notice of Availability of the Final Staff Guidance Section 1.0, Revision...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-12-09

    ... Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition,'' Section 1.0, Revision 2 on... Advanced Reactors and Rulemaking, Office of New Reactors, U.S. Nuclear Regulatory Commission, Washington... Nuclear Regulatory Commission. Amy E. Cubbage, Chief, Policy Branch, Division of Advanced Reactors...

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

    SciTech Connect

    Monti, S.; Toti, A.

    2013-07-01

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

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

    SciTech Connect

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

    2011-05-01

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

  13. Development of the cascade inertial-confinement-fusion reactor

    SciTech Connect

    Pitts, J.H.

    1985-04-15

    Cascade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670/sup 0/K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis and experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.

  14. Development of the cascade inertial-confinement-fusion reactor

    SciTech Connect

    Pitts, J.H.

    1985-07-01

    Caqscade, originally conceived as a football-shaped, steel-walled reactor containing a Li/sub 2/O granule blanket, is now envisaged as a double-cone-shaped reactor containing a two-layered (three-zone) flowing blanket of BeO and LiAlO/sub 2/ granules. Average blanket exit temperature is 1670 K and gross plant efficiency (net thermal conversion efficiency) using a Brayton cycle is 55%. The reactor has a low-activation SiC-tiled wall. It rotates at 50 rpm, and the granules are transported to the top of the heat exchanger using their peripheral speed; no conveyors or lifts are required. The granules return to the reactor by gravity. After considerable analysis and experimentation, we continue to regard Cascade as a promising reactor concept with the advantages of safety, efficiency, and low activation.

  15. Shielding considerations for advanced space nuclear reactor systems

    SciTech Connect

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO/sub 2/) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications.

  16. Introduction and overall description of nuclear power plant. Volume I

    SciTech Connect

    Not Available

    1986-01-01

    Topics covered in this volume include content and purpose of booklets; how to study; producing electricity; the fossil fuel power plant; the nuclear power plant; the nuclear reactor; generating steam in a nuclear power plant; using the steam in a nuclear power plant; nuclear power station facilities; and special features of nuclear power plants.

  17. Some views on nuclear reactor safety

    SciTech Connect

    Tanguy, P.Y.

    1995-04-01

    This document is the text of a speech given by Pierre Y. Tanguy (Electricite de France) at the 22nd Water Reactor Safety Meeting held in Bethesda, MD in 1994. He describes the EDF nuclear program in broad terms and proceeds to discuss operational safety results with EDF plants. The speaker also outlines actions to enhance safety planned for the future, and he briefly mentions French cooperation with the Chinese on the Daya Bay project.

  18. Nuclear reactor reflector

    DOEpatents

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

    1994-01-01

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

  19. 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.

  20. Microfluidic electrochemical reactors

    DOEpatents

    Nuzzo, Ralph G.; Mitrovski, Svetlana M.

    2011-03-22

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

  1. CONTROL FOR NEUTRONIC REACTOR

    DOEpatents

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

    1959-03-31

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

  2. REACTOR CONTROL SYSTEM

    DOEpatents

    MacNeill, J.H.; Estabrook, J.Y.

    1960-05-10

    A reactor control system including a continuous tape passing through a first coolant passageway, over idler rollers, back through another parallel passageway, and over motor-driven rollers is described. Discrete portions of fuel or poison are carried on two opposed active sections of the tape. Driving the tape in forward or reverse directions causes both active sections to be simultaneously inserted or withdrawn uniformly, tending to maintain a more uniform flux within the reactor. The system is particularly useful in mobile reactors, where reduced inertial resistance to control rod movement is important.

  3. COOLED NEUTRONIC REACTOR

    DOEpatents

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

    1958-03-18

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

  4. NUCLEAR REACTOR FUEL SYSTEMS

    DOEpatents

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

    1959-09-15

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

  5. Spherical torus fusion reactor

    DOEpatents

    Martin Peng, Y.K.M.

    1985-10-03

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

  6. Fast Breeder Reactor studies

    SciTech Connect

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

    1980-07-01

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

  7. NEUTRONIC REACTOR CONTROL

    DOEpatents

    Metcalf, H.E.

    1958-10-14

    Methods of controlling reactors are presented. Specifically, a plurality of neutron absorber members are adjustably disposed in the reactor core at different distances from the center thereof. The absorber members extend into the core from opposite faces thereof and are operated by motive means coupled in a manner to simultaneously withdraw at least one of the absorber members while inserting one of the other absorber members. This feature effects fine control of the neutron reproduction ratio by varying the total volume of the reactor effective in developing the neutronic reaction.

  8. Design and simulation of a plant control system for a GCFR demonstration plant

    SciTech Connect

    Estrine, E.A.; Greiner, H.G.

    1980-02-01

    A plant control system is being designed for a 300 MW(e) Gas Cooled Fast Breeder Reactor (GCFR) demonstration plant. Control analysis is being performed as an integral part of the plant design process to ensure that control requirements are satisfied as the plant design evolves. Plant models and simulations are being developed to generate information necessary to further define control system requirements for subsequent plant design iterations.

  9. Hot zero power reactor calculations using the Insilico code

    NASA Astrophysics Data System (ADS)

    Hamilton, Steven P.; Evans, Thomas M.; Davidson, Gregory G.; Johnson, Seth R.; Pandya, Tara M.; Godfrey, Andrew T.

    2016-06-01

    In this paper we describe the reactor physics simulation capabilities of the Insilico code. A description of the various capabilities of the code is provided, including detailed discussion of the geometry, meshing, cross section processing, and neutron transport options. Numerical results demonstrate that Insilico using an SPN solver with pin-homogenized cross section generation is capable of delivering highly accurate full-core simulation of various pressurized water reactor problems. Comparison to both Monte Carlo calculations and measured plant data is provided.

  10. 61. VIEW LOOKING NORTHWEST AT A SIGNAL REACTOR OR CHOKE ...

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

    61. VIEW LOOKING NORTHWEST AT A SIGNAL REACTOR OR CHOKE COIL. WITHIN THE PROTECTIVE ENCLOSURE IS AN AIR AND PORCELAIN INSULATED COIL OF 5/8' DIAMETER STRANDED COPPER WIRE. REACTOR COILS WERE PLACED IN SERIES WITH EACH LEG OF THREE PHASE GENERATORS. THEIR FUNCTION WAS TO MODERATE SURGES OF CURRENT CAUSED BY LIGHTNING STRIKES, OPEN OR SHORT CIRCUIT PROBLEMS ON THE LINE. - New York, New Haven & Hartford Railroad, Cos Cob Power Plant, Sound Shore Drive, Greenwich, Fairfield County, CT

  11. Reactor pressure vessel. Status report

    SciTech Connect

    Elliot, B.J.; Hackett, E.M.; Lee, A.D.

    1996-10-01

    This report describes the issues raised as a result of the staffs review of Generic Letter (GL) 92-01, Revision 1, responses and plant-specific reactor pressure vessel (RPV) assessments and the actions taken or work in progress to address these issues. In addition, the report describes actions taken by the staff and the nuclear industry to develop a thermal annealing process for use at U.S. commercial nuclear power plants. This process is intended to be used as a means of mitigating the effects of neutron radiation on the fracture toughness of RPV materials. The Nuclear Regulatory Commission (NRC) issued GL 92-01, Revision 1, Supplement 1, to obtain information needed to assess compliance with regulatory requirements and licensee commitments regarding RPV integrity. GL 92-01, Revision 1, Supplement 1, was issued as a result of generic issues that were raised in the NRC staff`s reviews of licensee responses to GL 92-01, Revision 1, and plant-specific RPV evaluations. In particular, an integrated review of all data submitted in response to GL 92-01, Revision 1, indicated that licensees may not have considered all relevant data in their RPV assessments. This report is representative of submittals to and evaluations by the staff as of September 30, 1996. An update of this report will be issued at a later date.

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

    SciTech Connect

    Loflin, Leonard; McRimmon, Beth

    2014-12-18

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

  13. Role of nuclear reactors in future military satellites

    SciTech Connect

    Buden, D.; Angelo, J.A. Jr.

    1982-01-01

    Future military capabilities will be profoundly influenced by emerging Shuttle Era space technology. Regardless of the specific direction or content of tomorrow's military space program, it is clear that advanced space transportation systems, orbital support facilities, and large-capacity power subsystems will be needed to create the generally larger, more sophisticated military space systems of the future. This paper explores the critical role that space nuclear reactors should play in America's future space program and reviews the current state of nuclear reactor power plant technology. Space nuclear reactor technologies have the potential of satisfying power requirements ranging from 10 kW/sub (e)/ to 100 MW/sub (e)/.

  14. Advanced Burner Reactor Preliminary NEPA Data Study.

    SciTech Connect

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

    2007-10-15

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

  15. Packed Bed Reactor Experiment

    NASA Video Gallery

    The purpose of the Packed Bed Reactor Experiment in low gravity is to determine how a mixture of gas and liquid flows through a packed bed in reduced gravity. A packed bed consists of a metal pipe ...

  16. NEUTRONIC REACTOR STRUCTURE

    DOEpatents

    Daniels, F.

    1961-10-24

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

  17. NEUTRONIC REACTOR FUEL COMPOSITION

    DOEpatents

    Thurber, W.C.

    1961-01-10

    Uranium-aluminum alloys in which boron is homogeneously dispersed by adding it as a nickel boride are described. These compositions have particular utility as fuels for neutronic reactors, boron being present as a burnable poison.

  18. Laser-driven fusion reactor

    DOEpatents

    Hedstrom, J.C.

    1973-10-01

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

  19. 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.

  20. Nuclear Plant Data Bank

    SciTech Connect

    Booker, C.P.; Turner, M.R.; Spore, J.W.

    1986-01-01

    The Nuclear Plant Data Bank (NPDB) is being developed at the Los Alamos National Laboratory to assist analysts in the rapid and accurate creation of input decks for reactor transient analysis. The NPDB will reduce the time and cost of the creation or modification of a typical input deck. This data bank will be an invaluable tool in the timely investigation of recent and ongoing nuclear reactor safety analysis. This paper discusses the status and plans for the NPDB development and describes its anticipated structure and capabilities.

  1. Molten metal reactors

    DOEpatents

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

    2013-11-05

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

  2. Compact power reactor

    DOEpatents

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

    1978-01-01

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

  3. Future reactor experiments

    SciTech Connect

    Wen, Liangjian

    2015-07-15

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

  4. Plant control impact on IFR power plant passive safety response

    SciTech Connect

    Vilim, R.B.

    1993-03-01

    A method is described for optimizing the closed-loop plant control strategy with respect to safety margins sustained in the unprotected upset response of a liquid metal reactor. The optimization is performed subject to the normal requirements for reactor startup, load change and compensation for reactivity changes over the cycle. The method provides a formal approach to the process of exploiting the innate self-regulating property of a metal fueled reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error.

  5. Plant control impact on IFR power plant passive safety response

    SciTech Connect

    Vilim, R.B.

    1993-01-01

    A method is described for optimizing the closed-loop plant control strategy with respect to safety margins sustained in the unprotected upset response of a liquid metal reactor. The optimization is performed subject to the normal requirements for reactor startup, load change and compensation for reactivity changes over the cycle. The method provides a formal approach to the process of exploiting the innate self-regulating property of a metal fueled reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error.

  6. 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.

  7. 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."

  8. F Reactor Inspection

    ScienceCinema

    Grindstaff, Keith; Hathaway, Boyd; Wilson, Mike

    2016-07-12

    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."

  9. 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.

  10. Design of virtual SCADA simulation system for pressurized water reactor

    NASA Astrophysics Data System (ADS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  11. NEUTRONIC REACTOR CONSTRUCTION AND OPERATION

    DOEpatents

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

    1960-03-15

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

  12. 75 FR 6224 - Northern States Power Company of Minnesota; Monticello Nuclear Generating Plant Environmental...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-08

    ... COMMISSION Northern States Power Company of Minnesota; Monticello Nuclear Generating Plant Environmental... Northern States Power Company of Minnesota (NSPM) for operation of Monticello Nuclear Generating Plant... Project Manager, Plant Licensing Branch III-1, Division of Operating Reactor Licensing, Office of...

  13. STEAM PLANT, TRA609. STANDBY DIESEL GENERATOR. INL NEGATIVE NO. 3589. ...

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

    STEAM PLANT, TRA-609. STANDBY DIESEL GENERATOR. INL NEGATIVE NO. 3589. Unknown Photographer, 10/29/1951 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  14. STEAM PLANT, TRA609. BOILERS AFTER INSTALLATION. CAMERA FACING WESTERLY. INL ...

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

    STEAM PLANT, TRA-609. BOILERS AFTER INSTALLATION. CAMERA FACING WESTERLY. INL NEGATIVE NO. 4131. Unknown Photographer, 1/29/1952 - Idaho National Engineering Laboratory, Test Reactor Area, Materials & Engineering Test Reactors, Scoville, Butte County, ID

  15. 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.

  16. Measurements of the reactor neutron power in absolute units

    SciTech Connect

    Lebedev, G. V.

    2015-12-15

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

  17. 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.

  18. Reactor Dosimetry State of the Art 2008

    NASA Astrophysics Data System (ADS)

    Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

    2009-08-01

    . Williams, A. P. Ribaric and T. Schnauber. Agile high-fidelity MCNP model development techniques for rapid mechanical design iteration / J. A. Kulesza.Extension of Raptor-M3G to r-8-z geometry for use in reactor dosimetry applications / M. A. Hunter, G. Longoni and S. L. Anderson. In vessel exposure distributions evaluated with MCNP5 for Atucha II / J. M. Longhino, H. Blaumann and G. Zamonsky. Atucha I nuclear power plant azimutal ex-vessel flux profile evaluation / J. M. Longhino ... [et al.]. UFTR thermal column characterization and redesign for maximized thermal flux / C. Polit and A. Haghighat. Activation counter using liquid light-guide for dosimetry of neutron burst / M. Hayashi ... [et al.]. Control rod reactivity curves for the annular core research reactor / K. R. DePriest ... [et al.]. Specification of irradiation conditions in VVER-440 surveillance positions / V. Kochkin ... [et al.]. Simulations of Mg-Ar ionisation and TE-TE ionisation chambers with MCNPX in a straightforward gamma and beta irradiation field / S. Nievaart ... [et al.]. The change of austenitic stainless steel elements content in the inner parts of VVER-440 reactor during operation / V. Smutný, J. Hep and P. Novosad. Fast neutron environmental spectrometry using disk activation / G. Lövestam ... [et al.]. Optimization of the neutron activation detector location scheme for VVER-lOOO ex-vessel dosimetry / V. N. Bukanov ... [et al.]. Irradiation conditions for surveillance specimens located into plane containers installed in the WWER-lOOO reactor of unit 2 of the South-Ukrainian NPP / O. V. Grytsenko. V. N. Bukanov and S. M. Pugach. Conformity between LRO mock-ups and VVERS NPP RPV neutron flux attenuation / S. Belousov. Kr. Ilieva and D. Kirilova. FLUOLE: a new relevant experiment for PWR pressure vessel surveillance / D. Beretz ... [et al.]. Transport of neutrons and photons through the iron and water layers / M. J. Kost'ál ... [et al.]. Condition evaluation of spent nuclear fuel assemblies

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

    SciTech Connect

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

    2009-10-01

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

  20. ATOMIC POWER PLANT

    DOEpatents

    Daniels, F.

    1957-11-01

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