Science.gov

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, 2006

    SciTech Connect

    Agnihotri, Newal

    2006-09-15

    The focus of the September-October issue is on plant maintenance and advanced reactors. Major articles/reports in this issue include: Advanced plants to meet rising expectations, by John Cleveland, International Atomic Energy Agency, Vienna; A flexible and economic small reactor, by Mario D. Carelli and Bojan Petrovic, Westinghouse Electric Company; A simple and passively safe reactor, by Yury N. Kuznetsov, Research and Development Institute of Power Engineering (NIKIET), Russia; Gas-cooled reactors, by Jeffrey S. Merrifield, U.S. Nuclear Regulatory Commission; ISI project managment in the PRC, by Chen Chanbing, RINPO, China; and, Fort Calhoun refurbishment, by Sudesh Cambhir, Omaha Public Power District.

  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. BN-800 advanced nuclear power plant with fast reactor

    SciTech Connect

    Shishkin, A.N.; Kuzavkov, N.G.; Sobolev, V.A.; Shestakov, G.V.; Bagdasarov, Yu.E.; Kochetkov, L.A.; Matveyev, V.I.; Poplavsky, V.M.

    1993-12-31

    Bn-800 reactor plant with fast reactor and sodium coolant in the primary and secondary circuits is designed for operation as part of the power units in the Yuzhno-Uralskaya nuclear power plant scheduled to be constructed in Chelyabinsk region and as part unit 4 in the Beloyarskaya nuclear power plant. Reactor operations are described.

  6. 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.

  7. 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.

  8. 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.

  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

    Energy Science and Technology Software Center (ESTSC)

    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. Development of a Heavy Water Detritiation Plant for PIK Reactor

    SciTech Connect

    Alekseev, I.A.; Bondarenko, S.D.; Fedorchenko, O.A.; Konoplev, K.A.; Vasyanina, T.V.; Arkhipov, E.A.; Uborsky, V.V

    2005-07-15

    The research reactor PIK should be supplied with a Detritiation Plant (DP) to remove tritium from heavy water in order to reduce operator radiation dose and tritium emissions. The original design of the reactor PIK Detritiation Plant was completed several years ago. A number of investigations have been made to obtain data for the DP design. Nowadays the design of the DP is being revised on a basis of our investigations. The Combined Electrolysis and Catalytic Exchange (CECE) process will be used at the Detritiation Plant instead of Vapor Phase Catalytic Exchange. The experimental industrial plant for hydrogen isotope separation on the basis of the CECE process is under operation in Petersburg Nuclear Physics Institute. The plant was updated to provide a means for heavy water detritiation. Very high detritiation factors have been achieved in the plant. The use of the CECE process will allow the development of a more compact and less expensive detritiation plant for heavy water reactor PIK.

  13. Clinch River Breeder Reactor Plant Project: construction schedule

    SciTech Connect

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

    1982-01-01

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

  14. Liquid metal reactor/Pressurized water reactor plant comparison study

    SciTech Connect

    Halverson, T.G.

    1986-01-01

    The selection between alternative electric power generating technologies is mainly based on their overall economics. Capital costs account for over 60% of the total busbar cost of nuclear plants. Estimates reported in the literature have shown capital cost ratios of LMRs to PWRs ranging from less than 1 to as high as 1.8. To reduce this range of uncertainty, the study selected a method for cataloging plant hardware and then performed comparisons using engineering judgment as to the anticipated and reasonable cost differences. The paper summarizes the resulting one-on-one comparisons of components, systems, and buildings and identifies the LMR-PWR similarities and differences which influence costs. The study leads to the conclusion that the capital cost of the most up-to-date large LMR design would be very close to that of the latest PWRs.

  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. Medium Power Lead Alloy Fast Reactor Balance of Plant Options

    SciTech Connect

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

    2004-09-01

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

  17. 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.

  18. 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.

  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. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 75 FR 7633 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the Subcommittee on Plant Operations...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-02-22

    ...: February 12, 2010. Antonio F. Dias, Chief, Reactor Safety Branch B, Advisory Committee on Reactor... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the Subcommittee on Plant Operations... October 14, 2009 (74 FR 58268-58269). ] Detailed meeting agendas and meeting transcripts are available...

  6. Fabrication of reactor shields at the Y-12 plant

    NASA Astrophysics Data System (ADS)

    Asbury, W. L.; Kosinski, F. E.; Royer, L. T.

    The Oak Ridge Y-12 Plant is the lead DOE facility for the large scale synthesis, manufacture, and fabrication of lithium hydride/deuteride (LiH/D) materials. Fabrication of large crack free LiH/D parts of uniform density and strength is best achieved by isostatic pressing, using powder metallurgy methods. The highly reactive and moisture sensitive nature of LiH/D has required a number of unique facilities and processing methods at Y-12 to process LiH/D in a safe manner and to maintain high quality products. The history of the fabrication of neutron shields for the SNAP 10A reactor is discussed. Powder metallurgical techniques were used to fabricate that shielding system at Y-12, and the units were tested for shielding efficiency at the Tower Shielding Facility at ORNL.

  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. 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.

  9. 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).

  10. 75 FR 58447 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-09-24

    ... published in the Federal Register on October 14, 2009, (74 FR 58268-58269). Detailed meeting agendas and... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant License... Branch A, Advisory Committee on Reactor Safeguards. BILLING CODE 7590-01-P...

  11. 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.

  12. 75 FR 82093 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2010-12-29

    ... 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 Renewal; Notice of Meeting The ACRS Subcommittee on Plant License Renewal will hold a meeting on January 12, 2011, Room T-2B1, 11545 Rockville...

  13. 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.

  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. 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.

  16. 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.

  17. 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.

  18. Helium circulator design considerations for modular high temperature gas-cooled reactor plant

    SciTech Connect

    McDonald, C.F.; Nichols, M.K.

    1986-12-01

    Efforts are in progress to develop a standard modular high temperature gas-cooled reactor (MHTGR) plant that is amenable to design certification and serial production. The MHTGR reference design, based on a steam cycle power conversion system, utilizes a 350 MW(t) annular reactor core with prismatic fuel elements. Flexibility in power rating is afforded by utilizing a multiplicity of the standard module. The circulator, which is an electric motor-driven helium compressor, is a key component in the primary system of the nuclear plant, since it facilitates thermal energy transfer from the reactor core to the steam generator; and, hence, to the external turbo-generator set. This paper highlights the helium circulator design considerations for the reference MHTGR plant and includes a discussion on the major features of the turbomachine concept, operational characteristics, and the technology base that exists in the US.

  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. 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... COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Future Plant Designs..., 2010. Antonio F. Dias, Chief, Reactor Safety Branch B, Advisory Committee on Reactor...

  1. 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).

  2. 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).

  3. 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.

  4. Remote reactor vessel inspections at Savannah River Plant

    SciTech Connect

    Lewis, W.I. III; Goodwin, B.W.; French, T.J.

    1988-01-01

    The Equipment Engineering Division of the Savannah River Laboratory has recently completed the development of a robotic inspection system for use in conducting full volumetric examinations of the site's reactor tanks. The development of the inspection equipment was necessary to obtain quantitative volumetric data on the structural integrity of the tanks such that service life could be evaluated. The development program included the design and fabrication of four, five-axis robotic inspection tools, six three-axis camera and lighting masts, video equipment, mobile control centers, complete mockup and testing facilities, and the integration of ultrasonic and eddy current data collection equipment. This paper gives a brief history of past reactor vessel inspections. The need, justification, and implementation of the development effort is then discussed. The various system components developed including the tools, controls, data acquisition systems, etc. are discussed along with the unique design constraints each had to meet and how each supports the overall inspection system. Implementation of the inspection system in the field and future development acres to increase the equipments versatility are also described. 2 refs., 9 figs.

  5. 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.

  6. 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...

  7. 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,…

  8. 78 FR 27442 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-10

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Plant License... October 18, 2012, (77 FR 64146- 64147). Detailed meeting agendas and meeting transcripts are available on..., Rockville, MD. After registering with security, please contact Mr. Theron Brown (Telephone 240-888-9835)...

  9. 78 FR 12800 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-02-25

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Plant License... participation in ACRS meetings were published in the Federal Register on October 18, 2012, (77 FR 64146- 64147... White Flint North building, 11555 Rockville Pike, Rockville, MD. After registering with security,...

  10. 76 FR 63329 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-10-12

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant License... October 21, 2010, (75 FR 65038-65039). Detailed meeting agendas and meeting transcripts are available on..., Rockville, MD. After registering with security, please contact Mr. Theron Brown (240-888-9835) to...

  11. 78 FR 31988 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-05-28

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant... Register on October 18, 2012, (77 FR 64146- 64147). Detailed meeting agendas and meeting transcripts are... Rockville Pike, Rockville, MD. After registering with security, please contact Mr. Theron Brown...

  12. Winding insulation in electromagnetic systems for Tokamak reactor plants

    NASA Astrophysics Data System (ADS)

    Maslov, V. V.; Trubachev, S. G.

    1985-01-01

    Magnetic containment of the plasma in nuclear fusion reactors of the Tokamak type requires electromagnets with insulation which must withstand high temperatures and thermal shocks as well as ionizing radiation in various forms and electric fields, and mechanical loads. Insulation materials to ensure adequate thermophysical and mechanical properties are evaluated, followed by design of insulation systems with satisfactory performance characteristics. Data on neutron fluence energy characteristics and radiation absorption doses during neutron interactions are essential for such an evaluation. Materials considered for insulation in electromagnets with superconductor and cryoresistance windings are glass mica tape with epoxy compound impregnation, glass cloth with epoxy compound impregnation (STE), polyimide glass cloth with adhesive coating (LSNL), glass Textolite with epoxy phenolic binder (STEN), epoxy resin paste with mineral fillers (PE), and polyurethane compound modified by epoxy resin with mineral filler (KPU).

  13. Containment vs confinement trade study, small HTGR plant PCRV [prestressed concrete reactor vessel] concept

    SciTech Connect

    1985-03-01

    This trade study has been conducted to evaluate the differences between four different HTGR nuclear power plants. All of the plants use a prestressed concrete reactor vessel (PCRV) to house the core and steam generation equipment. The reactor uses LEU U/Th fuel in prismatic carbon blocks. All plant concepts meet the utility/user requirements established for small HTGR plants. All plants will be evaluated with regard to their ability to produce safe, economical power to satisfy Goals 1, 2, and 3 of the HTGR program and by meeting the MUST criteria established in the concept evaluation plan. Capital costs for each plant were evaluated on a differential cost basis. These costs were developed according to the ``NUS`` code of accounts as defined in the Cost Estimating and Control Procedure, HP-20901. Accounts that were identical in scope for all four plants were not used for the comparison. Table 1-1 is a list of capital cost accounts that were evaluated for each plant.

  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. 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.

  16. 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

  17. Plant System Design of Supercritical CO{sub 2} Direct Cycle Gas Turbine Fast Reactor

    SciTech Connect

    Katsuhiro, Tozawa; Nobumasa, Tsuji; Yasushi, Muto; Yasuyoshi, Kato

    2006-07-01

    The conceptual plant design and preliminary safety analysis of SCDFR, Supercritical CO{sub 2} Direct Cycle Gas Turbine Fast Reactor, were performed. Plant thermal power is 600 MW. Core outlet/inlet pressure and temperature are 12.5/12.8 MPa and 527/388 deg C respectively. The core height and equivalent diameter are about 1.2 m and about 3.146 m respectively. The core can be burning for 10 years without refueling by adding 6.5% content of {sup 237}Np into the fuel as a burnable poison. Reactor pressure vessel height and inner diameter are about 19.3 m and about 6.55 m respectively. Steel containment vessel contains the reactor system and the gas turbine system. Preliminary analysis of core temperature behavior during the depressurization accident in SCDFR was performed. In the result of the analysis, core temperature is limited under 900 deg C, assumed limit temperature of the fuel clad, at the condition of minimum gas circulation flow rate of 2.0 m{sup 3}/s. On the other hand, gas circulator designed flow rate of the auxiliary core cooling system is over 11.6 m{sup 3}/s. These show that the integrity of the fuel clad during depressurization accident is maintained. We conclude that the plant concept of SCDFR is developed and the plant safety under depressurization accident conditions is confirmed by preliminary analysis. (authors)

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

    SciTech Connect

    S. Blaine Grover

    2009-09-01

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

  19. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    SciTech Connect

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J.; Seren, T.; Lipponen, M.; Kekki, T.

    2011-07-01

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  20. 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.

  1. 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

    2008-08-01

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

  2. 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.

  3. 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

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

    SciTech Connect

    Dostal, Vaclav; Hejzlar, Pavel; Todreas, Neil E.; Buongiorno, Jacopo

    2004-09-15

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

  5. 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

  6. 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

  7. Individual plant examination program: Perspectives on reactor safety and plant performance. Part 6, appendices A, B, and C

    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 (including internal flooding, but excluding internal fire). The U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, reviewed the WE 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. In particular, these results are assessed in relation to the design and operational characteristics of the various reactor and containment types, and by comparing the IPEs to probabilistic risk assessment characteristics. 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.

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

    DOE PAGESBeta

    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

  9. 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.

  10. 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)

  11. 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.

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

    SciTech Connect

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

    2003-05-15

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

  14. 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.

  15. 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.

  16. An integrated plant-life extension program for EBR-II (Experimental Breeder Reactor)

    SciTech Connect

    King, R.W.

    1986-01-01

    An integrated plant-life extension program is being developed and implemented at EBR-II. The program plan has five primary areas of focus, and is structured to take advantage of inherent features of the liquid-metal-cooled reactor that give it potential for very long life. The program is centered around development and increased use of computer-based software systems for surveillance, diagnostics, prognostics, data handling, and knowledge transfer. Even though the program is only partially implemented, benefits are already being realized in the form of increased understanding of plant system status and performance due to development of diagnostic data-handling software for manipulation of plant sensor data, and improved force monitoring and protection of the remotely operated fuel handling system. The eventual integration of the elements of the program is a key feature that is expected to enhance the overall effectiveness of the program.

  17. 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.

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

    SciTech Connect

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2002-07-01

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

  19. 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.

  20. 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.

  1. REACTORS

    DOEpatents

    Spitzer, L. Jr.

    1961-10-01

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

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

    SciTech Connect

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

    2012-07-01

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

  3. 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)

  4. Complete study of the pyrolysis and gasification of scrap tires in a pilot plant reactor.

    PubMed

    Conesa, Juan A; Martín-Gullón, I; Font, R; Jauhiainen, J

    2004-06-01

    The pyrolysis and gasification of tires was studied in a pilot plant reactor provided with a system for condensation of semivolatile matter. The study comprises experiments at 450, 750, and 1000 degrees C both in nitrogen and 10% oxygen atmospheres. Analysis of all the products obtained (gases, liquids, char, and soot) are presented. In the gas phase only methane and benzene yields increase with temperature until 1000 degrees C. In the liquids the main components are styrene, limonene, and isoprene. The solid fraction (including soot) increases with temperature. Zinc content of the char decreases with increasing temperature. PMID:15224754

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

    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.

  6. 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.

  7. Underground nuclear power plant with the VK-300 reactor as a substituting power source for Zheleznogorsk, Russia

    SciTech Connect

    Adamov, E.O.; Lebedev, V.A.; Kuznetsov, Yu.N.; Samarkin, A.A.; Tokarev, Yu.I.

    1996-07-01

    Zheleznogorsk is situated near the territorial center -- Krasnoyarsk on the Yenisei river. Mining and chemical complex is the main industrial enterprise of the town, which has been constructed for generation and used for isolation of weapons-grade plutonium. Heat supply to the chemical complex and town at the moment is largely provided by nuclear co-generation plant (NCGP) on the basis of the ADEh-2 dual-purpose reactor, generating 430 Gcal/h of heat and, partially, by coal backup peak-load boiler houses. NCGP also provides 73% of electric power consumed. In line with agreements between Russia and USA on strategic arms reduction and phasing out of weapons-grade plutonium production, decommissioning of the ADEh-2 reactor by 2000 is planned. Thus, a problem arises relative to compensation for electric and thermal power generation for the needs of the town and industrial enterprises, which is now supplied by the reactor. A nuclear power plant constructed on the same site as a substituting power source should be considered as the most practical option. Basic requirements to the reactor of substituting nuclear power plant are as follows. It is to be a new generation reactor on the basis of verified technologies, having an operating prototype optimal for underground siting and permitting utmost utilization of the available mining workings and those being disengaged. NCGP with the reactor is to be constructed in the time period required and is to become competitive with other possible power sources. Analysis has shown that the VK-300 simplified vessel-type boiling reactor meets the requirements made in the maximum extent. Its design is based on the experience of the VK-50 reactor operation for a period of 30 years in Dimitrovgrad (Russia) and allows for experience in the development of the SBWR type reactors. The design of the reactor is discussed.

  8. Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

    SciTech Connect

    Goldberg, A.; Streit, R.D.

    1981-05-01

    Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 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 and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads.

  9. Probability of pipe failure in the reactor coolant loops of Babcock and Wilcox PWR plants. Volume 1. Summary report

    SciTech Connect

    Holman, G.S.; Chou, C.K.

    1986-05-01

    As part of its reevaluation of the double-ended guillotine break (DEGB) of reactor coolant piping as a design basis event for nuclear power plants, the US Nuclear Regulatory Commission (NRC) contracted the Lawrence Livermore National Laboratory (LLNL) to estimate the probability of occurrence of a DEGB, and to assess the effect that earthquakes have on DEGB probability. This report describes an evaluation of reactor coolant loop piping in PWR plants having nuclear steam supply systems designed by Babcock and Wilcox. Two causes of pipe break were considered: pipe fracture due to the growth of cracks at welded joints (''direct'' DEGB), and pipe rupture indirectly caused by failure of heavy component supports due to an earthquake (''indirect'' DEGB). Unlike in earlier evaluations of Westinghouse and Combustion Engineering reactor coolant loop piping, in which the probability of direct DEGB had been explicitly estimated using a probabilistic fracture mechanics model, no detailed fracture mechanics calculations were performed. Instead, a comparison of relevant plant data, mainly reactor coolant loop stresses, for one representative B and W plant with equivalent information for Westinghouse and C-E systems inferred that the probability of direct DEGB should be similarly low (less than le-10 per reactor year). The probability of indirect DEGB, on the other hand, was explicitly estimated for two representative plants. The results of this study indicate that the probability of a DEGB form either cause is very low for reactor coolant loop piping in these specific plants and, because of similarity in design, infer that the probability of DEGB is generally very low in B and W reactor coolant loop piping. The NRC should therefore consider eliminating DEGB as a design basis event in favor of more realistic criteria. 13 refs., 9 tabs.

  10. Plutonium disposition in the BN-600 fast-neutron reactor at the Beloyarsk nuclear power plant

    NASA Astrophysics Data System (ADS)

    Moses, D. L.; Chebeskov, A. N.; Matveev, V. I.; Vasiliev, B. A.; Maltsev, V. V.

    In 1996, the United States and the Russian Federation completed an initial joint study that evaluated the candidate options for the disposition of surplus weapons-derived plutonium in both countries. While Russia advocates building new reactors for converting weapons-derived plutonium to spent fuel, the cost is high, and the continuing joint study of the Russian options is considering only the use of the existing VVER-1000 LWRs in Russia (and possibly in Ukraine) and the existing BN-600 fast-neutron reactor at the Beloyarsk Nuclear Power Plant in Russia. The BN-600 reactor, which currently uses enriched uranium fuel, is capable with certain design modifications of converting up to 1.3 metric tons (MT) of surplus weapons-derived plutonium to spent fuel each year. The steps needed to convert BN-600 to a plutonium-burner core will be discussed. The step involving the hybrid core allows an early and timely start that takes advantage of the limited capacity for fabricating uranium-plutonium mixed-oxide fuel early in the disposition program. The design lifetime of BN-600 must safely and reliably be extended by 10 yr to at least 2020 so that a sufficient amount of plutonium (˜20 MT) can be converted to spent fuel.

  11. 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)

  12. REACTOR

    DOEpatents

    Spitzer, L. Jr.

    1962-01-01

    The system conteraplates ohmically heating a gas to high temperatures such as are useful in thermonuclear reactors of the stellarator class. To this end the gas is ionized and an electric current is applied to the ionized gas ohmically to heat the gas while the ionized gas is confined to a central portion of a reaction chamber. Additionally, means are provided for pumping impurities from the gas and for further heating the gas. (AEC)

  13. 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.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

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

    SciTech Connect

    Louison, R.; Boardman, C.E.

    1981-05-29

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

  20. 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.

  1. 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

  2. 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.

  3. 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.

  4. A new impulse in the development of nuclear pool-type reactors for underground heating plant: Designing, running background and possible perspectives

    SciTech Connect

    Adamov, E.O.; Mikhan, V.I.; Romenkov, A.A.; Melnikov, N.N.; Konuhin, V.P.

    1996-07-01

    This paper considers the concept of energy supply with using ultimately safe pool-type integral nuclear reactors. Safety and reliability of these reactors has already been demonstrated to the public by the long-term operation of this type various research reactors. The reactor and power plant design features, new approach to the nuclear safety, the nuclear upgrading of existing energy system in a small Russian town are considered in the paper.

  5. 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..., (75 FR 65038-65039). Detailed meeting agendas and meeting transcripts are available on the NRC Web..., MD. After registering with security, please contact Mr. Theron Brown (Telephone 240-888-9835) to...

  6. 77 FR 36582 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Plant License...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2012-06-19

    ... COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Plant License... October 17, 2011, (76 FR 64126- 64127). Detailed meeting agendas and meeting transcripts are available on..., Rockville, MD. After registering with security, please contact Mr. Theron Brown (Telephone 240-888-9835)...

  7. 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.

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

    SciTech Connect

    Curtis Smith; Scott Beck; William Galyean

    2006-06-01

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

  9. 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.

  10. 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.

  11. 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.

  12. Remediation of dinitrotoluene contaminated soils from former ammunition plants: soil washing efficiency and effective process monitoring in bioslurry reactors.

    PubMed

    Zhang, C; Daprato, R C; Nishino, S F; Spain, J C; Hughes, J B

    2001-10-12

    A pilot-scale bioslurry system was used to test the treatment of soils highly contaminated with 2,4-dinitrotoluene (2,4-DNT) and 2,6-dinitrotoluene (2,6-DNT). The treatment scheme involved a soil-washing process followed by two sequential aerobic slurry reactors augmented with 2,4-DNT- and 2,6-DNT-mineralizing bacteria. Test soils were obtained from two former army ammunition plants, the Volunteer Army Ammunition Plant (VAAP, Chattanooga, TN) and the Badger Army Ammunition Plant (BAAP, Baraboo, WI). Soil washing was used to minimize operational problems in slurry reactors associated with large particulates. The Eimco slurry reactors were operated in a draw-and-fill mode for 3 months and were monitored for the biodegradation of 2,4-DNT and 2,6-DNT, nitrite production, NaOH consumption, and oxygen uptake rate. Results show that soil washing was very effective for the removal of sands and the recovery of soil fines containing 2,4-DNT and 2,6-DNT. Bioslurry reactors offered rapid and nearly complete degradation of both DNT isomers, but require real time monitoring to avoid long lag periods upon refeeding. Results found a significant discrepancy between the measured DNT concentrations and calculated DNT concentrations in the slurry reactors because of solids profiles in the slurry reactors and the presence of floating crystal of DNTs. Based on the actual amount of dinitrotoluene degradation, nitrite release, NaOH consumption, and oxygen uptake were close to the theoretical stoichiometric coefficients of complete DNT mineralization. Such stoichiometric relationships were not achieved if the calculation was based on the measured DNT concentrations due to the heterogeneity of DNT in the reactor. Results indicate that nitrite release, NaOH consumption, and oxygen uptake rates provide a fast assessment of 2,4-DNT degradation and microbial activity in a slurry reactor, but could not be extended to a second reactor in series where the degradation of a much lower concentration

  13. 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.

  14. 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.

  15. 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

  16. 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

  17. On the power and size of tokamak fusion pilot plants and reactors

    NASA Astrophysics Data System (ADS)

    Costley, A. E.; Hugill, J.; Buxton, P. F.

    2015-03-01

    It is generally accepted that the route to fusion power involves large devices of ITER scale or larger. However, we show, contrary to expectations, that for steady state tokamaks operating at fixed fractions of the density and beta limits, the fusion gain, Qfus, depends mainly on the absolute level of the fusion power and the energy confinement, and only weakly on the device size. Our investigations are carried out using a system code and also by analytical means. Further, we show that for the two qualitatively different global scalings that have been developed to fit the data contained in the ITER ELMy H-mode database, i.e. the normally used beta-dependent IPB98y2 scaling and the alternative beta-independent scalings, the power needed for high fusion performance differs substantially, typically by factors of three to four. Taken together, these two findings imply that lower power, smaller, and hence potentially lower cost, pilot plants and reactors than currently envisaged may be possible. The main parameters of a candidate low power (˜180 MW), high Qfus (˜5), relatively small (˜1.35 m major radius) device are given.

  18. 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)

  19. Electrochemical treatment of Procion Black 5B using cylindrical flow reactor--a pilot plant study.

    PubMed

    Raghu, S; Basha, C Ahmed

    2007-01-10

    The paper presents the results of an efficient electrochemical treatment of Procion Black 5B--a pilot plant study. Experiments were conducted at different current densities and selected electrolyte medium using Ti/RuO2 as anode, stainless-steel as cathode in a cylindrical flow reactor. By cyclic voltammetric analysis, the best condition for maximum redox reaction rate was found to be in NaCl medium. During the various stages of electrolysis, parameters such as COD, colour, FTIR, UV-vis spectra studies, energy consumption and mass transfer coefficient were computed and presented. The experimental results showed that the electrochemical oxidation process could effectively remove colour and the chemical oxygen demand (COD) from the synthetic dye effluent. The maximum COD reduction and colour removal efficiencies were 74.05% and 100%, respectively. Probable theory, reaction mechanism and modeling were proposed for the oxidation of dye effluent. The results obtained reveal the feasibilities of application of electrochemical treatment for the degradation of Procion Black 5B. PMID:17008006

  20. 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.

  1. Probability of pipe failure in the reactor coolant loops of Westinghouse PWR plants. Volume 4. Pipe failure induced by crack growth in west coast plants

    SciTech Connect

    Chinn, D.J.; Holman, G.S.; Lo, T.Y.; Mensing, R.W.

    1985-07-01

    The US Nuclear Regulatory Commission contracted with the Lawrence Livermore National Laboratory to conduct a study to determine if the probability of occurrence of a double-ended guillotine break (DEGB) in primary coolant piping warrants the current design requirements that safeguard against the effecs of such a break. This report assesses the reactor-coolant-loop piping system of west coast Westinghouse plants. The results indicate that directly induced DEGB is an unlikely event in the west coast Westinghouse plants. For the Trojan plant, leak is far more likely than a direct DEGB. Further, earthquakes have very little effect on the probabilities of leak and direct DEGB. At the Diablo Canyon plant, the increase in postulated seismic levels due to reevaluation of the site to account for the Hosgri Fault has caused directly induced DEGB failure probability to be dependent on earthquake occurrences. The resulting direct DEGB failure probability is still much lower than the indirect DEGB failure probability for Diablo Canyon.

  2. Verification of the Plant Dynamics Analytical Code CERES Using the Results of the Plant Trip Test of the Prototype Fast Breeder Reactor MONJU

    SciTech Connect

    Yoshihisa Nishi; Nobuyuki Ueda; Izumi Kinoshita; Akira Miyakawa; Mitsuya Kato

    2006-07-01

    CERES is plant system analysis code for LMRs (liquid metal cooled reactors) developed by the Central Research Institute of Electric Power Industry (CRIEPI). To verify the CERES code, analyses were performed by using the result of the plant trip test of the prototype FBR (fast breeder reactor) 'MONJU' at 40% rated power. The verification work was performed as a joint research of CRIEPI and JAEA (Japan Atomic Energy Agency). Following three verification analyses were performed mainly. (I) Analysis concerning the primary/ secondary/auxiliary cooling system (the plenum in the reactor vessel (R/V) was modeled in R-Z 2-dimension). (II) Analysis concerning the thermal-hydraulic characteristics in the plenum of R/V (the plenum was modeled in 3-dimension). (III) Analysis concerning the flow characteristics inside the intermediate heat exchanger (IHX) (the plenum in the IHX was modeled in 3-dimension). Analytical results by the CERES code showed good agreement with the results of the test of the 'MONJU'. Fundamental abilities of the CERES as a plant dynamics calculation code had been verified through these analyses. Additionally, some characteristic flows in plenums of 'MONJU' became clear by these analyses. (authors)

  3. 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.

  4. 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

  5. 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%.

  6. [Safety systems in nuclear power plants with pressurized water reactors and demonstration of their function in technical scale experiments. II].

    PubMed

    Hennies, H H

    1987-11-01

    Analyses and experiments carried out during the last decade on the sequence and consequences of accidents in German pressurized water reactors have shown that the functioning capability of the safety systems is guaranteed for the case of the MCA, the maximum credible accident. For the case of core meltdown, simulation experiments have also made it evident that the consequences remain largely restricted to the plant proper. PMID:3431585

  7. 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.

  8. Probability of pipe failure in the reactor coolant loops of Westinghouse PWR Plants. Volume 1. Summary report

    SciTech Connect

    Holman, G.S.; Chou, C.K.

    1985-07-01

    As part of its reevaluation of the double-ended guillotine break (DEGB) of reactor coolant loop piping as a design basis event for nuclear power plants, the US Nuclear Regulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to estimate the probability of occurrence probability. This report describes a probabilistic evaluation of reactor coolant loop piping in PWR plants having nuclear steam supply systems designed by Westinghouse. Two causes of pipe break were considered: pipe fracture due to the growth of cracks at welded joints (''direct'' DEGB), and pipe rupture indirectly caused by failure of component supports due to an earthquake (''indirect'' DEGB). The probability of direct DEGB was estimated using a probabilistic fracture mechanics model. The probability of indirect DEGB was estimated by estimating support fragility and then convolving fragility and seismic hazard. The results of this study indicate that the probability of a DEGB from either cause is very low for reactor coolant loop piping in these plants, and that NRC should therefore consider eliminating DEGB as a design basis event in favor of more realistic criteria. 17 refs., 15 figs., 11 tabs.

  9. 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

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. Environmental distribution and long-term dispersion of reactor 14CO2 around two German nuclear power plants.

    PubMed

    Levin, I; Kromer, B; Barabas, M; Münnich, K O

    1988-02-01

    Carbon-14 data on atmospheric CO2 as well as on plant material (tree leaves and wheat) from the vicinity of two German boiling water reactors (Philippsburg and Isar/Ohu) are reported. Atmospheric CO2 samples taken routinely with an integration time of one or two weeks 1.75 km downwind of the Philippsburg reactor (900 MW electrical power) show a maximum 14C excess concentration of delta 14C (excess) = 300 +/- 7%, corresponding to 12.7 mBq m-3 (STP air). The long-term average excess amounts to delta 14C (excess) = 47 +/- 3%, corresponding to 2.0 mBq m-3 (STP air). The concentrations observed with plant material at the same sampling site range between delta 14C (excess) = 0% and 120%, corresponding to 0 and 27 mBq (g carbon)-1. With the meteorological dispersion parameters actually measured at the nuclear power plants, the dispersion factors for the various sampling sites and for the individual periods of sampling were calculated on the basis of a one-dimensional Gaussian plume model. With the observed 14C excess concentrations and the dispersion factor, a "theoretical" (i.e. calculated) reactor 14C source strength is then determined. For the Philippsburg reactor, which is situated in the flat Rhine valley, the "theoretical" and the observed yearly mean 14C emissions compare rather well (within a factor of 2). A significant systematical deviation from the model was found in the concentration decrease with source distance: the decrease predicted between the 1.75-km and 3.25-km distances is steeper than actually observed. The 14C excess concentrations found in tree leaves around the Isar/Ohu reactor (907 MW electrical power) at 1-2 km distance fall into the same range as observed at Philippsburg. In the hilly terrain at this reactor site, the model calculations agree well with the observed values up to a distance of 1 km if the relative elevation of the sampling site is taken into account by introducing a "reduced stack height" in the model calculations. This method

  15. Probability of pipe failure in the reactor coolant loops of Combustion Engineering PWR plants. Volume 1. Summary report

    SciTech Connect

    Holman, G.S.; Lo, T.; Chou, C.K.

    1985-01-01

    As part of its reevaluation of the double-ended guillotine break (DEGB) as a design requirement for reactor coolant piping, the US Nuclear Regulatory Commission (NRC) contracted with the Lawrence Livermore National Laboratory (LLNL) to estimate the probability of occurrence of a DEGB, and to assess the effect that earthquakes have on DEGB probability. This report describes a probabilistic evaluation of reactor coolant loop piping in PWR plants having nuclear steam supply systems designed by Combustion Engineering. Two causes of pipe break were considered: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by failure of component supports due to an earthquake (indirect DEGB). The probability of direct DEGB was estimated using a probabilistic fracture mechanics model. The probability of indirect DEGB was estimated by estimating support fragility and then convolving fragility with seismic hazard. The results of this study indicate that the probability of a DEGB from either cause is very low for reactor coolant loop piping in these plants, and that NRC should therefore consider eliminating DEGB as a design basis in favor of more realistic criteria.

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

    SciTech Connect

    1998-09-01

    The purpose of this report is to provide the status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1995 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2, and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems.

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

    SciTech Connect

    1996-12-31

    The purpose of this report is to provide a status of the progress that was made towards Design Certification of System 80+{trademark} during the US government`s 1996 fiscal year. The System 80+ Advanced Light Water Reactor (ALWR) is a 3931 MW (1350 MWe) Pressurized Water Reactor (PWR). The design covers an essentially complete plant. It is based on EPRI ALWR Utility Requirements Document (URD) improvements to the Standardized System 80 Nuclear Steam Supply System (NSSS) in operation at Palo Verde Units 1, 2 and 3. The NSSS is a traditional two-loop arrangement with two steam generators, two hot legs and four cold legs, each with a reactor coolant pump. The System 80+ standard design houses the NSSS in a spherical steel containment vessel which is enclosed in a concrete shield building, thus providing the safety advantages of a dual barrier to radioactivity release. Other major features include an all-digital, human-factors-engineered control room, an alternate electrical AC power source, an In-Containment Refueling Water Storage Tank (IRWST), and plant arrangements providing complete separation of redundant trains in safety systems.

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

    SciTech Connect

    Not Available

    1993-12-31

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

  19. 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.

  20. 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. PMID:25043796

  1. The AVR nuclear power plant - a milestone in high-temperature reactor development

    SciTech Connect

    Schulten, R.

    1985-08-01

    For some time small high-temperature reactors, because of their special safety properties, have been of interest to various countries. The prototype of these facilities is the Arbeitsgemeinschaft Versuchsreaktor in Julich. During the past 17 years of operation, the components of the reactor, especially the fuel elements, have proved suitable for achieving permanent temperatures of the heat transfer medium of up to 950/sup 0/ C with low contamination of the loop. The convincing safety behavior of th reactor has been demonstrated. It is apparent that for this design an afterheat removal system is not required, since the afterheat can be adequately removed by heat conduction. In addition to the thorium/uranium fuel cycle originally planned, utilization of low-enrichment uranium in the fuel elements has also been fully developed.

  2. 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.

  3. Comparison of oxide- and metal-core behavior during CRBRP (Clinch River Breeder Reactor Plant) station blackout

    SciTech Connect

    Polkinghorne, S T; Atkinson, S A

    1986-01-01

    A resurrected concept that could significantly improve the inherently safe response of Liquid-Metal cooled Reactors (LMRs) during severe undercooling transients is the use of metallic fuel. Analytical studies have been reported on for the transient behavior of metal-fuel cores in innovative, inherently safe LMR designs. This paper reports on an analysis done, instead, for the Clinch River Breeder Reactor Plant (CRBRP) design with the only innovative change being the incorporation of a metal-fuel core. The SSC-L code was used to simulate a protected station blackout accident in the CRBRP with a 943 MWt Integral Fast Reactor (IFR) metal-fuel core. The results, compared with those for the oxide-fueled CRBRP, show that the margin to boiling is greater for the IFR core. However, the cooldown transient is more severe due to the faster thermal response time of metallic fuel. Some additional calculations to assess possible LMR design improvements (reduced primary system pressure losses, extended flow coastdown) are also discussed. 8 refs., 13 figs., 2 tabs.

  4. 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).

  5. Treatment of fresh leachate with high-strength organics and calcium from municipal solid waste incineration plant using UASB reactor.

    PubMed

    Ye, Jiexu; Mu, Yongjie; Cheng, Xiang; Sun, Dezhi

    2011-05-01

    Treatment of a fresh leachate with high-strength organics and calcium from municipal solid waste (MSW) incineration plant by an up-flow anaerobic sludge blanket (UASB) reactor was investigated under mesophilic conditions, emphasizing the influence of organic loading rate (OLR). When the reactor was fed with the raw leachate (COD as high as 70,390-75,480 mg/L) at an OLR of 12.5 kg COD/(m(3)d), up to ∼ 82.4% of COD was removed suggesting the feasibility of UASB process for treating fresh leachates from incineration plants. The ratio of volatile solids/total solids (VS/TS) of the anaerobic sludge in the UASB decreased significantly after a long-term operation due to the precipitation of calcium carbonate in the granules. Scanning electron microscopy (SEM) observation shows that Methanosaeta-like species were in abundance, accompanied by a variety of other species. The result was further confirmed by polymerase chain reaction-denaturing gradient gel electrophoresis (PCR-DGGE) and sequencing. PMID:21277197

  6. 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

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

    SciTech Connect

    Not Available

    1995-01-01

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

  8. 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.

  9. 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)

  10. 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).

  11. 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.

  12. 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.

  13. 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. PMID:24788329

  14. 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...

  15. 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.

  16. 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.

  17. 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.

  18. Exploratory design study on reactor configurations for carbon dioxide capture from conventional power plants employing regenerable solid sorbents

    SciTech Connect

    Yang, W.C.; Hoffman, J.

    2009-01-15

    Preliminary commercial designs were carried out for a fluidized bed as a CO{sub 2} adsorber and a moving bed as a CO{sub 2} regenerator. Reverse engineering methodology was employed on the basis of a commercial 500 MW supercritical PC power plant whereby the boundaries required for a particular reactor design and configuration could be set. Employing the proposed moving bed for regenerator is, however, not promising because of poor heat transfer, evolution of CO{sub 2} during regeneration, and high pressure drop when small particles are used. If regeneration kinetics is as slow as reported in tens of minutes, the bed height can be quite high and the reactor can be quite costly. In its place, a so-called assisted self-fluidization bed with embedded heat transfer surface was proposed. Theoretically, there is no reason why the fluidized bed cannot be successfully designed and operated both as an adsorber and a regenerator under proper adsorption and regeneration kinetics. Recent publications, where fluidized beds, circulating fluidized beds, or a combination of them were employed both as an adsorber and a regenerator, were cited. Staging may not be necessary employing the fluidized bed technology because of the capability to control reaction temperature at the optimum operating temperature through embedded heat transfer surface in the fluidized beds. Even if the staging is necessary, the implementation of staging in fluidized beds at ambient pressure and moderate temperature is relatively easy and with minimum cost penalty. Example designs are presented.

  19. 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. PMID:25725386

  20. Stability of cement-glass packages containing sodium borate salt generated from pressurized water reactor power plants

    SciTech Connect

    Izumida, T.; Kawamura, F.; Chino, K.; Kikuchi, M.

    1987-08-01

    A new solidification technique using cement-glass, which is a mixture of sodium silicate and cement, was studied for solidification of sodium borate salt of liquid waste generated from pressurized water reactor plants. When the sodium borate salt was solidified with the cement-glass, the resulting package contained eight times more sodium borate than was found in cement because it did not interact with sodium borate. The leaching ratio of cesium ion from the cement-glass package was one-tenth that of cement. Its low leaching ratio was due to the high cesium adsorption ability of cement-glass. The ratio could be theoretically evaluated by considering the cesium adsorption-desorption equilibrium.

  1. Bio-oil production via fast pyrolysis of biomass residues from cassava plants in a fluidised-bed reactor.

    PubMed

    Pattiya, Adisak

    2011-01-01

    Biomass residues from cassava plants, namely cassava stalk and cassava rhizome, were pyrolysed in a fluidised-bed reactor for production of bio-oil. The aims of this work were to investigate the yields and properties of pyrolysis products produced from both feedstocks as well as to identify the optimum pyrolysis temperature for obtaining the highest organic bio-oil yields. Results showed that the maximum yields of the liquid bio-oils derived from the stalk and rhizome were 62 wt.% and 65 wt.% on dry basis, respectively. The pyrolysis temperatures that gave highest bio-oil yields for both feedstocks were in the range of 475-510 °C. According to the analysis of the bio-oils properties, the bio-oil derived from cassava rhizome showed better quality than that derived from cassava stalk as the former had lower oxygen content, higher heating value and better storage stability. PMID:20864338

  2. 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.

  3. 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.

  4. 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.

  5. Probability of pipe failure in the reactor coolant loops of Combustion Engineering PWR Plants. Volume 3. Double-ended guillotine break indirectly induced by earthquakes

    SciTech Connect

    Ravindra, M.K.; Campbell, R.D.; Kennedy, R.P.; Banon, H.

    1985-01-01

    The requirements to design nuclear power plants for the effects of an instantaneous double-ended guillotine break (DEGB) of reactor coolant loop (RCL) piping have led to excessive design costs, interference of normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel. This report describes an aspect of the NRC/Lawrence Livermore National Laboratory sponsored research program aimed at investigating whether the probability of DEGB in RCL Piping of nuclear power plants is acceptably small and the requirements to design for the DEGB effects (e.g., provision of pipe whip restraints) may be removed. This study estimated the probability of indirect DEGB in RCL piping as a consequence of seismic-induced structural failures within the containment of Combustion Engineering supplied pressurized water reactor nuclear power plants in the United States. The median probability of indirect DEGB was estimated to be in the range of 10/sup -6/ per year for older plants, and less than 10/sup -8/ per year for modern plants; using very conservative assumptions, the 90% subjective probability value (confidence) of P/sub DEGB/ was found to be less than 5 x 10/sup -5/ per year for older plants and less than 3 x 10/sup -7/ per year for modern plants.

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

    SciTech Connect

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

    1987-10-01

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

  7. 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.

  8. 78 FR 2695 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2013-01-14

    ... ACRS meetings were published in the Federal Register on October 18, 2012, (77 FR 64146- 64147... Operations and Fire Protection; Notice of Meeting The ACRS Subcommittee on Plant Operations and Fire... onsite emergency response capabilities such as Emergency Operating Procedures (EOPs), Severe...

  9. 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

    ..., (74 FR 58268-58269). Detailed meeting agendas and meeting transcripts are available on the NRC Web... Operations and Fire Protection The ACRS Subcommittee on Plant Operations and Fire Protection will hold a..., October 6, 2010--8:30 a.m. until 12 p.m. The subcommittee will discuss the construction, inspection,...

  10. 76 FR 38212 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Plant...

    Federal Register 2010, 2011, 2012, 2013, 2014

    2011-06-29

    ... published in the Federal Register on October 21, 2010, (75 FR 65038-65039). Detailed meeting agendas and... Operations and Fire Protection; Notice of Meeting The ACRS Subcommittee on Plant Operations and Fire... Region II to discuss the construction inspection program at the Vogtle site and discuss other items...

  11. 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.

  12. 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".

  13. Rheology measurements for online monitoring of solids in activated sludge reactors of municipal wastewater treatment plant.

    PubMed

    Piani, Luciano; Rizzardini, Claudia Bruna; 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

  14. 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

  15. A model for the release, dispersion and environmental impact of a postulated reactor accident from a submerged commercial nuclear power plant

    NASA Astrophysics Data System (ADS)

    Bertch, Timothy Creston

    1998-12-01

    Nuclear power plants are inherently suitable for submerged applications and could provide power to the shore power grid or support future underwater applications. The technology exists today and the construction of a submerged commercial nuclear power plant may become desirable. A submerged reactor is safer to humans because the infinite supply of water for heat removal, particulate retention in the water column, sedimentation to the ocean floor and inherent shielding of the aquatic environment would significantly mitigate the effects of a reactor accident. A better understanding of reactor operation in this new environment is required to quantify the radioecological impact and to determine the suitability of this concept. The impact of release to the environment from a severe reactor accident is a new aspect of the field of marine radioecology. Current efforts have been centered on radioecological impacts of nuclear waste disposal, nuclear weapons testing fallout and shore nuclear plant discharges. This dissertation examines the environmental impact of a severe reactor accident in a submerged commercial nuclear power plant, modeling a postulated site on the Atlantic continental shelf adjacent to the United States. This effort models the effects of geography, decay, particle transport/dispersion, bioaccumulation and elimination with associated dose commitment. The use of a source term equivalent to the release from Chernobyl allows comparison between the impacts of that accident and the postulated submerged commercial reactor plant accident. All input parameters are evaluated using sensitivity analysis. The effect of the release on marine biota is determined. Study of the pathways to humans from gaseous radionuclides, consumption of contaminated marine biota and direct exposure as contaminated water reaches the shoreline is conducted. The model developed by this effort predicts a significant mitigation of the radioecological impact of the reactor accident release

  16. 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.

  17. 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.

  18. 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

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

    SciTech Connect

    Oxstrand, Johanna; Bly, Aaron; LeBlanc, Katya

    2014-09-01

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

  20. 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

  1. DEVELOPMENT OF NUCLEAR POWER PLANT SIMULATORS FOR SOVIET-DESIGNED NUCLEAR REACTORS.

    SciTech Connect

    Kohut, P.; Tutu, N.K.; Cleary, E.J.; Erickson, K.G.; Yoder, J.; Kroshilin, A.

    2001-01-07

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

  2. 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

  3. 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

  4. 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

  5. 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

  6. Anaerobic side-stream reactor for excess sludge reduction: 5-year management of a full-scale plant.

    PubMed

    Velho, V F; Foladori, P; Andreottola, G; Costa, R H R

    2016-07-15

    The long-term performances of a full-scale anaerobic side-stream reactor (ASSR) aimed at sludge reduction have been monitored for the first time, in comparison with a conventional activated sludge process (CAS). The plant was integrated with an ASSR treatment of 2293-3293 m(3). Operational parameters in the ASSR were: ORP -250 mV, interchange ratio of 7-10%, hydraulic retention time of 7 d. No worsening of effluent quality was observed in the ASSR configuration and removal efficiency of COD and NH4 was above 95%. A slight increase in the Sludge Volume Index did not cause worsening in effluent solids concentration. The observed sludge yield (Yobs) passed from 0.44 kgTSS/kgCOD in the CAS to 0.35 in the ASSR configuration. The reduction of Yobs by 20% is lower than expected from the literature where sythetic wastewater is used, indicating that sludge reduction efficiency is largely affected by inert mass fed with influent real wastewater. An increase by 45% of the ASSR volume did not promote a further reduction of Yobs, because sludge reduction is affected not solely by endogenous decay but also by other factors such as interchange ratio and aerobiosis/anaerobiosis alternation. PMID:27107390

  7. 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.

  8. 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.

  9. Reactor safety assessment system

    SciTech Connect

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

    1987-01-01

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

  10. Probability of pipe failure in the reactor coolant loops of Babcock and Wilcox PWR plants. Volume 2. Guillotine break indirectly induced by earthquakes

    SciTech Connect

    Ravindra, M.K.; Campbell, R.D.; Kipp, T.R.; Sues, R.H.

    1985-07-01

    The requirements to design nuclear power plants for the effects of an instantaneous double-ended guillotine break (DEGB) of the reactor coolant loop (RCL) piping have led to excessive design costs, interference with normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel. This report describes an aspect of the NRC/Lawrence Livermore National Laboratory sponsored research program aimed at exploring whether the probability of DEGB in RCL Piping of nuclear power plants is acceptably small and the requirements to design for the DEGB effects (e.g., provision of pipe whip restraints) may be removed. This study estimates the probability of indirect DEGB in RCL piping as a consequence of seismic-induced structural failures within the containment of Babcock and Wilcox supplied pressurized water reactor nuclear power plants in the United States. The median probability of indirect DEGB was estimated to range between 6 x 10/sup -11/ and 1 x 10/sup -7/ per year. Using very conservative assumptions, the 90% subjective probability value (confidence) of P/sub DEGB/ was found to be less than 1 x 10/sup -5/ per year. 19 refs., 19 figs., 11 tabs.

  11. Probability of pipe failure in the reactor coolant loop of Westinghouse PWR plants. Volume 3. Guillotine break indirectly induced by earthquakes

    SciTech Connect

    Ravindra, M.K.; Campbell, R.D.; Kennedy, R.P.; Banon, H.

    1985-02-01

    The requirements to design nuclear power plants for the effects of an instantaneous double-ended guillotine break (DEGB) of reactor coolant loop (RCL) piping have led to excessive design costs, interference of normal plant operation and maintenance, and unnecessary radiation exposure of plant maintenance personnel. This report describes an aspect of the NRC/Lawrence Livermore National Laboratory sponsored research program aimed at investigating whether the probability of DEGB in RCL piping of nuclear power plants is acceptably small and the requirements to design for the DEGB effects (e.g., provision of pipe whip restraints) may be removed. This study estimated the probability of indirect DEGB in RCL piping as a consequence of seismic-induced structural failures within the containment of Westinghouse supplied pressurized water reactor nuclear power plants in the United States. The median probability of indirect DEGB was estimated to be about 3x10/sup -6/ per year with a 10% to 90% subjective probability range approximately from 1x10/sup -7/ per year to 5x10/sup -5/ per year.

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

    PubMed

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

    2002-12-01

    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. PMID:12365490

  13. 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.

  14. 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.

  15. 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. PMID:26923717

  16. A novel membrane reactor design for controlled studies of interacting populations (simulation of the interaction between microorganism and plant suspension cultures).

    PubMed

    Pestchanker, L J; Ercoli, E C

    1997-08-20

    The design of a reactor in which two interacting cell populations (microorganisms and plants) could grow under controlled conditions was considered. In this reactor, the cell populations are separated by a membrane which permits semi-in vivo study of induced interaction-specific changes in metabolism. In this paper, the interaction of suspension culture of Nicotiana tabacum (tobacco) and the Oomycete, Phytophthora nicotiana was simulated. The results of the computer simulation show the induced metabolic changes as a consequence of the biological interaction. The paper introduces a novel approach in the strategy for the study of interacting population in suspension cultures. This type of system has potential applications in studies of the regulation of secondary metabolism and for the production of high values pharmaceuticals. (c) 1997 John Wiley & Sons, Inc. Biotechnol Bioeng 55: 609-615, 1997. PMID:18636571

  17. 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.

  18. 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.

  19. 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. PMID:16887646

  20. Evaluation of a upflow anaerobic sludge blanket reactor with partial recirculation of effluent used to treat wastewaters from pulp and paper plants.

    PubMed

    Buzzini, A P; Pires, E C

    2007-07-01

    The main purpose of this study was to evaluate the performance of a UASB reactor treating diluted black liquor from a Kraft pulp mill, which simulates an unbleached Kraft plant wastewater, under different operational conditions, including partial recycling of the effluent. The reactor's performance was evaluated from the standpoint of COD, pH, volatile acid concentration, alkalinity, concentration of methane in the biogas, and microbiological examinations of the sludge. Without recirculation the reduction of the HRT from 36 to 30h did not significantly affect the average COD removal efficiency. The parameter displaying the greatest variation was the average concentration of effluent volatile acids, which increased by 16%. With recirculation the reduction of the HRT from 30 to 24h increased the average COD removal efficiency from 75% to 78%. In this case, the average effluent alkalinity also showed an increase. The use of partial recirculation of the effluent did not improve significantly the COD removal under the operational conditions tested in this work, but it was possible to operate the reactor with lower hydraulic retention time without disintegration of the granules. PMID:17008096

  1. 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.

  2. 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. PMID:24727280

  3. 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

  4. Using a multi-parameter monitoring methodology to predict failures in the cryogenic plant of the cold neutron source at Australia's OPAL reactor

    NASA Astrophysics Data System (ADS)

    Lu, Weijian; Thiering, Russell

    2012-06-01

    A 5 kW Brayton-cycle helium refrigeration plant provides cooling at 20 K to the Cold Neutron Source (CNS) at Australia's OPAL Reactor. During several years of operation to the present day, the plant has experienced an unusually high number of turbine and compressor failures. The root cause for some of the failures is known, but for others remains to be determined. All of the failures were catastrophic without any prior warning from standard industrial monitoring based on singular process variables such as temperature, pressure and vibration. The failures and the down time they caused have been very costly. As the operator of the plant, we have developed a multi-parameter monitoring (MPM) methodology to track the performance of the plant. The methodology utilises indicators obtained from a combination of process variables based on their thermodynamic relations. By studying the historical trends of appropriate indicators, especially during the past failures, we have found some indicators that would be able to improve our predictive capability so that we can avoid similar failures in the future.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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

  10. 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

    ... Generation Nuclear Plant (NGNP) fuel and source term research and development of risk-informed performance... were published in the Federal Register on October 18, 2012, (77 FR 64146-64147). Detailed...

  11. 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

    ... participation in ACRS meetings were published in the Federal Register on October 21, 2010, (75 FR 65038-65039... Designs The ACRS Subcommittee on Future Plant Designs will hold a meeting on February 9, 2011, at...

  12. 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

    ... ACRS meetings were published in the Federal Register on October 21, 2010 (75 FR 65038-65039). Detailed... Designs The ACRS Subcommittee on Future Plant Designs will hold a meeting on April 5, 2011, at...

  13. 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.

  14. 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)

  15. 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. PMID:24527646

  16. 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. PMID:26894292

  17. Probability of pipe failure in the reactor coolant loops of Combustion Engineering PWR plants. Volume 2. Pipe failure induced by crack growth

    SciTech Connect

    Lo, T.Y.; Mensing, R.W.; Woo, H.H.; Holman, G.S.

    1984-09-01

    A study was conducted to determine if the probability of occurrence of a double-ended guillotine break (DEGB) in the primary coolant piping warrants the current design requirements that safeguard against the effect of DEGB. This report describes the results of an assessment of reactor coolant loop piping systems designed by Combustion Engineering, Inc. A probabilistic fracture mechanics approach was used to estimate the crack growth and to assess the crack stability in the piping throughout the lifetime of the plant. The results of the assessment indicate that the probability of occurrence of DEGB due to crack growth and instability is extremely small, which supports the argument that the postulation of DEGB in design should be eliminated and replaced with more reasonable criteria.

  18. 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.

  19. 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

    ... participation in ACRS meetings were published in the Federal Register on October 14, 2009, (74 FR 58268-58269... 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...

  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

    ...The NRC staff is soliciting public comment on its proposed Interim Staff Guidance (ISG) DC/COL-ISG-021 titled ``Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby Emergency Alternating Current Power System,'' (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092640035). This ISG provides new guidance information for......

  1. 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.

  2. 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.

  3. 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

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

    SciTech Connect

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

    1990-01-01

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

  5. 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

    ...The NRC staff is issuing its Final Interim Staff Guidance (ISG) DC/COL-ISG-021 titled ``Interim Staff Guidance on the Review of Nuclear Power Plant Designs Using a Gas Turbine Driven Standby Emergency Alternating Current Power System,'' Agencywide Documents Access and Management System (ADAMS) Accession No. ML102510119 for DC/ COL-ISG-021 and ADAMS Accession No. ML102510164 for Attachment 1 to......

  6. 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.

  7. 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.

  8. 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)

  9. 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. PMID:21878762

  10. 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)

  11. 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. PMID:23662694

  12. 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

  13. 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.

  14. 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. PMID:19043740

  15. 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

  16. 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.

  17. Breeder reactors in France

    SciTech Connect

    Zaleski, C.P.

    1980-04-11

    France relies on nuclear power as an important part of her energy program. Anticipating problems with the availability of natural uranium before the year 2020, the French have been pursuing a three-stage program of development of breeder reactors. The third reactor in this program, the near-commercial plant Super Phenix Mark I, is expected to reach power operation in 1983. Although there are still some uncertainties, particularly about the date when the breeder will become competitive with other energy sources, the outlook is considered favorable and preliminary designs for commercial plants are under way.

  18. The Fast-spectrum Transmutation Experimental Facility FASTEF: Main design achievements (part 2: Reactor building design and plant layout) within the FP7-CDT collaborative project of the European Commission

    SciTech Connect

    De Bruyn, D.; Engelen, J.; Ortega, A.; Aguado, M. P.

    2012-07-01

    MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) is the flexible experimental accelerator-driven system (ADS) in development at SCK-CEN in replacement of its material testing reactor BR2. SCK-CEN in association with 17 European partners from industry, research centres and academia, responded to the FP7 (Seventh Framework Programme) call from the European Commission to establish a Central Design Team (CDT) for the design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) able to demonstrate efficient transmutation and associated technology through a system working in subcritical and/or critical mode. The project has started on April 01, 2009 for a period of three years. In this paper, we present the latest concept of the reactor building and the plant layout. The FASTEF facility has evolved quite a lot since the intermediate reporting done at the ICAPP'10 and ICAPP'11 conferences 1,2. Many iterations have been performed to take into account the safety requirements. The present configuration enables an easy operation and maintenance of the facility, including the possibility to change large components of the reactor. In a companion paper 3, we present the latest configuration of the reactor core and primary system. (authors)

  19. 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

    SciTech Connect

    Vickers, Lisa R.

    2002-07-01

    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 {sup 239}Pu 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 personnel and public safety, environment, and operating life of the reactor. The primary conclusion of this research was that the addition of the maximum fraction of 1/3 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 personnel and public safety, environment, and operating life of the reactor. (author)

  20. 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.

  1. 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.

  2. Performance evaluation of a granular activated carbon-sequencing batch biofilm reactor pilot plant system used in treating real wastewater from recycled paper industry.

    PubMed

    Muhamad, Mohd Hafizuddin; Sheikh Abdullah, Siti Rozaimah; Mohamad, Abu Bakar; Rahman, Rakmi Abdul; Kadhum, Abdul Amir Hasan

    2012-01-01

    A pilot scale granular activated carbon-sequencing batch biofilm reactor with a capacity of 2.2 m3 was operated for over three months to evaluate its performance treating real recycled paper industry wastewater under different operational conditions. In this study, dissolved air floatation (DAF) and clarifier effluents were used as influent sources of the pilot plant. During the course of the study, the reactor was able to biodegrade the contaminants in the incoming recycled paper mill wastewater in terms of chemical oxygen demand (COD), adsorbable organic halides (AOX; specifically 2,4-dichlorophenol (2,4-DCP)) and ammoniacal nitrogen (NH3-N) removal efficiencies at varying hydraulic retention times (HRTs) of 1-3 days, aeration rates (ARs) of 2.1-3.4 m3/min and influent feed concentration of 40-950 mg COD/l. Percentages of COD, 2,4-DCP and NH3-N removals increased with increasing HRT, resulting in more than 90% COD, 2,4-DCP and NH3-N removals at HRT values above two days. Degradation of COD, 2,4-DCP and NH3-N were seriously affected by variation of ARs, which resulted in significant decrease of COD, 2,4-DCP and NH3-N removals by decreasing ARs from 3.4 m3/min to 2.1 m3/min, varying in the ranges of 24-80%, 6-96% and 5-42%, respectively. In comparison to the clarifier effluent, the treatment performance of DAF effluent, containing high COD concentration, resulted in a higher COD removal of 82%. The use of diluted DAF effluent did not improve significantly the COD removal. Higher NH3-N removal efficiency of almost 100% was observed during operation after maintenance shutdown compared to normal operation, even at the same HRT of one day due to the higher dissolved oxygen concentrations (1-7 mg/l), while no significant difference in COD removal efficiency was observed. PMID:22720416

  3. 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)

  4. Hanford plots reactor move

    SciTech Connect

    King, H.

    1993-10-04

    Anxious to show skeptics some bang for the mounting cleanup bucks, the US Dept. of Energy has taken steps to get a large and visible project under way at its Hanford weapon plant-moving eight old nuclear reactors to permanent burial at an inland dump site. The effort, conservatively budgeted at $235 million, will be the eastern Washington site's largest [open quotes]D D[close quotes]-decontamination and decommissioning-project yet. Last month, DOE unveiled its final record of decision for the plants that spells out D D options-from doing nothing to immediate removal of entire reactor blocks. At issue are reactors built from 1943 to 1963 along the Columbia River. Defunct since 1971, they once produced plutonium.

  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. BOILING REACTORS

    DOEpatents

    Untermyer, S.

    1962-04-10

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

  7. Structural mechanics in reactor technology

    SciTech Connect

    Wittmann, F.H.

    1987-01-01

    This series consists of 14 volumes. Each contains several papers. The volume subtitles are: Indexes, Abbreviations, Supplement; Computational Mechanics and Computer-Aided Engineering; Fuel Elements and Assemblies; Experience with Structures and Components in Operating Reactors; Fast Reactor Core and Coolant Circuit Structures; LWR Pressure Components; Fracture Mechanics and NDE; Concrete and Concrete Structures; Extreme Loading and Response of Reactor Containments; Seismic Response Analysis of Nuclear Power Plant Systems; Mechanical and Thermal Problems of Fusion Reactors; Structural Reliability Probabilistic Safety Assessment; and Inelastic Behaviour of Metals and Constitutive Equations.

  8. 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.

  9. 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

  10. 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.