WorldWideScience

Sample records for reactor design status

  1. Status of Preconceptual Design of the Advanced High-Temperature Reactor (AHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Ingersoll, D.T.

    2004-07-29

    A new reactor plant concept is presented that combines the benefits of ceramic-coated, high-temperature particle fuel with those of clean, high-temperature, low-pressure molten salt coolant. The Advanced High-Temperature Reactor (AHTR) concept is a collaboration of Oak Ridge National Laboratory, Sandia National Laboratories, and the University of California at Berkeley. The purpose of the concept is to provide an advanced design capable of satisfying the top-level functional requirements of the U.S. Department of Energy Next Generation Nuclear Plant (NGNP), while also providing a technology base that is sufficiently robust to allow future development paths to higher temperatures and larger outputs with highly competitive economics. This report summarizes the status of the AHTR preconceptual design. It captures the results from an intense effort over a period of 3 months to (1) screen and examine potential feasibility concerns with the concept; (2) refine the conceptual design of major systems; and (3) identify research, development, and technology requirements to fully mature the AHTR design. Several analyses were performed and are presented to quantify the AHTR performance expectations and to assist in the selection of several design parameters. The AHTR, like other NGNP reactor concepts, uses coated particle fuel in a graphite matrix. But unlike the other NGNP concepts, the AHTR uses molten salt rather than helium as the primary system coolant. The considerable previous experience with molten salts in nuclear environments is discussed, and the status of high-temperature materials is reviewed. The large thermal inertia of the system, the excellent heat transfer and fission product retention characteristics of molten salt, and the low-pressure operation of the primary system provide significant safety attributes for the AHTR. Compared with helium coolant, a molten salt cooled reactor will have significantly lower fuel temperatures (150-200-C lower) for the

  2. Status of French reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ballagny, A. [Commissariat a l`Energie Atomique, Saclay (France)

    1997-08-01

    The status of French reactors is reviewed. The ORPHEE and RHF reactors can not be operated with a LEU fuel which would be limited to 4.8 g U/cm{sup 3}. The OSIRIS reactor has already been converted to LEU. It will use U{sub 3}Si{sub 2} as soon as its present stock of UO{sub 2} fuel is used up, at the end of 1994. The decision to close down the SILOE reactor in the near future is not propitious for the start of a conversion process. The REX 2000 reactor, which is expected to be commissioned in 2005, will use LEU (except if the fast neutrons core option is selected). Concerning the end of the HEU fuel cycle, the best option is reprocessing followed by conversion of the reprocessed uranium to LEU.

  3. Status of the design concepts for a high fluence fast pulse reactor (HFFPR)

    Energy Technology Data Exchange (ETDEWEB)

    Philbin, J.S.; Nelson, W.E.; Rosenstroch, B.

    1978-10-01

    The report describes progress that has been made on the design of a High Fluence Fast Pulse Reactor (HFFPR) through the end of calendar year 1977. The purpose of this study is to present design concepts for a test reactor capable of accommodating large scale reactor safety tests. These concepts for reactor safety tests are adaptations of reactor concepts developed earlier for DOE/OMA for the conduct of weapon effects tests. The preferred driver core uses fuel similar to that developed for Sandia's ACPR upgrade. It is a BeO/UO/sub 2/ fuel that is gas cooled and has a high volumetric heat capacity. The present version of the design can drive large (217) pin bundles of prototypically enriched mixed oxide fuel well beyond the fuel's boiling point. Applicability to specific reactor safety accident scenarios and subsequent design improvements will be presented in future reports on this subject.

  4. STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

    Energy Technology Data Exchange (ETDEWEB)

    Davenport, Michael; Petti, D. A.; Palmer, Joe

    2016-11-01

    The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven 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). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the 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 several independent 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 was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control

  5. Status report on the conceptual design of a commercial tokamak hybrid reactor (CTHR)

    Energy Technology Data Exchange (ETDEWEB)

    None

    1979-09-01

    A preliminary conceptual design is presented for an early twenty-first century fusion hybrid reactor called the Commercial Tokamak Hybrid Reactor (CTHR). This design was developed as a first generation commercial plant producing fissile fuel to support a significant number of client Light Water Reactor (LWR) plants. The study has been made in sufficient depth to indicate no insurmountable technical problems exist and has provided a basis for valid cost estimates of the hybrid plants as well as the hybrid/LWR system busbar electricity costs. This energy system can be optimized to have a net cost of busbar electricity that is equivalent to the conventional LWR plant, yet is not dependent on uranium ore prices or standard enrichment costs, since the fusion hybrid can be fueled by numerous fertile fuel resources.

  6. Nuclear reactor design

    CERN Document Server

    2014-01-01

    This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

  7. INVAP's Research Reactor Designs

    Directory of Open Access Journals (Sweden)

    Eduardo Villarino

    2011-01-01

    Full Text Available INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors.

  8. Current status of fast reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, H.H.

    1979-01-01

    The subject of calculation of reactivity coefficients for fast reactors is developed, starting with a discussion of the status of relevant nuclear data and proceeding to the subjects of group cross section generation and of methods of obtaining reactivity coefficients from group cross sections. Reactivity coefficients measured in critical experiments are compared with calculated values. Dependence of reactivity coefficients on reactor design is discussed. Finally, results of the recent international comparison of calculated reactivity coefficients are presented.

  9. Preliminary conceptual design of a Demonstration Tokamak Hybrid Reactor (DTHR). Status report, January 1978--March 1978

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, J.L. (ed.)

    1978-03-01

    The DTHR preliminary conceptual design consists of a magnetically confined fusion reactor fitted with a fertile thorium blanket. The fusion driver concept is based on a beam driven plasma, but at sufficiently high plasma densities that neutrons originating from the interactions of bulk plasma ions contribute significantly to the wall loading. The tokamak has a major radius of 5.2 m, a minor radius of 1.2 m, and the elongation is 1.6. All of the magnetic coil systems are superconducting Nb/sub 3/Sn based on the Large Coil Project (LCP) technology. The toroidal field (TF) coils employ an innovative concept, the ''compact D'' configuration. An engineered bundle divertor concept has been developed based on the bundle divertor design techniques developed for TNS and ISX-B. A thermal power of 150MW of 200 keV deuterium is injected into the plasma through six ducts of a positive ion, neutral beam injection system (NBIS). A water cooled, 316 stainless steel vacuum vessel concept was developed and initial scoping analyses look encouraging. The fusile fuel handling system was evaluated and defined. Details of the tritium injection system remain to be developed. Tritium breeding will be assessed in subsequent phases of the DTHR operation. The fusion driver provides a neutron first wall loading of 2MW/m/sup 2/ for fissile production in the blanket.

  10. Design and Status of the NGNP Fuel Experiment AGR-3/4 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven 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). 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 several years to demonstrate and qualify new TRISO coated 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 was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in November 2013. Since the purpose of this experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  11. SNTP program reactor design

    Science.gov (United States)

    Walton, Lewis A.; Sapyta, Joseph J.

    1993-06-01

    The Space Nuclear Thermal Propulsion (SNTP) program is evaluating the feasibility of a particle bed reactor for a high-performance nuclear thermal rocket engine. Reactors operating between 500 MW and 2,000 MW will produce engine thrusts ranging from 20,000 pounds to 80,000 pounds. The optimum reactor arrangement depends on the power level desired and the intended application. The key components of the reactor have been developed and are being tested. Flow-to-power matching considerations dominate the thermal-hydraulic design of the reactor. Optimal propellant management during decay heat cooling requires a three-pronged approach. Adequate computational methods exist to perform the neutronics analysis of the reactor core. These methods have been benchmarked to critical experiment data.

  12. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  13. Slurry reactor design studies

    Energy Technology Data Exchange (ETDEWEB)

    Fox, J.M.; Degen, B.D.; Cady, G.; Deslate, F.D.; Summers, R.L. (Bechtel Group, Inc., San Francisco, CA (USA)); Akgerman, A. (Texas A and M Univ., College Station, TX (USA)); Smith, J.M. (California Univ., Davis, CA (USA))

    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.

  14. INVAP's Research Reactor Designs

    OpenAIRE

    Eduardo Villarino; Alicia Doval

    2011-01-01

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper ...

  15. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-08-01

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

  16. Computer analyses for the design, operation and safety of new isotope production reactors: A technology status review

    Energy Technology Data Exchange (ETDEWEB)

    Wulff, W.

    1990-01-01

    A review is presented on the currently available technologies for nuclear reactor analyses by computer. The important distinction is made between traditional computer calculation and advanced computer simulation. Simulation needs are defined to support the design, operation, maintenance and safety of isotope production reactors. Existing methods of computer analyses are categorized in accordance with the type of computer involved in their execution: micro, mini, mainframe and supercomputers. Both general and special-purpose computers are discussed. Major computer codes are described, with regard for their use in analyzing isotope production reactors. It has been determined in this review that conventional systems codes (TRAC, RELAP5, RETRAN, etc.) cannot meet four essential conditions for viable reactor simulation: simulation fidelity, on-line interactive operation with convenient graphics, high simulation speed, and at low cost. These conditions can be met by special-purpose computers (such as the AD100 of ADI), which are specifically designed for high-speed simulation of complex systems. The greatest shortcoming of existing systems codes (TRAC, RELAP5) is their mismatch between very high computational efforts and low simulation fidelity. The drift flux formulation (HIPA) is the viable alternative to the complicated two-fluid model. No existing computer code has the capability of accommodating all important processes in the core geometry of isotope production reactors. Experiments are needed (heat transfer measurements) to provide necessary correlations. It is important for the nuclear community, both in government, industry and universities, to begin to take advantage of modern simulation technologies and equipment. 41 refs.

  17. Thermionic Reactor Design Studies

    Energy Technology Data Exchange (ETDEWEB)

    Schock, Alfred

    1994-06-01

    During the 1960's and early 70's the author performed extensive design studies, analyses, and tests aimed at thermionic reactor concepts that differed significantly from those pursued by other investigators. Those studies, like most others under Atomic Energy Commission (AEC and DOE) and the National Aeronautics and Space Administration (NASA) sponsorship, were terminated in the early 1970's. Some of this work was previously published, but much of it was never made available in the open literature. U.S. interest in thermionic reactors resumed in the early 80's, and was greatly intensified by reports about Soviet ground and flight tests in the late 80's. This recent interest resulted in renewed U.S. thermionic reactor development programs, primarily under Department of Defense (DOD) and Department of Energy (DOE) sponsorship. Since most current investigators have not had an opportunity to study all of the author's previous work, a review of the highlights of that work may be of value to them. The present paper describes some of the author's conceptual designs and their rationale, and the special analytical techniques developed to analyze their performance. The basic designs, first published in 1963, are based on single-cell converters, either double-ended diodes extending over the full height of the reactor core or single-ended diodes extending over half the core height. In that respect they are similar to the thermionic fuel elements employed in the Topaz-2 reactor subsequently developed in the Soviet Union, copies of which were recently imported by the U.S. As in the Topaz-2 case, electrically heated steady-state performance tests of the converters are possible before fueling. Where the author's concepts differed from the later Topaz-2 design was in the relative location of the emitter and the collector. Placing the fueled emitter on the outside of the cylindrical diodes permits much higher axial conductances to reduce ohmic

  18. Turning points in reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Beckjord, E.S.

    1995-09-01

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

  19. Mirror Advanced Reactor Study interim design report

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

    The status of the design of a tenth-of-a-kind commercial tandem-mirror fusion reactor is described at the midpoint of a two-year study. When completed, the design is to serve as a strategic goal for the mirror fusion program. The main objectives of the Mirror Advanced Reactor Study (MARS) are: (1) to design an attractive tandem-mirror fusion reactor producing electricity and synfuels (in alternate versions), (2) to identify key development and technology needs, and (3) to exploit the potential of fusion for safety, low activation, and simple disposal of radioactive waste. In the first year we have emphasized physics and engineering of the central cell and physics of the end cell. Design optimization and trade studies are continuing, and we expect additional modifications in the end cells to further improve the performance of the final design.

  20. Jules Horowitz Reactor, basic design

    Energy Technology Data Exchange (ETDEWEB)

    Bergamaschi, Y.; Bouilloux, Y.; Chantoin, P.; Guigon, B.; Bravo, X.; Germain, C.; Rommens, M.; Tremodeux, P

    2003-07-01

    Since the shutdown of the SILOE reactor in 1997, the OSIRIS reactor has ensured the needs regarding technological irradiation at CEA including those of its industrial partners and customers. The Jules Horowitz Reactor will replace it. It has the ambition to provide the necessary nuclear data and maintain a fission research capacity in Europe after 2010. This capacity should be service-oriented. It will be established in Cadarache. The Jules Horowitz reactor will also: - represent a significant step in term of performances and experimental capabilities, - be designed with a high flexibility, in order to satisfy the current demand from European industry, research and be able to accommodate future requirements, - reach a high level of safety, according to the best current practice. This paper will present the main functionalities and the design options resulting from the 'preliminary design' studies. (authors)

  1. Reactor antineutrino fluxes - status and challenges

    CERN Document Server

    Huber, Patrick

    2016-01-01

    In this contribution we describe the current understanding of reactor antineutrino fluxes and point out some recent developments. This is not intended to be a complete review of this vast topic but merely a selection of observations and remarks, which despite their incompleteness, will highlight the status and the challenges of this field.

  2. Reactor antineutrino fluxes – Status and challenges

    Energy Technology Data Exchange (ETDEWEB)

    Huber, Patrick, E-mail: pahuber@vt.edu

    2016-07-15

    In this contribution we describe the current understanding of reactor antineutrino fluxes and point out some recent developments. This is not intended to be a complete review of this vast topic but merely a selection of observations and remarks, which despite their incompleteness, will highlight the status and the challenges of this field.

  3. Simplifying Microbial Electrosynthesis Reactor Design

    Directory of Open Access Journals (Sweden)

    Cloelle G.S. Giddings

    2015-05-01

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

  4. Status and problems of fusion reactor development.

    Science.gov (United States)

    Schumacher, U

    2001-03-01

    Thermonuclear fusion of deuterium and tritium constitutes an enormous potential for a safe, environmentally compatible and sustainable energy supply. The fuel source is practically inexhaustible. Further, the safety prospects of a fusion reactor are quite favourable due to the inherently self-limiting fusion process, the limited radiologic toxicity and the passive cooling property. Among a small number of approaches, the concept of toroidal magnetic confinement of fusion plasmas has achieved most impressive scientific and technical progress towards energy release by thermonuclear burn of deuterium-tritium fuels. The status of thermonuclear fusion research activity world-wide is reviewed and present solutions to the complicated physical and technological problems are presented. These problems comprise plasma heating, confinement and exhaust of energy and particles, plasma stability, alpha particle heating, fusion reactor materials, reactor safety and environmental compatibility. The results and the high scientific level of this international research activity provide a sound basis for the realisation of the International Thermonuclear Experimental Reactor (ITER), whose goal is to demonstrate the scientific and technological feasibility of a fusion energy source for peaceful purposes.

  5. Reference mirror hybrid fusion-fission reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Bender, D.J.; Lee, J.D.; Neef, W.S.

    1977-06-08

    The status of the reference mirror hybrid reactor design being performed by LLL and General Atomic is summarized. The reactor parameters have been chosen to minimize the cost of producing fissile fuel for consumption in fission power reactors. The design draws on the experience developed at LLL in previous hybrid reactor conceptual designs and on GA expertise in gas-cooling technology and fission reactor mechanical design. As in the past, we have emphasized the use of existing technology where possible and a minimum extrapolation of technology otherwise. We consider our projections for the plasma physics parameters to be conservative, in that they are well-founded on the experiments in 2XIIB and the interpretation of these experiments.

  6. Design for the WWR-M reactor vessel removal

    Energy Technology Data Exchange (ETDEWEB)

    Lobach, Yu.N., E-mail: lobach@kinr.kiev.ua [Institute for Nuclear Research of NASU, Prospekt Nauki 47, 03680 Kiev (Ukraine); Toth, G., E-mail: gtoth@aeki.kfki.hu [Centre for Energy Research of HAS, Konkoly Thege Miklós út 29-33, Budapest 1121 (Hungary)

    2013-05-15

    Highlights: ► The current status of the decommissioning planning for the WWR-M reactor is outlined. ► The general dismantling strategy consists of the dismantling and removal of the separate bulky elements as whole pieces without preliminary segmentation. ► The design for the reactor vessel extraction was selected and developed. -- Abstract: The final decommissioning planning for the Kiev's research reactor WWR-M is in progress now. The general dismantling strategy consists of the dismantling and removal of the separate bulky elements as whole pieces without preliminary segmentation. The reactor vessel removal is considered as the key element in the sequence of dismantling works; a separate design for the reactor vessel extraction was developed. The extensive analysis focused on the optimization of this technical task has been performed. The outline of the design is presented in this paper.

  7. The Status of the US High-Temperature Gas Reactors

    Directory of Open Access Journals (Sweden)

    Andrew C. Kadak

    2016-03-01

    Full Text Available In 2005, the US passed the Energy Policy Act of 2005 mandating the construction and operation of a high-temperature gas reactor (HTGR by 2021. This law was passed after a multiyear study by national experts on what future nuclear technologies should be developed. As a result of the Act, the US Congress chose to develop the so-called Next-Generation Nuclear Plant, which was to be an HTGR designed to produce process heat for hydrogen production. Despite high hopes and expectations, the current status is that high temperature reactors have been relegated to completing research programs on advanced fuels, graphite and materials with no plans to build a demonstration plant as required by the US Congress in 2005. There are many reasons behind this diminution of HTGR development, including but not limited to insufficient government funding requirements for research, unrealistically high temperature requirements for the reactor, the delay in the need for a “hydrogen” economy, competition from light water small modular light water reactors, little utility interest in new technologies, very low natural gas prices in the US, and a challenging licensing process in the US for non-water reactors.

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

    Science.gov (United States)

    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.

  9. Small modular reactors: current status, economic aspects and licensing

    Energy Technology Data Exchange (ETDEWEB)

    Zimbron E, E. [Instituto Tecnologico de Estudios Superiores de Monterrey, Campus Santa Fe, Av. Carlos Lazo No. 100, Santa Fe, 01389 Mexico D. F. (Mexico); Puente E, F., E-mail: erick.zimbron@gmail.com [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    Interest for nuclear energy had resurgence since the beginning of the new century. This was a consequence of the new world conditions and needs: increasing energy demands (mainly from developing countries), awareness of the importance of energetic security and the necessity of limiting carbon emissions. In this nuclear boom, Small Modular Reactors (SMRs) develop and start to consolidate as a viable option for the present energy market. Their modular characteristics, lower initial capital cost, passive safety features and their niche applications, situate them as a technology with various advantages. The following study will present and analysis that will help to comprehend the SMRs present status. Information will show planned reactors, reactors in construction and in operation, advantages and challenges of their implementation, relevant economic aspects and important licensing factors that need to be highlighted. The analysis showed that the SMR technology is still in an initial stage that could reach and important development point in the next ten years. In this period, many of the reactors that are in design stage or that are through their licensing process might be constructed and could be getting ready for a commercial status. On the other hand, it has been observed that there are two main economic factors that need to be considered for any SMRs implementation project. First, the costs (initial, operation, maintenance, fuel and decommissioning) and second their possible niche market applications. Additionally, it has been noted that the licensing process is one of the greatest challenges for SMR general development. Licensing is mainly related to topic such as Emergency Planning Zone, first-of-a-kind engineering, passive safety features, proliferation resistance, multiple module designs and staffing. Previous information will serve as a base for carrying out a feasibility assessment analysis for SMR in Mexico. This part will be the last section of the project

  10. Design of an organic simplified nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

    2016-08-15

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  11. Advanced tokamak concepts and reactor designs

    NARCIS (Netherlands)

    Oomens, A. A. M.

    2000-01-01

    From a discussion of fusion reactor designs based on today's well-established experience gained in the operation of large tokamaks, it is concluded that such reactors are economically not attractive. The physics involved in the various options for concept improvement is described, some examples

  12. Design of an Organic Simplified Nuclear Reactor

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-08-01

    Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

  13. Indian fast reactor technology: Current status and future programme

    Indian Academy of Sciences (India)

    S C Chetal; P Chellapandi

    2013-10-01

    The paper brings out the advantages of fast breeder reactor and importance of developing closed nuclear fuel cycle for the large scale energy production, which is followed by its salient safety features. Further, the current status and future strategy of the fast reactor programme since the inception through 40 MWt/13 MWe Fast Breeder Test Reactor (FBTR), is highlighted. The challenges and achievements in science and technology of FBRs focusing on safety are described with the particular reference to 500 MWe capacity Prototype Fast Breeder Reactor (PFBR), being commissioned at Kalpakkam. Roadmap with comprehensive R&D for the large scale deployment of Sodium Cooled Fast Reactor (SFRs) and timely introduction of metallic fuel reactors with emphasis on breeding gain and enhanced safety are being brought out in this paper.

  14. Seismic design of reactors in NUCEF

    Energy Technology Data Exchange (ETDEWEB)

    Kurosaki, Akira [Mitsui Shipbuilding and Engineering Co. Ltd., Tokyo (Japan); Kuchiya, Masao; Yasuda, Naomitsu; Kitanaka, Tsutomu; Ogawa, Kazuhiko; Sakuraba, Koichi; Izawa, Naoki; Takeshita, Isao

    1997-03-01

    Basic concept and calculation method for the seismic design of the main equipment of the reactors in NUCEF (Nuclear Fuel Cycle Safety Engineering Research Facility) are described with actual calculation examples. The present paper is published to help the seismic design of the equipment and application of the authorization for the design and constructing of facilities. (author)

  15. Generic small modular reactor plant design.

    Energy Technology Data Exchange (ETDEWEB)

    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.

  16. Status Report on Scoping Reactor Physics and Sensitivity/Uncertainty Analysis of LR-0 Reactor Molten Salt Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Mueller, Donald E. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Patton, Bruce W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Reactor and Nuclear Systems Division

    2016-08-31

    Experiments are being planned at Research Centre Rež (RC Rež) to use the FLiBe (2 7LiF-BeF2) salt from the Molten Salt Reactor Experiment (MSRE) to perform reactor physics measurements in the LR-0 low power nuclear reactor. These experiments are intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems utilizing FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL) is performing sensitivity/uncertainty (S/U) analysis of these planned experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objective of these analyses is to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a status update on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. The S/U analyses will be used to inform design of FLiBe-based experiments using the salt from MSRE.

  17. Design of an Organic Simplified Nuclear Reactor

    OpenAIRE

    Koroush Shirvan; Eric Forrest

    2016-01-01

    Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attr...

  18. Design options for a bunsen reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Robert Charles

    2013-10-01

    This work is being performed for Matt Channon Consulting as part of the Sandia National Laboratories New Mexico Small Business Assistance Program (NMSBA). Matt Channon Consulting has requested Sandia's assistance in the design of a chemical Bunsen reactor for the reaction of SO2, I2 and H2O to produce H2SO4 and HI with a SO2 feed rate to the reactor of 50 kg/hour. Based on this value, an assumed reactor efficiency of 33%, and kinetic data from the literature, a plug flow reactor approximately 1%E2%80%9D diameter and and 12 inches long would be needed to meet the specification of the project. Because the Bunsen reaction is exothermic, heat in the amount of approximately 128,000 kJ/hr would need to be removed using a cooling jacket placed around the tubular reactor. The available literature information on Bunsen reactor design and operation, certain support equipment needed for process operation and a design that meet the specification of Matt Channon Consulting are presented.

  19. Ultra high temperature particle bed reactor design

    Science.gov (United States)

    Lazareth, Otto; Ludewig, Hans; Perkins, K.; Powell, J.

    1990-01-01

    A direct nuclear propulsion engine which could be used for a mission to Mars is designed. The main features of this reactor design are high values for I(sub sp) and very efficient cooling. This particle bed reactor consists of 37 cylindrical fuel elements embedded in a cylinder of beryllium which acts as a moderator and reflector. The fuel consists of a packed bed of spherical fissionable fuel particles. Gaseous H2 passes over the fuel bed, removes the heat, and is exhausted out of the rocket. The design was found to be neutronically critical and to have tolerable heating rates. Therefore, this particle bed reactor design is suitable as a propulsion unit for this mission.

  20. Spheromak reactor-design study

    Energy Technology Data Exchange (ETDEWEB)

    Les, J.M.

    1981-06-30

    A general overview of spheromak reactor characteristics, such as MHD stability, start up, and plasma geometry is presented. In addition, comparisons are made between spheromaks, tokamaks and field reversed mirrors. The computer code Sphero is also discussed. Sphero is a zero dimensional time independent transport code that uses particle confinement times and profile parameters as input since they are not known with certainty at the present time. More specifically, Sphero numerically solves a given set of transport equations whose solutions include such variables as fuel ion (deuterium and tritium) density, electron density, alpha particle density and ion, electron temperatures.

  1. The current status of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tri Wulan Tjiptono; Syarip

    1998-10-01

    The Kartini reactor reached the first criticality on January 25, 1979. In the first three years, the reactor power is limited up to 50 kW thermal power and on July 1, 1982 has been increased to 100 kW. It has been used as experiments facility by researcher of Atomic Energy National Agency and students of the Universities. Three beam tubes used as experiments facilities, the first, is used as a neutron source for H{sub 2}O-Natural Uranium Subcritical Assembly, the second, is developed for neutron radiography facility and the third, is used for gamma radiography facility. The other facilities are rotary rack and two pneumatic transfer systems, one for delayed neutron counting system and the other for the new Neutron Activation Analysis (NAA) facility. The rotary rack used for isotope production for NAA purpose (for long time irradiation), the delayed neutron counting system used for analysis the Uranium contents of the ores and the new NAA is provided for short live elements analysis. In the last three years the Reactor Division has a joint use program with the Nuclear Component and Engineering Center in research reactor instrumentation and control development. (author)

  2. Design method of divertor in tokamak reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ueda, Noriaki (Mitsubishi Atomic Power Industries, Inc., Tokyo (Japan)); Itoh, Sanae; Tanaka, Masaaki; Itoh, Kimitaka

    1991-03-01

    Computational method to design the efficient divertor configuration in tokamak reactor is presented. The two-dimensional code has been developed to analyze the distributions of the plasma and neutral particles for realistic configurations. Using this code, a method to design the efficient divertor configuration is developed. An example of new divertor, which consists of the baffle and fin plates, is analyzed. (author).

  3. Status and Prospects of Reactor Neutrino Experiments

    CERN Document Server

    Kim, Soo-Bong

    2015-01-01

    New generation of three reactor neutrino experiments have made definitive measurements of the smallest neutrino mixing angle theta13 in 2012, based on the disappearance of electron antineutrinos. More precise measurements of the mixing angle have been made as well as the squared mass difference between electron neutrinos. A rather large value of theta13 has opened a new window to find the CP violation phase and to determine the neutrino mass hierarchy. Future reactor experiments, JUNO and RENO50, are proposed to determine the neutrino mass hierarchy and to make highly precise measurements of theta12, the squared mass difference between neutrino masses 2 and 1, and the squared mass difference between electron neutrinos.

  4. Advanced Small Modular Reactor Economics Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Harrison, Thomas J [ORNL

    2014-10-01

    This report describes the data collection work performed for an advanced small modular reactor (AdvSMR) economics analysis activity at the Oak Ridge National Laboratory. The methodology development and analytical results are described in separate, stand-alone documents as listed in the references. The economics analysis effort for the AdvSMR program combines the technical and fuel cycle aspects of advanced (non-light water reactor [LWR]) reactors with the market and production aspects of SMRs. This requires the collection, analysis, and synthesis of multiple unrelated and potentially high-uncertainty data sets from a wide range of data sources. Further, the nature of both economic and nuclear technology analysis requires at least a minor attempt at prediction and prognostication, and the far-term horizon for deployment of advanced nuclear systems introduces more uncertainty. Energy market uncertainty, especially the electricity market, is the result of the integration of commodity prices, demand fluctuation, and generation competition, as easily seen in deregulated markets. Depending on current or projected values for any of these factors, the economic attractiveness of any power plant construction project can change yearly or quarterly. For long-lead construction projects such as nuclear power plants, this uncertainty generates an implied and inherent risk for potential nuclear power plant owners and operators. The uncertainty in nuclear reactor and fuel cycle costs is in some respects better understood and quantified than the energy market uncertainty. The LWR-based fuel cycle has a long commercial history to use as its basis for cost estimation, and the current activities in LWR construction provide a reliable baseline for estimates for similar efforts. However, for advanced systems, the estimates and their associated uncertainties are based on forward-looking assumptions for performance after the system has been built and has achieved commercial operation

  5. A Course in Chemical Reactor Design.

    Science.gov (United States)

    Takoudis, Christos G.

    1983-01-01

    Presents course outline, topics covered, and final project (doubling as a take home final exam) for a one-semester, interdisciplinary course on the design and behavior of chemical reactors. Interplay of chemical and physical rate processes is stressed in the course. (JM)

  6. Advanced burner test reactor preconceptual design report.

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-12-16

    advanced fuel cycle; (2) To qualify the transuranics-containing fuels and advanced structural materials needed for a full-scale ABR; and (3) To support the research, development and demonstration required for certification of an ABR standard design by the U.S. Nuclear Regulatory Commission. The ABTR should also address the following additional objectives: (1) To incorporate and demonstrate innovative design concepts and features that may lead to significant improvements in cost, safety, efficiency, reliability, or other favorable characteristics that could promote public acceptance and future private sector investment in ABRs; (2) To demonstrate improved technologies for safeguards and security; and (3) To support development of the U.S. infrastructure for design, fabrication and construction, testing and deployment of systems, structures and components for the ABRs. Based on these objectives, a pre-conceptual design of a 250 MWt ABTR has been developed; it is documented in this report. In addition to meeting the primary and additional objectives listed above, the lessons learned from fast reactor programs in the U.S. and worldwide and the operating experience of more than a dozen fast reactors around the world, in particular the Experimental Breeder Reactor-II have been incorporated into the design of the ABTR to the extent possible.

  7. An Overview of Reactor Concepts, a Survey of Reactor Designs.

    Science.gov (United States)

    1985-02-01

    Public Affairs Office and is releasaole to the National Technical Information Services (NTIS). At NTIS, it will be available to the general public...Reactors that use deu- terium (heavy water) as a coolant can use natural uranium as a fuel. The * Canadian reactor, CANDU , utilizes this concept...reactor core at the top and discharged at the Dotton while the reactor is in operation. The discharged fuel can then b inspected to see if it can De used

  8. Conceptual design of inherently safe integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. I.; Chang, M. H.; Lee, D. J. and others

    1999-03-01

    The design concept of a 300 MWt inherently safe integral reactor(ISIR) for the propulsion of extra large and superhigh speed container ship was developed in this report. The scope and contents of this report are as follows : 1. The state of the art of the technology for ship-mounted reactor 2. Design requirements for ISIR 3. Fuel and core design 4. Conceptual design of fluid system 5. Conceptual design of reactor vessel assembly and primary components 6. Performance analyses and safety analyses. Installation of two ISIRs with total thermal power of 600MWt and efficiency of 21% is capable of generating shaft power of 126,000kW which is sufficient to power a container ship of 8,000TEU with 30knot cruise speed. Larger and speedier ship can be considered by installing 4 ISIRs. Even though the ISIR was developed for ship propulsion, it can be used also for a multi-purpose nuclear power plant for electricity generation, local heating, or seawater desalination by mounting on a movable floating barge. (author)

  9. Fusion reactor blanket/shield design study

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.; Clemmer, R.G.; Harkness, S.D.

    1979-07-01

    A joint study of tokamak reactor first-wall/blanket/shield technology was conducted by Argonne National Laboratory (ANL) and McDonnell Douglas Astronautics Company (MDAC). The objectives of this program were the identification of key technological limitations for various tritium-breeding-blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium-breeding-blanket concepts were evaluated according to the proposed coolant. The ANL effort concentrated on evaluation of lithium- and water-cooled blanket designs while the MDAC effort focused on helium- and molten salt-cooled designs. A joint effort was undertaken to provide a consistent set of materials property data used for analysis of all blanket concepts. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first-wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented.

  10. Jordan's First Research Reactor Project: Driving Forces, Present Status and the Way Ahead

    Energy Technology Data Exchange (ETDEWEB)

    Xoubi, Ned, E-mail: Ned@Xoubi.co [Jordan Atomic Energy Commission (JAEC), P.O.Box 70, Shafa Badran, 11934 Amman (Jordan)

    2011-07-01

    In a gigantic step towards establishing Jordan's nuclear power program, Jordan Atomic Energy Commission (JAEC) is building the first nuclear research and test reactor in the Kingdom. The new reactor will serve as the focal point for Jordan Center for Nuclear Research (JCNR), a comprehensive state of the art nuclear center not only for Jordan but for the whole region, the center will include in addition to the reactor a radioisotopes production plant, a nuclear fuel fabrication plant, a cold neutron source (CNS), a radioactive waste treatment facility, and education and training center. The JRTR reactor is the only research reactor new build worldwide in 2010, it is a 5 MW light water open pool multipurpose reactor, The reactor core is composed of 18 fuel assemblies, MTR plate type, with 19.75% enriched uranium silicide (U{sub 3}Si{sub 2}) in an aluminum matrix. It is reflected on all sides by beryllium and graphite blocks. Reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45x10{sup 14} cm{sup -2}s{sup -1}. The reactor reactivity is controlled by four Hafnium Control Absorber Rods (CAR). Jordan Center for Nuclear Research is located in Ramtha city, it is owned by Jordan Atomic Energy Commission (JAEC), and is contracted to Korea Atomic Energy Research Institute (KAERI) and Daewoo E and C. The JCNR project is a 56 months EPC fixed price contract for the design engineering, construction, and commissioning the JCNR reactor, and other nuclear facilities. The project presents many challenges for both the owner and the contractor, being the first nuclear reactor for Jordan, and the first nuclear export for Korea. The driving forces, present status and the way ahead will be presented in this paper. (author)

  11. The Traveling Wave Reactor: Design and Development

    Directory of Open Access Journals (Sweden)

    John Gilleland

    2016-03-01

    Full Text Available The traveling wave reactor (TWR is a once-through reactor that uses in situ breeding to greatly reduce the need for enrichment and reprocessing. Breeding converts incoming subcritical reload fuel into new critical fuel, allowing a breed-burn wave to propagate. The concept works on the basis that breed-burn waves and the fuel move relative to one another. Thus either the fuel or the waves may move relative to the stationary observer. The most practical embodiments of the TWR involve moving the fuel while keeping the nuclear reactions in one place−sometimes referred to as the standing wave reactor (SWR. TWRs can operate with uranium reload fuels including totally depleted uranium, natural uranium, and low-enriched fuel (e.g., 5.5% 235U and below, which ordinarily would not be critical in a fast spectrum. Spent light water reactor (LWR fuel may also serve as TWR reload fuel. In each of these cases, very efficient fuel usage and significant reduction of waste volumes are achieved without the need for reprocessing. The ultimate advantages of the TWR are realized when the reload fuel is depleted uranium, where after the startup period, no enrichment facilities are needed to sustain the first reactor and a chain of successor reactors. TerraPower's conceptual and engineering design and associated technology development activities have been underway since late 2006, with over 50 institutions working in a highly coordinated effort to place the first unit in operation by 2026. This paper summarizes the TWR technology: its development program, its progress, and an analysis of its social and economic benefits.

  12. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  13. Safety status of space radioisotope and reactor power sources

    Science.gov (United States)

    Bennett, Gary L.

    1990-01-01

    The current overall safety criterion for both radioisotope and reactor power sources is containment or immobilization in the case of a reentry accident. In addition, reactors are designed to remain subcritical under conditions of land impact or water immersion. A very extensive safety test and analysis program was completed on the radioisotope thermoelectric generators (RTGs) in use on the Galileo spacecraft and planned for use on the Ulysses spacecraft. The results of this work show that the RTGs will pose little or no risk for any credible accident. The SP-100 space nuclear reactor program has begun addressing its safety criteria, and the design is planned to be such as to ensure meeting the various safety criteria. Preliminary mission risk analyses on SP-100 show the expected value population dose from postulated accidents on the reference mission to be very small. It is concluded that the current US nuclear power sources are the safest flown.

  14. Status of neutron beam utilization at the Dalat nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dien, Nguyen Nhi; Hai, Nguyen Canh [Nuclear Research Institute, Dalat (Viet Nam)

    2003-03-01

    The 500-kW Dalat nuclear research reactor was reconstructed from the USA-made 250-kW TRIGA Mark II reactor. After completion of renovation and upgrading, the reactor has been operating at its nominal power since 1984. The reactor is used mainly for radioisotope production, neutron activation analysis, neutron beam researches and reactor physics study. In the framework of the reconstruction and renovation project of the 1982-1984 period, the reactor core, the control and instrumentation system, the primary and secondary cooling systems, as well as other associated systems were newly designed and installed by the former Soviet Union. Some structures of the reactor, such as the reactor aluminum tank, the graphite reflector, the thermal column, horizontal beam tubes and the radiation concrete shielding have been remained from the previous TRIGA reactor. As a typical configuration of the TRIGA reactor, there are four neutron beam ports, including three radial and one tangential. Besides, there is a large thermal column. Until now only two-neutron beam ports and the thermal column have been utilized. Effective utilization of horizontal experimental channels is one of the important research objectives at the Dalat reactor. The research program on effective utilization of these experimental channels was conducted from 1984. For this purpose, investigations on physical characteristics of the reactor, neutron spectra and fluxes at these channels, safety conditions in their exploitation, etc. have been carried out. The neutron beams, however, have been used only since 1988. The filtered thermal neutron beams at the tangential channel have been extracted using a single crystal silicon filter and mainly used for prompt gamma neutron activation analysis (PGNAA), neutron radiography (NR) and transmission experiments (TE). The filtered quasi-monoenergetic keV neutron beams using neutron filters at the piercing channel have been used for nuclear data measurements, study on

  15. NRC policy on future reactor designs

    Energy Technology Data Exchange (ETDEWEB)

    None

    1985-07-01

    On April 13, 1983, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation'' (48 FR 16014). This report presents and discusses the Commission's final version of that policy statement now entitled, ''Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants.'' It provides an overview of comments received from the public and the Advisory Committee on Reactor Safeguards and the staff response to these. In addition to the Policy Statement, the report discusses how the policies of this statement relate to other NRC programs including the Severe Accident Research Program; the implementation of safety measures resulting from lessons learned in the accident at Three Mile Island; safety goal development; the resolution of Unresolved Safety Issues and other Generic Safety Issues; and possible revisions of rules or regulatory requirements resulting from the Severe Accident Source Term Program. Also discussed are the main features of a generic decision strategy for resolving Regulatory Questions and Technical Issues relating to severe accidents; the development and regulatory use of new safety information; the treatment of uncertainty in severe accident decision making; and the development and implementation of a Systems Reliability Program for both existing and future plants to ensure that the realized level of safety is commensurate with the safety analyses used in regulatory decisions.

  16. NRC policy on future reactor designs

    Energy Technology Data Exchange (ETDEWEB)

    None

    1985-07-01

    On April 13, 1983, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Commission Policy Statement on Severe Accidents and Related Views on Nuclear Reactor Regulation'' (48 FR 16014). This report presents and discusses the Commission's final version of that policy statement now entitled, ''Policy Statement on Severe Reactor Accidents Regarding Future Designs and Existing Plants.'' It provides an overview of comments received from the public and the Advisory Committee on Reactor Safeguards and the staff response to these. In addition to the Policy Statement, the report discusses how the policies of this statement relate to other NRC programs including the Severe Accident Research Program; the implementation of safety measures resulting from lessons learned in the accident at Three Mile Island; safety goal development; the resolution of Unresolved Safety Issues and other Generic Safety Issues; and possible revisions of rules or regulatory requirements resulting from the Severe Accident Source Term Program. Also discussed are the main features of a generic decision strategy for resolving Regulatory Questions and Technical Issues relating to severe accidents; the development and regulatory use of new safety information; the treatment of uncertainty in severe accident decision making; and the development and implementation of a Systems Reliability Program for both existing and future plants to ensure that the realized level of safety is commensurate with the safety analyses used in regulatory decisions.

  17. FASTER Test Reactor Preconceptual Design Report

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-31

    The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  18. FASTER test reactor preconceptual design report summary

    Energy Technology Data Exchange (ETDEWEB)

    Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-02-29

    The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

  19. Composite nozzle design for reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Marlatt, G.R.; Allison, D.K.

    1984-01-24

    A composite nozzle is described for a fuel assembly adapted for installation on the upper or lower end thereof and which is constructed from two components. The first component includes a casting weldment or forging designed to carry handling loads, support fuel assembly weight and flow loads, and interface with structural members of both the fuel assembly and reactor internal structures. The second component of the nozzle consists of a thin stamped bore machine flow plate adapted for attachment to the casting body. The plate is designed to prevent fuel rods from being ejected from the core and provide orifices for coolant flow to a predetermined value and pressure drop which is consistent with the flow at other locations in the core.

  20. Status of reactor shielding research in the United States

    Energy Technology Data Exchange (ETDEWEB)

    Bartine, D.E.

    1983-01-01

    Shielding research in the United States continues to place emphasis on: (1) the development and refinement of shielding design calculational methods and nuclear data; and (2) the performance of confirmation experiments, both to evaluate specific design concepts and to verify specific calculational techniques and input data. The successful prediction of the radiation levels observed within the now-operating Fast Flux Test Facility (FFTF) has demonstrated the validity of this two-pronged approach, which has since been applied to US fast breeder reactor programs and is now being used to determine radiation levels and possible further shielding needs at operating light water reactors, especially under accident conditions. A similar approach is being applied to the back end of the fission fuel cycle to verify that radiation doses at fuel element storage and transportation facilities and within fuel reprocessing plants are kept at acceptable levels without undue economic penalties.

  1. Reactor and process design in sustainable energy technology

    CERN Document Server

    Shi, Fan

    2014-01-01

    Reactor Process Design in Sustainable Energy Technology compiles and explains current developments in reactor and process design in sustainable energy technologies, including optimization and scale-up methodologies and numerical methods. Sustainable energy technologies that require more efficient means of converting and utilizing energy can help provide for burgeoning global energy demand while reducing anthropogenic carbon dioxide emissions associated with energy production. The book, contributed by an international team of academic and industry experts in the field, brings numerous reactor design cases to readers based on their valuable experience from lab R&D scale to industry levels. It is the first to emphasize reactor engineering in sustainable energy technology discussing design. It provides comprehensive tools and information to help engineers and energy professionals learn, design, and specify chemical reactors and processes confidently. Emphasis on reactor engineering in sustainable energy techn...

  2. Design and performance of subgrade biogeochemical reactors.

    Science.gov (United States)

    Gamlin, Jeff; Downey, Doug; Shearer, Brad; Favara, Paul

    2017-02-18

    Subgrade biogeochemical reactors (SBGRs), also commonly referred to as in situ bioreactors, are a unique technology for treatment of contaminant source areas and groundwater plume hot spots. SBGRs have most commonly been configured for enhanced reductive dechlorination (ERD) applications for chlorinated solvent treatment. However, they have also been designed for other contaminant classes using alternative treatment media. The SBGR technology typically consists of removal of contaminated soil via excavation or large-diameter augers, and backfill of the soil void with gravel and treatment amendments tailored to the target contaminant(s). In most cases SBGRs include installation of infiltration piping and a low-flow pumping system (typically solar-powered) to recirculate contaminated groundwater through the SBGR for treatment. SBGRs have been constructed in multiple configurations, including designs capable of meeting limited access restrictions at heavily industrialized sites, and at sites with restrictions on surface disturbance due to sensitive species or habitat issues. Typical performance results for ERD applications include 85 to 90 percent total molar reduction of chlorinated volatile organic compounds (CVOCs) near the SBGR and rapid clean-up of adjacent dissolved contaminant source areas. Based on a review of the literature and CH2M's field-scale results from over a dozen SBGRs with a least one year of performance data, important site-specific design considerations include: 1) hydraulic residence time should be long enough for sufficient treatment but not too long to create depressed pH and stagnant conditions (e.g., typically between 10 and 60 days), 2) reactor material should balance appropriate organic mulch as optimal bacterial growth media along with other organic additives that provide bioavailable organic carbon, 3) a variety of native bacteria are important to the treatment process, and 4) biologically mediated generation of iron sulfides along with

  3. Design guide for Category III reactors: pool type reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-11-01

    The Department of Energy (DOE) in the ERDA Manual requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification, operation, maintenance, and decommissioning of DOE-owned reactors be in accordance with generally uniform standards, guides, and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirement of Category III reactor structures, components, and systems.

  4. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

    Brynda, W J; Lobner, P R; Powell, R W; Straker, E A

    1978-05-01

    The Department of Energy (DOE), in the ERDA Manual, requires that all DOE-owned reactors be sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that gives adequate consideration to health and safety factors. Specific guidance pertinent to the safety of DOE-owned reactors is found in Chapter 0540 of the ERDA Manual. The purpose of this Design Guide is to provide additional guidance to aid the DOE facility contractor in meeting the requirement that the siting, design, construction, modification operation, maintainance, and decommissioning of DOW-owned reactors be in accordance with generally uniform standards, guide and codes which are comparable to those applied to similar reactors licensed by the Nuclear Regulatory Commission (NRC). This Design Guide deals principally with the design and functional requirements of Category II reactor structure, components, and systems.

  5. Advanced Reactor Fuels Irradiation Experiment Design Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Chichester, Heather Jean MacLean [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hayes, Steven Lowe [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dempsey, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Harp, Jason Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

  6. Status of research and development on reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

  7. Development of mechanical design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Keun Bae; Choi, Suhn; Kim, Kang Soo; Kim, Tae Wan; Jeong, Kyeong Hoon; Lee, Gyu Mahn

    1999-03-01

    While Korean nuclear reactor strategy seems to remain focused on the large capacity power generation, it is expected that demand of small and medium size reactor will arise for multi-purpose application such as small capacity power generation, co-generation and sea water desalination. With this in mind, an integral reactor SMART is under development. Design concepts, system layout and types of equipment of integral reactor are significantly different from those of loop type reactor. Conceptual design development of mechanical structures of integral reactor SMART is completed through the first stage of the project. Efforts were endeavored for the establishment of design basis and evaluation of applicable codes and standards. Design and functional requirements of major structural components were setup, and three dimensional structural modelling of SMART reactor vessel assembly was prepared. Also, maintenance and repair scheme as well as preliminary fabricability evaluation were carried out. Since small integral reactor technology includes sensitive technologies and know-how's, it is hard to achieve systematic and comprehensive technology transfer from nuclear-advanced countries. Thus, it is necessary to develop the related design technology and to verify the adopted methodologies through test and experiments in order to assure the structural integrity of reactor system. (author)

  8. Reactor core design and characteristics of the Fugen

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Mitsuo; Kowata, Yasuki; Sugawara, Satoru; Deshimaru, Takehide

    1988-03-01

    The heavy water moderated, boiling light water cooled pressure tube type reactor Fugen uses plutonium-uranium mixed oxide as a fuel. Heavy water as the moderator and the light water of coolant are separated by the pressure tubes and calandria tubes. Thereby, the reactor core is heterogenes compared with that of LWRs. This paper describes the development of reactor core design procedure based on the feature of the Fugen type reactor, the feasibility test and the validity of nuclear and thermalhydraulic design based on the operating experience.

  9. Development status and potential program for development of proliferation-resistant molten-salt reactors

    Energy Technology Data Exchange (ETDEWEB)

    Engel, J.R.; Bauman, H.F.; Dearing, J.F.; Grimes, W.R.; McCoy, H.E. Jr.

    1979-03-01

    Preliminary studies of existing and conceptual molten-salt reactor (MSR) designs have led to the identification of conceptual systems that are technologically attractive when operated with denatured uranium as the principal fissile fuel. These denatured MSRs would also have favorable resource-utilization characteristics and substantial resistance to proliferation of weapons-usable nuclear materials. The report presents a summary of the current status of technology and a discussion of the major technical areas of a possible base program to develop commercial denatured MSRs. The general areas treated are (1) reactor design and development, (2) safety and safety related technology, (3) fuel-coolant behavior and fuel processing, and (4) reactor materials. A substantial development effort could lead to authorization for construction of a molten-salt test reactor about 5 years after the start of the program and operation of the unit about 10 years later. A prototype commercial denatured MSR could be expected to begin operating 25 years from the start of the program. The postulated base program would extend over 32 years and would cost about $700 million (1978 dollars, unescalated). Additional costs to construct the MSTR, $600 million, and the prototype commercial plant, $1470 million, would bring the total program cost to about $2.8 billion. Additional allowances probably should be made to cover contingencies and incidental technology areas not explicitly treated in this preliminary review.

  10. The Design of a Nuclear Reactor

    Indian Academy of Sciences (India)

    2016-09-01

    The aim of this largely pedagogical article is toemploy pre-college physics to arrive at an understanding of a system as complex as a nuclear reactor. We focus on three key issues: the fuelpin, the moderator, and lastly the dimensions ofthe nuclear reactor.

  11. The Status of the US High-Temperature Gas Reactors

    OpenAIRE

    2016-01-01

    In 2005, the US passed the Energy Policy Act of 2005 mandating the construction and operation of a high-temperature gas reactor (HTGR) by 2021. This law was passed after a multiyear study by national experts on what future nuclear technologies should be developed. As a result of the Act, the US Congress chose to develop the so-called Next-Generation Nuclear Plant, which was to be an HTGR designed to produce process heat for hydrogen production. Despite high hopes and expectations, the current...

  12. Parametric design study of tandem mirror fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Carlson, G.A.

    1977-05-27

    The parametric design study of the tandem mirror reactor (TMR) is described. The results of this study illustrate the variation of reactor characteristics with changes in the independent design parameters, reveal the set of design parameters which minimizes the cost of the reactor, and show the sensitivity of the optimized design to physics and technological uncertainties. The total direct capital cost of an optimized 1000 MWe TMR is estimated to be $1300/kWe. The direct capital cost of a 2000 MWe plant is less than $1000/kWe.

  13. Conceptual Design of a Nuclear Reactor Dedicated for Desalination

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yong Hun; Moon, Jang Sik; Jeong, Yong Hoon [Korea Adavanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    The many advantages of nuclear desalination, the nuclear safety issues still remain a perennial problem today. To respond to such needs, the development of a desalination-dedicated nuclear reactor with maximized safety features was proposed. From the feasibility study, the desalination-dedicated reactor was found to be a good solution for meeting future water demand during the winter season in some countries like UAE by decoupling water and electricity supply. The economic analysis results indicated that under certain conditions, the desalination-dedicated reactor can produce freshwater at lower cost than the target nuclear cogeneration reactor using steam extraction technologies. A conceptual design of the desalination-dedicated nuclear reactor is in progress. The design features of the desalination-dedicated nuclear reactor could significantly enhance safety, reliability, and simplicity, and facilitate the extensive use of innovative passive safety systems. These maximized safety features of desalination-dedicated reactor could provide advanced capabilities for passive reactor shutdown and residual heat removal, and eventually prevent radioactivity release into the environment. The conceptual design achieved will provide a foothold for the future commercialization of the desalination-dedicated nuclear reactor and eventually help to address both a serious water crisis and nuclear safety issues.

  14. DDE-MURR Status Report of Conceptual Design Activities

    Energy Technology Data Exchange (ETDEWEB)

    N.E. Woolstenhulme; R.B. Nielson; M.H. Sprenger; G.K. Housley

    2012-09-01

    The Design Demonstration Experiment for the University of Missouri Research Reactor (DDE-MURR) is intended to facilitate Low Enriched Uranium (LEU) conversion of the MURR by demonstrating the performance and fabrication of the LEU fuel element design through an irradiation test in a 200mm channel at the Belgium Reactor 2. At the time this report was prepared the resources for furthering DDE design work were expected to be postponed. As such, the conceptual design effort to date is summarized herein in order to provide the status of key objectives, notable results, and provisions for future design work. These demonstrate that the DDE-MURR design effort is well on the path to producing a suitable irradiation experiment, but also exhibits several challenges for which timely resolution is recommend in order to facilitate success of the irradiation campaign and ultimate conversion of the MURR.

  15. DDE-NBSR Status Report of Conceptual Design Activities

    Energy Technology Data Exchange (ETDEWEB)

    N.E. Woolstenhulme; R.B. Nielson; B.P. Durtschi; C.R. Glass; G.A. Roth; D.T. Clark

    2012-09-01

    The Design Demonstration Experiment for the National Bureau of Standard Reactor (DDE-NBSR) is intended to facilitate Low Enriched Uranium (LEU) conversion of the NBSR by demonstrating the performance and fabrication of the LEU fuel element design through an irradiation test in the Advanced Test Reactor center flux trap. At the time this report was prepared the resources for furthering DDE design work were expected to be postponed. As such, the conceptual design effort to date is summarized herein in order to provide the status of key objectives, notable results, and provisions for future design work. These demonstrate that the DDE-NBSR design effort is well on the path to producing a suitable irradiation experiment, but also exhibits several challenges for which timely resolution is recommend in order to facilitate success of the irradiation campaign and ultimate conversion of the NBSR.

  16. DDE-MITR Status Report of Conceptual Design Activities

    Energy Technology Data Exchange (ETDEWEB)

    N.E. Woolstenhulme; R.B. Nielson; J.D. Wiest; J.W. Nielsen; G.A. Roth; S.D. Snow

    2012-09-01

    The Design Demonstration Experiment for the Massachusetts Institute of Technology Reactor (DDE-MITR) is intended to facilitate Low Enriched Uranium (LEU) conversion of the MITR by demonstrating the performance and fabrication of the LEU fuel element design through an irradiation test in the Advanced Test Reactor center flux trap. At the time this report was prepared the resources for furthering DDE design work were expected to be postponed. As such, the conceptual design effort to date is summarized herein in order to provide the status of key objectives, notable results, and provisions for future design work. These demonstrate that the DDE-MITR design effort is well on the path to producing a suitable irradiation experiment, but also exhibits several challenges for which timely resolution is recommend in order to facilitate success of the irradiation campaign and ultimate conversion of the MITR.

  17. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  18. Design of A solar Thermophilic Anaerobic Reactor for Small Farms

    NARCIS (Netherlands)

    Mashad, El H.; Loon, van W.K.P.; Zeeman, G.; Bot, G.P.A.; Lettinga, G.

    2004-01-01

    A 10 m(3) completely stirred tank reactor has been designed for anaerobic treatment of liquid cow manure under thermophilic conditions (50degreesC), using a solar heating system mounted on the reactor roof. Simulation models for two systems have been developed. The first system consists of loose com

  19. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  20. Animal Guts as Ideal Reactors: An Open-Ended Project for a Course in Kinetics and Reactor Design.

    Science.gov (United States)

    Carlson, Eric D.; Gast, Alice P.

    1998-01-01

    Presents an open-ended project tailored for a senior kinetics and reactor design course in which basic reactor design equations are used to model the digestive systems of several animals. Describes the assignment as well as the results. (DDR)

  1. FRIB Front End Design Status

    CERN Document Server

    Pozdeyev, E; Machicoane, G; Morgan, G; Rao, X; Zhao, Q; Stovall, J; Vorozhtsov, S; Sun, L

    2013-01-01

    The Facility for Rare Isotope Beams (FRIB) will provide a wide range of primary ion beams for nuclear physics research with rare isotope beams. The FRIB SRF linac will be capable of accelerating medium and heavy ion beams to energies beyond 200 MeV/u with a power of 400 kW on the fragmentation target. This paper presents the status of the FRIB Front End designed to produce uranium and other medium and heavy mass ion beams at world-record intensities. The paper describes the FRIB high performance superconducting ECR ion source, the beam transport designed to transport two-charge state ion beams and prepare them for the injection in to the SRF linac, and the design of a 4-vane 80.5 MHz RFQ. The paper also describes the integration of the front end with other accelerator and experimental systems.

  2. U.S. Nuclear Power Reactor Plant Status

    Data.gov (United States)

    Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

  3. Present status and future perspectives of research and test reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-10-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  4. DDE Design Status Report Nov 2011

    Energy Technology Data Exchange (ETDEWEB)

    N.E. Woolstenhulme; R.B. Nielson

    2011-11-01

    The National Nuclear Security Agency Global Threat Reduction Initiative employs the Reduced Enrichment for Research and Test Reactors (RERTR) Fuel Development program to facilitate maturation of Low Enriched Uranium (LEU) fuel technology in order to enable conversion of High Power Research Reactors (HPRR) to LEU fuels. The RERTR Fuel Development program has overseen design, fabrication, irradiation, and examination of numerous tests on small to medium sized specimens containing LEU fuels. To enable the three nearest term HPRR conversions, including the Massachusetts Institutes of Technology Reactor (MITR), University of Missouri Research Reactor (MURR), and National Bureau of Standard Reactor (NBSR), the FD pillar is currently focused on qualification of the 'Base Monolithic Design'. The Base Monolithic Design consists of uranium-10 wt% molybdenum alloy (U-10Mo) in the form of a monolithic foil, with thin zirconium interlayers, clad in aluminum by hot isostatic press. The licensing basis of the aforementioned HPRR's restricts them from testing lead test elements of their respective LEU fuel element designs. In order to provide the equivalent of a lead test assembly, one Design Demonstration Experiment (DDE) is planned for each of the three NRC licensed reactors.

  5. Improved Pyrolysis Micro reactor Design via Computational Fluid Dynamics Simulations

    Science.gov (United States)

    2017-05-23

    NUMBER (Include area code) 23 May 2017 Briefing Charts 25 April 2017 - 23 May 2017 Improved Pyrolysis Micro-reactor Design via Computational Fluid... PYROLYSIS MICRO-REACTOR DESIGN VIA COMPUTATIONAL FLUID DYNAMICS SIMULATIONS Ghanshyam L. Vaghjiani* DISTRIBUTION A: Approved for public release...Approved for public release, distribution unlimited. PA Clearance 17247 Chen-Source (>240 references from SciFinder as of 5/1/17): Flash pyrolysis

  6. Preliminary Reactor Head Bolt Design of Prototype Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Han, Insu; Koo, Gyeonghoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    As structural requirements, the reactor head is designed to withstand all of the pressure, temperatures and forces which are likely to be imposed on it. The bolts that fasten the head to the vessel flange. Design of the reactor head bolts so as to withstand the loads applied should be designed. Currently, preliminary design of the PGSFR reactor bolts is progressed. So far, we have designed and evaluated example. The number and cross-sectional areas of bolts were determined using the procedure given in ASME BPVC Section III, Division 1, Appendix E. The purpose of this study is to conduct design the number and cross-sectional area of bolts attaching the PGSFR reactor head to the reactor vessel, using the ASME procedure. In this paper, preliminary bolt design for PGSFR was carried out according to the ASME procedure. Detailed calculations were carried out for bolt root diameter = 80 mm and number of bolts Nb = 45. It should be noted that the seating pressure recommended in the ASME code is only a suggested value, not mandatory appendix E. It does not guarantee a leak-tight joint. So these quantities are needed to carry out fatigue analysis of the bolts and to assure leak tightness of the joint during operation. For the future work, the fatigue and seismic analysis will be performed.

  7. Status of the NGNP Fuel Experiment AGR-2 Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Blaine Grover

    2012-10-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven 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). 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 several years to demonstrate and qualify new TRISO coated 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 was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and support systems will be briefly discussed, followed by the progress and status of the experiment to date.

  8. Design and development of small and medium integral reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Chang, M. H.; Lee, C. C.; Song, J. S.; Cho, B. O.; Kim, K. Y.; Kim, S. J.; Park, S. Y.; Lee, K. B.; Lee, C. H.; Chun, T. H.; Oh, D. S.; In, W. K.; Kim, H. K.; Lee, C. B.; Kang, H. S.; Song, K. N.

    1997-07-01

    Recently, the role of small and medium size integral reactors is remarkable in the heat applications rather than the electrical generations. Such a range of possible applications requires extensive used of inherent safety features and passive safety systems. It also requires ultra-longer cycle operations for better plant economy. Innovative and evolutionary designs such as boron-free operations and related reactor control methods that are necessary for simple reactor system design are demanded for the small and medium reactor (SMR) design, which are harder for engineers to implement in the current large size nuclear power plants. The goals of this study are to establish preliminary design criteria, to perform the preliminary conceptual design and to develop core specific technology for the core design and analysis for System-integrated Modular Advanced ReacTor (SMART) of 330 MWt power. Based on the design criteria of the commercial PWR`s, preliminary design criteria will be set up. Preliminary core design concept is going to be developed for the ultra-longer cycle and boron-free operation and core analysis code system is constructed for SMART. (author). 100 refs., 40 tabs., 92 figs.

  9. Civilian Power Program. Part 1, Summary, Current status of reactor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Author, Not Given

    1959-09-01

    This study group covered the following: delineation of the specific objectives of the overall US AEC civilian power reactor program, technical objectives of each reactor concept, preparation of a chronological development program for each reactor concept, evaluation of the economic potential of each reactor type, a program to encourage the the development, and yardsticks for measuring the development. Results were used for policy review by AEC, program direction, authorization and appropriation requests, etc. This evaluation encompassed civilian power reactors rated at 25 MW(e) or larger and related experimental facilities and R&D. This Part I summarizes the significant results of the comprehensive effort to determine the current technical and economic status for each reactor concept; it is based on the 8 individual technical status reports (Part III).

  10. Conceptual design of Indian molten salt breeder reactor

    Indian Academy of Sciences (India)

    P K Vijayan; A Basak; I V Dulera; K K Vaze; S Basu; R K Sinha

    2015-09-01

    The third stage of Indian nuclear power programme envisages the use of thorium as the fertile material with 233U, which would be obtained from the operation of Pu/Th-based fast reactors in the later part of the second stage. Thorium-based reactors have been designed in many configurations, from light water-cooled designs to high-temperature liquid metal-cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian molten salt breeder reactor (IMSBR). Presently, various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel, fundamental studies on natural circulation and corrosion behaviour of various molten salts have also been initiated.

  11. Design activities on helical DEMO reactor FFHR-d1

    Energy Technology Data Exchange (ETDEWEB)

    Sagara, A., E-mail: sagara.akio@LHD.nifs.ac.jp [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Goto, T.; Miyazawa, J.; Yanagi, N.; Tanaka, T.; Tamura, H.; Sakamoto, R.; Tanaka, M.; Tsumori, K. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan); Mitarai, O. [Tokai University, 9-1-1 Toroku, Kumamoto 862-8652 (Japan); Imagawa, S.; Muroga, T. [National Institute for Fusion Science, Toki, Gifu 509-5292 (Japan)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer Conceptual design studies of the helical DEMO reactor FFHR-d1 have been conducted. Black-Right-Pointing-Pointer Design window analyses with core plasma and engineering designs have been carried out. Black-Right-Pointing-Pointer R and Ds on blanket, magnet, tritium control, fuelling and heating systems are discussed. - Abstract: Based on high-density and high-temperature plasma experiments in the large helical device (LHD), conceptual design studies of the LHD-type helical DEMO reactor FFHR-d1 have been conducted by integrating wide-ranged R and D activities on core plasmas and reactor technologies through cooperative researches under the fusion engineering research project, which has been launched newly in NIFS. Current activities for the FFHR-d1 in this project are presented on design window analyses with designs on core plasma, neutronics for liquid blankets, continuous helical magnets, pellet fueling, tritium systems and plasma heating devices.

  12. Dismantling design for the loop rooms on the MR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Craig, D.; Fecitt, L. [NUKEM Limited, Dounreay (United Kingdom); Gorlinsky, Yu.E. [RRC Kurchatov Institute, Moscow (Russian Federation); Harman, N.F.; Jackson, R. [Serco Technical and Assurance Services, Warrington (United Kingdom); Kolyadin, V.I. [RRC Kurchatov Institute, Moscow (Russian Federation); Lobach, Yu.N., E-mail: lobach@kinr.kiev.u [Institute for Nuclear Research of NASU, pr.Nauki, 47, 03680 Kiev (Ukraine); Pavlenko, V.I. [RRC Kurchatov Institute, Moscow (Russian Federation)

    2009-12-15

    The recently completed international co-operation project was aimed at planning for decommissioning the MR reactor identified as a pilot plant for the decommissioning of the other shutdown reactors on the site. The MR reactor was a pool-type, materials testing reactor with the total thermal power of 50 MW which incorporated pressure tubes containing fuel under test. The MR facility includes the reactor with its nine loop rig rooms containing pumps, heat exchangers and experimental equipment as well as systems and equipment located in other buildings in the complex. The objective of the MR reactor decommissioning project was to identify dismantling equipment and the decommissioning methodology for the reactor, loop rooms and redundant services to permit the refit and re-use of the building for a different nuclear related purpose. The dismantling design comprises two separate, but combined, tasks, namely, the dismantling of reactor installation itself and dismantling of experimental loops. The techniques proposed to undertake the dismantling operations within the loop rooms are described. Two options have been developed for removing contaminated equipment from the high radiation field loop rooms and packaging the waste into approved waste containers. The benefits and detriments of both methods have been identified, which allows implementing the safe, timely and cost-effective decommissioning.

  13. Reference design for the standard mirror hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bender, D.J.; Fink, J.H.; Galloway, T.R.; Kastenberg, W.E.; Lee, J.D.; Devoto, R.S.; Neef, W.S. Jr.; Schultz, K.R.; Culver, D.W.; Rao, S.B.; Rao, S.R.

    1978-05-22

    This report describes the results of a two-year study by Lawrence Livermore Laboratory and General Atomic Co. to develop a conceptual design for the standard (minimum-B) mirror hybrid reactor. The reactor parameters have been chosen to minimize the cost of producing nuclear fuel (/sup 239/Pu) for consumption in fission power reactors (light water reactors). The deuterium-tritium plasma produces approximately 400 MW of fusion power with a plasma Q of 0.64. The fast-fission blanket, which is fueled with depleted uranium and lithium, generates sufficient tritium to run the reactor, has a blanket energy multiplication of M = 10.4, and has a net fissile breeding ratio of Pu/n = 1.51. The reactor has a net electrical output of 600 MWe, a fissile production of 2000 kg of plutonium per year (at a capacity factor of 0.74), and a net plant efficiency of 0.18. The plasma-containment field is generated by a Yin-Yang magnet using NbTi superconductor, and the neutral beam system uses positive-ion acceleration with beam direct conversion. The spherical blanket is based on gas-cooled fast reactor technology. The fusion components, blanket, and primary heat-transfer loop components are all contained within a prestressed-concrete reactor vessel, which provides magnet restraint and supports the primary heat-transfer loop and the blanket.

  14. The current status of nuclear research reactor in Thailand

    Energy Technology Data Exchange (ETDEWEB)

    Sittichai, C.; Kanyukt, R.; Pongpat, P. [Office of Atomic Energy for Peace, Bangkok (Thailand)

    1998-10-01

    Since 1962, the Thai Research Reactor has been serving for various kinds of activities i.e. the production of radioisotopes for medical uses and research and development on nuclear science and technology, for more than three decades. The existing reactor site should be abandoned and relocated to the new suitable site, according to Thai cabinet`s resolution on the 27 December 1989. The decommissioning project for the present reactor as well as the establishment of new nuclear research center were planned. This paper discussed the OAEP concept for the decommissioning programme and the general description of the new research reactor and some related information were also reported. (author)

  15. Design requirement for electrical system of an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

  16. Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game

    Science.gov (United States)

    Orbey, Nese; Clay, Molly; Russell, T.W. Fraser

    2014-01-01

    An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…

  17. Introduction to Chemical Engineering Reactor Analysis: A Web-Based Reactor Design Game

    Science.gov (United States)

    Orbey, Nese; Clay, Molly; Russell, T.W. Fraser

    2014-01-01

    An approach to explain chemical engineering through a Web-based interactive game design was developed and used with college freshman and junior/senior high school students. The goal of this approach was to demonstrate how to model a lab-scale experiment, and use the results to design and operate a chemical reactor. The game incorporates both…

  18. SABR fusion-fission hybrid transmutation reactor design concept

    Science.gov (United States)

    Stacey, Weston

    2009-11-01

    A conceptual design has been developed for a sub-critical advanced burner reactor (SABR) consisting of i) a sodium cooled fast reactor fueled with the transuranics (TRU) from spent nuclear fuel, and ii) a D-T tokamak fusion neutron source based on ITER physics and technology. Subcritical operation enables more efficient transmutation fuel cycles in TRU fueled reactors (without compromising safety), which may be essential for significant reduction in high-level waste repository requirements. ITER will serve as the prototype for the fusion neutron source, which means SABRs could be implemented to help close the nuclear fuel cycle during the 2^nd quarter of the century.

  19. Modelling and control design for SHARON/Anammox reactor sequence

    DEFF Research Database (Denmark)

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work presents a complete model of the SHARON/Anammox reactor sequence. The dynamics of the reactor were explored pointing out the different scales of the rates in the system: slow microbial...... metabolism against fast chemical reaction and mass transfer. Likewise, the analysis of the dynamics contributed to establish qualitatively the requirements for control of the reactors, both for regulation and for optimal operation. Work in progress on quantitatively analysing different control structure...

  20. Light Water Reactor Sustainability Program Status of Silicon Carbide Joining Technology Development

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton

    2013-09-01

    Advanced, accident tolerant nuclear fuel systems are currently being investigated for potential application in currently operating light water reactors (LWR) or in reactors that have attained design certification. Evaluation of potential options for accident tolerant nuclear fuel systems point to the potential benefits of silicon carbide (SiC) relative to Zr-based alloys, including increased corrosion resistance, reduced oxidation and heat of oxidation, and reduced hydrogen generation under steam attack (off-normal conditions). If demonstrated to be applicable in the intended LWR environment, SiC could be used in nuclear fuel cladding or other in-core structural components. Achieving a SiC-SiC joint that resists corrosion with hot, flowing water, is stable under irradiation and retains hermeticity is a significant challenge. This report summarizes the current status of SiC-SiC joint development work supported by the Department of Energy Light Water Reactor Sustainability Program. Significant progress has been made toward SiC-SiC joint development for nuclear service, but additional development and testing work (including irradiation testing) is still required to present a candidate joint for use in nuclear fuel cladding.

  1. Decontamination and decommissioning project status of the TRIGA Mark-II and III reactors in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Paik, S. T.; Park, S. K.; Chung, K. W.; Chung, U. S.; Jung, K. J. [TRIGA Research Reactor D and D Project Team, Korea Atomic Energy Research Insitutute, Taejon (Korea, Republic of)

    1999-07-01

    The decontamination and decommissioning (D and D) project of the TRIGA Mark-II and Mark-III was started in January 1997, after their shutdown in 1995 due to their life and the operation of a new research reactor, HANARO, at the KAERI site in Taejon. Preparation of the decommissioning plan and environmental impact assessment, and setting up of licensing procedure and documentation for the project were performed in 1997. At the end of 1997, Hyundai Engineering Company (HEC) was selected as the main contractor to do design and licensing documentation for the D and D of both reactors. British Nuclear Fuels Plc. (BNFL) was the technical assisting partner to Heck. Licensing documents were submitted to the Ministry of Science and Technology (MOST) at the end of 1998. And the Korea Institute of Nuclear Safety (KINS) is reviewing the documents. Practical work of the D and D will start at the end of 1999 upon the government issues the license. In the meantime, July 1998, all spent fuels from the TRIGA Mark-II and III were safely transported to the US. The foremost part of the D and D work will be the TRIGA Mark-III reactor hall that will be used as a temporary storage of radioactive waste produced during the D and D work, and followed by the TRIGA Mark-II and auxiliary facilities. This paper summarizes the current status and future plans for the D and D work. (author)

  2. Design Concept of Advanced Sodium-Cooled Fast Reactor and Related R&D in Korea

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

    Full Text Available Korea imports about 97% of its energy resources due to a lack of available energy resources. In this status, the role of nuclear power in electricity generation is expected to become more important in future years. In particular, a fast reactor system is one of the most promising reactor types for electricity generation, because it can utilize efficiently uranium resources and reduce radioactive waste. Acknowledging the importance of a fast reactor in a future energy policy, the long-term advanced SFR development plan was authorized by KAEC in 2008 and updated in 2011 which will be carried out toward the construction of an advanced SFR prototype plant by 2028. Based upon the experiences gained during the development of the conceptual designs for KALIMER, KAERI recently developed advanced sodium-cooled fast reactor (SFR design concepts of TRU burner that can better meet the generation IV technology goals. The current status of nuclear power and SFR design technology development program in Korea will be discussed. The developments of design concepts including core, fuel, fluid system, mechanical structure, and safety evaluation have been performed. In addition, the advanced SFR technologies necessary for its commercialization and the basic key technologies have been developed including a large-scale sodium thermal-hydraulic test facility, super-critical Brayton cycle system, under-sodium viewing techniques, metal fuel development, and developments of codes, and validations are described as R&D activities.

  3. Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    S. Blaine Grover; David A. Petti; Michael E. Davenport

    2013-07-01

    The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven 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). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the 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 several independent 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 was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is

  4. Design of slurry reactor for indirect liquefaction applications

    Energy Technology Data Exchange (ETDEWEB)

    Prakash, A.; Bendale, P.G.

    1991-01-01

    The objective of this project is to design and model a conceptual slurry reactor for two indirect liquefaction applications; (1) production of methanol and (2) production of hydrocarbon fuels via Fischer-Tropsch route. A slurry reactor is defined here as a three-phase bubble column reactor using a fine catalyst particle suspension in a high molecular weight liquid. The feed gas is introduced through spargers. It then bubbles through the column providing the agitation necessary for catalyst suspension and mass transfer. The reactor models for the two processes have been formulated using computer simulation. Process data, kinetic and thermodynamic data, heat and mass transfer data and hydrodynamic data have been used in the mathematical models to describe the slurry reactor for each of the two processes. Available data from process development units and demonstration units were used to test and validate the models. Commercial size slurry reactors for methanol and Fischer-Tropsch synthesis were sized using reactor models developed in this report.

  5. Design of slurry reactor for indirect liquefaction applications. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Prakash, A.; Bendale, P.G.

    1991-12-31

    The objective of this project is to design and model a conceptual slurry reactor for two indirect liquefaction applications; (1) production of methanol and (2) production of hydrocarbon fuels via Fischer-Tropsch route. A slurry reactor is defined here as a three-phase bubble column reactor using a fine catalyst particle suspension in a high molecular weight liquid. The feed gas is introduced through spargers. It then bubbles through the column providing the agitation necessary for catalyst suspension and mass transfer. The reactor models for the two processes have been formulated using computer simulation. Process data, kinetic and thermodynamic data, heat and mass transfer data and hydrodynamic data have been used in the mathematical models to describe the slurry reactor for each of the two processes. Available data from process development units and demonstration units were used to test and validate the models. Commercial size slurry reactors for methanol and Fischer-Tropsch synthesis were sized using reactor models developed in this report.

  6. High flux isotope reactor cold source preconceptual design study report

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Bucholz, J.A.; Burnette, S.E. [and others

    1995-12-01

    In February 1995, the deputy director of Oak Ridge National Laboratory (ORNL) formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced Neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. The anticipated cold source will consist of a cryogenic LH{sub 2} moderator plug, a cryogenic pump system, a refrigerator that uses helium gas as a refrigerant, a heat exchanger to interface the refrigerant with the hydrogen loop, liquid hydrogen transfer lines, a gas handling system that includes vacuum lines, and an instrumentation and control system to provide constant system status monitoring and to maintain system stability. The scope of this project includes the development, design, safety analysis, procurement/fabrication, testing, and installation of all of the components necessary to produce a working cold source within an existing HFIR beam tube. This project will also include those activities necessary to transport the cold neutron beam to the front face of the present HFIR beam room. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and research and development (R and D), (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the preconceptual phase and establishes the concept feasibility. The information presented includes the project scope, the preliminary design requirements, the preliminary cost and schedule, the preliminary performance data, and an outline of the various plans for completing the project.

  7. Status of the Prediction of Reactor Anti-neutrino Spectra

    Science.gov (United States)

    Fallot, Muriel

    2015-04-01

    New generation neutrino physics experiments at reactors have recently determined the value of the θ13 mixing angle. Even though their principle is to use multiple detectors allowing to minimize the influence of reactor and nuclear physics ingredients on their results, these ingredients cannot be totally eliminated. They include reactor simulations, but also new computations of reactor anti-neutrino energy spectra. Recently, after a new computation of the reactor anti-neutrino energy spectra, based on the conversion of integral data of the beta spectra from 235U, and 239;241Pu, a deficit of reactor anti-neutrinos measured by short baseline experiments was pointed out. This is called the reactor anomaly, a new puzzle in the neutrino physics area. Since then, numerous new experimental neutrino projects have emerged. In parallel, computations of the anti-neutrino spectra independent from the ILL data would be desirable. One possibility is the use of the summation method, summing all the contributions of the fission product beta decay branches that can be found in nuclear databases. Studies have shown that in order to obtain reliable summation anti-neutrino energy spectra, new nuclear physics measurements of selected fission product beta decay properties are required. Lately, the first integral measurement of the beta spectrum associated to fast fission of 238U has been performed. Even more recently, the question of the influence of forbidden decays in the determination of reactor anti-neutrino energy spectrum has been raised. At this conference, we will present the methods used to compute reactor anti-neutrino energy spectra, the recent published developments on the topic, remaining open questions and some experimental outlooks.

  8. Summary of the Advanced Reactor Design Criteria (ARDC) Phase 2 Activities

    Energy Technology Data Exchange (ETDEWEB)

    Holbrook, Mark Raymond [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    This report provides an end-of-year summary reflecting the progress and status of proposed regulatory design criteria for advanced non-LWR designs in accordance with the Level 3 milestone in M3AT-15IN2001017 in work package AT-15IN200101. These criteria have been designated as ARDC, and they provide guidance to future applicants for addressing the GDC that are currently applied specifically to LWR designs. The report provides a summary of Phase 2 activities related to the various tasks associated with ARDC development and the subsequent development of example adaptations of ARDC for Sodium Fast Reactor (SFR) and modular High Temperature Gas-cooled Reactor (HTGR) designs.

  9. Current status, research progress and future plan of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sardjono, Y.; Syarip; Tjiptono, T.W. [Yogyakarta Nuclear Research Center, Batan (Indonesia)

    1999-10-01

    The current status, research progress and future plan of the Kartini Research Reactor (KRR) is presented. The measurements of axial burn-up distributions for each fuel element by gamma scanning techniques, core axial power distribution display, fuel management for safeguards purpose as well as some research progress activities i.e.; utilization of beamport for: neutron radiography, application neutron activation analysis and history record of KRR power operations is also presented. The KRR is 100 kW pool water reactor type which uses natural circulation and provided by: five beamports in which one of them already coupled with natural uranium subcritical assembly, two thermalizing columns in which one of them is prepared for developing Boron Neutron Capture Therapy (BNCT), two rabbit systems utilized for special analysis uranium ore by delayed neutron counting techniques, one center timbre and 40 irradiation rack (lazy susan) for neutron activation analysis. The KRR was constructed as a second research reactor in Indonesia with special purpose for training and education, high safety margin with involve in high negative temperature coefficient which achieved its first criticality on January 25, 1979. The maximum power level on first criticality is 50 kW and since August 1981 up to now is operating 100 kW. Base on the KRR design limit, it is planned to increase the power level up to 250 kW in the future plan. The preliminary activities such as Non Destructive Testing (NDT) for some reactor components especially water tank and thermal column should be done before decided to increase power level. (author)

  10. Design of megawatt power level heat pipe reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Dasari, Venkateswara Rao [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Reid, Robert Stowers [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-12

    An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors. The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.

  11. Westinghouse Small Modular Reactor nuclear steam supply system design

    Energy Technology Data Exchange (ETDEWEB)

    Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

    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 first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

  12. Basis for NGNP Reactor Design Down-Selection

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2011-11-01

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  13. Basis for NGNP Reactor Design Down-Selection

    Energy Technology Data Exchange (ETDEWEB)

    L.E. Demick

    2010-08-01

    The purpose of this paper is to identify the extent of technology development, design and licensing maturity anticipated to be required to credibly identify differences that could make a technical choice practical between the prismatic and pebble bed reactor designs. This paper does not address a business decision based on the economics, business model and resulting business case since these will vary based on the reactor application. The selection of the type of reactor, the module ratings, the number of modules, the configuration of the balance of plant and other design selections will be made on the basis of optimizing the Business Case for the application. These are not decisions that can be made on a generic basis.

  14. Prevention of ageing of research reactors by design

    Energy Technology Data Exchange (ETDEWEB)

    Boado, J. [INVAP, Bariloche (Argentina); Lolich, J. [INVAP, Bariloche (Argentina)

    1995-12-31

    However, it is our experience as designers and builders of research reactors, that the most important cause of ageing of any experimental installation, is the loss of motivation of the personnel involved in the operation and maintenance, when the objectives for the utilisation of the facility change or research programs are abandoned for whatever reason. We therefore have endeavoured to design and construct research reactors with several engineering features and with an untraditional approach to the training of the future operator of the facility. During all phases of the design and construction of the reactor, we develop in the future operator of the facility the capacity, not only to operate it properly, but to innovate and to adapt the installation to the daily operating problems due to new requirements and options that might not have been foreseen when the facility was ordered. The versatility of the operator is thus a further guarantee against ageing by abandonment. (orig./HP)

  15. Development of core design and analyses technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  16. Review of the status of low power research reactors and considerations for its development

    Energy Technology Data Exchange (ETDEWEB)

    Lim, In Cheol; Wu, Sang Ik; Lee, Byung Chul; Ha, Jae Joo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    At present, 232 research reactors in the world are in operation and two thirds of them have a power less than 1 MW. Many countries have used research reactors as the tools for educating and training students or engineers and for scientific service such as neutron activation analysis. As the introduction of a research reactor is considered a stepping stone for a nuclear power development program, many newcomers are considering having a low power research reactor. The IAEA has continued to provide forums for the exchange of information and experiences regarding low power research reactors. Considering these, the Agency is recently working on the preparation of a guide for the preparation of technical specification possibly for a member state to use when wanting to purchase a low power research reactor. In addition, ANS has stated that special consideration should be given to the continued national support to maintain and expand research and test reactor programs and to the efforts in identifying and addressing the future needs by working toward the development and deployment of next generation nuclear research and training facilities. Thus, more interest will be given to low power research reactors and its role as a facility for education and training. Considering these, the status of low power research reactors was reviewed, and some aspects to be considered in developing a low power research reactor were studied.

  17. Designed porosity materials in nuclear reactor components

    Science.gov (United States)

    Yacout, A. M.; Pellin, Michael J.; Stan, Marius

    2016-09-06

    A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

  18. Designed porosity materials in nuclear reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Yacout, A. M.; Pellin, Michael J.; Stan, Marius

    2016-09-06

    A nuclear fuel pellet with a porous substrate, such as a carbon or tungsten aerogel, on which at least one layer of a fuel containing material is deposited via atomic layer deposition, and wherein the layer deposition is controlled to prevent agglomeration of defects. Further, a method of fabricating a nuclear fuel pellet, wherein the method features the steps of selecting a porous substrate, depositing at least one layer of a fuel containing material, and terminating the deposition when the desired porosity is achieved. Also provided is a nuclear reactor fuel cladding made of a porous substrate, such as silicon carbide aerogel or silicon carbide cloth, upon which layers of silicon carbide are deposited.

  19. Design Features of Regional Energy Reactor, REX-10

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Won; Lee, Yeon Gun; Lim, Sung Won; Park, Goon Cherl [Seoul National Univ., Seoul (Korea, Republic of); Joo, Hyeong Min; Jang, Byeong Il [Hanyang Univ., Seoul (Korea, Republic of); Kim, Moo Hwan [POSTECH, Pohang (Korea, Republic of)

    2009-05-15

    Recently, today's global pattern of energy supply is not sustainable. The provision of affordable energy services is a fundamental prerequisite for economic growth and development. In addition, the environmental problems such as energy crisis, global warming and acid rain are issued and more demand for reliable electricity supply increases. As one of the realistic solutions, the extension of the peaceful uses of nuclear energy has been suggested. Small and medium nuclear reactors with non-electric applications arise as an alternative energy source. The main non-electric applications are defined as district heating, desalination (of sea, brackish and waste water), industrial heat supply, ship propulsion and the energy supply for spacecraft. RERI (Regional Energy Research Institute for the Next Generation) is to develop a small-scale electric power system by an environmentally-friendly and stable small nuclear reactor. The newly designed REX-10 (Regional Energy Reactor, 10MWth) has been developed to maintain system safety in order to be placed in a densely populated region, or island, etc. The REX-10 reactor system was designed based on SMART, however, the operation mode and the system pressure and capacity were determined properly for a regional energy reactor. In this study, the design characteristics of REX-10 and the related researches will be introduced.

  20. Current status of neutron activation analysis in HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min [Korea Atomic Energy Research Institute, Daejeon (Korea)

    2003-03-01

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10{sup 13} - 1 x 10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  1. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

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

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; K. Hamman

    2009-09-01

    Suggested for Track 7: Advances in Reactor Core Design and In-Core Management _____________________________________________________________________________________ Fast Reactor Subassembly Design Modifications for Increasing Electricity Generation Efficiency R. Wigeland and K. Hamman Idaho National Laboratory Given the ability of fast reactors to effectively transmute the transuranic elements as are present in spent nuclear fuel, fast reactors are being considered as one element of future nuclear power systems to enable continued use and growth of nuclear power by limiting high-level waste generation. However, a key issue for fast reactors is higher electricity cost relative to other forms of nuclear energy generation. The economics of the fast reactor are affected by the amount of electric power that can be produced from a reactor, i.e., the thermal efficiency for electricity generation. The present study is examining the potential for fast reactor subassembly design changes to improve the thermal efficiency by increasing the average coolant outlet temperature without increasing peak temperatures within the subassembly, i.e., to make better use of current technology. Sodium-cooled fast reactors operate at temperatures far below the coolant boiling point, so that the maximum coolant outlet temperature is limited by the acceptable peak temperatures for the reactor fuel and cladding. Fast reactor fuel subassemblies have historically been constructed using a large number of small diameter fuel pins contained within a tube of hexagonal cross-section, or hexcan. Due to this design, there is a larger coolant flow area next to the hexcan wall as compared to flow area in the interior of the subassembly. This results in a higher flow rate near the hexcan wall, overcooling the fuel pins next to the wall, and a non-uniform coolant temperature distribution. It has been recognized for many years that this difference in sodium coolant temperature was detrimental to achieving

  3. Design of a Compact and Versatile Bench Scale Tubular Reactor

    Directory of Open Access Journals (Sweden)

    R. Prasad

    2009-06-01

    Full Text Available A compact and versatile laboratory tubular reactor has been designed and fabricated keeping in view of reducing capital cost and minimising energy consumption for gas/vapor-phase heterogeneous catalytic reactions. The reactor is consisted of two coaxial corning glass tubes with a helical coil of glass tube in between the coaxial tubes serving as vaporiser and pre-heater, the catalyst bed is in the inner tube. A schematic diagram of the reactor with detailed dimensions and working principles are described. The attractive feature of the reactor is that the vaporiser, pre-heater and fixed bed reactor are merged in a single compact unit. Thus, the unit minimises separate vaporiser and pre-heater, also avoids separate furnaces used for them and eliminate auxiliary instrumentation such as temperature controller etc. To demonstrate the system operation and illustrate the key features, catalyst screening data and the efficient collection of complete, and accurate intrinsic kinetic data are provided for oxidation of CO over copper chromite catalyst. CO oxidation is an important reaction for auto-exhaust pollution control. The suitability of the versatile nature of the reactor has been ascertained for catalytic reactions where either volatile or vaporizable feeds can be introduced to the reaction zone, e.g. oxidation of iso-octane, reduction of nitric oxide, dehydrogenation of methanol, ethanol and iso-propanol, hydrogenation of nitrobenzene to aniline, etc. Copyright (c 2009 by BCREC. All Rights reserved.[Received: 10 February 2009, Accepted: 9 May 2009][How to Cite: R. Prasad, G. Rattan. (2009. Design of a Compact and Versatile Bench Scale Tubular Reactor. Bulletin of Chemical Reaction Engineering and Catalysis, 4(1: 5-9.  doi:10.9767/bcrec.4.1.1250.5-9][How to Link/ DOI: http://dx.doi.org/10.9767/bcrec.4.1.1250.5-9

  4. Concept design on RH maintenance of CFETR Tokamak reactor

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yuntao, E-mail: songyt@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Wu, Songtao; Wan, Yuanxi; Li, Jiangang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China); University of Science and Technology of China, Hefei (China); Ye, Minyou [University of Science and Technology of China, Hefei (China); Zheng, Jinxing; Cheng, Yong; Zhao, Wenlong; Wei, Jianghua [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei (China)

    2014-10-15

    Highlights: •We discussed the concept design of the RH maintenance system based on the main design work of the key components for CFETR. •The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. •The technical problems encountered in the design process were discussed. •The present concept design of remote maintenance system in this paper can meet the physical and engineering requirement of CFETR. -- Abstract: CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.

  5. Enzymatic reactors for biodiesel synthesis: Present status and future prospects.

    Science.gov (United States)

    Poppe, Jakeline Kathiele; Fernandez-Lafuente, Roberto; Rodrigues, Rafael C; Ayub, Marco Antônio Záchia

    2015-01-01

    Lipases are being extensively researched for the production of biodiesel as a "silver bullet" in order to avoid the drawbacks of the traditional alkaline transesterification. In this review, we analyzed the main factors involved in the enzymatic synthesis of biodiesel, focusing in the choice of the immobilization protocol, and the parameters involved in the choice and configuration of the reactors. An extensive discussion is presented about the advantages and disadvantages of each type of reactor and their mode of operation. The current scenario of the market for enzymatic biodiesel and some future prospects and necessary developments are also briefly presented.

  6. Research Reactor Design for Export to Myanmar

    Energy Technology Data Exchange (ETDEWEB)

    Win Naing, Lay Lay Myint and Myung-Hyun Kim [Kyunghee Univ. Yongin (Korea, Republic of)

    2006-07-01

    Myanmar is striving to acquire the innovative technology in all field areas including maritime, aerospace and nuclear engineering. There is a high intention to construct a new research reactor for peaceful purposes. The Ministry of Science and Technology (MOST) and Ministry of Education (MOE) are the important government organizations for Myanmar's education and they control most of institutes, universities and colleges. The Department of Atomic Energy (DAE), one of the departments under MOST, leads research projects such as for radiation protection as well as radiation application and coordinates government departments and institutions regarding nuclear energy and its applications. Myanmar's Scientific and Technological Research Department (MSTRD) under MOST guides researches in metallurgy, polymer, pharmacy and biotechnology and so on, and acts as an official body for Myanmar industrial standard. The Department of Higher Education (DHE) under MOE controls art and science universities and colleges including research centers such as Asia Research Center (ARC), Universities Research Center (URC), Microbiology Research Center and so on and does to expand research areas and to utilize advanced technology in science. The wide use of radiation and radioisotopes is developed in Myanmar especially for the field areas such as Medical Science and Agricultural Science. Co{sup 60}, I{sup 131} and Tc{sup 99} are the major use of radioisotopes in diagnosis and therapy. In Agricultural Science, H{sup 3}, C{sup 14}, C{sup 60} etc are used to provide biological effects of radiations on plants, radio-isotopic study of soil physics and tracer studies.

  7. Portfolio Assessment on Chemical Reactor Analysis and Process Design Courses

    Science.gov (United States)

    Alha, Katariina

    2004-01-01

    Assessment determines what students regard as important: if a teacher wants to change students' learning, he/she should change the methods of assessment. This article describes the use of portfolio assessment on five courses dealing with chemical reactor and process design during the years 1999-2001. Although the use of portfolio was a new…

  8. Status report on the indonesian multipurpose reactor, RSG-GAS

    Energy Technology Data Exchange (ETDEWEB)

    Arbie, B.; Supadi, S. (Badan Tenaga Atom National (BATAN), Jakarta (Indonesia))

    1994-01-01

    RSG-GAS, a 30 MW multipurpose reactor, and located at Serpong Atomic Energy Research Center, went to its first criticality on July 29, 1987. Its nominal power was just achieved in March, 1992 by operating the full core configuration (core-VI) after successive five transition core commissioning operation. By then, the reactor performed as expected to produce thermal neutron flux in the order of 10[sup 14] n/(cm[sup 2].sec) safely to support research, radioisotope production as well as material testing including man power development. The long commissioning activities and experiences are outlined in the paper. Main characteristics of the reactor is also presented. Nowadays, the reactor is operating at core no. 9, in weekly modes, mainly for radioisotope production. Some facilities for neutronic experiment, material and fuel testing are being commissioned. Study of silicide fuels to replace the oxide ones is in progress. Several silicide fuel elements with similar specification to the existing oxide fuels have been inserted in the core since 1989. In five years, the RSG-GAS is expected to operate with the silicide fuel. (author).

  9. Commercial US nuclear reactors and waste: the current status

    Energy Technology Data Exchange (ETDEWEB)

    Platt, A.M.; Robinson, J.V.

    1980-09-01

    Between March 1 and June 15, 1980, the declared size of the commercial light waste reactor (LWR) nuclear power industry in the US has decreased another 9 GWe. For the presently declared size: the 165 declared reactors will peak at a capacity of 153 GWe in 2001 and will consume about 870,000 MTU as enrichment feed; the theoretical rate of enrichment requirements will peak at about 19,000,000 SWUs/y in the year 2014; as few as two repositories each with capacity equivalent to 100,000 MTU would hold the waste; and predisposal storage reactor basins and AFRs (away-from-reactor basins) would peak at <85,000 MTU in the year 2020 if the two respositories were commissioned in the years 1997 and 2020. It should be noted that the number of declared LWRs has dropped from 226 on December 31, 1974 to 165 as of this writing. The oil equivalent of the energy loss, assuming a 50% efficiency in use as in cars, is 17,000 million barrels. This is about 10 years of the current rate of US consumption of OPEC oil.

  10. Safety design maps: an early evaluation of safety to support reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Zanocco, P. E-mail: zanoccop@cab.cnea.gov.ar; Gimenez, M.; Delmastro, D

    2003-11-01

    A new tendency in nuclear reactor conceptual design is to include safety criteria through accident analysis to address the selection of the engineering solutions and the value of the main design parameters. In this work, the concept of design map is used to correlate reactor safety performance with the design parameters. The effect of different design parameters on characteristic safety variables, referred to as 'observable variables', extracted from reactor evolution during accidents, is analyzed, and the concept of 'safety design maps' is introduced. The sensitivity of these 'observables variables' regarding changes in design parameters is visualized. Several safety design maps are built from the performance of an integral type reactor during loss of heat sink (LOHS) and main steam line break sequences without SCRAM to show the technique potentiality. Maximum reactor pressure vessel (RPV) pressure and minimum departure from nucleate boiling ratio are chosen as 'observable variables' and their sensitivity to geometry-related parameters and reactivity coefficient is studied. Multiple-parameter single design maps and combined design maps for both accidental sequences are built as examples. The results show the usefulness of this technique to balance and optimize reactor design through an early engineering step. A computer code (HUARPE) has been developed in order to simulate these transients. The cooling circuit, the steam dome, the pressure vessel structures and core models are considered.

  11. Design of gas-inducing reactors

    Energy Technology Data Exchange (ETDEWEB)

    Patwardhan, A.W.; Joshi, J.B. [Univ. of Mumbai (India). Dept. of Chemical Technology

    1999-01-01

    A gas-inducing impeller enables efficient recycling of gas from the headspace into the liquid. Historically, these impellers were used for the first time in froth flotation machines. The various designs of gas-inducing impellers (including those used in froth flotation) could be classified into three categories, depending on the flow pattern coming into and leaving the impeller zone. Correlations for the power consumption, fractional gas holdup, mass-transfer coefficient, and so forth are also available in the literature, although these studies are not comprehensive. A process design algorithm has been presented for the design of gas-inducing impellers. The algorithm consists of the determination of the rate-controlling step, selection of geometry and the operating conditions, and an economic analysis to choose the optimum design. Guidelines have been given about the desired geometry of gas-inducing impellers for achieving different design objectives such as heat transfer, mass transfer, mixing, solid suspension, froth flotation, and so forth. It has been shown that the use of a gas-inducing impeller in a conventional stirred vessel can lead to a substantial increase in the productivity. It has been shown that the optimum geometry may not correspond to the maintenance of equal power consumption per unit volume, or equal tip speed on scale-up. Suggestions have been made for future work in this area.

  12. Defining New Parameters for Green Engineering Design of Treatment Reactors

    Directory of Open Access Journals (Sweden)

    Susana Boeykens

    2016-06-01

    Full Text Available This study proposes a green way to design Plug Flow Reactors (PFR that use biodegradable polymer solutions, capable of contaminant retaining, for industrial wastewater treatment. Usually, to the design of a PFR, the reaction rate is determined by tests on a Continuous Stirred-Tank Reactor (CSTR, these generate toxic effluents and also increase the cost of the design. In this work, empirical expressions (called “slip functions”, in terms of the average concentration of the contaminant, were developed through the study of the transport behaviour of CrVI into solutions of xanthan gum. “In situ” XRµF was selected as a no-invasive micro-technique to determine local concentrations. Slip functions were used with laboratory PFR experiments planned in similar conditions, to obtain useful dimensionless parameters for the industrial design

  13. DDE-MURR Status Report of Conceptual Design Activities

    Energy Technology Data Exchange (ETDEWEB)

    N.E. Woolstenhulme; R.B. Nielson; M.H. Sprenger; G.K. Housley

    2013-09-01

    The Design Demonstration Experiment for the University of Missouri Research Reactor (DDE-MURR) is intended to facilitate Low Enriched Uranium (LEU) conversion of the MURR by demonstrating the performance and fabrication of the LEU fuel element design through an irradiation test in a 200mm channel at the Belgium Reactor 2 (BR2). Revision 0 of this report was prepared at the end of government fiscal year 2012 when most of the resources for furthering DDE design work were expected to be postponed. Hence, the conceptual design efforts were summarized to provide the status of key objectives, notable results, and provisions for future design work. Revision 1 of this report was prepared at the end of fiscal year 2013 in order to include results from a neutronic study performed by BR2, to incorporate further details that had been achieved in the engineering sketches of the irradiation devices, and to provide an update of the DDE-MURR campaign in relation to program objectives and opportunities for its eventual irradiation. These updates were purposed to bring the DDE-MURR conceptual design to level of maturity similar to that of the other two DDE efforts (DDE-MITR and DDE-NBSR). This report demonstrates that the DDE-MURR design effort is well on the path to producing a suitable irradiation experiment, but also puts forth several recommendations in order to facilitate success of the irradiation campaign.

  14. The Virtual Environment for Reactor Applications (VERA). Design and architecture☆

    Science.gov (United States)

    Turner, John A.; Clarno, Kevin; Sieger, Matt; Bartlett, Roscoe; Collins, Benjamin; Pawlowski, Roger; Schmidt, Rodney; Summers, Randall

    2016-12-01

    VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). CASL was established for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both software and numerical perspectives, along with the goals and constraints that drove major design decisions, and their implications. We explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of VERA tools for a variety of challenging applications within the nuclear industry.

  15. Design of virtual SCADA simulation system for pressurized water reactor

    Science.gov (United States)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-02-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  16. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  17. The Virtual Environment for Reactor Applications (VERA): Design and architecture

    Energy Technology Data Exchange (ETDEWEB)

    Turner, John A., E-mail: turnerja@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Clarno, Kevin; Sieger, Matt; Bartlett, Roscoe; Collins, Benjamin [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Pawlowski, Roger; Schmidt, Rodney; Summers, Randall [Sandia National Laboratories, Albuquerque, NM 87185 (United States)

    2016-12-01

    VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). CASL was established for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both software and numerical perspectives, along with the goals and constraints that drove major design decisions, and their implications. We explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of VERA tools for a variety of challenging applications within the nuclear industry.

  18. Conceptual design of the D- sup 3 He reactor artemis

    Energy Technology Data Exchange (ETDEWEB)

    Momoto, H.; Tomita, Y. (National Inst. for Fusion Science, Nagoya (JP)); Ishida, A. (Niigata Univ. (Japan)); Miley, G.H. (Illinois Univ., Urbana, IL (United States)); Kohzaki, Y. (Inst. for Future Technology, Tokyo (JP)); Ohi, S. (Osaka Univ., Suita (Japan)); Ohnishi, M. (Kyoto Univ. (Japan)); Sato, H. (Himeji Inst. of Tech., Hyogo (Japan)); Steinhauer, L.C. (STI Optronics, Inc., Bellevue, WA (US)); Tuszewski, M. (Los Alamos National Lab., NM (United States))

    1992-07-01

    In this paper, a comprehensive design study of the D-{sup 3}He-fueled field-reversed configuration (FRC) reactor Artemis is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of preferential trapping of D-{sup 3}He fusion-produced energetic protons. A novel direct energy converter for 15-MeV protons is also presented. On the basis of consistent fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The D-{sup 3}He FRC power plant offers a most attractive prospect for energy development. It is environmentally acceptable in terms of radioactivity and fuel resources, and the estimated cost of electricity is low compared with a light water reactor. Critical physics and engineering issues in the development of the D-{sup 3}He FRC reactor are clarified.

  19. SCW Pressure-Channel Nuclear Reactor Some Design Features

    Science.gov (United States)

    Pioro, Igor L.; Khan, Mosin; Hopps, Victory; Jacobs, Chris; Patkunam, Ruban; Gopaul, Sandeep; Bakan, Kurtulus

    Concepts of nuclear reactors cooled with water at supercritical pressures were studied as early as the 1950s and 1960s in the USA and Russia. After a 30-year break, the idea of developing nuclear reactors cooled with SuperCritical Water (SCW) became attractive again as the ultimate development path for water cooling. The main objectives of using SCW in nuclear reactors are: 1) to increase the thermal efficiency of modern Nuclear Power Plants (NPPs) from 30-35% to about 45-48%, and 2) to decrease capital and operational costs and hence decrease electrical energy costs (˜1000 US/kW or even less). SCW NPPs will have much higher operating parameters compared to modern NPPs (pressure about 25 MPa and outlet temperature up to 625°C), and a simplified flow circuit, in which steam generators, steam dryers, steam separators, etc., can be eliminated. Also, higher SCW temperatures allow direct thermo-chemical production of hydrogen at low cost, due to increased reaction rates. Pressure-tube or pressure-channel SCW nuclear reactor concepts are being developed in Canada and Russia for some time. Some design features of the Canadian concept related to fuel channels are discussed in this paper. The main conclusion is that the development of SCW pressure-tube nuclear reactors is feasible and significant benefits can be expected over other thermal-energy systems.

  20. Designing visual displays and system models for safe reactor operations

    Energy Technology Data Exchange (ETDEWEB)

    Brown-VanHoozer, S.A.

    1995-12-31

    The material presented in this paper is based on two studies involving the design of visual displays and the user`s prospective model of a system. The studies involve a methodology known as Neuro-Linguistic Programming and its use in expanding design choices from the operator`s perspective image. The contents of this paper focuses on the studies and how they are applicable to the safety of operating reactors.

  1. Optimized Design and Discussion on Middle and Large CANDLE Reactors

    Directory of Open Access Journals (Sweden)

    Xiaoming Chai

    2012-08-01

    Full Text Available CANDLE (Constant Axial shape of Neutron flux, nuclide number densities and power shape During Life of Energy producing reactor reactors have been intensively researched in the last decades [1–6]. Research shows that this kind of reactor is highly economical, safe and efficiently saves resources, thus extending large scale fission nuclear energy utilization for thousands of years, benefitting the whole of society. For many developing countries with a large population and high energy demands, such as China and India, middle (1000 MWth and large (2000 MWth CANDLE fast reactors are obviously more suitable than small reactors [2]. In this paper, the middle and large CANDLE reactors are investigated with U-Pu and combined ThU-UPu fuel cycles, aiming to utilize the abundant thorium resources and optimize the radial power distribution. To achieve these design purposes, the present designs were utilized, simply dividing the core into two fuel regions in the radial direction. The less active fuel, such as thorium or natural uranium, was loaded in the inner core region and the fuel with low-level enrichment, e.g. 2.0% enriched uranium, was loaded in the outer core region. By this simple core configuration and fuel setting, rather than using a complicated method, we can obtain the desired middle and large CANDLE fast cores with reasonable core geometry and thermal hydraulic parameters that perform safely and economically; as is to be expected from CANDLE. To assist in understanding the CANDLE reactor’s attributes, analysis and discussion of the calculation results achieved are provided.

  2. Promising design options for the encapsulated nuclear heat source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Conway, L.; Carelli, M.D.; Dzodzo, M. [Westinghouse Science and Technology, Pittsburgh, PA (United States); Hossain, Q.; Brown, N.W. [Lawrence Livermore National Lab., CA (United States); Wade, D.C.; Sienick, J.J. [Argonne National Lab., IL (United States); Greenspan, E.; Kastenberg, W.E.; Saphier, D. [University of California Dept of Nuclear Engineering, Berkeley, CA (United States)

    2001-07-01

    Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

  3. Design and analysis of multicavity prestressed concrete reactor vessels. [HTGR

    Energy Technology Data Exchange (ETDEWEB)

    Goodpasture, D.W.; Burdette, E.G.; Callahan, J.P.

    1977-01-01

    During the past 25 years, a rather rapid evolution has taken place in the design and use of prestressed concrete reactor vessels (PCRVs). Initially the concrete vessel served as a one-to-one replacement for its steel counterpart. This was followed by the development of the integral design which led eventually to the more recent multicavity vessel concept. Although this evolution has seen problems in construction and operation, a state-of-the-art review which was recently conducted by the Oak Ridge National Laboratory indicated that the PCRV has proven to be a satisfactory and inherently safe type of vessel for containment of gas-cooled reactors from a purely functional standpoint. However, functionalism is not the only consideration in a demanding and highly competitive industry. A summary is presented of the important considerations in the design and analysis of multicavity PCRVs together with overall conclusions concerning the state of the art of these vessels.

  4. Contribution of CAD and PLM Research Reactors Design and Construction

    Energy Technology Data Exchange (ETDEWEB)

    Bonnetain, Xavier [AREVA TA, Paris (France)

    2013-07-01

    As all the reactors, the main stakes in the engineering of design and construction of the research reactors consist of the management and sharing of the technical data, the functional, physical and contractual interfaces data between the various contributors on the whole designs and construction cycle project. For 40 years, AREVA TA designs and builds reactors. Computer Aided Design (CAD) tools were introduced for 30 years into the engineering processes of AREVA TA, completed for 15 years by Product Lifecycle Management (PLM) tools. For 15 years AREVA TA pursues the integration since the feasibility of its newest Information Technologies (IT). In the first part, the paper presents IN the second part, the paper presents how the schematics and CAD tools support the engineering processes during the different phases of the project. CAD was used during the studies and now supports the management of the layout and design studies, including interfaces between suppliers, up to the constitution of the as built CAD mock-up. In the third part, the paper presents the relations between the various tools and the PLM solution implemented by AREVA TA to ensure the consistency between all tools and data for the benefit of the project.

  5. Development of reactor design aid tool using virtual reality technology

    Energy Technology Data Exchange (ETDEWEB)

    Mizuguchi, N. [National Institute for Fusion Science, 322-6 Orochi-cho, Toki, Gifu 509-5292 (Japan)]. E-mail: mizu@nifs.ac.jp; Tamura, Y. [National Institute for Fusion Science, 322-6 Orochi-cho, Toki, Gifu 509-5292 (Japan); Imagawa, S. [National Institute for Fusion Science, 322-6 Orochi-cho, Toki, Gifu 509-5292 (Japan); Sagara, A. [National Institute for Fusion Science, 322-6 Orochi-cho, Toki, Gifu 509-5292 (Japan); Hayashi, T. [National Institute for Fusion Science, 322-6 Orochi-cho, Toki, Gifu 509-5292 (Japan)

    2006-11-15

    A new type of aid system for fusion reactor design, to which the virtual reality (VR) visualization and sonification techniques are applied, is developed. This system provides us with an intuitive interaction environment in the VR space between the observer and the designed objects constructed by the conventional 3D computer-aided design (CAD) system. We have applied the design aid tool to the heliotron-type fusion reactor design activity FFHR2m [A. Sagara, S. Imagawa, O. Mitarai, T. Dolan, T. Tanaka, Y. Kubota, et al., Improved structure and long -life blanket concepts for heliotron reactors, Nucl. Fusion 45 (2005) 258-263] on the virtual reality system CompleXcope [Y. Tamura, A. Kageyama, T. Sato, S. Fujiwara, H. Nakamura, Virtual reality system to visualize and auralize numerical imulation data, Comp. Phys. Comm. 142 (2001) 227-230] of the National Institute for Fusion Science, Japan, and have evaluated its performance. The tool includes the functions of transfer of the observer, translation and scaling of the objects, recording of the operations and the check of interference.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-10-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code. Because PEBBED-THERMIX has not been extensively validated, confirmatory calculations were also performed with RELAP5-3D for the pebble-bed design. During normal operation, the predicted axial profiles in reactor vessel temperature were similar with both codes and the predicted maximum values were within 2 °C. The trends of the calculated vessel temperatures were similar during the depressurized conduction cooldown accident. The maximum value predicted with RELAP5-3D during the depressurized conduction cooldown accident was about 40 °C higher than that predicted with PEBBED. This agreement is considered reasonable based on the expected uncertainty in either calculation. The differences between the PEBBED and RELAP5-3D calculations were not large enough to affect conclusions concerning comparisons between calculated and allowed maximum temperatures during normal operation and the depressurized conduction cooldown accident.

  7. Conceptual design of main coolant pump for integral reactor SMART

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Seok; Kim, Jong In; Kim, Min Hwan [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    The conceptual design for MCP to be installed in the integral reactor SMART was carried out. Canned motor pump was adopted in the conceptual design of MCP. Three-dimensional modeling was performed to visualize the conceptual design of the MCP and to check interferences between the parts. The theoretical design procedure for the impeller was developed. The procedures for the flow field and structural analysis of impeller was also developed to assess the design validity and to verify its structural integrity. A computer program to analyze the dynamic characteristics of the rotor shaft of MCP was developed. The rotational speed sensor was designed and its performance test was conducted to verify the possibility of operation. A prototypes of the canned motor was manufactured and tested to confirm the validity of the design concept. The MCP design concept was also investigated for fabricability by establishing the manufacturing procedures. 41 refs., 96 figs., 10 tabs. (Author)

  8. Designing a mini subcritical nuclear reactor; Diseno de un mini reactor nuclear subcritico

    Energy Technology Data Exchange (ETDEWEB)

    Escobedo G, C. R.; Vega C, H. R.; Davila H, V. M., E-mail: rafelaescobedo@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Jardin Juarez 147, Col. Centro, 98000 Zacatecas, Zac. (Mexico)

    2015-10-15

    In this work the design of a mini subcritical nuclear reactor formed by means of light water moderator, uranium as fuel, and isotopic neutron source of {sup 239}PuBe was carried out. The design was done by Monte Carlo methods with the code MCNP5 in which uranium was modeled in an array of concentric holes cylinders of 8.5, 14.5, 20.5, 26.5, 32.5 cm of internal radius and 3 cm of thickness, 36 cm of height. Different models were made from a single fuel cylinder (natural uranium) to five. The neutron source of {sup 239}PuBe was situated in the center of the mini reactor; in each arrangement was used water as moderator. Cross sections libraries Endf/Vi were used and the number of stories was large enough to ensure less uncertainty than 3%. For each case the effective multiplication factor k{sub e}-f{sub f}, the amplification factor and the power was calculated. Outside the mini reactor the ambient dose equivalent H (10) was calculated for different cases. The value of k{sub eff}, the amplification factor and power are directly related to the number of cylinders of uranium as fuel. Although the average energy of the neutrons {sup 239}PuBe is between 4.5 and 5 MeV in the case of the mini reactor for a cylinder, in the neutron spectrum the presence of thermal neutrons does not exist, so that produced fissions are generated with fast neutrons, and in designs of two and three rings the neutron spectra shows the presence of thermal neutrons, however the fissions are being generated with fast neutrons. Finally in the four and five cases the amount of moderator is enough to thermalized the neutrons and thereby produce the fission. The maximum value for k{sub eff} was 0.82; this value is very close to the assembly of Universidad Autonoma de Zacatecas generating a k{sub eff} of 0.86. According to the safety and radiation protection standards for the design of mini reactor of one, two and three cylinders they comply with the established safety, while designs of four and five

  9. Preliminary design concept of a subcritical reactor using available resources

    Energy Technology Data Exchange (ETDEWEB)

    Churnetski, E.L. [Oak Ridge Y-12 Plant, TN (United States); Hoyny, V.; Chaudhuri, B.R.; Taprantzis, A.; Yavas, A. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering

    1993-12-31

    During the Fall 1993 semester, a project was initiated within the Nuclear Engineering Department of the University of Tennessee with the objective of developing a design for a subcritical reactor with maximized multiplication factor using materials currently available. Such a device, if constructed, would serve as a teaching tool for the Department of Nuclear Engineering. Design work was conducted as a large number of computer calculations, with trial pile configurations based on fundamental nuclear engineering principles, in an effort to maximize multiplication factor through fuel element geometry, moderator type, fissile/moderator ratio, and reflector character. The principal objective of the design group for the early phase of this project was to present several possible ``baseline`` reactor designs and identify directions for improvements. For the sake of calculational ease, the cores analyzes to date have been of nearly cubic shape. The SCALE CSAS25 software which runs KENO.Va, a Monte Carlo code, was used for all neutronics calculations. The baseline reactors resulting from work to date are cuboidal in shape and graphite reflected. Two types of fuel element geometries are proposed, a typical triangular pitch rod lattice and an arrangement of discrete fuel slugs placed in a lattice corresponding to body centered cubic packing. The latter arrangement provides slightly higher multiplication factors than the former. Calculations were performed for both graphite and heavy water moderation with heavy water moderation producing considerably higher multiplication factors, as expected. In general, the maximum k{sub eff} for the reactors are in the range of 0.5 to 0.9, well subcritical, except in the cases of the extreme possible values of fuel assay where critical configurations are possible. In these cases, designs with reduced fuel loading are recommended to assure a subcritical multiplication factor.

  10. 76 FR 78096 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-12-16

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design.... Advanced Boiling Water Reactor (U.S. ABWR) standard plant design to comply with the NRC's aircraft impact...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis A...

  11. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis A...

  12. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  13. Preconceptual design of the new production reactor circulator test facility

    Energy Technology Data Exchange (ETDEWEB)

    Thurston, G.

    1990-06-01

    This report presents the results of a study of a new circulator test facility for the New Production Reactor Modular High-Temperature Gas-Cooled Reactor. The report addresses the preconceptual design of a stand-alone test facility with all the required equipment to test the Main Circulator/shutoff valve and Shutdown Cooling Circulator/shutoff valve. Each type of circulator will be tested in its own full flow, full power helium test loop. Testing will cover the entire operating range of each unit. The loop will include a test vessel, in which the circulator/valve will be mounted, and external piping. The external flow piping will include a throttle valve, flowmeter, and heat exchanger. Subsystems will include helium handling, helium purification, and cooling water. A computer-based data acquisition and control system will be provided. The estimated costs for the design and construction of this facility are included. 2 refs., 15 figs.

  14. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  15. Non-linear analysis in Light Water Reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, Y.R.; Sharabi, M.N.; Nickell, R.E.; Esztergar, E.P.; Jones, J.W.

    1980-03-01

    The results obtained from a scoping study sponsored by the US Department of Energy (DOE) under the Light Water Reactor (LWR) Safety Technology Program at Sandia National Laboratories are presented. Basically, this project calls for the examination of the hypothesis that the use of nonlinear analysis methods in the design of LWR systems and components of interest include such items as: the reactor vessel, vessel internals, nozzles and penetrations, component support structures, and containment structures. Piping systems are excluded because they are being addressed by a separate study. Essentially, the findings were that nonlinear analysis methods are beneficial to LWR design from a technical point of view. However, the costs needed to implement these methods are the roadblock to readily adopting them. In this sense, a cost-benefit type of analysis must be made on the various topics identified by these studies and priorities must be established. This document is the complete report by ANATECH International Corporation.

  16. Saphyr: a code system from reactor design to reference calculations

    Energy Technology Data Exchange (ETDEWEB)

    Akherraz, B.; Baudron, A.M.; Buiron, L.; Coste-Delclaux, M.; Fedon-Magnaud, C.; Lautard, J.J.; Moreau, F.; Nicolas, A.; Sanchez, R.; Zmijarevic, I. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service d' Etudes des Reacteurs et de Modelisation Avancee (DENDMSS/SERMA), 91 - Gif sur Yvette (France); Bergeron, A.; Caruge, D.; Fillion, P.; Gallo, D.; Royer, E. [CEA Saclay, Direction de l' Energie Nucleaire, Departement de Modelisation des Systemes et Structures, Service Fluides numeriques, Modelisations et Etudes (DEN/DMSS/SFNME), 91 - Gif sur Yvette (France); Loubiere, S. [CEA Saclay, Direction de l' Energie Nucleaire, Direction de la Simulation et des Outils Experimentaux, 91- Gif sur Yvette (France)

    2003-07-01

    In this paper we briefly present the package SAPHYR (in French Advanced System for Reactor Physics) which is devoted to reactor calculations, safety analysis and design. This package is composed of three main codes: APOLLO2 for lattice calculations, CRONOS2 for whole core neutronic calculations and FLICA4 for thermohydraulics. Thanks to a continuous development effort, the SAPHYR system is an outstanding tool covering a large domain of applications, from sophisticated 'research and development' studies that need state-of-the-art methodology to routine industrial calculations for reactor and criticality analysis. SAPHYR is powerful enough to carry out calculations for all types of reactors and is invaluable to understand complex phenomena. SAPHYR components are in use in various nuclear companies such as 'Electricite de France', Framatome-ANP, Cogema, SGN, Transnucleaire and Technicatome. Waiting for the next generation tools (DESCARTES for neutronics and NEPTUNE for thermohydraulics) to be available for such a variety of use, with a better level of flexibility and at least equivalent validation and qualification level, the improvement of SAPHYR is going on, to acquire new functions constantly required by users and to improve current performance levels.

  17. Decommissioning of the Astra research reactor: Review and status on July 2003

    Directory of Open Access Journals (Sweden)

    Meyer Franz

    2003-01-01

    Full Text Available The paper describes work on the decommissioning of the ASTRA research reactor at the Austrian Research Centers Seibersdorf. Organizational, planning, and dismantling work done until July 2003 including radiation protection and waste management procedures as well as the current status of the project are presented. Completion of the decommissioning activities is planned for 2006.

  18. Status and some safety philosophies of the China advanced research reactor CARR

    Energy Technology Data Exchange (ETDEWEB)

    Luzheng Yuan [China Inst. of Atomic Energy, Beijing, BJ (China). Reactor Engineering Research and Design Dept.

    2001-07-01

    The existing two research reactors, HWRR (heavy water research reactor) and SPR (swimming pool reactor), have been operated by China Institute of Atomic Energy (CIAE) since, respectively, 1958 and 1964, and are both in extending service and facing the aging problem. It is expected that they will be out of service successively in the beginning decade of the 21{sup st} century. A new, high performance and multipurpose research reactor called China advanced research reactor (CARR) will replace these two reactors. This new reactor adopts the concept of inverse neutron trap compact core structure with light water as coolant and heavy water as the outer reflector. Its design goal is as follows: under the nuclear power of 60MW, the maximum unperturbed thermal neutron flux in peripheral D{sub 2}O reflector not less than 8 x 10{sup 14} n/cm{sup 2}. s while in central experimental channel, if the central cell to be replaced by an experimental channel, the corresponding value not less than 1 x 10{sup 15} n/cm{sup 2}. s. The main applications for this research reactor will cover RI production, neutron scattering experiments, NAA and its applications, neutron photography, NTD for monocrystaline silicon and applications on reactor engineering technology. By the end of 1999, the preliminary design of CARR was completed, then the draft of preliminary safety analysis report (PSAR) was submitted to the relevant authority at the end of 2000 for being reviewed. Now, the CARR project has entered the detail design phase and safety reviewing procedure for obtaining the construction permit from the relevant licensing authority. This paper will only briefly introduce some aspects of safety philosophy of CARR design and PSAR. (orig.)

  19. Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T. and others

    2006-01-15

    A research reactor design is a kind of integral engineering project and a process to obtain a concrete shape through several years of concept development, conceptual design, basic design and detail design. So it requires close cooperation in various areas as well as lots of manpower and cost. The overall process at each stage may be said to be similar except for some stage-specific works. In 2005 as last year of a concept development stage, investigations on the various concepts of the fuel, reactor structure and systems which can meet the requirements established. The requirements for the process systems and I and C systems have also been embodied. The major tasks planned at the early of 2005 have been performed for each area of reactor design as follows: Establishment of the fuel and reactor core concept, and the core analysis, Preliminary thermal-hydraulic and safety analyses for the conceptual cores, Establishment and improvement of analysis system, Concept developments of the reactor structures and major systems, Test and test plan to verify the developed concepts, International cooperation to establish the foundations for exporting a research reactor.

  20. Development of fluid system design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, D. J.; Chang, M. H.; Kang, D. J. and others

    1999-03-01

    This study presents the technology development of the system design concepts of SMART, a multi-purposed integral reactor with enhanced safety and operability, for use in diverse usages and applications of the nuclear energy. This report contains the following; - Design characteristics - Performance and safety related design criteria - System description: Primary system, Secondary system, Residual heat removal system, Make-up system, Component cooling system, Safety system - Development of design computer code: Steam generator performance(ONCESG), Pressurizer performance(COLDPZR), Steam generator flow instability(SGINS) - Development of component module and modeling using MMS computer code - Design calculation: Steam generator thermal sizing, Analysis of feed-water temperature increase at a low flow rate, Evaluation of thermal efficiency in the secondary system, Inlet orifice throttling coefficient for the prevention of steam generator flow instability, Analysis of Nitrogen gas temperature in the pressurizer during heat-up process, evaluation of water chemistry and erosion etc. The results of this study can be utilized not only for the foundation technology of the next phase basic system design of the SMART but also for the basic model in optimizing the system concepts for future advanced reactors. (author)

  1. Conceptual design of a large Spectral Shift Controlled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Matzie, R A; Menzel, G P

    1979-08-01

    Within the framework of the Nonproliferation Alternative Systems Assessment Program (NASAP), the US Department of Energy (DOE) has sponsored the development of a conceptual design of a large Spectral Shift Controlled Reactor (SSCR). This report describes the results of the development program and assesses the performance of the conceptual SSCR on the basis of fuel resource utilization and total power costs. The point of departure of the design study was a 1270 MW(e) PWR using Combustion Engineering's System 80/sup TM/ reactor and Stone and Webster's Reference Plant Design. The initial phase of the study consisted of establishing an optimal core design for both the once-through uranium cycle and the denatured U-235/thorium cycle with uranium recycle. The performance of the SSCR was then also assessed for the denatured U-233/thorium cycle with uranium recycle and for the plutonium/thorium cycle with plutonium recycle. After the optimal core design was established, the design of the NSSS and balance of plant was developed.

  2. Conceptual design study of a scyllac fusion test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomassen, K.I. (comp.)

    1975-07-01

    The report describes a conceptual design study of a fusion test reactor based on the Scyllac toroidal theta-pinch approach to fusion. It is not the first attempt to describe the physics and technology required for demonstrating scientific feasibility of the approach, but it is the most complete design in the sense that the physics necessary to achieve the device goals is extrapolated from experimentally tested MHD theories of toroidal systems,and it uses technological systems whose engineering performance has been carefully calculated to ensure that they meet the machine requirements.

  3. Design of Controllers for Continuous Stirred Tank Reactor

    Directory of Open Access Journals (Sweden)

    Somasundaram Deepa

    2015-02-01

    Full Text Available The objective of the project is to design various controllers for temperature control in Continuous Stirred Tank Reactor (CSTR systems. Initially Zeigler-Nichols, modified Zeigler-Nichols, Tyreus-Luyben, Shen-Yu and IMC based method of tuned Proportional Integral (PI controller is designed and comparisons are made with Fuzzy Logic Controller. Simulations are carried out and responses are obtained for the above controllers. Maximum peak overshoot, Settling time, Rise time,  ISE, IAE  are chosen as performance index. From the analysis it is found that the Fuzzy Logic Controller  is a promising controller than the conventional controllers.

  4. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  5. High-Density Plasma Reactors: Simulations for Design

    Science.gov (United States)

    Hash, David B.; Meyyappan, Meyya; Arnold, James O. (Technical Monitor)

    1998-01-01

    The development of improved and more efficient plasma reactors is a costly process for the semiconductor industry. Until five years ago, the Industry made most of its advancements through a trial and error approach. More recently, the role of computational modeling in the design process has increased. Both conventional computational fluid dynamics (CFD) techniques like Navier-Stokes solvers as well as particle simulation methods are used to model plasma reactor flowfields. However, since high-density plasma reactors generally operate at low gas pressures on the order of 1 to 10 mTorr, a particle simulation may be necessary because of the failure of CFD techniques to model rarefaction effects. The direct simulation Monte Carlo method is the most widely accepted and employed particle simulation tool and has previously been used to investigate plasma reactor flowfields. A plasma DSMC code is currently under development at NASA Ames Research Center with its foundation as the object-oriented parallel Cornell DSMC code, MONACO. The present investigation is a follow up of a neutral flow investigation of the effects of process parameters as well as reactor design on etch rate and etch rate uniformity. The previous work concentrated on silicon etch of a chlorine flow in a configuration typical of electron cyclotron resonance (ECR) or helical resonator type reactors. The effects of the plasma on the dissociation chemistry were modeled by making assumptions about the electron temperature and number density. The electrons or ions themselves were not simulated.The present work extends these results by simulating the charged species.The electromagnetic fields are calculated such that power deposition is modeled self-consistently. Electron impact reactions are modeled along with mechanisms for charge exchange. An bipolar diffusion assumption is made whereby electrons remain tied to the ions. However, the velocities of tile electrons are allowed to be modified during collisions

  6. The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

    Energy Technology Data Exchange (ETDEWEB)

    Iracane, Daniel; Bignan, Gilles [CEA Atomic Energy Commission Saclay Batiment 121- 91191 Gif Sur Yvette (France); Lindbaeck, Jan-Erik; Blomgren, Jan [VATTENFALL AB Nuclear Power Jaemtlandsgatan 99 SE-16287 Stockholm (Sweden)

    2010-07-01

    The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

  7. Status of reduced enrichment programs for research reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Nishihara, Hedeaki [Kyoto Univ., Osaka (Japan); Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo [Japan Atomic Energy Research Institute, Tokyo (Japan)

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.

  8. Simplifying Chemical Reactor Design by using Molar Quantities Instead of Fractional Conversion.

    Science.gov (United States)

    Brown, Lee F.; Falconer, John L.

    1987-01-01

    Explains the advantages of using molar quantities in chemical reactor design. Advocates the use of differential versions of reactor mass balances rather than the integrated forms. Provides specific examples and cases to illustrate the principles. (ML)

  9. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR) standard...

  10. Design of Upelow Anaerobic Sludge Blanket reactor for treatment of organic wastewaters.

    Science.gov (United States)

    Ghangrekar, M M; Kahalekar, U J; Takalkar, S V

    2003-04-01

    The Upflow Anaerobic Sludge Blanket (UASB) Reactor is widely applied anaerobic wastewater treatment method all over the world. Uniform distribution of wastewater at reactor bottom is necessary to establish proper contact between sludge and wastewater. In addition, proper functioning of Gas-Liquid-Solid (GLS) separator is crucial to ensure maximum sludge retention in the reactor and to achieve maximum COD removal rate in the reactor. Hence, proper design of reactor is necessary for appropriate functioning of various components for a given wastewater flow rate and COD concentration. The design procedure for UASB reactor taking due consideration to the GLS design and design of inlet arrangement is discussed in this paper for various wastewater strength and flow rates. A software is developed to make economical design of UASB reactor for different type of wastewater by adopting maximum loading conditions, based on literature recommendations, and at the same time to satisfy all design recommendation, as far as possible.

  11. Robust reactor power control system design by genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Joon; Cho, Kyung Ho; Kim, Sin [Cheju National University, Cheju (Korea, Republic of)

    1997-12-31

    The H{sub {infinity}} robust controller for the reactor power control system is designed by use of the mixed weight sensitivity. The system is configured into the typical two-port model with which the weight functions are augmented. Since the solution depends on the weighting functions and the problem is of nonconvex, the genetic algorithm is used to determine the weighting functions. The cost function applied in the genetic algorithm permits the direct control of the power tracking performances. In addition, the actual operating constraints such as rod velocity and acceleration can be treated as design parameters. Compared with the conventional approach, the controller designed by the genetic algorithm results in the better performances with the realistic constraints. Also, it is found that the genetic algorithm could be used as an effective tool in the robust design. 4 refs., 6 figs. (Author)

  12. The Design of Reactor Internals Hold-Down Spring

    Directory of Open Access Journals (Sweden)

    Guohong Xue

    2016-01-01

    Full Text Available Hold down spring(HDS, clamped between the upper support plate flange and the core barrel flange inside a pressurized reactor vessel, is used to provide a downward force to keep the core barrel stable and from being lifted off from the vessel ledge during reactor normal operation. Spring designer used to think under certain extreme operating conditions, the spring could be lifted off from the ledge because the spring is not stiff enough to prevent the “lift-off”. Therefore, the spring was designed as stiff as it practically can be. However, finite element study indicated that the magnitude of preload is a strong function of friction coefficient. To find the magnitude of the friction coefficient, a series of tests were conducted on 1/10 scale hold down spring samples of three different spring designs. From the test results, it was found that the friction coefficient increases rapidly as the number of loading cycles increases. This implies that the spring preload would increase rapidly during plant operation. Therefore, it is concluded that, because of the strong frictional effect, not only the current design is more than adequate, but also should be made even softer.

  13. Packaging design criteria for the N Reactor/single pass reactor fuel characterization shipments

    Energy Technology Data Exchange (ETDEWEB)

    Stevens, P.F.

    1994-08-31

    The majority of the spent fuel from the N Reactor and the single pass reactors (SPR) is presently being stored at the basins in the 100 K Area. Characterization of these fuels is essential to formulate a safe and efficient processing/disposal method for the spent fuel. Consequently, it is necessary to transport a cross section of spent fuel from the K Basins to the hot cells at the 327 Building in the 300 Area for analysis. The CNS 1-13G cask, a US Nuclear Regulatory Commission (NRC) certified cask manufactured by the ChemNuclear company, will be utilized for the transportation for irradiated fuel elements from the K Basins to the 327 Laboratories for characterization. The cask will utilize an inner container to compensate for the possibility of failed fuel cladding and to reduce the chances of contaminating the cask or the off loading facility. The Packaging Design Criteria (PDC) for these shipments establishes the acceptance criteria for the cask and for the design of an inner container that will be used in the Safety Evaluation for Packaging (SEP).

  14. Interim status report on lead-cooled fast reactor (LFR) research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G.; Li, N.; Hosemann, P.; Zhang, J.; Bolind, A.; LLNL; LANL; Univ. of Illinois

    2008-03-31

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup

  15. Design of a rotary reactor for chemical-looping combustion. Part 1: Fundamentals and design methodology

    KAUST Repository

    Zhao, Zhenlong

    2014-04-01

    Chemical-looping combustion (CLC) is a novel and promising option for several applications including carbon capture (CC), fuel reforming, H 2 generation, etc. Previous studies demonstrated the feasibility of performing CLC in a novel rotary design with micro-channel structures. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet, and depleted air and product streams at exit. The rotary wheel consists of a large number of micro-channels with oxygen carriers (OC) coated on the inner surface of the channel walls. In the CC application, the OC oxidizes the fuel while the channel is in the fuel zone to generate undiluted CO2, and is regenerated while the channel is in the air zone. In this two-part series, the effect of the reactor design parameters is evaluated and its performance with different OCs is compared. In Part 1, the design objectives and criteria are specified and the key parameters controlling the reactor performance are identified. The fundamental effects of the OC characteristics, the design parameters, and the operating conditions are studied. The design procedures are presented on the basis of the relative importance of each parameter, enabling a systematic methodology of selecting the design parameters and the operating conditions with different OCs. Part 2 presents the application of the methodology to the designs with the three commonly used OCs, i.e., nickel, copper, and iron, and compares the simulated performances of the designs. © 2013 Elsevier Ltd. All rights reserved.

  16. Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A L; Cetiner, M S; Wilson, Jr, T L

    2012-04-30

    The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

  17. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  18. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  19. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  20. SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control

    OpenAIRE

    Lindström, Tobias

    2015-01-01

    In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristic...

  1. Rotary Bed Reactor for Chemical-Looping Combustion with Carbon Capture. Part 1: Reactor Design and Model Development

    KAUST Repository

    Zhao, Zhenlong

    2013-01-17

    Chemical-looping combustion (CLC) is a novel and promising technology for power generation with inherent CO2 capture. Currently, almost all of the research has been focused on developing CLC-based interconnected fluidized-bed reactors. In this two-part series, a new rotary reactor concept for gas-fueled CLC is proposed and analyzed. In part 1, the detailed configuration of the rotary reactor is described. In the reactor, a solid wheel rotates between the fuel and air streams at the reactor inlet and exit. Two purging sectors are used to avoid the mixing between the fuel stream and the air stream. The rotary wheel consists of a large number of channels with copper oxide coated on the inner surface of the channels. The support material is boron nitride, which has high specific heat and thermal conductivity. Gas flows through the reactor at elevated pressure, and it is heated to a high temperature by fuel combustion. Typical design parameters for a thermal capacity of 1 MW have been proposed, and a simplified model is developed to predict the performances of the reactor. The potential drawbacks of the rotary reactor are also discussed. © 2012 American Chemical Society.

  2. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  3. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  4. Reactor design for minimizing product inhibition during enzymatic lignocellulose hydrolysis: II. Quantification of inhibition and suitability of membrane reactors.

    Science.gov (United States)

    Andrić, Pavle; Meyer, Anne S; Jensen, Peter A; Dam-Johansen, Kim

    2010-01-01

    Product inhibition of cellulolytic enzymes affects the efficiency of the biocatalytic conversion of lignocellulosic biomass to ethanol and other valuable products. New strategies that focus on reactor designs encompassing product removal, notably glucose removal, during enzymatic cellulose conversion are required for alleviation of glucose product inhibition. Supported by numerous calculations this review assesses the quantitative aspects of glucose product inhibition on enzyme-catalyzed cellulose degradation rates. The significance of glucose product inhibition on dimensioning of different ideal reactor types, i.e. batch, continuous stirred, and plug-flow, is illustrated quantitatively by modeling different extents of cellulose conversion at different reaction conditions. The main operational challenges of membrane reactors for lignocellulose conversion are highlighted. Key membrane reactor features, including system set-up, dilution rate, glucose output profile, and the problem of cellobiose are examined to illustrate the quantitative significance of the glucose product inhibition and the total glucose concentration on the cellulolytic conversion rate. Comprehensive overviews of the available literature data for glucose removal by membranes and for cellulose enzyme stability in membrane reactors are given. The treatise clearly shows that membrane reactors allowing continuous, complete, glucose removal during enzymatic cellulose hydrolysis, can provide for both higher cellulose hydrolysis rates and higher enzyme usage efficiency (kg(product)/kg(enzyme)). Current membrane reactor designs are however not feasible for large scale operations. The report emphasizes that the industrial realization of cellulosic ethanol requires more focus on the operational feasibility within the different hydrolysis reactor designs, notably for membrane reactors, to achieve efficient enzyme-catalyzed cellulose degradation. (c) 2010 Elsevier Inc. All rights reserved.

  5. Liquid metal reactor development. Development of LMR design technology

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Cheol; Kim, Y. I.; Kim, Y. G.; Kim, E. K.; Song, H.; Chung, H. T.; Sim, Y. S.; Min, B. T.; Kim, Y. S.; Wi, M. H.; Yoo, B.; Lee, J. H.; Lee, H. Y.; Kim, J. B.; Koo, G. H.; Hahn, D. H.; Na, B. C.; Hwang, W.; Nam, C.; Ryu, W. S.; Lim, G. S.; Kim, D. H.; Kim, J. D.; Gil, C. S.

    1997-07-01

    This project was performed in five parts, the scope and contents of which are as follows: The nuclear data processing system was established and the KFS group constant library was improved and verified. Basic computation system was constructed by either developing or adding its function. Input/output (I/O) interface processing was developed to establish an integrated calculation system for LMR core nuclear rand thermal-hydraulic design and analysis. An experimental data analysis was performed to validate the constructed core neutronic calculation system. Using the established core calculation system and design technology, preliminary core design and performance analysis on the domestic LMR core design concept were carried out. To develop the basic technology of the LMR system analysis, LMR system behavior characteristics evaluation, thermal -fluid system analysis in the reactor pool, preliminary overall plant analysis and computer codes development have been performed. A porous model and simple one-dimensional model have been evaluated for the reactor pool analysis. The evaluation of the residual heat removal system on different design concepts has been also conducted. For the development of high temperature structural analysis, the heat transfer and thermal stress analyses were performed using finite element program with user subroutine that has been developed with an implementation of the Chaboche constitutive model for inelastic analysis capability, and the evaluation of creep-fatigue and ratcheting behavior of high temperature structure was carried out using this program. for development of the seismic isolation system and to predict the shear behavior for the laminated rubber bearing were established. And the behavior tests of isolation bearing and rubber specimens were carried out, and the seismic response tests for the isolation model structure were performed using the 30 ton shaking table. (author). 369 refs., 119 tabs., 320 figs.

  6. Use of computational fluid dynamics simulations for design of a pretreatment screw conveyor reactor.

    Science.gov (United States)

    Berson, R Eric; Hanley, Thomas R

    2005-01-01

    Computational fluid dynamics simulations were employed to compare performance of various designs of a pretreatment screw conveyor reactor. The reactor consisted of a vertical screw used to create cross flow between the upward conveying solids and the downward flow of acid. Simulations were performed with the original screw design and a modified design in which the upper flights of the screw were removed. Results of the simulations show visually that the modified design provided favorable plug flow behavior within the reactor. Pressure drop across the length of the reactor without the upper screws in place was predicted by the simulations to be 5 vs 40 kPa for the original design.

  7. Preconceptual design and assessment of a Tokamak Hybrid Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Teofilo, V.L.; Leonard, B.R. Jr.; Aase, D.T.

    1980-09-01

    The preconceptual design of a commercial Tokamak Hybrid Reactor (THR) power plant has been performed. The tokamak fusion driver for this hybrid is operated in the ignition mode. The D-T fusion plasma, which produces 1140 MW of power, has a major radius of 5.4 m and a minor radius of 1.0 m with an elongation of 2.0. Double null poloidal divertors are assumed for impurity control. The confining toroidal field is maintained by D-shaped Nb/sub 3/Sn superconducting magnets with a maximum field of 12T at the coil. Three blankets with four associated fuel cycle alternatives have been combined with the ignited tokamak fusion driver. The engineering, material, and balance of plant design requirements for the THR are briefly described. Estimates of the capital, operating and maintenance, and fuel cycle costs have been made for the various driver/blanket combinations and an assessment of the market penetrability of hybrid systems is presented. An analysis has been made of the nonproliferation aspects of the hybrid and its associated fuel cycles relative to fission reactors. The current and required level of technology for both the fusion and fission components of the hybrid system has been reviewed. Licensing hybrid systems is also considered.

  8. Design and Construction of Operation Bridge for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Kwangsub; Choi, Jinbok; Lee, Jongmin; Oh, Jinho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    The operation bridge contains a lower working deck mounted on a saddle that travels on rails. Upright members are mounted on the saddle to support the upper structure and two hoist monorails. The saddle contains an anti-derail system that is composed of seismic lugs and guide rollers. The operation bridge travels along the rails to transport the fuel assembly, irradiated object, and reactor components in the pools by using tools. Hoists are installed at the top girder. The hoist is suspended from the monorail by means of a motor driven trolley that runs along the monorail. Movements of hoist and trolley are controlled by using the control pendant switch. Processes of design and construction of the operation bridge for the research reactor are introduced. The operation bridge is designed under consideration of functions of handling equipment in the pool and operational limits for safety. Structural analysis is carried out to evaluate the structural integrity in the seismic events. Tests and inspections are also performed during fabrication and installation to confirm the function and safety of the operation bridge.

  9. Moving bed biofilm reactor technology: process applications, design, and performance.

    Science.gov (United States)

    McQuarrie, James P; Boltz, Joshua P

    2011-06-01

    The moving bed biofilm reactor (MBBR) can operate as a 2- (anoxic) or 3-(aerobic) phase system with buoyant free-moving plastic biofilm carriers. These systems can be used for municipal and industrial wastewater treatment, aquaculture, potable water denitrification, and, in roughing, secondary, tertiary, and sidestream applications. The system includes a submerged biofilm reactor and liquid-solids separation unit. The MBBR process benefits include the following: (1) capacity to meet treatment objectives similar to activated sludge systems with respect to carbon-oxidation and nitrogen removal, but requires a smaller tank volume than a clarifier-coupled activated sludge system; (2) biomass retention is clarifier-independent and solids loading to the liquid-solids separation unit is reduced significantly when compared with activated sludge systems; (3) the MBBR is a continuous-flow process that does not require a special operational cycle for biofilm thickness, L(F), control (e.g., biologically active filter backwashing); and (4) liquid-solids separation can be achieved with a variety of processes, including conventional and compact high-rate processes. Information related to system design is fragmented and poorly documented. This paper seeks to address this issue by summarizing state-of-the art MBBR design procedures and providing the reader with an overview of some commercially available systems and their components.

  10. Status of Knowledge of Radiation Embrittlement in USA Reactor Pressure Vessel Steels.

    Science.gov (United States)

    1982-02-01

    NRC-FIN-5528 UNCLASSIFIED NRL-R-,737 NUREG -CR-2511 Ni." EEEIIEIIIEI mEE/hhhE AD ~ NUREG /CR-251 1 0 AD A I INRL Memo Rpt 4737 Status of Knowledge of...J2.75 ti a Tec nica for ati Irv Ce Spr ~~~ Sid 1 NUREG /CR-2511 NRL Memo Rpt 4737 R5 Status of Knowledge of Radiation Embrittlement in USA Reactor...vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence. The following documents in the NUREG

  11. Titer-plate formatted continuous flow thermal reactors: Design and performance of a nanoliter reactor.

    Science.gov (United States)

    Chen, Pin-Chuan; Park, Daniel S; You, Byoung-Hee; Kim, Namwon; Park, Taehyun; Soper, Steven A; Nikitopoulos, Dimitris E; Murphy, Michael C

    2010-08-06

    Arrays of continuous flow thermal reactors were designed, configured, and fabricated in a 96-device (12 × 8) titer-plate format with overall dimensions of 120 mm × 96 mm, with each reactor confined to a 8 mm × 8 mm footprint. To demonstrate the potential, individual 20-cycle (740 nL) and 25-cycle (990 nL) reactors were used to perform the continuous flow polymerase chain reaction (CFPCR) for amplification of DNA fragments of different lengths. Since thermal isolation of the required temperature zones was essential for optimal biochemical reactions, three finite element models, executed with ANSYS (v. 11.0, Canonsburg, PA), were used to characterize the thermal performance and guide system design: (1) a single device to determine the dimensions of the thermal management structures; (2) a single CFPCR device within an 8 mm × 8 mm area to evaluate the integrity of the thermostatic zones; and (3) a single, straight microchannel representing a single loop of the spiral CFPCR device, accounting for all of the heat transfer modes, to determine whether the PCR cocktail was exposed to the proper temperature cycling. In prior work on larger footprint devices, simple grooves between temperature zones provided sufficient thermal resistance between zones. For the small footprint reactor array, 0.4 mm wide and 1.2 mm high fins were necessary within the groove to cool the PCR cocktail efficiently, with a temperature gradient of 15.8°C/mm, as it flowed from the denaturation zone to the renaturation zone. With temperature tolerance bands of ±2°C defined about the nominal temperatures, more than 72.5% of the microchannel length was located within the desired temperature bands. The residence time of the PCR cocktail in each temperature zone decreased and the transition times between zones increased at higher PCR cocktail flow velocities, leading to less time for the amplification reactions. Experiments demonstrated the performance of the CFPCR devices as a function of flow

  12. Structural Design Challenges in Design Certification Applications for New Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Miranda, M.; Braverman, J.; Wei, X.; Hofmayer, C.; Xu, J.

    2011-07-17

    The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are confined within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of structural design chal- lenges encountered in recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

  13. DOE/NNSA perspective safeguard by design: GEN III/III+ light water reactors and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Pan, Paul Y [Los Alamos National Laboratory

    2010-12-10

    An overview of key issues relevant to safeguards by design (SBD) for GEN III/IV nuclear reactors is provided. Lessons learned from construction of typical GEN III+ water reactors with respect to SBD are highlighted. Details of SBD for safeguards guidance development for GEN III/III+ light water reactors are developed and reported. This paper also identifies technical challenges to extend SBD including proliferation resistance methodologies to other GEN III/III+ reactors (except HWRs) and GEN IV reactors because of their immaturity in designs.

  14. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  15. On-Line NDE for Advanced Reactor Designs

    Science.gov (United States)

    Nakagawa, N.; Inanc, F.; Thompson, R. B.; Junker, W. R.; Ruddy, F. H.; Beatty, J. M.; Arlia, N. G.

    2003-03-01

    This expository paper introduces the concept of on-line sensor methodologies for monitoring the integrity of components in next generation power systems, and explains general benefits of the approach, while describing early conceptual developments of suitable NDE methodologies. The paper first explains the philosophy behind this approach (i.e. the design-for-inspectability concept). Specifically, we describe where and how decades of accumulated knowledge and experience in nuclear power system maintenance are utilized in Generation IV power system designs, as the designs are being actively developed, in order to advance their safety and economy. Second, we explain that Generation IV reactor design features call for the replacement of the current outage-based maintenance by on-line inspection and monitoring. Third, the model-based approach toward design and performance optimization of on-line sensor systems, using electromagnetic, ultrasonic, and radiation detectors, will be explained. Fourth, general types of NDE inspections that are considered amenable to on-line health monitoring will be listed. Fifth, we will describe specific modeling developments to be used for radiography, EMAT UT, and EC detector design studies.

  16. Conceptual Design of Passive Safety System for Lead-Bismuth Cooled Fast Reactor

    Science.gov (United States)

    Abdullah, A. G.; Nandiyanto, A. B. D.

    2016-04-01

    This paper presents the results of the conceptual design of passive safety systems for reactor power 225 MWth using Pb-Bi coolant. Main purpose of this research is to design of heat removal system from the reactor wall. The heat from the reactor wall is removed by RVACS system using the natural circulation from the atmosphere around the reactor at steady state. The calculation is performed numerically using Newton-Raphson method. The analysis involves the heat transfer systems in a radiation, conduction and natural convection. Heat transfer calculations is performed on the elements of the reactor vessel, outer wall of guard vessel and the separator plate. The simulation results conclude that the conceptual design is able to remove heat 1.33% to 4.67% from the thermal reactor power. It’s can be hypothesized if the reactor had an accident, the system can still overcome the heat due to decay.

  17. Conceptual design of a large heavy water reactor for US siting

    Energy Technology Data Exchange (ETDEWEB)

    Shapiro, N. L.; Jesick, J. F.; Molin, A. T.; Daniel, J. A.

    1979-09-01

    Information on the PHWR type reactor is presented concerning design characteristics; fuel management and resource utilization; economic evaluations; safety, licensing, and environmental impact; and commercial introduction.

  18. Innovative fusion reactor design analysis: Annual performance report, May 15, 1988--January 31, 1989

    Energy Technology Data Exchange (ETDEWEB)

    Klein, A.C.

    1989-01-31

    This report discusses the following topics on fusion reactor component design: FLiBe intermediate heat exchanger design analysis; FLiBe properties; design methodology; FLiBe system steam generator freezeup; FLiBe reactor systems studies; tritium breeding ratio control; analysis of original objectives; and budget analysis. 15 refs., 13 figs., 3 tabs. (LSP)

  19. Magnetic Fustion Reactor Design Studies Program final report, 1 July 1986--30 September 1986

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-09-30

    This report presents progress reported during the period, 7/1/86 - 9/30/86 for the Technical Support Services (TSS) for the Magnetic Fusion Reactor Design Studies Program. Tasks reported include: systems studies work plan, normalization of reactor design studies, interpretation of design study activities, research and development plan, conference support, and reports generated.

  20. Preliminary design study of the Tandem Mirror Reactor (TMR)

    Energy Technology Data Exchange (ETDEWEB)

    Moir, R.W.; Barr, W.L.; Carlson, G.A.

    1978-07-15

    This report describes work done in Fiscal Year 1977 by the Fusion Reactor Studies Group of LLL on the conceptual design of a 1000-MW(e) Tandem Mirror Reactor (TMR). The high Q (defined as the ratio of fusion power to injection power) predicted for the TMR (approximately 5) reduces the recirculating power to a nondominant problem and results in an attractive mirror fusion power plant. The fusion plasma of the TMR is contained in the 100-m-long central cell where the magnetic field strength is a modest 2 T. The blanket for neutron energy recovery and tritium breeding is cylindrical and, along with the solenoidal magnet, is divided into 3-m-long modules to facilitate maintenance. The central cell is fueled (but not heated) by the injection of low-energy neutral beams near its ends. Thus, the central cell is simple and of low technology. The end-cell plasmas must be of high density and high energy in order to plug and heat (via the electrons) the central-cell plasma. The present conceptual design uses 1.2-MeV neutral-beam injection for the end plugs and a cryogenic-aluminum, Yin-Yang magnet that produces an incremental field of about 1 T over a field of 16 T produced by a pair of Nb/sub 3/Sn superconducting solenoids. Important design problems remain in both the neutral-beam injector and in the end-plug magnet. Also remaining are important physics questions such as alpha-beam particle transport and end-plug stability. These questions are discussed at length in the report and suggestions for future work are given.

  1. Design parameters for sludge reduction in an aquatic worm reactor

    NARCIS (Netherlands)

    Hendrickx, T.L.G.; Temmink, B.G.; Elissen, H.J.H.; Buisman, C.J.N.

    2010-01-01

    Reduction and compaction of biological waste sludge from waste water treatment plants (WWTPs) can be achieved with the aquatic worm Lumbriculus variegatus. In our reactor concept for a worm reactor, the worms are immobilised in a carrier material. The size of a worm reactor will therefore mainly be

  2. Model based design of biochemical micro-reactors

    Directory of Open Access Journals (Sweden)

    Tobias eElbinger

    2016-02-01

    Full Text Available Mathematical modelling of biochemical pathways is an important resource in Synthetic Biology, as the predictive power of simulating synthetic pathways represents an important step in the design of synthetic metabolons. In this paper, we are concerned with the mathematical modeling, simulation and optimization of metabolic processes in biochemical micro-reactors able to carry out enzymatic reactions and to exchange metabolites with their surrounding medium. The results of the reported modeling approach are incorporated in the design of the first micro-reactor prototypes that are under construction. These microreactors consist of compartments separated by membranes carrying specific transporters for the input of substrates and export of products. Inside the compartments multi-enzyme complexes assembled on nano-beads by peptide adapters are used to carry out metabolic reactions.The spatially resolved mathematical model describing the ongoing processes consists of a system of diffusion equations together with boundary and initial conditions. The boundary conditions model the exchange of metabolites with the neighboring compartments and the reactions at the surface of the nano-beads carrying the multi-enzyme complexes. Efficient and accurate approaches for numerical simulation of the mathematical model and for optimal design of the micro-reactor are developed. As a proof-of-concept scenario, a synthetic pathway for the conversion of sucrose to glucose-6-phosphate (G6P was chosen. In this context, the mathematical model is employed to compute the spatio-temporal distributions of the metabolite concentrations, as well as application relevant quantities like the outflow rate of G6P. These computations are performed for different scenarios, where the number of beads as well as their loading capacity are varied. The computed metabolite distributions show spatial patterns which differ for different experimental arrangements. Furthermore, the total output

  3. Status of the irradiation test vehicle for testing fusion materials in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States); Palmer, A.J.; Ingram, F.W. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States); Wiffen, F.W. [Dept. of Energy, Germantown, MD (United States). Office of Fusion Energy

    1998-09-01

    The design of the irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) has been completed. The main application for the ITV is irradiation testing of candidate fusion structural materials, including vanadium-base alloys, silicon carbide composites, and low-activation steels. Construction of the vehicle is underway at the Lockheed Martin Idaho Technology Company (LMITCO). Dummy test trains are being built for system checkout and fine-tuning. Reactor insertion of the ITV with the dummy test trains is scheduled for fall 1998. Barring unexpected difficulties, the ITV will be available for experiments in early 1999.

  4. Circuits design of action logics of the protection system of nuclear reactor IAN-R1 of Colombia; Diseno de los circuitos de la logica de actuacion del sistema de proteccion del reactor nuclear IAN-R1 de Colombia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J. L.; Rivero G, T.; Sainz M, E., E-mail: joseluis.gonzalez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    Due to the obsolescence of the instrumentation and control system of the nuclear research reactor IAN-R1, the Institute of Geology and Mining of Colombia, IngeoMinas, launched an international convoking for renewal it which was won by the Instituto Nacional de Investigaciones Nucleares (ININ). Within systems to design, the reactor protection system is described as important for safety, because this carried out, among others two primary functions: 1) ensuring the reactor shutdown safely, and 2) controlling the interlocks to protect against operational errors if defined conditions have not been met. To fulfill these functions, the various subsystems related to the safety report the state in which they are using binary signals and are connected to the inputs of two redundant logic wiring circuits called action logics (Al) that are part of the reactor protection system. These Al also serve as logical interface to indicate at all times the status of subsystems, both the operator and other systems. In the event that any of the subsystems indicates a state of insecurity in the reactor, the Al generate signals off (or scram) of the reactor, maintaining the interlock until the operator sends a reset signal. In this paper the design, implementation, verification and testing of circuits that make up the Al 1 and 2 of IAN-R1 reactor is described, considering the fulfillment of the requirements that the different international standards imposed on this type of design. (Author)

  5. Designing a SCADA system simulator for fast breeder reactor

    Science.gov (United States)

    Nugraha, E.; Abdullah, A. G.; Hakim, D. L.

    2016-04-01

    SCADA (Supervisory Control and Data Acquisition) system simulator is a Human Machine Interface-based software that is able to visualize the process of a plant. This study describes the results of the process of designing a SCADA system simulator that aims to facilitate the operator in monitoring, controlling, handling the alarm, accessing historical data and historical trend in Nuclear Power Plant (NPP) type Fast Breeder Reactor (FBR). This research used simulation to simulate NPP type FBR Kalpakkam in India. This simulator was developed using Wonderware Intouch software 10 and is equipped with main menu, plant overview, area graphics, control display, set point display, alarm system, real-time trending, historical trending and security system. This simulator can properly simulate the principle of energy flow and energy conversion process on NPP type FBR. This SCADA system simulator can be used as training media for NPP type FBR prospective operators.

  6. The design and safety features of the IRIS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carelli, Mario D. E-mail: carellmd@westinghouse.com; Conway, L.E.; Oriani, L.; Petrovic, B.; Lombardi, C.V.; Ricotti, M.E.; Barroso, A.C.O.; Collado, J.M.; Cinotti, L.; Todreas, N.E.; Grgic, D.; Moraes, M.M.; Boroughs, R.D.; Ninokata, H.; Ingersoll, D.T.; Oriolo, F

    2004-05-01

    Salient features of the International Reactor Innovative and Secure (IRIS) are presented here. IRIS, an integral, modular, medium size (335 MWe) PWR, has been under development since the turn of the century by an international consortium led by Westinghouse and including over 20 organizations from nine countries. Described here are the features of the integral design which includes steam generators, pumps and pressurizer inside the vessel, together with the core, control rods, and neutron reflector/shield. A brief summary is provided of the IRIS approach to extended maintenance over a 48-month schedule. The unique IRIS safety-by-design approach is discussed, which, by eliminating accidents, at the design stage, or decreasing their consequences/probabilities when outright elimination is not possible, provides a very powerful first level of defense in depth. The safety-by-design allows a significant reduction and simplification of the passive safety systems, which are presented here, together with an assessment of the IRIS response to transients and postulated accidents.

  7. Regulatory Risk Reduction for Advanced Reactor Technologies – FY2016 Status and Work Plan Summary

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne Leland [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-08-01

    Millions of public and private sector dollars have been invested over recent decades to realize greater efficiency, reliability, and the inherent and passive safety offered by advanced nuclear reactor technologies. However, a major challenge in experiencing those benefits resides in the existing U.S. regulatory framework. This framework governs all commercial nuclear plant construction, operations, and safety issues and is highly large light water reactor (LWR) technology centric. The framework must be modernized to effectively deal with non-LWR advanced designs if those designs are to become part of the U.S energy supply. The U.S. Department of Energy’s (DOE) Advanced Reactor Technologies (ART) Regulatory Risk Reduction (RRR) initiative, managed by the Regulatory Affairs Department at the Idaho National Laboratory (INL), is establishing a capability that can systematically retire extraneous licensing risks associated with regulatory framework incompatibilities. This capability proposes to rely heavily on the perspectives of the affected regulated community (i.e., commercial advanced reactor designers/vendors and prospective owner/operators) yet remain tuned to assuring public safety and acceptability by regulators responsible for license issuance. The extent to which broad industry perspectives are being incorporated into the proposed framework makes this initiative unique and of potential benefit to all future domestic non-LWR applicants

  8. Status of deuterium nuclear data for the simulation of heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S.; Roubtsov, D.; Rao, R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Svenne, J.P. [Univ. of Manitoba, Winnipeg, Manitoba (Canada); Winnipeg Inst. for Theoretical Physics, Winnipeg, Manitoba (Canada); Canton, L. [Inst. Nazionale de Fisica Nucleare, Sezione di Padova, Padova (Italy); Univ. di Padova, Dipartimento di Fisica, Padova (Italy); Plompen, A.J.M. [EC-JRC, Inst. for Reference Materials and Measurements, Retieseweg, Geel (Belgium); Stanoiu, M. [Horia Hulubei National Inst. for Physics and Nuclear Engineering, Magurele (Romania); Nankov, N.; Rouki, C. [EC-JRC, Inst. for Reference Materials and Measurement, Retieseweg, Geel (Belgium)

    2011-07-01

    An overview is presented of the status of the deuterium nuclear data used in reactor physics simulations of heavy water (D{sub 2}O) reactors and of ongoing activities to improve their accuracy. The main subjects having noticeable reactivity impact for critical systems involving D{sub 2}O are the degree of backscatter in D(n,n)D elastic scattering at neutron energies <3.2 MeV, the value of the elastic scattering cross section at thermal neutron energies and the adequacy of their numerical representation in evaluated nuclear data libraries. The scope includes fundamental nuclear-data measurements; three-body nuclear-theory calculations; and MCNP5 simulations of experiments involving D{sub 2}O or deuterated targets. (author)

  9. A brief history of design studies on innovative nuclear reactors

    Science.gov (United States)

    Sekimoto, Hiroshi

    2014-09-01

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970's the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980's the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  10. A brief history of design studies on innovative nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sekimoto, Hiroshi, E-mail: hsekimot@gmail.com [Emeritus Professor, Tokyo Institute of Technology (Japan)

    2014-09-30

    In a short period after the success of CP1, many types of nuclear reactors were proposed and investigated. However, soon only a small number of reactors were selected for practical use. Around 1970, only LWRs with small number of CANDUs were operated in the western world, and FBRs were under development. It was about the time when Apollo moon landing was accomplished. However, at the same time, the future of human being was widely considered pessimistic and Limits to Growth was published. In the end of 1970’s the TMI accident occurred and many nuclear reactor contracts were cancelled in USA and any more contracts had not been concluded until recent years. From the reflection of this accident, many Inherent Safe Reactors (ISRs) were proposed, though none of them were constructed. A common idea of ISRs is smallness of their size. Tokyo Institute of Technology (TokyoTech) held a symposium on small reactors, SR/TIT, in 1991, where many types of small ISRs were presented. Recently small reactors attract interest again. The most ideas employed in these reactors were the same discussed in SR/TIT. In 1980’s the radioactive wastes from fuel cycle became a severe problem around the world. In TokyoTech, this issue was discussed mainly from the viewpoint of nuclear transmutations. The neutron economy became inevitable for these innovative nuclear reactors especially small long-life reactors and transmutation reactors.

  11. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  12. Continuous supercritical water gasification of isooctane: A promising reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Susanti, Ratna F.; Kim, Jae-Duck; Kim, Jaehoon [Supercritical Fluid Research Laboratory, Clean Energy Center, Energy Division, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seoungbuk-gu, Seoul 136-791 (Korea); Department of Green Process and System Engineering, University of Science and Technology (UST), 113 Gwahangno, Yuseong-gu, Daejeon 305-333 (Korea); Veriansyah, Bambang [Supercritical Fluid Research Laboratory, Clean Energy Center, Energy Division, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seoungbuk-gu, Seoul 136-791 (Korea); Lee, Youn-Woo [School of Chemical and Biological Engineering, Seoul National University, Gwanangro 599, Gwanak-gu, Seoul 151-744 (Korea)

    2010-03-15

    A new design of supercritical water gasification system was developed to achieve high hydrogen gas yield and good gas-liquid flow stability. The apparatus consisted of a reaction zone, an insulation zone and a cooling zone that were directly connected to the reaction zone. The reactor was set up at an inclination of 75 from vertical position, and feed and water were introduced at the bottom of the reactor. The performances of this new system were investigated with gasification of isooctane at various experimental conditions - reaction temperatures of 601-676 C, residence times of 6-33 s, isooctane concentrations of 5-33 wt%, and oxidant (hydrogen peroxide) concentrations up to 4507 mmol/L without using catalysts. A significant increase in hydrogen gas yield, almost four times higher than that from the previous up-down gasifier configuration (B. Veriansyah, J. Kim, J.D. Kim, Y.W. Lee, Hydrogen Production by Gasification of Isooctane using Supercritical Water, Int. J. Green Energy. 5 (2008) 322-333) was observed with the present gasifier configuration. High hydrogen gas yield (6.13 mol/mol isooctane) was obtained at high reaction temperature of 637 C, a low feed concentration of 9.9 wt% and a long residence time of 18 s in the presence of 2701.1 mmol/L hydrogen peroxide. At this condition, the produced gases mainly consisted of hydrogen (59.5 mol%), methane (14.8 mol%) and carbon dioxide (22.0 mol%), and a small amount of carbon monoxide (1.6 mol%) and C{sub 2}-C{sub 3} species (2.1 mol%). Reaction mechanisms of supercritical water gasification of isooctane were also presented. (author)

  13. Shielding designs for pressurized water reactors in France

    Energy Technology Data Exchange (ETDEWEB)

    Champion, G.; Forestier, J.; Vergnaud, T.

    1986-07-01

    The efforts made by Electricite de France to reduce exposure from the two-component neutron-gamma radiation fields inside the pressurized water reactor (PWR) building are described. Most of the attention had been focused on the problem of neutron exposure relative to the problem of achieving a highly efficient confinement within the reactor cavity and the state of the art of personnel neutron dosimetry. A description of the general neutron calculation scheme that links the characteristics of the neutron fields escaping from the reactor vessel to the dose equivalent rate cartographies inside the reactor building is provided.

  14. Documentation control process of Brazilian multipurpose reactor: conceptual design and basic design

    Energy Technology Data Exchange (ETDEWEB)

    Kibrit, Eduardo; Prates, Jose Eduardo; Longo, Guilherme Carneiro; Salvetti, Tereza Cristina, E-mail: ekibrit@ipen.br, E-mail: jeprates@ipen.br, E-mail: glongo@ipen.br, E-mail: salvetti@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Established in the scope of Plan of Action of the Ministry of Science, Technology and Innovation (PACTI/MCTI) in 2007, the construction of the Brazilian Multipurpose Reactor (RMB) is on the way. This type of reactor has a broad spectrum of applications in the nuclear field and related technologies such as the radioisotopes used as supplies in the production of radiopharmaceuticals, with very much benefit to the Brazilian society being, therefore, the main goal of the Project. RMB Project consists of the following stages: site selection and site evaluation; design (conceptual design, basic design, detailed design and experimental design); construction (procurement, manufacturing; civil construction; electromechanical construction and assembling); commissioning; operation and decommissioning. Each stage requires adaptation of human resources for the stage schedule execution. The implementation of a project of this magnitude requires a complex project management, which covers not only technical, but also administrative areas. Licensing, financial resources, quality and document control systems, engineering are some of the areas involved in project success. The development of the conceptual and basic designs involved the participation of three main engineering companies. INTERTECHNE Consultores S.A. was in charge of conceptual and basic designs for conventional systems of buildings and infrastructure. INVAP S.E. was responsible for preparing the basic design of the reactor core and annexes. MRS Estudos Ambientais Ltda. has prepared documents for environmental licensing. This paper describes the procedures used during conceptual and basic design stages to control design documentation and flow of this documentation, involving the analysis and incorporation of comments from experts, control and storage of a volume of approximately 15,000 documents. (author)

  15. Simulation of Reactor Transient and Design Criteria of Sodium-cooled Fast Reactors

    OpenAIRE

    Gottfridsson, Filip

    2010-01-01

    The need for energy is growing in the world and the market of nuclear power is now once more expanding. Some issues of the current light-water reactors can be solved by the next generation of nuclear power, Generation IV, where sodium-cooled reactors are one of the candidates. Phénix was a French prototype sodium-cooled reactor, which is seen as a success. Although it did encounter an earlier unexperienced phenomenon, A.U.R.N., in which a negative reactivity transient followed by an oscillati...

  16. Conceptual design of a commercial tokamak hybrid reactor fueling system

    Energy Technology Data Exchange (ETDEWEB)

    Matney, K.D.; Donnert, H.J.; Yang, T.F.

    1979-12-01

    A conceptual design of a fuel injection system for CTHR (Commercial Tokamak Hybrid Reactor) is discussed. Initially, relative merits of the cold-fueling concept are compared with those of the hot-fueling concept; that is, fueling where the electron is below 1 eV is compared with fueling where the electron temperature exceeds 100 eV. It is concluded that cold fueling seems to be somewhat more free of drawbacks than hot fueling. Possible implementation of the cold-fueling concept is exploited via frozen-pellet injection. Several methods of achieving frozen-pellet injection are discussed and the light-gas-gun approach is chosen from these possibilities. A modified version of the ORNL Neutral Gas Shielding Model is used to simulate the pellet injection process. From this simulation, the penetration-depth dependent velocity requirement is determined. Finally, with the velocity requirement known, a gas-pressure requirement for the proposed conceptual design is established. The cryogenic fuel-injection and fuel-handling systems are discussed. A possible way to implement the conceptual device is examined along with the attendant effects on the total system.

  17. Conceptual design of a commercial tokamak hybrid reactor fueling system

    Energy Technology Data Exchange (ETDEWEB)

    Matney, K D; Donnert, H J; Yang, T F

    1979-12-01

    A conceptual design of a fuel injection system for CTHR (Commercial Tokamak Hybrid Reactor) is discussed. Initially, relative merits of the cold-fueling concept are compared with those of the hot-fueling concept; that is, fueling where the electron temperature is below 1 eV is compared with fueling where the electron temperature exceeds 100 eV. It is concluded that cold fueling seems to be somewhat more free of drawbacks than hot fueling. Possible implementation of the cold-fueling concept is exploited via frozen-pellet injection. Several methods of achieving frozen-pellet injection are discussed and the light-gas-gun approach is chosen from these possibilities. A modified version of the ORNL Neutral Gas Shielding Model is used to simulate the pellet injection process. From this simulation, the penetration-depth dependent velocity requirement is determined. Finally, with the velocity requirement known, a gas-pressure requirement for the proposed conceptual design is established. The cryogenic fuel-injection and fuel-handling systems are discussed. A possible way to implement the conceptual device is examined along with the attendant effects on the total system.

  18. The economic impact of the proposed demonstration plant for the Pebble Bed Modular Reactor design

    OpenAIRE

    Thomas, Stephen

    2005-01-01

    This report examines the history of the South African Pebble Bed Modular Reactor, a new design of nuclear power plant. It illustrates the serious delays in developing the design and the huge increases in cost of the technology.

  19. A Basic LEGO Reactor Design for the Provision of Lunar Surface Power

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2008-06-01

    A final design has been established for a basic Lunar Evolutionary Growth-Optimized (LEGO) Reactor using current and near-term technologies. The LEGO Reactor is a modular, fast-fission, heatpipe-cooled, clustered-reactor system for lunar-surface power generation. The reactor is divided into subcritical units that can be safely launched with lunar shipments from Earth, and then emplaced directly into holes drilled into the lunar regolith to form a critical reactor assembly. The regolith would not just provide radiation shielding, but serve as neutron-reflector material as well. The reactor subunits are to be manufactured using proven and tested materials for use in radiation environments, such as uranium-dioxide fuel, stainless-steel cladding and structural support, and liquid-sodium heatpipes. The LEGO Reactor system promotes reliability, safety, and ease of manufacture and testing at the cost of an increase in launch mass per overall rated power level and a reduction in neutron economy when compared to a single-reactor system. A single unshielded LEGO Reactor subunit has an estimated mass of approximately 448 kg and provides approximately 5 kWe. The overall envelope for a single subunit with fully extended radiator panels has a height of 8.77 m and a diameter of 0.50 m. Six subunits could provide sufficient power generation throughout the initial stages of establishing a lunar outpost. Portions of the reactor may be neutronically decoupled to allow for reduced power production during unmanned periods of base operations. During later stages of lunar-base development, additional subunits may be emplaced and coupled into the existing LEGO Reactor network, subject to lunar base power demand. Improvements in reactor control methods, fuel form and matrix, shielding, as well as power conversion and heat rejection techniques can help generate an even more competitive LEGO Reactor design. Further modifications in the design could provide power generative opportunities for

  20. What went Right: Resilience of Existing Reactors to - for Generation III+ Reactor Design

    Science.gov (United States)

    Garwin, Richard L.

    2014-07-01

    To quote Tolstoy's Anna Karenina, "All happy families are alike; each unhappy family is unhappy in its own way." So the reactors that have been working well in the world don't get a lot of attention...

  1. Incorporation of statistical distribution of particle properties in chemical reactor design and operation: the cooled tubular reactor

    NARCIS (Netherlands)

    Wijngaarden, R.J.; Westerterp, K.R.

    1992-01-01

    Pellet heat and mass transfer coefficients inside packed beds do not have definite deterministic values, but are stochastic quantities with a certain distribution. Here, a method is presented to incorporate the stochastic distribution of pellet properties in reactor design and operation models. The

  2. First observation of a reactor-status effect on the beta+ decay rate of 22Na

    CERN Document Server

    de Meijer, Robert; Stegenga, Jan; Steyn, Steph; Lindsay, Robbie; van Rooy, Milton

    2016-01-01

    In the search for an electron antineutrino detection method with sensitivity below the 1.8 MeV threshold for the inverse beta decay reaction, beta-decay counting experiments with ca. 3 kBq 22Na and 60Co sources were conducted at unit #1 (2.775 GWth) of the Koeberg Nuclear Power Station in South Africa. The goal was to determine if the rate of decay is measurably influenced by a change between the ON-OFF status of such a reactor. The experimental setup consisted of a single NaI crystal to measure de-excitation and annihilation photons associated with beta-decay. Its volume and well shape were purposely chosen to use coincidence summing in the interval 170-2452 keV to differentiate between electron capture and beta+ emission in 22Na. The Pb-shielded setup was placed in the seismic vault underneath the containment building, thereby shielded from the reactor core by 8 m of uninterrupted concrete. For 22Na two measurement series were made, each covering an ON-OFF-ON cycle of the reactor. The following fractional c...

  3. Fusion reactor design-III. Report on the third IAEA technical committee meeting and workshop, Tokyo, Japan, 5-16 October 1981

    Energy Technology Data Exchange (ETDEWEB)

    Iso, Y. (Japan Atomic Energy Research Inst., Tokyo); Stacey, W.M. Jr. (Georgia Inst. of Tech., Atlanta (USA)); Kulcinski, G.L. (Wisconsin Univ., Madison (USA)); Krakowski, R.A. (Los Alamos National Lab., NM (USA)); Carlson, G.A. (Lawrence Livermore National Lab., CA (USA)); Yamanaka, C. (Osaka Univ., Suita, Japan); Casini, G. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Igata, N. (Tokyo Univ. (Japan))

    1982-05-01

    A brief summary is given of the plenary sessions of the Third IAEA Technical Committee Meeting and Workshop on Fusion Reactor Design and Technology. The large tokamak experiments now under construction have brought fusion research to the threshold of fusion reactor power. A number of major in-depth reactor design studies are reported such as the 650-MW INTOR near-term experimental fusion reactor based upon essentially current technology. Also reported are studies of commercial reactor designs. Results are given of the discussions of seven workshops that were based on the papers presented. These groups evaluated the current status and identified key problem areas for each in the following areas: near-term tokamaks, long-term tokamaks, toroidal systems (alternative concepts), open systems, inertial confinement systems (advanced fuels, hybrids, etc.), and fusion reactor materials. The importance of beginning now to prepare for the next major step in fusion reactor development was emphasized and the benefits of international co-operation were evident in the consensus reached in the INTOR results and the strong influence it has had on the direction of the leading national design efforts.

  4. Designing reverse-flow packed bed reactors for stable treatment of volatile organic compounds.

    Science.gov (United States)

    Chan, Fan Liang; Keith, Jason M

    2006-02-01

    Reverse-flow packed bed reactors can be used to treat gaseous pollutants from chemical plants. This article describes the design and operation of a modified reverse-flow reactor (MRFR) which has a recuperator on each end of the reactor and a reaction zone in the middle. The recuperators have low thermal dispersion and the reaction zone has a high thermal dispersion, obtained by placing metal inserts into the bed, parallel with the gas flow. Performance of the MRFR during extended lean and rich conditions is determined with analytical analysis and compares well with numerical simulations of CO oxidation; however, the theory is expected to be useful for any reaction kinetics. A major advantage of this MRFR design is an extended time for the reactor to extinguish during lean conditions. This work also describes MRFR performance with internal reactor cooling, which can be used as a control mechanism to maintain reactor temperature for proper removal of volatile organic compounds.

  5. Status on the Searches of Neutrino Magnetic Moment at the Kuo-Sheng Power Reactor

    CERN Document Server

    Wong, H T

    2002-01-01

    The TEXONO collaboration has been built up among scientists from Taiwan and China to pursue an experimental program in neutrino and astro-particle physics. The flagship efforts have been the study of low energy neutrino physics at the Kuo-Sheng Power Reactor Plant in Taiwan. The Reactor Laboratory is equipped with flexibly-designed shieldings, cosmic veto systems, electronics and data acquisition systems which can function with different detector schemes. Data are taken during the Reactor Period June-01 till April-02 with a high purity germanium detector and 46 kg of CsI(Tl) crystal scintillator array operating in parallel. A threshold of 5 keV has been achieved for the germanium detector, and the background level comparable to those of Dark Matter experiments underground is achieved. Based on 62/46 days of analyzed Reactor ON/OFF data, a preliminary result of $\\rm{(\\munue / 10^{-10} \\mub)^2 = - 1.1 \\pm 2.5}$ can be derived for neutrino magnetic moment $\\munue$. Sensitivity region on neutrino radiative decay ...

  6. Design of a micro-channel reactor for decomposition of organic pollutants in waste water treatment

    OpenAIRE

    Charles, Guillaume; Corbel, Serge; Carré, Marie-Christiane; Roques-Carmes, Thibault; Zahraa, Orfan

    2009-01-01

    International audience; Photocatalytic micro-channel reactor was built by using stereolithography process. A reactor with a micro-channel as a support of TiO2 photocatalyst was designed in order to reduce dimensions while improving the efficiency. Photocatalytic activity of the micro-reactor at various flow rates was evaluated by the inlet and outlet concentrations of salicylic acid as a model of pollutant. Influence of the initial pollutant concentration, the irradiation intensity on the rat...

  7. Design of micro-reactors and solar photocatalytic prototypes; Diseno de micro-reactores y prototipos fotocataliticos solares

    Energy Technology Data Exchange (ETDEWEB)

    Flores E, R.M.; Hernandez H, M.; Perusquia del Cueto, M.R.; Bonifacio M, J.; Jimenez B, J.; Ortiz O, H.B.; Castaneda J, G.; Lugo H, M. [ININ, Km. 36.5 Carr. Mexico-Toluca, 52750 La Marquesa, Ocoyoacac (Mexico)]. e-mail: rmfe@nuclear.inin.mx

    2007-07-01

    In the ININ is carried out research in heterogeneous photocatalysis using artificial light for to degrade organic compounds. In this context, it is sought to use the solar radiation as energy source to knock down costs. Of equal form it requires to link the basic and applied research. For it, a methodology that allows to design and to build micro-reactors and plants pilot has been developed, like previous step, to request external supports and to a future commercialization. The beginning of these works gave place to the partial construction of a prototype of photocatalytic reactor of the cylinder-parabolic composed type (CPC)

  8. Design of an Actinide Burning, Lead-Bismuth Cooled Reactor That Produces Low Cost Electricity

    Energy Technology Data Exchange (ETDEWEB)

    C. Davis; S. Herring; P. MacDonald; K. McCarthy; V. Shah; K. Weaver (INEEL); J. Buongiorno; R. Ballinger; K. Doyoung; M. Driscoll; P. Hejzler; M. Kazimi; N. Todreas (MIT)

    1999-07-01

    The purpose of this project is to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The choice of lead-bismuth for the reactor coolant is an actinide burning fast reactor offers enhanced safety and reliability. The advantages of lead-bismuth over sodium as a coolant are related to the following material characteristics: chemical inertness with air and water; higher atomic number; lower vapor pressure at operating temperatures; and higher boiling temperature. Given the status of the field, it was agreed that the focus of this investigation in the first two years will be on the assessment of approaches to optimize core and plant arrangements in order to provide maximum safety and economic potential in this type of reactor.

  9. Tokamak experimental power reactor conceptual design. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    1976-08-01

    A conceptual design has been developed for a tokamak Experimental Power Reactor to operate at net electrical power conditions with a plant capacity factor of 50 percent for 10 years. The EPR operates in a pulsed mode at a frequency of approximately 1/min., with an approximate 75 percent duty cycle, is capable of producing approximately 72 MWe and requires 42 MWe. The annual tritium consumption is 16 kg. The EPR vacuum chamber is 6.25 m in major radius and 2.4 m in minor radius, is constructed of 2-cm thick stainless steel, and has 2-cm thick detachable, beryllium-coated coolant panels mounted on the interior. An 0.28 m stainless steel blanket and a shield ranging from 0.6 to 1.0 m surround the vacuum vessel. The coolant is H/sub 2/O. Sixteen niobium-titanium superconducting toroidal-field coils provide a field of 10 T at the coil and 4.47 T at the plasma. Superconducting ohmic-heating and equilibrium-field coils provide 135 V-s to drive the plasma current. Plasma heating is accomplished by 12 neutral beam-injectors, which provide 60 MW. The energy transfer and storage system consists of a central superconducting storage ring, a homopolar energy storage unit, and a variety of inductor-converters.

  10. Fast reactor programme in India

    Indian Academy of Sciences (India)

    P Chellapandi; P R Vasudeva Rao; Prabhat Kumar

    2015-09-01

    Role of fast breeder reactor (FBR) in the Indian context has been discussed with appropriate justification. The FBR programme since 1985 till 2030 is highlighted focussing on the current status and future direction of fast breeder test reactor (FBTR), prototype fast breeder reactor (PFBR) and FBR-1 and 2. Design and technological challenges of PFBR and design and safety targets with means to achieve the same are the major highlights of this paper.

  11. Design of fuzzy PID controller for high temperature pebble bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Badgujar, Kushal D.; Satpute, Satchidanand R.; Revankara, Shripad T.; Lee, John C.; Kim, Moo Hwan [POSTECH, Pohang (Korea, Republic of)

    2012-10-15

    Control system is most important characteristic to be considered to control spontaneous fission reaction in the design of the nuclear reactor. Recently fuzzy based control systems have been designed and applied as control system for nuclear plants. This article emphasize on controlling the power of the high temperature pebble bed reactor (HTPBR) with the design of Fuzzy proportional integral derivative (PID) controller. A simplified reactor model with point kinetics equation and reactor heat balance equation is used. The reactivity feedback arising from power coefficient of reactivity and Xenon poisoning is also considered. The reactor is operated at various power levels by using fuzzy PID controller. The fuzzy logic eliminates the necessity of the tuning the gains of PID controller each time by extending the finite sets of the PID controller gains.

  12. Effect of design parameters on enhancement of hydrogen charging in metal hydride reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kaplan, Y. [Mechanical Engineering Department, Nigde University, 51100 Nigde (Turkey)

    2009-03-15

    The effects of heat transfer mechanisms on the charging process in metal hydride reactors are studied under various charging pressures. Three different cylindrical reactors with the same base dimensions are designed and manufactured. The first one is a closed cylinder cooled with natural convection, the fins are manufactured around the second reactor and the third reactor is cooled with water circulating around the reactor. The temperatures of the reactor at several locations are measured during charging with a range of pressure of 1-10 bar. The third reactor shows the lowest temperature increase with the fastest charging time under all charging pressures investigated. The effective heat transfer coefficients of the reactors are also calculated according to the experimental results and they are found to be 5.5 {+-} 1 W m{sup -2} K{sup -1}, 35 {+-} 2 W m{sup -2} K{sup -1} and 113 {+-} 1 W m{sup -2} K{sup -1}, respectively. The experimental results showed that the charging of hydride reactors is mainly heat transfer dependent and the reactor with better cooling exhibits the fastest charging characteristics. (author)

  13. Status of R&D Activities on Materials for Fusion Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baluc, N.; Abe, K.; Boutard, J. L.; Chernov, V. M.; Diegele, E.; Jitsukawa, S.; Kimura, Akihiko; Klueh, R. L.; Kohyama, Akira; Kurtz, Richard J.; Lasser, R.; Matsui, H.; Moslang, A.; Muroga, T.; Odette, George R.; Tran, M. Q.; van der Schaaf, B.; Wu, Y.; Yu, J.; Zinkle, Steven J.

    2007-09-19

    Current R&D activities on materials for fusion power reactors are mainly focused on plasma facing, structural and tritium breeding materials for plasma facing (first wall, divertor) and breeding blanket components. Most of these activities are being performed in Europe, Japan, P.R. China, Russia and the USA. They relate to development of new high temperature, radiation resistant materials, development of coatings that shall act as erosion, corrosion, permeation or electrical/MHD barriers, characterization of the whole candidate materials in terms of mechanical and physical properties, assessment of irradiation effects, compatibility experiments, development of reliable joints, and development and/or validation of design rules. Priorities defined worldwide in the field of materials for fusion power reactors are summarized, as well as the main achievements obtained during the last few years and the near-term perspectives in the different investigation areas.

  14. Conceptual design of a fluidized bed nuclear reactor: statics, dynamics and safety-related aspects

    NARCIS (Netherlands)

    Agung, A.

    2007-01-01

    In this thesis a conceptual design of an innovative high temperature reactor based on the fluidization principle (FLUBER) is proposed. The reactor should satisfy the following requirements: (a) modular and low power, (b)) large shutdown margin, (c) able to produce power when the bed of particles exp

  15. A fuel management study and cycle nuclear design for PW reactors

    Energy Technology Data Exchange (ETDEWEB)

    Minguez, E.; Ahnert, C.; Aragones, J. M.; Corella, M. R.

    1975-07-01

    A reference reactor was chosen to do a general study involving Fuel Management Evaluations of several cycles, and Design Calculations of cycles already performed, according to a calculation scheme set up in the Reactor Technology Division of the J.E.N., using some computer codes acquired to foreign sources and other ones developed in the J.E.N. (Author) 5 refs.

  16. Conceptual design of a fluidized bed nuclear reactor: statics, dynamics and safety-related aspects

    NARCIS (Netherlands)

    Agung, A.

    2007-01-01

    In this thesis a conceptual design of an innovative high temperature reactor based on the fluidization principle (FLUBER) is proposed. The reactor should satisfy the following requirements: (a) modular and low power, (b)) large shutdown margin, (c) able to produce power when the bed of particles

  17. Production of Advanced Biofuels via Liquefaction - Hydrothermal Liquefaction Reactor Design: April 5, 2013

    Energy Technology Data Exchange (ETDEWEB)

    Knorr, D.; Lukas, J.; Schoen, P.

    2013-11-01

    This report provides detailed reactor designs and capital costs, and operating cost estimates for the hydrothermal liquefaction reactor system, used for biomass-to-biofuels conversion, under development at Pacific Northwest National Laboratory. Five cases were developed and the costs associated with all cases ranged from $22 MM/year - $47 MM/year.

  18. Design criteria of integrated reactors based on transients; Criterios de diseno de reactores integrados basados en transitorios

    Energy Technology Data Exchange (ETDEWEB)

    Zanocco, P.; Gimenez, M.; Delmastro, D. [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche

    1999-11-01

    A new tendency in integrated reactors conceptual design is to include safety criteria through accident analysis. In this work, the effect of design parameters in a Loss of Heat Sink transient using design maps is analyzed. Particularly, geometry related parameters and reactivity coefficients are studied. Also the effect of primary relief/safety valve during the transient is evaluated. A design map for valve area vs. coolant density reactivity coefficient is obtained. A computer code (HUARPE) is developed in order to simulate these transients. Coolant, steam dome, pressure vessel structures and core models are implemented. This code is checked against TRAC with satisfactory results. (author) 5 refs., 13 figs.

  19. The use of PRA in the development of ALWR (advanced light water reactor) design requirements. [Advanced Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Summitt, R.L. (Safety and Reliability Optimization Services, Inc., Knoxville, TN (USA)); Additon, S.L. (TENERA, L.P., Bethesda, MD (USA)); Pasedag, W.F. (USDOE, Washington, DC (USA))

    1989-01-01

    The current hiatus in nuclear power plant orders provides an opportunity for the development of advanced light water reactor (ALWR) design concepts and regulatory requirements which incorporate the insights gained from the application of the probabilistic risk assessment. The US Department of Energy is assisting the Electric Power Research Institute (EPRI) in the incorporation of PRA insights into the specification of the utility requirements, and reactor vendors in support of the conceptual design of safety systems, for such advanced plants. This paper reviews the applications of PRA methods in this development of specifications for, and the design of simplified, rugged ALWRs with a significantly improved risk profile. Specific examples of the impact of utilizing published PRA insights, construction and use of functional PRA models, and feedback of PRA experience into the specification of the key assumptions and groundrules for ALWR PRAs are presented. 13 refs., 3 tabs.

  20. Improve Design of Fuel Shear for Fast Reactor

    Institute of Scientific and Technical Information of China (English)

    GAO; Wei; OUYANG; Ying-gen; LI; Wei-min

    2012-01-01

    <正>Due to the deeper burnup and higher fuel swelling, fast reactor metal fuel rod using 316 stainless steel cladding, replacing the traditional zirconia cladding. The diameter of fuel rod of fast reactor is much longer than that of PWR, and the cladding of stainless steel has better ductility than zirconia cladding. Using the existing shear still will cause several aspects of problem: 1) Longer diameter of rod leads to

  1. Conceptual design of a new homogeneous reactor for medical radioisotope Mo-99/Tc-99m production

    Energy Technology Data Exchange (ETDEWEB)

    Liem, Peng Hong [Nippon Advanced Information Service (NAIS Co., Inc.) Scientific Computational Division, 416 Muramatsu, Tokaimura, Ibaraki (Japan); Tran, Hoai Nam [Chalmers University of Technology, Dept. of Applied Physics, Div. of Nuclear Engineering, SE-412 96 Gothenburg (Sweden); Sembiring, Tagor Malem [National Nuclear Energy Agency (BATAN), Center for Reactor Technology and Nuclear Safety, Kawasan Puspiptek, Serpong, Tangerang Selatan, Banten (Indonesia); Arbie, Bakri [PT MOTAB Technology, Kedoya Elok Plaza Blok DA 12, Jl. Panjang, Kebun Jeruk, Jakarta Barat (Indonesia)

    2014-09-30

    To partly solve the global and regional shortages of Mo-99 supply, a conceptual design of a nitrate-fuel-solution based homogeneous reactor dedicated for Mo-99/Tc-99m medical radioisotope production is proposed. The modified LEU Cintichem process for Mo-99 extraction which has been licensed and demonstrated commercially for decades by BATAN is taken into account as a key design consideration. The design characteristics and main parameters are identified and the advantageous aspects are shown by comparing with the BATAN's existing Mo-99 supply chain which uses a heterogeneous reactor (RSG GAS multipurpose reactor)

  2. Gas core reactor power plants designed for low proliferation potential

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, L.L. (comp.)

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF/sub 6/ and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on /sup 233/U born from thorium. Fission product removal was continuous. Newly born /sup 233/U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of /sup 233/U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors.

  3. Reactor Neutrino Experiments

    OpenAIRE

    Cao, Jun

    2007-01-01

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

  4. Design characteristics and requirements of irradiation holes for research reactor experimental facilities

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T.; Lee, C. S.; Seo, C. G

    2003-07-01

    In order to be helpful for the design of a new research reactor with high performance, are summarized the applications of research reactors in various fields and the design characteristics of experimental facility such as vertical irradiation holes and beam tubes. Basic requirements of such experimental facilities are also described. Research reactor has been widely utilized in various fields such as industry, engineering, medicine, life science, environment etc., and now the application fields are gradually being expanded together with the development of technology. Looking into the research reactors which are recently constructed or in plan, it seems that to develop a multi-purpose research reactor with intensive neutron beam research capability has become tendency. In the layout of the experimental facilities, the number and configuration of irradiation and beam holes should be optimized to meet required test conditions such as neutron flux at the early design stage. But, basically high neutron flux is required to perform experiments efficiently. In this aspect, neutron flux is regarded as one of important parameters to judge the degree of research reactor performance. One of main information for a new research reactor design is utilization demands and requirements of experimental holes. So basic requirements which should be considered in a new research reactor design were summarized from the survey of experimental facilities characteristics of various research reactors with around 20 MW thermal power and the experiences of HANARO utilization. Also is suggested an example of the requirements of experimental holes such as size, number and neutron flux, which are thought as minimum, in a new research reactor for exporting to developing countries such as Vietnam.

  5. Current design efforts for the gas-cooled fast reactor (GFR)

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3850 (United States)]. e-mail: Kevan.Weaver@inl.gov

    2005-07-01

    Current research and development on the Gas-Cooled Fast Reactor (GCFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFC I) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GCFR: a helium-cooled, direct Brayton cycle power conversion system that will operate with an outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in participating in research related to the development of the GCFR. These are EURATOM (European Commission), France, Japan, South Africa, South Korea, Switzerland, and the United Kingdom. Of these, EURATOM (including the United Kingdom), France, Japan, and Switzerland have active research activities with respect to the GCFR. The research includes GCFR design and safety, and fuels/in-core materials/fuel cycle projects. This paper outlines the current design status of the GCFR, and includes work done in the areas mentioned above. (Author)

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

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  7. Conceptual Nuclear Design of a 20 MW Multipurpose Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, Hak Sung; Park, Cheol [KAERI, Daejeon (Korea, Republic of); Nghiem, Huynh Ton; Vinh, Le Vinh; Dang, Vo Doan Hai [Dalat Nuclear Research Reactor, Hanoi (Viet Nam)

    2007-08-15

    A conceptual nuclear design of a 20 MW multi-purpose research reactor for Vietnam has been jointly done by the KAERI and the DNRI (VAEC). The AHR reference core in this report is a right water cooled and a heavy water reflected open-tank-in-pool type multipurpose research reactor with 20 MW. The rod type fuel of a dispersed U{sub 3}Si{sub 2}-Al with a density of 4.0 gU/cc is used as a fuel. The core consists of fourteen 36-element assemblies, four 18-element assemblies and has three in-core irradiation sites. The reflector tank filled with heavy water surrounds the core and provides rooms for various irradiation holes. Major analyses have been done for the relevant nuclear design parameters such as the neutron flux and power distributions, reactivity coefficients, control rod worths, etc. For the analysis, the MCNP, MVP, and HELIOS codes were used by KAERI and DNRI (VAEC). The results by MCNP (KAERI) and MVP (DNRI) showed good agreements and can be summarized as followings. For a clean, unperturbed core condition such that the fuels are all fresh and there are no irradiation holes in the reflector region, the fast neutron flux (E{sub n}{>=}1.0 MeV) reaches 1.47x10{sup 14} n/cm{sup 2}s and the maximum thermal neutron flux (E{sub n}{<=}0.625 eV) reaches 4.43x10{sup 14} n/cm{sup 2}s in the core region. In the reflector region, the thermal neutron peak occurs about 28 cm far from the core center and the maximum thermal neutron flux is estimated to be 4.09x10{sup 14} n/cm{sup 2}s. For the analysis of the equilibrium cycle core, the irradiation facilities in the reflector region were considered. The cycle length was estimated as 38 days long with a refueling scheme of replacing three 36-element fuel assemblies or replacing two 36-element and one 18-element fuel assemblies. The excess reactivity at a BOC was 103.4 mk, and 24.6 mk at a minimum was reserved at an EOC. The assembly average discharge burnup was 54.6% of initial U-235 loading. For the proposed fuel management

  8. PRELIMINARY DESIGN OF OSCILLATORY FLOW BIODIESEL REACTOR FOR CONTINUOUS BIODIESEL PRODUCTION FROM JATROPHA TRIGLYCERIDES

    Directory of Open Access Journals (Sweden)

    AZHARI T. I. MOHD. GHAZI

    2008-08-01

    Full Text Available The concept of a continuous process in producing biodiesel from jatropha oil by using an Oscillatory Flow Biodiesel Reactor (OFBR is discussed in this paper. It has been recognized that the batch stirred reactor is a primary mode used in the synthesis of biodiesel. However, pulsatile flow has been extensively researcehed and the fundamental principles have been successfully developed upon which its hydrodynamics are based. Oscillatory flow biodiesel reactor offers precise control of mixing by means of the baffle geometry and pulsation which facilitates to continuous operation, giving plug flow residence time distribution with high turbulence and enhanced mass and heat transfer. In conjunction with the concept of reactor design, parameters such as reactor dimensions, the hydrodynamic studies and physical properties of reactants must be considered prior to the design work initiated recently. The OFBR reactor design involves the use of simulation software, ASPEN PLUS and the reactor design fundamentals. Following this, the design parameters shall be applied in fabricating the OFBR for laboratory scale biodiesel production.

  9. Design and properties of marine reactors and associated R and D

    Energy Technology Data Exchange (ETDEWEB)

    Gagarinski, A.; Ignatiev, V. [Russian Research Centre Kurchatov Inst., Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1996-05-01

    The report is a review of open information available in the USA, UK, France, Russia and other countries on the design and properties of marine reactors and associated R and D. First, a short discussion is given of the milestones and main trends for the development of nuclear-powered ships. Then a brief review is presented of features for ship reactor design. Light water and liquid metal cooled reactor technologies are described and reactor operating experiences for Russian ice-breakers assessed. Traditional and alternative civil uses of submarine and surface shipboard reactor technology in Russia and Japan are also treated. Finally, some problems connected with radioactive waste by the nuclear-powered fleet are briefly considered. 41 refs, 27 figs, 19 tabs.

  10. The present status of iodine chemistry research in Canada and its application to reactor safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.R. [Ontario Hydro Nuclear, Toronto (Canada); Kupferschmid, W.C.H.; Wren, J.C.; Ball, J.M. [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

    1996-12-01

    The current need to understand iodine chemistry in a reactor safety context has become more sharply focussed as the level of that understanding has advanced. At the same time, the situations of most concern within containment, from an iodine perspective, are also being redefined in the light of that understanding. The present paper summarises these developments. Over the past five years, considerable advances have occurred in our understanding of iodine chemistry under conditions of interest in reactor accidents. A number of key experiments have yielded important results in the areas of solution chemistry, the role of surfaces, the importance of organics and the effects of impurities. This understanding supplements the already substantial gains made in characterising the key roles of pH and the effects of radiation. All these factors underline the now evident fact that the kinetics of iodine are the controlling factor when radiation is involved, and that a number of reactive species, not present in thermal reactions, effectively control the observed volatility of iodine. In this paper, recent advances are summarised and the present status of our understanding of iodine chemistry is reviewed. Specifically, an attempt is made to identify those areas where our understanding appears to be relatively complete, and to flag the remaining critical areas where our attention is currently focussed. The state of our modelling capability is reviewed, as is the significance or related areas such as the role of mass transfer. Finally, an overview is presented of the significance of this work for reactor safety, and our expectations for its application over the near term future. (author) 2 figs., 12 refs.

  11. Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 2. Conceptual balance of plant design

    Energy Technology Data Exchange (ETDEWEB)

    1979-11-01

    This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. This volume describes the conceptual balance-of-plant (BOP) design and was prepared by United Engineers and Constructors, Inc. of Philadelphia, Pennsylvania. The major emphasis of the BOP study was a preliminary design of an overall plant to provide a basis for future studies.

  12. Reactor design for minimizing product inhibition during enzymatic lignocellulose hydrolysis II. Quantification of inhibition and suitability of membrane reactors

    DEFF Research Database (Denmark)

    Andric, Pavle; Meyer, Anne S.; Jensen, Peter Arendt;

    2010-01-01

    hydrolysis rates and higher enzyme usage efficiency (kg(product/)kg(enzyme)). Current membrane reactor designs are however not feasible for large scale operations. The report emphasizes that the industrial realization of cellulosic ethanol requires more focus on the operational feasibility within......Product inhibition of cellulolytic enzymes affects the efficiency of the biocatalytic conversion of lignocellulosic biomass to ethanol and other valuable products. New strategies that focus on reactor designs encompassing product removal, notably glucose removal, during enzymatic cellulose...... conversion are required for alleviation of glucose product inhibition. Supported by numerous calculations this review assesses the quantitative aspects of glucose product inhibition on enzyme-catalyzed cellulose degradation rates. The significance of glucose product inhibition on dimensioning of different...

  13. Reactor

    Science.gov (United States)

    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.

  14. The final status of the decommissioning of research reactors in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. W.; Hong, S. B.; Park, J. H.; Chung, U. S. [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon (Korea, Republic of)

    2010-10-15

    A decommissioning project for the Korean Research Reactors (KRR-1 and 2) was started in 1997 and had been carried out with the goal of completion being by the end of 2008. All the facilities were dismantled and the building surfaces decontaminated. The radioactive waste was packed into 200 liter drums and 4 m{sup 3} containers and temporarily stored on site until their final disposal at the national repository facility. Some of the releasable waste was freely released and utilized for the non-nuclear industries. The assessment of the residual radioactivity was carried out according to Multi Agency Radiation Site Survey and Investigation Manual guidance, and accordingly, the safety of the site release was verified. The site and the buildings will be cleared for a reuse for non nuclear purposes after a review of the assessment. In this paper, the final status of the decommissioning of research reactors in Korea including dismantlement processes, waste management and a final assessment for unrestricted use of the site and buildings for the final goal of the decommissioning project that will be described. (Author)

  15. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  16. The role of integral experiments and nuclear cross section evaluations in space nuclear reactor design

    Science.gov (United States)

    Moses, David L.; McKnight, Richard D.

    The importance of the nuclear and neutronic properties of candidate space reactor materials to the design process has been acknowledged as has been the use of benchmark reactor physics experiments to verify and qualify analytical tools used in design, safety, and performance evaluation. Since June 1966, the Cross Section Evaluation Working Group (CSEWG) has acted as an interagency forum for the assessment and evaluation of nuclear reaction data used in the nuclear design process. CSEWG data testing has involved the specification and calculation of benchmark experiments which are used widely for commercial reactor design and safety analysis. These benchmark experiments preceded the issuance of the industry standards for acceptance, but the benchmarks exceed the minimum acceptance criteria for such data. Thus, a starting place has been provided in assuring the accuracy and uncertainty of nuclear data important to space reactor applications.

  17. Introduction to the special issue on the technical status of materials for a fusion reactor

    Science.gov (United States)

    Stork, D.; Zinkle, S. J.

    2017-09-01

    Materials determine in a fundamental way the performance and environmental attractiveness of a fusion reactor: through the size (power fluxes to the divertor, neutron fluxes to the first wall); economics (replacement lifetime of critical in-vessel components, thermodynamic efficiency through operating temperature etc); plasma performance (erosion by plasma fluxes to the divertor surfaces); robustness against off-normal accidents (safety); and the effects of post-operation radioactivity on waste disposal and maintenance. The major philosophies and methodologies used to formulate programmes for the development of fusion materials are outlined, as the basis for other articles in this special issue, which deal with the fundamental understanding of the issues regarding these materials and their technical status and prospects for development.

  18. Residual stress diffractometer KOWARI at the Australian research reactor OPAL: Status of the project

    Science.gov (United States)

    Brule, Alain; Kirstein, Oliver

    2006-11-01

    Neutron scattering using diffraction techniques is now recognized as the most precise and reliable method of mapping sub-surface residual stresses in materials or even components, which are not only of academic but also of industrial-economic relevance. The great potential of neutrons in the field of residual stresses was recognized by ANSTO and its external Beam Instrument Advisory Group for the new research reactor OPAL. The recommendation was to build the dedicated strain scanner KOWARI among the first suite of instruments available to users. We give an update on the overall project and present the current status of the diffractometer. It is anticipated that the instrument will be commissioned in mid 2006 and available to users at the end of the OPAL project.

  19. Design of Continuous Reactor Systems for API Production

    DEFF Research Database (Denmark)

    Pedersen, Michael Jønch

    in continuous reactor setups. Grignard chemistry encompasses a very powerful reaction type frequently applied in the pharmaceutical industry, for the formation of new carbon-carbon bonds. Three Grignard addition reactions have been studied, all having very different behaviors related to aspects of reaction......-scale production equipment enabled complete replacement of the existing batch production of this intermediate. The crowning achievement in this work was the realization of continuous laboratory reactor setups capable of manufacturing the entire GMP portion of the synthesis of melitracen HCl at H. Lundbeck A...... able to be coupled to construct a single continuous reactor facility for manufacturing melitracen HCl. The study of Grignard addition reactions to the three different substrates investigated in this thesis has culminated in a methodology by which reaction engineering decisions can be guided...

  20. Development Status for a Combined Solid Oxide Co-Electrolyzer and Carbon Formation Reactor System for Oxygen Regeneration

    Science.gov (United States)

    Green, Robert D.; Matter, Paul H.; Holt, Chris; Beachy, Michael; Gaydos, James; Farmer, Serene C.; Setlock, John

    2016-01-01

    A critical component in spacecraft life support loop closure is the removal of carbon dioxide (CO2, produced by the crew) from the cabin atmosphere and chemical reduction of this CO2 to recover the oxygen. In 2015, we initiated development of an oxygen recovery system for life support applications consisting of a solid oxide co-electrolyzer (SOCE) and a carbon formation reactor (CFR). The SOCE electrolyzes a combined stream of carbon dioxide (CO2) and water (H2O) gas mixtures to produce synthesis gas (e.g., CO and H2 gas) and pure dry oxygen as separate products. This SOCE is being developed from a NASA GRC solid oxide fuel cell and stack design originally developed for aeronautics long-duration power applications. The CFR, being developed by pHMatter LLC, takes the CO and H2 output from the SOCE, and converts it primarily to solid carbon (C(s)) and H2O and CO2. Although the solid carbon accumulates in the CFR, the innovative design allows easy removal of the carbon product, requiring minimal crew member (CM) time and low resupply mass (1.0 kg/year/CM) for replacement of the solid carbon catalyst, a significant improvement over previous Bosch reactor approaches. In this work, we will provide a status of our Phase I efforts in the development and testing of both the SOCE and CFR prototype units, along with an initial assessment of the combined SOCE-CFR system, including a mass and power projections, along with an estimate of the oxygen recovery rate.

  1. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    Science.gov (United States)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  2. Proposed design for the PGAA facility at the TRIGA IPR-R1 research reactor

    OpenAIRE

    Guerra, Bruno T.; Jacimovic, Radojko; Menezes, Maria Angela BC; Leal,Alexandre S.

    2013-01-01

    Background This work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed. Findings The proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below t...

  3. A Study on Conceptual Design of Fischer-Tropsch Reactors in GTL Applications

    OpenAIRE

    2016-01-01

    GTL (Gas-to-liquid) process is becoming an attractive technology which can produce liquid petroleum products using natural gas. As a part of preliminary design of GTL-FPSO application, process simulation analysis for conceptual design and optimization of reformers and F-T reactors are performed in GTL-FPSO applications by implementing the user made subroutine programs of kinetic equations into PRO/II PROVISION simulator. As for the F-T reactors, Plug Flow Reactor (PFR) model is used with deta...

  4. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    OpenAIRE

    Galvez, Cristhian

    2011-01-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the pa...

  5. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  6. An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Gyu; Lee, Jae Han

    2007-03-15

    This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

  7. An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Gyu; Lee, Jae Han

    2007-03-15

    This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    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.

  9. Development of the Plan for Improvement of the Competitiveness of KALIMER Design against the Competing Reactor Design in Gen IV

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Yoon Sub; Kim, Yeong Il; Hong, Ser Gi (and others)

    2005-11-15

    A fast reactor has a good transmutation capability and it enables breeding of fuel and use of a closed fuel cycle. By these characteristics of a fast reactor, the limited uranium resources of the world can be much more effectively utilized and the nuclear wastes of a high level of radioactivity and toxicity from the current nuclear power reactors of LWRs and HWRs can be drastically reduced in its volume and the management of the wastes can be easily treated. Also electricity can be generated more effectively since a fast reactor has the feature of high operation temperature. These features of a fast reactor makes it inevitable on a long term basis to construct fast reactors in Korea. The domestic fast reactor technology level, however, is at the level of coming out of a beginning stage and needs utilization of international expertise. Recently an international cooperation program called GIF has been formulated and our KALIMER was selected as one of the two reference designs for the international joint R and D works with JSFR of Japan. In the current frame of the GIF program, the two selected reference designs are supposed to be evaluated against each other in future and one design is to be finally selected. To make the international cooperation program directed more useful to our fast reactor technology development, it is required to strengthen the competitiveness of KALIMER so that it can be selected. To meet the necessity, a study was made to enhance the competitiveness of KALIMER. The design characteristics of JSFR was investigated and relatively weak points and strong points of KALIMER were identified. Also the GenIV SFR design requirements were applied to two designs to evaluate the level of requirements satisfaction. Based on these investigation results, a plan for improving the weak features and strengthening the strong points of the KALIMER design has been developed.

  10. Modeling and Design Optimization of Multifunctional Membrane Reactors for Direct Methane Aromatization.

    Science.gov (United States)

    Fouty, Nicholas J; Carrasco, Juan C; Lima, Fernando V

    2017-08-29

    Due to the recent increase of natural gas production in the U.S., utilizing natural gas for higher-value chemicals has become imperative. Direct methane aromatization (DMA) is a promising process used to convert methane to benzene, but it is limited by low conversion of methane and rapid catalyst deactivation by coking. Past work has shown that membrane separation of the hydrogen produced in the DMA reactions can dramatically increase the methane conversion by shifting the equilibrium toward the products, but it also increases coke production. Oxygen introduction into the system has been shown to inhibit this coke production while not inhibiting the benzene production. This paper introduces a novel mathematical model and design to employ both methods in a multifunctional membrane reactor to push the DMA process into further viability. Multifunctional membrane reactors, in this case, are reactors where two different separations occur using two differently selective membranes, on which no systems studies have been found. The proposed multifunctional membrane design incorporates a hydrogen-selective membrane on the outer wall of the reaction zone, and an inner tube filled with airflow surrounded by an oxygen-selective membrane in the middle of the reactor. The design is shown to increase conversion via hydrogen removal by around 100%, and decrease coke production via oxygen addition by 10% when compared to a tubular reactor without any membranes. Optimization studies are performed to determine the best reactor design based on methane conversion, along with coke and benzene production. The obtained optimal design considers a small reactor (length = 25 cm, diameter of reaction tube = 0.7 cm) to subvert coke production and consumption of the product benzene as well as a high permeance (0.01 mol/s·m²·atm(1/4)) through the hydrogen-permeable membrane. This modeling and design approach sets the stage for guiding further development of multifunctional membrane reactor

  11. DESIGN RULE FOR PUMPING DEVICES IN A LOOP POLYMERIZATION REACTOR

    NARCIS (Netherlands)

    MOZARYN, W; JANSSEN, LPBM; VANDIERENDONCK, LL

    1994-01-01

    The mutual influence on the pumping capacity of independently driven pumping-stirring devices in a polymerization loop reactor is presented. A simple method for the calculation of the minimum rotational speed for an axial pumping stirrer in relation to the pumping action of a helical ribbon is descr

  12. DESIGN AND APPLICATION OF FLUIDIZED BED PHOTOCATALYTIC REACTOR

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    @@Photocatalytic degradation of organic pollutant is a new and potential method to transform it to harmless inorganic material, such as CO2 and H2O. So far, most of photocatalytic reactors were cylinder or tabulate photoreactor. The relevant photocatalyst was TiO2 nanometer powder. Although a few investigators had aimed their research field to fluidized bed reactor, their reaction systems were of biphase, such as solid-liquid or solid-gas. Few people focused their research on the triphasic fluidized bed photocatalytic reactor[1]. Compared with traditional photoreactors, a triphasic fluidized bed photoreactor has more advantages[2]: (1) The solid photocatalyst can be separated easily. (2) Its configuration meets the requirement of higher surface area-to-volume ratio of photocatalytic, which is much lower in a fixed bed or a plate photoreactor. (3) The UV light can be used more efficiently. (4) The mass transfer conditions can be controlled and improved easily. (5) It suited to pilot-scale or large-scale operations. For the UV light penetration and photon efficiency should be considered, the photocatalytic reactor differed greatly from a typical fluidized bed reactor.

  13. Microwave radiation and reactor design influence microbial communities during methane fermentation.

    Science.gov (United States)

    Cydzik-Kwiatkowska, Agnieszka; Zieliński, Marcin; Jaranowska, Paulina

    2012-09-01

    The effect of reactor design and method of heating on the efficiency of methane fermentation and composition of microbial communities, especially methanogenic Archaea, were determined. The research was carried out using submerge- and trickling-bed reactors fed with wastewater and the heat supply into the reactors included a convection heating method and microwave radiation. The polymerase chain reaction-denaturing gradient gel electrophoresis and relative real-time PCR were used in order to assess the biofilm communities. The best fermentation results and the highest abundance of methanogenic Archaea in biomass were observed in microwave heated trickling-bed reactors. The research proved that in reactors of identical design, the application of microwaves enabled a higher fermentation efficiency to be obtained and simultaneously increased the diversity of methanogenic Archaea communities that favors process stability. All the identified sequences of Archaea belonged to Methanosarcina sp., suggesting that species from this genera are susceptible to non-thermal effects of microwaves. There were no effects from microwaves on the bacterial communities in both types of reactors, however, the bacterial species composition varied in the reactors of different design.

  14. Counter-current membrane reactor for WGS process: Membrane design

    Energy Technology Data Exchange (ETDEWEB)

    Piemonte, Vincenzo; Favetta, Barbara [Department of Chemical Engineering Materials and Environment, University of Rome ' ' La Sapienza' ' , via Eudossiana 18, 00184 Rome (Italy); De Falco, Marcello [Faculty of Engineering, University Campus Bio-Medico of Rome, via Alvaro del Portillo 21, 00128 Rome (Italy); Basile, Angelo [CNR-ITM, c/o University of Calabria, Via Pietro Bucci, Cubo 17/C, 87030 Rende (CS) (Italy)

    2010-11-15

    Water gas shift (WGS) is a thermodynamically limited reaction which has to operate at low temperatures, reducing kinetics rate and increasing the amount of catalyst required to reach valuable CO conversions. It has been widely demonstrated that the integration of hydrogen selective membranes is a promising way to enhance WGS reactors performance: a Pd-based MR operated successfully overcoming the thermodynamic constraints of a traditional reactor thanks to the removal of hydrogen from reaction environment. In the first part of a MR, the H{sub 2} partial pressure starts from a minimum value since the reaction has not started. As a consequence, if the carrier gas in the permeation zone is sent in counter-current, which is the most efficient configuration, in the first reactor section the H{sub 2} partial pressure in reaction zone is low while in the permeation zone is high, potentially implying back permeation. This means a bad utilization of the first part of the membrane area and thus, a worsening of the MR performance with lower H{sub 2} recovery and lower CO conversion with respect to the case in which the whole selective surface is properly used. To avoid this problem different MR configurations were evaluated by a 1-D pseudo-homogeneous model, validated with WGS industrial data reported in scientific literature. It was demonstrated that the permeated H{sub 2} flow rate per membrane surface, i.e. the membrane flux, strongly improves if selective membrane is placed only in the second part of the reactor: in fact, if the membrane is placed at L{sub m}/L{sub tot} = 0.5, the membrane flux is 0.2 kmol/(m{sup 2}h) about, if it is placed along all reactor tube (L{sub m}/L{sub tot} = 1), flux is 0.05 kmol/(m{sup 2}h). The effect of the L/D reactor ratio and of the reactor wall temperature on the CO conversion were also assessed. (author)

  15. Design of slurry reactor for indirect liquefaction applications. Quarterly technical progress report, January 1991--March 1991

    Energy Technology Data Exchange (ETDEWEB)

    Prakash, A.; Bendale, P.G.

    1991-04-08

    The objective of this project is to design a conceptual slurry reactor for two indirect liquefaction applications; production of methanol and production of hydrogen fuels via Fischer-Tropsch route. The work will be accomplished by the formulation of reactor models for both the processes and use computer simulation. Process data, kinetic and thermodynamic data, heat and mass transfer data and hydrodynamic data will be used in the mathematical models to describe the slurry reactor for each of the two processes. The cost of current vapor phase reactor systems will be compared with cost estimated for the slurry reactor systems. For the vapor phase systems, upstream and downstream processing equipments may have to be included during cost analysis for a meaningful cost comparison.

  16. Design of Reactor Coolant Pump Seal Online Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2008-05-15

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

  17. Safe design of cooled tubular reactors for exothermic, multiple reactions; parallel reactions—II: The design and operation of an ethylene oxide reactor

    NARCIS (Netherlands)

    Westerterp, K.R.; Ptasiński, K.J.

    1984-01-01

    In part I a model and criteria have been developed for the safe design and operation of cooled tubular reactors for multiple reactions of the parallel type. In this Part II the model is extended to parallel reactions with an arbitrary stoichiometry. The results are applied to the industrial process

  18. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. Rempe; D. Knudson; J. Daw; T. Unruh; B. Chase; R. Schley; J. Palmer; K. Condie

    2014-01-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

  19. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

  20. Design and status of the AGS booster accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Forsyth, E.B.; Lee, Y.Y.

    1987-01-01

    Comments are given on some areas of the design considered for the AGS Booster Accelerator, including lattice design, energy and repetition rate, injection, radio frequency system, and the vacuum system. The current status is then briefly described. (LEW)

  1. Inertial Fusion Energy reactor design studies: Prometheus-L, Prometheus-H. Volume 2, Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-03-01

    This report contains a review of design studies for Inertial Confinement reactor. This second of three volumes discussions is some detail the following: Objectives, requirements, and assumptions; rationale for design option selection; key technical issues and R&D requirements; and conceptual design selection and description.

  2. 78 FR 28896 - Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components

    Science.gov (United States)

    2013-05-16

    ... COMMISSION Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components... Regulatory Commission (NRC) is issuing Revision 2 to Regulatory Guide (RG) 1.57, ``Design Limits and Loading... clarity. This guide describes a method that the NRC staff considers acceptable for design limits...

  3. Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

  4. Design and analysis of a single stage to orbit nuclear thermal rocket reactor engine

    Energy Technology Data Exchange (ETDEWEB)

    Labib, Satira, E-mail: Satira.Labib@duke-energy.com; King, Jeffrey, E-mail: kingjc@mines.edu

    2015-06-15

    Graphical abstract: - Highlights: • Three NTR reactors are optimized for the single stage launch of 1–15 MT payloads. • The proposed rocket engines have specific impulses in excess of 700 s. • Reactivity and submersion criticality requirements are satisfied for each reactor. - Abstract: Recent advances in the development of high power density fuel materials have renewed interest in nuclear thermal rockets (NTRs) as a viable propulsion technology for future space exploration. This paper describes the design of three NTR reactor engines designed for the single stage to orbit launch of payloads from 1 to 15 metric tons. Thermal hydraulic and rocket engine analyses indicate that the proposed rocket engines are able to reach specific impulses in excess of 800 s. Neutronics analyses performed using MCNP5 demonstrate that the hot excess reactivity, shutdown margin, and submersion criticality requirements are satisfied for each NTR reactor. The reactors each consist of a 40 cm diameter core packed with hexagonal tungsten cermet fuel elements. The core is surrounded by radial and axial beryllium reflectors and eight boron carbide control drums. The 40 cm long reactor meets the submersion criticality requirements (a shutdown margin of at least $1 subcritical in all submersion scenarios) with no further modifications. The 80 and 120 cm long reactors include small amounts of gadolinium nitride as a spectral shift absorber to keep them subcritical upon submersion in seawater or wet sand following a launch abort.

  5. Development of guidelines for inelastic analysis in design of fast reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Kyotada [Japan Atomic Energy Agency (JAEA), 4002, Narita, O-arai, Ibaraki 311-1393 (Japan); Kasahara, Naoto [Japan Atomic Energy Agency (JAEA), 4002, Narita, O-arai, Ibaraki 311-1393 (Japan)], E-mail: kasahara.naoto@jaea.go.jp; Morishita, Masaki [Japan Atomic Energy Agency (JAEA), 4002, Narita, O-arai, Ibaraki 311-1393 (Japan); Shibamoto, Hiroshi; Inoue, Kazuhiko [The Japan Atomic Power Company, Dispatched to JAEA, 4002, Narita, O-arai, Ibaraki 311-1393 (Japan); Nakayama, Yasunari [Kawasaki Heavy Industries, Ltd., 1-1, Kawasaki, Akashi, Hyogo 673-8666 (Japan)

    2008-02-15

    The interim guidelines for the application of inelastic analysis to design of fast reactor components were developed. These guidelines are referred from 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)'. The basic policies of the guidelines are more rational predictions compared with elastic analysis approach and a guarantee of conservative results for design conditions. The guidelines recommend two kinds of constitutive equations to estimate strains conservatively. They also provide the methods for modeling load histories and estimating fatigue and creep damage based on the results of inelastic analysis. The guidelines were applied to typical design examples and their results were summarized as exemplars to support users.

  6. Preliminary conceptual design for electrical and I and C system of a new research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, Y. K.; Kim, M. J.; Kim, H. K.; Ryu, J. S

    2004-01-01

    The core type and the process system design will be varied according to the reactor's application and capacity. A New research reactor is being designed by KAERI since 2002 and the process systems are not fixed yet. But control and instrument systems are similar to each other even though the application and the size are not same. So the C and I system that encompasses reactor protection system, reactor control system, and computer system was designed conceptually according to the requirements based on new digital technology and HANARO's proven design. The plant electrical system consists of off-site system that delivers bulk electrical power to the reactor site and on-site system that distributes and controls electrical power at the facility. The electrical system includes building service system that consist of lighting, communication, fire detection, grounding, cathodic protection, etc. also. This report describes the design requirements of on-site and off-site electric power system that set up from the codes and standards and the conceptual design based on the design requirements.

  7. Reactor safeguards system assessment and design. Volume I

    Energy Technology Data Exchange (ETDEWEB)

    Varnado, G.B.; Ericson, D.M. Jr.; Daniel, S.L.; Bennett, H.A.; Hulme, B.L.

    1978-06-01

    This report describes the development and application of a methodology for evaluating the effectiveness of nuclear power reactor safeguards systems. Analytic techniques are used to identify the sabotage acts which could lead to release of radioactive material from a nuclear power plant, to determine the areas of a plant which must be protected to assure that significant release does not occur, to model the physical plant layout, and to evaluate the effectiveness of various safeguards systems. The methodology was used to identify those aspects of reactor safeguards systems which have the greatest effect on overall system performance and which, therefore, should be emphasized in the licensing process. With further refinements, the methodology can be used by the licensing reviewer to aid in assessing proposed or existing safeguards systems.

  8. IAEA designated international centre based on research reactors (ICERR)

    Energy Technology Data Exchange (ETDEWEB)

    Di Tigliole, Andrea Borio; Bradley, Edward; Khoroshev, Mikhail; Marshall, Frances; Morris, Charles; Tozser, Sandor [International Atomic Energy Agency, Vienna (Austria). Dept. of Nuclear Energy

    2016-04-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. These programmes will soon have achieved their goals. However, the needs of the nuclear community dictate that the majority of the research reactors continues to operate using low enriched uranium (LEU) fuel in order to meet the varied mission objectives. As a result, inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. In view of this fact, the IAEA drew up a report presenting available reprocessing and recycling services for RR SNF.

  9. Seismic design technology for Breeder Reactor structures. Volume 3: special topics in reactor structures

    Energy Technology Data Exchange (ETDEWEB)

    Reddy, D.P. (ed)

    1983-04-01

    This volume is divided into six chapters: analysis techniques, equivalent damping values, probabilistic design factors, design verifications, equivalent response cycles for fatigue analysis, and seismic isolation. (JDB)

  10. Conceptual design based on scale laws and algorithms for sub-critical transmutation reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwang Gu; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    In order to conduct the effective integration of computer-aided conceptual design for integrated nuclear power reactor, not only is a smooth information flow required, but also decision making for both conceptual design and construction process design must be synthesized. In addition to the aboves, the relations between the one step and another step and the methodologies to optimize the decision variables are verified, in this paper especially, that is, scaling laws and scaling criteria. In the respect with the running of the system, the integrated optimization process is proposed in which decisions concerning both conceptual design are simultaneously made. According to the proposed reactor types and power levels, an integrated optimization problems are formulated. This optimization is expressed as a multi-objective optimization problem. The algorithm for solving the problem is also presented. The proposed method is applied to designing a integrated sub-critical reactors. 6 refs., 5 figs., 1 tab. (Author)

  11. Investigation of biological reactor designs for treatment of methanol and thiodiglycol waste streams

    Energy Technology Data Exchange (ETDEWEB)

    Sines, B.J.; Teather, E.W.; Weigand, W.A. (Univ. of Maryland, College Park, MD (United States)); Harvey, S.P. (Army Edgewood, Aberdeen Proving Ground, MD (United States))

    Biological reactor designs for the degradation of the toxic compounds methanol and thiodiglycol are compared to determine the smallest volume. Both compounds exhibit substrate-inhibited cell growth behavior. Design equations were used to simulate a continuous stirred tanks in series, and an optimized repeated fed-batch reactor. Thiodiglycol is the primary hydrolysis product of sulfur mustard (2,2[prime]-dichlorodiethyl sulfide), commonly referred to as [open quotes]mustard gas[close quotes]. Experimental data for the growth of Alcaligenes xylosoxidans xylosoxidans (SH42) on thiodiglycol was fit by an Andrews type inhibition equation, while the data and model for the growth of methanol was taken from the literature. The simulation results indicate that the repeated fed-batch reactor leads to significant volume reduction compared to the other two reactor configurations. 11 refs., 7 figs., 7 tabs.

  12. Status of the spent fuel in the reactor buildings of Fukushima Daiichi 1–4

    Energy Technology Data Exchange (ETDEWEB)

    Jäckel, Bernd S., E-mail: bernd.jaeckel@psi.ch

    2015-03-15

    The ratios of the radionuclides Cs-134g and Cs-137 deduced from measurements of liquid samples from the spent fuel pools in Fukushima Daiichi 1–4 are used to interpret the status of the spent fuel assemblies in the pools of the damaged reactor buildings. The different natures of the production of Cs-134g (neutron capture product of Cs-133) and Cs-137 (cumulative fission product from mass chain 137) and the different half-lives (2.06 years and 30.17 years respectively) require a complicated calculation of the mass and activity of the two nuclides. These masses are depending on the local burn up of the fuel, the burn up history and the radioactive decay. Calculation of the neutron capture product Cs-134g is particularly complicated, because the production of Cs-133 (stable cumulative fission product from mass chain 133) has to be taken into account. The neutron capture cross section for Cs-133 for thermal neutrons is well known, but the energy spectrum of the neutrons in a reactor includes higher energies according to the degree of moderation. Therefore the cross section was fitted from a gamma scan of spent fuel rods in a hot cell. The method of the calculation of the nuclide activities and the interpretation of the gamma measurements of the spent fuel pool samples from Fukushima Daiichi 1–4 are described in detail. It could be shown that at most only very minor mechanical damage of some spent fuel elements occurred during the accident and the later phase of the clearing work.

  13. STATUS OF HIGH FLUX ISOTOPE REACTOR IRRADIATION OF SILICON CARBIDE/SILICON CARBIDE JOINTS

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [ORNL; Koyanagi, Takaaki [ORNL; Kiggans, Jim [ORNL; Cetiner, Nesrin [ORNL; McDuffee, Joel [ORNL

    2014-09-01

    Development of silicon carbide (SiC) joints that retain adequate structural and functional properties in the anticipated service conditions is a critical milestone toward establishment of advanced SiC composite technology for the accident-tolerant light water reactor (LWR) fuels and core structures. Neutron irradiation is among the most critical factors that define the harsh service condition of LWR fuel during the normal operation. The overarching goal of the present joining and irradiation studies is to establish technologies for joining SiC-based materials for use as the LWR fuel cladding. The purpose of this work is to fabricate SiC joint specimens, characterize those joints in an unirradiated condition, and prepare rabbit capsules for neutron irradiation study on the fabricated specimens in the High Flux Isotope Reactor (HFIR). Torsional shear test specimens of chemically vapor-deposited SiC were prepared by seven different joining methods either at Oak Ridge National Laboratory or by industrial partners. The joint test specimens were characterized for shear strength and microstructures in an unirradiated condition. Rabbit irradiation capsules were designed and fabricated for neutron irradiation of these joint specimens at an LWR-relevant temperature. These rabbit capsules, already started irradiation in HFIR, are scheduled to complete irradiation to an LWR-relevant dose level in early 2015.

  14. Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe; D. L. Knudson; J. E. Daw

    2011-03-01

    The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support U.S. leadership in nuclear science and technology. By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort is to prove new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this review, recommendations were made with respect to what instrumentation is needed at the ATR; and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. In 2009, a report was issued documenting this instrumentation development strategy and initial progress toward accomplishing instrumentation development program objectives. This document reports progress toward implementing this strategy in 2010.

  15. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    Science.gov (United States)

    Muhamad, Shalina Sheik; Hamzah, Mohd Arif Arif B.

    2014-02-01

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP).

  16. Design of unique pins for irradiation of higher actinides in a fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Basmajian, J.A.; Birney, K.R.; Weber, E.T.; Adair, H.L.; Quinby, T.C.; Raman, S.; Butler, J.K.; Bateman, B.C.; Swanson, K.M.

    1982-03-01

    The actinides produced by transmutation reactions in nuclear reactor fuels are a significant factor in nuclear fuel burnup, transportation and reprocessing. Irradiation testing is a primary source of data of this type. A segmented pin design was developed which provides for incorporation of multiple specimens of actinide oxides for irradiation in the UK's Prototype Fast Reactor (PFR) at Dounreay Scotland. Results from irradiation of these pins will extend the basic neutronic and material irradiation behavior data for key actinide isotopes.

  17. Reactor Design and Decommissioning - An Overview of International Activities in Post Fukushima Era1 - 12396

    Energy Technology Data Exchange (ETDEWEB)

    Devgun, Jas S. [Nuclear Power Technologies, Sargent and Lundy LLC, Chicago, IL (United States); Laraia, Michele [private consultant, formerly from IAEA, Kolonitzgasse 10/2, 1030, Vienna (Austria); Pescatore, Claudio [OECD, Nuclear Energy Agency, Issy-les-Moulineaux, Paris (France); Dinner, Paul [International Atomic Energy Agency, Wagramerstrasse 5, A-1400 Vienna (Austria)

    2012-07-01

    Accidents at the Fukushima Dai-ichi reactors as a result of the devastating earthquake and tsunami of March 11, 2011 have not only dampened the nuclear renaissance but have also initiated a re-examination of the design and safety features for the existing and planned nuclear reactors. Even though failures of some of the key site features at Fukushima can be attributed to events that in the past would have been considered as beyond the design basis, the industry as well as the regulatory authorities are analyzing what features, especially passive features, should be designed into the new reactor designs to minimize the potential for catastrophic failures. It is also recognized that since the design of the Fukushima BWR reactors which were commissioned in 1971, many advanced safety features are now a part of the newer reactor designs. As the recovery efforts at the Fukushima site are still underway, decisions with respect to the dismantlement and decommissioning of the damaged reactors and structures have not yet been finalized. As it was with Three Mile Island, it could take several decades for dismantlement, decommissioning and clean up, and the project poses especially tough challenges. Near-term assessments have been issued by several organizations, including the IAEA, the USNRC and others. Results of such investigations will lead to additional improvements in system and site design measures including strengthening of the anti-tsunami defenses, more defense-in-depth features in reactor design, and better response planning and preparation involving reactor sites. The question also arises what would the effect be on the decommissioning scene worldwide, and what would the effect be on the new reactors when they are eventually retired and dismantled. This paper provides an overview of the US and international activities related to recovery and decommissioning including the decommissioning features in the reactor design process and examines these from a new

  18. Overview of pool hydraulic design of Indian prototype fast breeder reactor

    Indian Academy of Sciences (India)

    K Velusamy; P Chellapandi; S C Chetal; Baldev Raj

    2010-04-01

    Thermal hydraulics plays an important role in the design of liquid metal cooled fast breeder reactor components, where thermal loads are dominant. Detailed thermal hydraulic investigations of reactor components considering multi-physics heat transfer are essential for choosing optimum designs among the various possibilities. Pool hydraulics is multi-dimensional in nature and simple one-dimensional treatment for the same is often inadequate. Computational Fluid Dynamics (CFD) plays a critical role in the design of pool type reactors and becomes an increasingly popular tool, thanks to the advancements in computing technology. In this paper, thermal hydraulic characteristics of a fast breeder reactor, design limits and challenging thermal hydraulic investigations carried out towards successful design of Indian Prototype Fast Breeder Reactor (PFBR) that is under construction, are highlighted. Special attention is paid to phenomena like thermal stratification, thermal stripping, gas entrainment, inter-wrapper flow in decay heat removal and multiphysics cellular convection. The issues in these phenomena and the design solutions to address them satisfactorily are elaborated. Experiments performed for special phenomena, which are not amenable for CFD treatment and experiments carried out for validation of the computer codes have also been described.

  19. A breed and burn reshuffling scheme for an Astrid-like reactor concept design

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois L, J. L., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2016-09-15

    The greenhouse emissions has a serious impact on environmental terms, reason why energy supply needs to be provided by low carbon emission technologies. Nuclear power is and environmentally friendly energy with future concepts and designs that ensure safety, and in this paper is taken as a solution to be considered in the energy supply mix. This study presents an extension on the operation of the Advanced Sodium Technological Reactor for Industrial demonstration (Astrid) nuclear reactor through a breed and burn operation to extend the operation of the reactor. The Astrid nuclear reactor is a fourth generation sodium-cooled fast reactor of 1500 MWt h, and it considers an innovative design: the low void effect core (Cfv: Coeur a Faible effet de Vidange sodium) due to the reactor configuration and the radial and axial position of the fuel subassemblies. Previous research in the Astrid-like cores aimed the model validation of a conventional oxide-fueled core and its comparison with a proposed metallic-fueled core. Taking into account the amount of fissile material (mainly 239-Pu) after the first cycle, a reshuffling scheme was suggested, which consists in changing strategically the position of the nuclear fuel assemblies when the reactivity drops near to the critical state of the reactor. Two different reshuffling schemes were simulated in every developed model, having operation extensions of 805 days and 1775 days for the oxide and metallic designs respectively. The implementation of the reshuffling schemes in the developed models enhanced the fuel utilization and could save up to 2.20 and 5.96 tons of plutonium for oxide and metallic designs respectively, which has an economic impact. The breeding of 239-Pu achieved in the fertile zone of the metallic design reached half of the initial concentration of the 239-Pu in the fissile zone and for the oxide design, the breeding reached one third of the initial concentration of the 239-Pu in the fissile zone. (Author)

  20. Safe design of cooled tubular reactors for exothermic, multiple reactions; parallel reactions—I: Development of criteria

    NARCIS (Netherlands)

    Westerterp, K.R.; Ptasiński, K.J.

    1984-01-01

    Previously reported design criteria for cooled tubular reactors are based on the prevention of reactor temperature run away and were developed for single reactions only. In this paper it is argued that such criteri a should be based on the reactor selectivity, from which eventually a maximum

  1. KJRR-FAI Status Report of Conceptual Design Activities

    Energy Technology Data Exchange (ETDEWEB)

    N.E. Woolstenhulme; B.P. Nielson; D.B. Chapman; J.W. Nielsen; P.E. Murray; D.S. Crawford; S.D. Snow

    2013-12-01

    The Korea Atomic Energy Research Institute has initiated the Ki-Jang Research Reactor (KJRR) project to construct a new dedicated radio-isotope production facility in the KiJang province of South Korea. The KJRR will employ a uranium-molybdenum dispersion plate-type fuel clad in aluminum. The KJRR fuel assembly design will undergo irradiation in the Advanced Test Reactor (ATR) as part of the regulatory qualification of the fuel. The Idaho National Laboratory performed a multi-disciplined conceptual design effort and found that one full-size KJRR fuel assembly can be irradiated in the ATR’s north east flux trap. The analyses accomplished during the conceptual design phase are sufficient to prove viability of the overall design and irradiation campaign. Requirements for fission power can be met. The desired burnup can be achieved well within 15% depending on reactor operating availability. Mechanical design and structural analysis show that structural integrity of the irradiation test is maintained. It is recommended that future detailed design efforts be based on the concept described in this report.

  2. Status of the Neutrino Factory accelerator design studies

    CERN Document Server

    Prior, Gersende

    2013-01-01

    This document is a review of the present status of the Neutrino Factory design study, after the publication of the Interim Design Report and before the publication of the Reference Design Report. The different components of the accelerator as well as their current design stage and future tasks are described here.

  3. Design and axial optimization of nuclear fuel for BWR reactors; Diseno y optimizacion axial de combustible nuclear para reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Garcia V, M.A

    2006-07-01

    In the present thesis, the modifications made to the axial optimization system based on Tabu Search (BT) for the axial design of BWR fuel type are presented, developed previously in the Nuclear Engineering Group of the UNAM Engineering Faculty. With the modifications what is mainly looked is to consider the particular characteristics of the mechanical design of the GE12 fuel type, used at the moment in the Laguna Verde Nucleo electric Central (CNLV) and that it considers the fuel bars of partial longitude. The information obtained in this thesis will allow to plan nuclear fuel reloads with the best conditions to operate in a certain cycle guaranteeing a better yield and use in the fuel burnt, additionally people in charge in the reload planning will be favored with the changes carried out to the system for the design and axial optimization of nuclear fuel, which facilitate their handling and it reduces their execution time. This thesis this developed in five chapters that are understood in the following way in general: Chapter 1: It approaches the basic concepts of the nuclear energy, it describes the physical and chemical composition of the atoms as well as that of the uranium isotopes, the handling of the uranium isotope by means of the nuclear fission until arriving to the operation of the nuclear reactors. Chapter 2: The nuclear fuel cycle is described, the methods for its extraction, its conversion and its enrichment to arrive to the stages of the nuclear fuel management used in the reactors are described. Beginning by the radial design, the axial design and the core design of the nuclear reactor related with the fuel assemblies design. Chapter 3: the optimization methods of nuclear fuel previously used are exposed among those that are: the genetic algorithms method, the search methods based on heuristic rules and the application of the tabu search method, which was used for the development of this thesis. Chapter 4: In this part the used methodology to the

  4. DESIGN AND LAYOUT CONCEPTS FOR COMPACT, FACTORY-PRODUCED, TRANSPORTABLE, GENERATION IV REACTOR SYSTEMS

    Energy Technology Data Exchange (ETDEWEB)

    Mynatt Fred R.; Townsend, L.W.; Williamson, Martin; Williams, Wesley; Miller, Laurence W.; Khan, M. Khurram; McConn, Joe; Kadak, Andrew C.; Berte, Marc V.; Sawhney, Rapinder; Fife, Jacob; Sedler, Todd L.; Conway, Larry E.; Felde, Dave K.

    2003-11-12

    The purpose of this research project is to develop compact (100 to 400 MWe) Generation IV nuclear power plant design and layout concepts that maximize the benefits of factory-based fabrication and optimal packaging, transportation and siting. The reactor concepts selected were compact designs under development in the 2000 to 2001 period. This interdisciplinary project was comprised of three university-led nuclear engineering teams identified by reactor coolant type (water, gas, and liquid metal) and a fourth Industrial Engineering team. The reactors included a Modular Pebble Bed helium-cooled concept being developed at MIT, the IRIS water-cooled concept being developed by a team led by Westinghouse Electric Company, and a Lead-Bismuth-cooled concept developed by UT. In addition to the design and layout concepts this report includes a section on heat exchanger manufacturing simulations and a section on construction and cost impacts of proposed modular designs.

  5. Mechanical Design Concept of Fuel Assembly for Prototype GEN-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, K. H.; Lee, C. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The prototype GEN-IV sodium-cooled fast reactor (PGSFR) is an advanced fast reactor plant design that utilizes compact modular pool-type reactors sized to enable factory fabrication and an affordable prototype test for design certification at minimum cost and risk. The design concepts of the fuel assembly (FA) were introduced for a PGSFR. Unlike that for the pressurized water reactor, there is a neutron shielding concept in the FA and recycling metal fuel. The PGSFR core is a heterogeneous, uranium-10% zirconium (U-10Zr) metal alloy fuel design with 112 assemblies: 52 inner core fuel assemblies, 60 outer core fuel assemblies, 6 primary control assemblies, 3 secondary control assemblies, 90 reflector assemblies and 102 B4C shield assemblies. This configuration is shown in Fig. 1. The core is designed to produce 150 MWe with an average temperature rise of 155 .deg. C. The inlet temperature is 390 .deg. C and the bulk outlet temperature is 545 .deg. C. The core height is 900 mm and the gas plenum length is 1,250 mm. A mechanical design of a fuel assembly for a PGSFR was established. The mechanical design concepts are well realized in the design. In addition to this, the analytical and experimental works will be carries out for verifying the design soundness.

  6. Design and construction of a cascading pressure reactor prototype for solar-thermochemical hydrogen production

    Science.gov (United States)

    Ermanoski, Ivan; Grobbel, Johannes; Singh, Abhishek; Lapp, Justin; Brendelberger, Stefan; Roeb, Martin; Sattler, Christian; Whaley, Josh; McDaniel, Anthony; Siegel, Nathan P.

    2016-05-01

    Recent work regarding the efficiency maximization for solar thermochemical fuel production in two step cycles has led to the design of a new type of reactor—the cascading pressure reactor—in which the thermal reduction step of the cycle is completed in multiple stages, at successively lower pressures. This approach enables lower thermal reduction pressures than in single-staged reactors, and decreases required pump work, leading to increased solar to fuel efficiencies. Here we report on the design and construction of a prototype cascading pressure reactor and testing of some of the key components. We especially focus on the technical challenges particular to the design, and their solutions.

  7. Design and R&D Progress of China Lead-Based Reactor for ADS Research Facility

    Directory of Open Access Journals (Sweden)

    Yican Wu

    2016-03-01

    Full Text Available In 2011, the Chinese Academy of Sciences launched an engineering project to develop an accelerator-driven subcritical system (ADS for nuclear waste transmutation. The China Lead-based Reactor (CLEAR, proposed by the Institute of Nuclear Energy Safety Technology, was selected as the reference reactor for ADS development, as well as for the technology development of the Generation IV lead-cooled fast reactor. The conceptual design of CLEAR-I with 10 MW thermal power has been completed. KYLIN series lead-bismuth eutectic experimental loops have been constructed to investigate the technologies of the coolant, key components, structural materials, fuel assembly, operation, and control. In order to validate and test the key components and integrated operating technology of the lead-based reactor, the lead alloy-cooled non-nuclear reactor CLEAR-S, the lead-based zero-power nuclear reactor CLEAR-0, and the lead-based virtual reactor CLEAR-V are under realization.

  8. Adaptive Controller Design for Continuous Stirred Tank Reactor

    Directory of Open Access Journals (Sweden)

    K. Prabhu

    2014-09-01

    Full Text Available Continues Stirred Tank Reactor (CSTR is an important issue in chemical process and a wide range of research in the area of chemical engineering. Temperature Control of CSTR has been an issue in the chemical control engineering since it has highly non-linear complex equations. This study presents problem of temperature control of CSTR with the adaptive Controller. The Simulation is done in MATLAB and result shows that adaptive controller is an efficient controller for temperature control of CSTR than PID controller.

  9. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-01

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

  11. NEET Enhanced Micro Pocket Fission Detector for High Temperature Reactors - FY15 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ugorowski, Phil [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ito, Takashi [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    A new project, that is a collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core parallel plate fission chamber and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year one of this three year project. Highlights from research accomplishments include: A joint collaboration was initiated between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. An updated HT MPFD design was developed. New high temperature-compatible materials for HT MPFD construction were procured. Construction methods to support the new design were evaluated at INL. Laboratory evaluations of HT MPFD were initiated. Electrical contact and fissile material plating has been performed at KSU. Updated detector electronics are undergoing evaluations at KSU. A

  12. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

  13. Safe design and operation of tank reactors for multiple reactions: Uniqueness and multiplicity

    NARCIS (Netherlands)

    Westerterp, K.R.; Jansma, E.

    1985-01-01

    A method is developed to design a tank reactor for unique operation and for two simultaneous or consecutive reactions of the first order. Dimensionless groups are introduced which are either exclusively representative for the properties of the reaction system or exclusively for the design and

  14. Reactor Neutrinos

    OpenAIRE

    Soo-Bong Kim; Thierry Lasserre; Yifang Wang

    2013-01-01

    We review the status and the results of reactor neutrino experiments. Short-baseline experiments have provided the measurement of the reactor neutrino spectrum, and their interest has been recently revived by the discovery of the reactor antineutrino anomaly, a discrepancy between the reactor neutrino flux state of the art prediction and the measurements at baselines shorter than one kilometer. Middle and long-baseline oscillation experiments at Daya Bay, Double Chooz, and RENO provided very ...

  15. NEET Enhanced Micro-Pocket Fission Detector for High Temperature Reactors - FY16 Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Unruh, Troy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Reichenberger, Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States); Stevenson, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tsai, Kevin [Idaho National Lab. (INL), Idaho Falls, ID (United States); McGregor, Douglas [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Atomic Energy Agency, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), has been initiated by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program for developing and testing High Temperature Micro-Pocket Fission Detectors (HT MPFD), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package for temperatures up to 800 °C. The MPFD technology utilizes a small, multi-purpose, robust, in-core fission chambers and thermocouple. As discussed within this report, the small size, variable sensitivity, and increased accuracy of the MPFD technology represent a revolutionary improvement over current methods used to support irradiations in US Material Test Reactors (MTRs). Previous research conducted through NEET ASI1-3 has shown that the MPFD technology could be made robust and was successfully tested in a reactor core. This new project will further the MPFD technology for higher temperature regimes and other reactor applications by developing a HT MPFD suitable for temperatures up to 800 °C. This report summarizes the research progress for year two of this three year project. Highlights from research accomplishments include: • Continuation of a joint collaboration between INL, KSU, and CEA. Note that CEA is participating at their own expense because of interest in this unique new sensor. • An updated parallel wire HT MPFD design was developed. • Program support for HT MPFD deployments was given to Accident Tolerant Fuels (ATF) and Advanced Gas-cooled Reactor (AGR) irradiation test programs. • Quality approved materials for HT MPFD construction were procured by irradiation test programs for upcoming deployments. • KSU improved and performed electrical contact and fissile material plating.

  16. Gas-cooled thorium reactor with fuel block of the unified design

    Directory of Open Access Journals (Sweden)

    I.V. Shamanin

    2015-11-01

    Analysis of information materials pertaining to the use of thorium as fuel element in rector facilities of the new generation and of its future potential was performed in the present study. Results of the first phase of neutronics studies of 3D model of high-temperatures gas-cooled reactor facility on the basis of unified design of the fuel block are presented. Calculation 3D model was developed using the software code of the MCU-5 series. Several optimal configurations of the reactor core were selected according to the results of comparison of neutronics characteristics of the examined options for the purpose of development of small-size modular nuclear power installations with power up to 60MW. Results of calculations of reactivity margin of the reactor, neutron flux distribution and power density profiles are presented for the selected options of reactor core configuration.

  17. Dynamic Modeling for the Design and Cyclic Operation of an Atomic Layer Deposition (ALD Reactor

    Directory of Open Access Journals (Sweden)

    Curtisha D. Travis

    2013-08-01

    Full Text Available A laboratory-scale atomic layer deposition (ALD reactor system model is derived for alumina deposition using trimethylaluminum and water as precursors. Model components describing the precursor thermophysical properties, reactor-scale gas-phase dynamics and surface reaction kinetics derived from absolute reaction rate theory are integrated to simulate the complete reactor system. Limit-cycle solutions defining continuous cyclic ALD reactor operation are computed with a fixed point algorithm based on collocation discretization in time, resulting in an unambiguous definition of film growth-per-cycle (gpc. A key finding of this study is that unintended chemical vapor deposition conditions can mask regions of operation that would otherwise correspond to ideal saturating ALD operation. The use of the simulator for assisting in process design decisions is presented.

  18. A Virtual Reality Framework to Optimize Design, Operation and Refueling of GEN-IV Reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Rizwan-uddin; Nick Karancevic; Stefano Markidis; Joel Dixon; Cheng Luo; Jared Reynolds

    2008-04-23

    many GEN-IV candidate designs are currently under investigation. Technical issues related to material, safety and economics are being addressed at research laboratories, industry and in academia. After safety, economic feasibility is likely to be the most important crterion in the success of GEN-IV design(s). Lessons learned from the designers and operators of GEN-II (and GEN-III) reactors must play a vital role in achieving both safety and economic feasibility goals.

  19. Material Control and Accounting Design Considerations for High-Temperature Gas Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Trond Bjornard; John Hockert

    2011-08-01

    The subject of this report is domestic safeguards and security by design (2SBD) for high-temperature gas reactors, focusing on material control and accountability (MC&A). The motivation for the report is to provide 2SBD support to the Next Generation Nuclear Plant (NGNP) project, which was launched by Congress in 2005. This introductory section will provide some background on the NGNP project and an overview of the 2SBD concept. The remaining chapters focus specifically on design aspects of the candidate high-temperature gas reactors (HTGRs) relevant to MC&A, Nuclear Regulatory Commission (NRC) requirements, and proposed MC&A approaches for the two major HTGR reactor types: pebble bed and prismatic. Of the prismatic type, two candidates are under consideration: (1) GA's GT-MHR (Gas Turbine-Modular Helium Reactor), and (2) the Modular High-Temperature Reactor (M-HTR), a derivative of Areva's Antares reactor. The future of the pebble-bed modular reactor (PBMR) for NGNP is uncertain, as the PBMR consortium partners (Westinghouse, PBMR [Pty] and The Shaw Group) were unable to agree on the path forward for NGNP during 2010. However, during the technology assessment of the conceptual design phase (Phase 1) of the NGNP project, AREVA provided design information and technology assessment of their pebble bed fueled plant design called the HTR-Module concept. AREVA does not intend to pursue this design for NGNP, preferring instead a modular reactor based on the prismatic Antares concept. Since MC&A relevant design information is available for both pebble concepts, the pebble-bed HTGRs considered in this report are: (1) Westinghouse PBMR; and (2) AREVA HTR-Module. The DOE Office of Nuclear Energy (DOE-NE) sponsors the Fuel Cycle Research and Development program (FCR&D), which contains an element specifically focused on the domestic (or state) aspects of SBD. This Material Protection, Control and Accountancy Technology (MPACT) program supports the present work

  20. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Oktamuliani, Sri, E-mail: srioktamuliani@ymail.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id [Nuclear and Reactor Physics Laboratory, FMIPA, ITB, Physics Buildings, Jl. Ganesha 10, Bandung 40132 (Indonesia)

    2015-09-30

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.

  1. Standard Practice for Design of Surveillance Programs for Light-Water Moderated Nuclear Power Reactor Vessels

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice covers procedures for designing a surveillance program for monitoring the radiation-induced changes in the mechanical properties of ferritic materials in light-water moderated nuclear power reactor vessels. This practice includes the minimum requirements for the design of a surveillance program, selection of vessel material to be included, and the initial schedule for evaluation of materials. 1.2 This practice was developed for all light-water moderated nuclear power reactor vessels for which the predicted maximum fast neutron fluence (E > 1 MeV) at the end of license (EOL) exceeds 1 × 1021 neutrons/m2 (1 × 1017 n/cm2) at the inside surface of the reactor vessel. 1.3 This practice applies only to the planning and design of surveillance programs for reactor vessels designed and built after the effective date of this practice. Previous versions of Practice E185 apply to earlier reactor vessels. 1.4 This practice does not provide specific procedures for monitoring the radiation induced cha...

  2. Pebble bed modular reactor safeguards: developing new approaches and implementing safeguards by design

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Brian David [Los Alamos National Laboratory; Beddingfield, David H [Los Alamos National Laboratory; Durst, Philip [INL; Bean, Robert [INL

    2010-01-01

    The design of the Pebble Bed Modular Reactor (PBMR) does not fit or seem appropriate to the IAEA safeguards approach under the categories of light water reactor (LWR), on-load refueled reactor (OLR, i.e. CANDU), or Other (prismatic HTGR) because the fuel is in a bulk form, rather than discrete items. Because the nuclear fuel is a collection of nuclear material inserted in tennis-ball sized spheres containing structural and moderating material and a PBMR core will contain a bulk load on the order of 500,000 spheres, it could be classified as a 'Bulk-Fuel Reactor.' Hence, the IAEA should develop unique safeguards criteria. In a multi-lab DOE study, it was found that an optimized blend of: (i) developing techniques to verify the plutonium content in spent fuel pebbles, (ii) improving burn-up computer codes for PBMR spent fuel to provide better understanding of the core and spent fuel makeup, and (iii) utilizing bulk verification techniques for PBMR spent fuel storage bins should be combined with the historic IAEA and South African approaches of containment and surveillance to verify and maintain continuity of knowledge of PBMR fuel. For all of these techniques to work the design of the reactor will need to accommodate safeguards and material accountancy measures to a far greater extent than has thus far been the case. The implementation of Safeguards-by-Design as the PBMR design progresses provides an approach to meets these safeguards and accountancy needs.

  3. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  4. Multi-physics design and analyses of long life reactors for lunar outposts

    Science.gov (United States)

    Schriener, Timothy M.

    Future human exploration of the solar system is likely to include establishing permanent outposts on the surface of the Moon. These outposts will require reliable sources of electrical power in the range of 10's to 100's of kWe to support exploration and resource utilization activities. This need is best met using nuclear reactor power systems which can operate steadily throughout the long ˜27.3 day lunar rotational period, irrespective of location. Nuclear power systems can potentially open up the entire lunar surface for future exploration and development. Desirable features of nuclear power systems for the lunar surface include passive operation, the avoidance of single point failures in reactor cooling and the integrated power system, moderate operating temperatures to enable the use of conventional materials with proven irradiation experience, utilization of the lunar regolith for radiation shielding and as a supplemental neutron reflector, and safe post-operation decay heat removal and storage for potential retrieval. In addition, it is desirable for the reactor to have a long operational life. Only a limited number of space nuclear reactor concepts have previously been developed for the lunar environment, and these designs possess only a few of these desirable design and operation features. The objective of this research is therefore to perform design and analyses of long operational life lunar reactors and power systems which incorporate the desirable features listed above. A long reactor operational life could be achieved either by increasing the amount of highly enriched uranium (HEU) fuel in the core or by improving the neutron economy in the reactor through reducing neutron leakage and parasitic absorption. The amount of fuel in surface power reactors is constrained by the launch safety requirements. These include ensuring that the bare reactor core remains safely subcritical when submerged in water or wet sand and flooded with seawater in the unlikely

  5. Conceptual design for the superconducting magnet system of a pulsed DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Duchateau, J.-L., E-mail: jean-luc.duchateau@cea.fr [CEA/IRFM, 13108 St. Paul lez Durance Cedex (France); Hertout, P.; Saoutic, B.; Magaud, P.; Artaud, J.-F.; Giruzzi, G.; Bucalossi, J.; Johner, J.; Sardain, P.; Imbeaux, F.; Ané, J.-M.; Li-Puma, A. [CEA/IRFM, 13108 St. Paul lez Durance Cedex (France)

    2013-10-15

    Highlights: ► A 1D design approach of a pulsed DEMO reactor is presented. ► The main CS and TF conductor design criteria are presented. ► A typical major radius for a 2 GW DEMO is 9 m. ► A typical plasma magnetic field is 4.9 T. ► The pulse duration is 1.85 h for an aspect ratio of 3. -- Abstract: A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing. Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin. The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force. Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios. The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified.

  6. A Miniature Membrane Reactor for Evaluation of Process Design Options on the Enzymatic Degradation of Pectin

    DEFF Research Database (Denmark)

    Zainal Alam, Muhd Nazrul Hisham; Pinelo, Manuel; Arnous, Anis

    2011-01-01

    The objective of this paper is to assess if a membrane microbioreactor system could potentially be used to diagnose consequences of different process design and reactor operation options relevant for larger-scale enzymatic degradation of pectin reactions. The membrane microbioreactor prototype...... was demonstrated by performing a continuous enzymatic degradation of pectin experiment for a range of reactor conditions: different membrane molecular weight cutoff (MWCO) values, enzyme-to-substrate ratios (E/S), and substrate feeding rates (F) were assessed. Based on the experimental data, it was found...... design affected the membrane rejection profile. The results obtained thus underlined the suitability of a miniature membrane reactor system for evaluating different process design options that are relevant for larger-scale reactions of enzymatic pectin degradation....

  7. Application of objective provision tree to development of standard review plan for sodium-cooled fast reactor nuclear design

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Moo-Hoon; Suh, Namduk; Choi, Yongwon; Shin, Andong [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2016-06-15

    A systematic methodology was developed for the standard review plan for sodium-cooled fast reactor nuclear design. The process is first to develop an objective provision tree of sodium-cooled fast reactor for the reactivity control safety function. The provision tree is generally developed by designer to confirm whether the design satisfies the defense-in-depth concept. Then applicability of the current standard review plan of nuclear design for light water reactor to sodium-cooled fast reactor was evaluated and complemented by the developed objective provision tree.

  8. Applicability of GALE-86 Codes to Integral Pressurized Water Reactor designs

    Energy Technology Data Exchange (ETDEWEB)

    Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Rishel, Jeremy P. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2012-06-01

    This report describes work that Pacific Northwest National Laboratory is doing to assist the U.S. Nuclear Regulatory Commission (NRC) Office of New Reactors (NRO) staff in their reviews of applications for nuclear power plants using new reactor core designs. These designs include small integral PWRs (IRIS, mPower, and NuScale reactor designs), HTGRs, (pebble-bed and prismatic-block modular reactor designs) and SFRs (4S and PRISM reactor designs). Under this specific task, PNNL will assist the NRC staff in reviewing the current versions of the GALE codes and identify features and limitations that would need to be modified to accommodate the technical review of iPWR and mPower® license applications and recommend specific changes to the code, NUREG-0017, and associated NRC guidance. This contract is necessary to support the licensing of iPWRs with a near-term focus on the B&W mPower® reactor design. While the focus of this review is on the mPower® reactor design, the review of the code and the scope of recommended changes consider a revision of the GALE codes that would make them universally applicable for other types of integral PWR designs. The results of a detailed comparison between PWR and iPWR designs are reported here. Also included is an investigation of the GALE code and its basis and a determination as to the applicability of each of the bases to an iPWR design. The issues investigated come from a list provided by NRC staff, the results of comparing the PWR and iPWR designs, the parameters identified as having a large impact on the code outputs from a recent sensitivity study and the main bases identified in NUREG-0017. This report will provide a summary of the gaps in the GALE codes as they relate to iPWR designs and for each gap will propose what work could be performed to fill that gap and create a version of GALE that is applicable to integral PWR designs.

  9. Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Center

    2015-11-15

    This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

  10. Treatment of irradiation effects in structural design criteria for fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S. [Argonne National Lab., IL (United States). Fusion Power Program; Smith, P. [San Diego Joint Work Site, CA (United States). ITER Joint Central Team

    1997-03-01

    The irradiation environment experienced by the in-vessel components of fusion reactors such as the International Thermonuclear Experimental Reactor (ITER) presents structural design challenges not envisioned in the development of existing structural design criteria such as the ASME Code or RCC-MR. From the standpoint of structural design criteria, the most significant issues stem from the irradiation-induced changes in material properties, specifically the reduction of ductility, strain hardening capability, and fracture toughness with neutron irradiation. These effects call into question the basis of the design rules in existing structural design criteria which assume that only code-approved materials with high toughness, ductility and strain hardening capability will be used. The present paper reviews the basis of new rules that address these issues in Draft 5 of the interim ITER structural design criteria (ISDC) which was released recently for trial use by the ITER designers.

  11. Status of NDE research in the US-contributions of NDE to reactor safety and implementation of NDE technology

    Energy Technology Data Exchange (ETDEWEB)

    Ammirato, F. [EPRI, Charlotte, NC (United States)

    1999-08-01

    Power plant designers, plant owners, and regulators have developed inservice inspection (ISI) programs as part of their comprehensive approach to ensuring nuclear safety. This paper examines the role of ISI in reactor safety through several examples drawn from recent industry initiatives to address implementation of effective examination technology for nuclear power plant piping, and BWR and PWR reactor pressure vessels. These examples also illustrate the importance of well designed performance demonstration activities to support application of effective ISI. Finally, the efforts required to implement effective ISI technology for field inspection is addressed. (orig./DGE)

  12. Current status and future prospective of advanced radiation resistant oxide dispersion strengthened steel (ARROS) development for nuclear reactor system applications

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Kyu; Noh, Sang Hoon; Kang, Suk Hoon; Park, Jin Ju; Jin, Hyun Ju; Lee, Min Ku; Jang, Jin Sugn; Rhee, Chang Kyu [Nuclear Materials Development Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    As one of the Gen-IV nuclear energy systems, a sodium-cooled fast reactor (SFR) is being developed at the Korea Atomic Energy Research Institute. As a long-term national research project, advanced radiation resistant oxide dispersion strengthened steel (ARROS) is being developed as an in-core fuel cladding tube material for a SFR in the future. In this paper, the current status of ARROS development is reviewed and its future prospective is discussed.

  13. Current status and future plan of the G.A. Siwabessy Multipurpose Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hastowo, Hudi; Tarigan, Alim; Mardi, Alfahari [Centre for Multipurpose Reactor, National Atomic Energy Agency, Serpong (Indonesia)

    1998-10-01

    Since the first criticality in July 1987, the G.A. Siwabessy Multipurpose Reactor, RSG-GAS, in Serpong-Indonesia has been operated about 26.000 hours up to now. In the last two years the reactor is operated more than 5000 hours per year or equivalent to seven cycles a year. The reactor is utilized for conducting research studies and isotopes production. In the near future the core will be converted to silicide fuel to improve the core performance instead of oxide fuel. The planned maintenance activities are performed according to the schedule. The modifications and remedial maintenance are also performed to ensure that all structures, systems and important to safety are capable of performing as intended. The future activities of the G.A. Siwabessy reactor will be stressed to maintain the reliability and availability of the reactor operation and to optimize the reactor utilization. (author)

  14. Fluid modeling and design of gas channels of solar non-stoichiometric redox reactor

    Science.gov (United States)

    Kedlaya, Aditya

    The present numerical study in FLUENT analyzes the fluid flow field within a solar powered reactor designed for syngas production. The thermochemical reactor is based on continuous cycling of cerium oxide (ceria) in a non-stoichiometric reduction/oxidation cycle. The reactor uses a hollow cylinder of porous ceria which rotates through a high-temperature zone, by exposure to concentrated sunlight and partially reduced in an inert atmosphere due to flow of the sweep gas (N2), and then through a lower temperature zone where the reduced ceria is re-oxidized with a flow of CO2 and/or H2O, to produce CO and/or H2. In terms of fluid flow modeling, the issue of crossover of species (leakage) within the reactor is critical for proper functioning of the reactor. The first part of the work relates to the geometry and placement of the inlet/outlet gas channels for the reactor optimized to minimize crossover of the species. This is done by conducting a parametric study of geometric variables associated with the inlet/outlet geometry. A simplified 2D fluid flow reactor model which incorporates multi-species flow is used for this study. Further, 2D and 3D reactor models which capture the internal structure more accurately are used to refine the inlet/outlet design. The optimized reactor model is found to have an O2 crossover of 2%-6% and oxidizer crossover of 8%-21% at different flow rates of the sweep gas and the oxidizer studied. In the second part of the work, the reactor model is simulated under varying test conditions. Different working conditions include morphologies of the reactive material, rotational speed of the ceria ring and the recuperator, flow rates of sweep gas and the oxidizer, types of oxidizer (CO2, H2O). The 3D reactor model is also tested using one, two and three discrete inlet/outlet ports and compared with slot configuration.

  15. Application of Decomposition Methodology to Solve Integrated Process Design and Controller Design Problems for Reactor-Separator-Recycle System

    DEFF Research Database (Denmark)

    Abd.Hamid, Mohd-Kamaruddin; Sin, Gürkan; Gani, Rafiqul

    2010-01-01

    This paper presents the integrated process design and controller design (IPDC) for a reactor-separator-recycle (RSR) system and evaluates a decomposition methodology to solve the IPDC problem. Accordingly, the IPDC problem is solved by decomposing it into four hierarchical stages: (i) pre-analysi...... to the solution of IPDC problems for RSR systems.......This paper presents the integrated process design and controller design (IPDC) for a reactor-separator-recycle (RSR) system and evaluates a decomposition methodology to solve the IPDC problem. Accordingly, the IPDC problem is solved by decomposing it into four hierarchical stages: (i) pre...... the design of a RSR system involving consecutive reactions, A B -> C and shown to provide effective solutions that satisfy design, control and cost criteria. The advantage of the proposed methodology is that it is systematic, makes use of thermodynamic-process knowledge and provides valuable insights...

  16. Design of a three-layer hot-wall horizontal flow MOCVD reactor

    Institute of Scientific and Technical Information of China (English)

    Gu Chengyan; Lee Chengming; Liu Xianglin

    2012-01-01

    A new three-layer hot-wall horizontal flow metal-organic chemical vapor deposition (MOCVD) reactor is proposed.When the susceptor is heated,the temperature of the wall over the susceptor also increases to the same temperature.Furthermore,the flowing speed of the top layer is also increased by up to four times that of the bottom layer.Both methods effectively decrease the convection and make most of the metal organic (MO) gas and the reactive gas distribute at the bottom surface of the reactor.By selecting appropriate shapes,sizes,nozzles array,and heating area of the walls,the source gases are kept in a laminar flow state.Results of the numeric simulation indicate that the nitrogen is a good carrier to reduce the diffusion among the precursors before arriving at the substrate,which leads to the reduction of pre-reaction.To get a good comparison with the conventional MOCVD horizontal reactor,the two-layer horizontal MOCVD reactor is also investigated.The results indicate that a twolayer reactor cannot control the gas flow effectively when its size and shape are the same as that of the three-layer reactor,so that the concentration distributions of the source gases in the susceptor surface are much more uniform in the new design than those in the conventional one.

  17. Design of a management information system for the Shielding Experimental Reactor ageing management

    Energy Technology Data Exchange (ETDEWEB)

    He Jie, E-mail: hejiejoe@163.co [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2010-01-15

    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  18. An autonomous long-term fast reactor system and the principal design limitations of the concept

    Science.gov (United States)

    Tsvetkova, Galina Valeryevna

    The objectives of this dissertation were to find a principal domain of promising and technologically feasible reactor physics characteristics for a multi-purpose, modular-sized, lead-cooled, fast neutron spectrum reactor fueled with an advanced uranium-transuranic-nitride fuel and to determine the principal limitations for the design of an autonomous long-term multi-purpose fast reactor (ALM-FR) within the principal reactor physics characteristic domain. The objectives were accomplished by producing a conceptual design for an ALM-FR and by analysis of the potential ALM-FR performance characteristics. The ALM-FR design developed in this dissertation is based on the concept of a secure transportable autonomous reactor for hydrogen production (STAR-H2) and represents further refinement of the STAR-H2 concept towards an economical, proliferation-resistant, sustainable, multi-purpose nuclear energy system. The development of the ALM-FR design has been performed considering this reactor within the frame of the concept of a self-consistent nuclear energy system (SCNES) that satisfies virtually all of the requirements for future nuclear energy systems: efficient energy production, safety, self-feeding, non-proliferation, and radionuclide burning. The analysis takes into consideration a wide range of reactor design aspects including selection of technologically feasible fuels and structural materials, core configuration optimization, dynamics and safety of long-term operation on one fuel loading, and nuclear material non-proliferation. Plutonium and higher actinides are considered as essential components of an advanced fuel that maintains long-term operation. Flexibility of the ALM-FR with respect to fuel compositions is demonstrated acknowledging the principal limitations of the long-term burning of plutonium and higher actinides. To ensure consistency and accuracy, the modeling has been performed using state-of-the-art computer codes developed at Argonne National

  19. Design status of heavy ion injector program

    Energy Technology Data Exchange (ETDEWEB)

    Ballard, E.O.; Meyer, E.A.; Rutkowski, H.L.; Shurter, R.P.; Van Haaften, F.W.; Riepe, K.B.

    1985-01-01

    Design and development of a sixteen beam, heavy ion injector is in progress at Los Alamos National Laboratory (LANL) to demonstrate the injector technology for the High Temperature Experiment (HTE) proposed by Lawrence Livermore Laboratory (LBL). The injector design provides for individual ion sources mounted to a support plate defining the sixteen beam array. The beamlets are electrostatically accelerated through a series of electrodes inside an evacuated (10/sup -7/ torr) high voltage (HV) accelerating column.

  20. Preliminary Neutronic Design of High Burnup OTTO Cycle Pebble Bed Reactor

    Directory of Open Access Journals (Sweden)

    T. Setiadipura

    2015-04-01

    Full Text Available The pebble bed type High Temperature Gas-cooled Reactor (HTGR is among the interesting nuclear reactor designs in terms of safety and flexibility for co-generation applications. In addition, the strong inherent safety characteristics of the pebble bed reactor (PBR which is based on natural mechanisms improve the simplicity of the PBR design, in particular for the Once-Through-Then-Out (OTTO cycle PBR design. One of the important challenges of the OTTO cycle PBR design, and nuclear reactor design in general, is improving the nuclear fuel utilization which is shown by attaining a higher burnup value. This study performed a preliminary neutronic design study of a 200 MWt OTTO cycle PBR with high burnup while fulfilling the safety criteria of the PBR design.The safety criteria of the design was represented by the per-fuel-pebble maximum power generation of 4.5 kW/pebble. The maximum burnup value was also limited by the tested maximum burnup value which maintained the integrity of the pebble fuel. Parametric surveys were performed to obtain the optimized parameters used in this study, which are the fuel enrichment, per-pebble heavy metal (HM loading, and the average axial speed of the fuel. An optimum design with burnup value of 131.1 MWd/Kg-HM was achieved in this study which is much higher compare to the burnup of the reference design HTR-MODUL and a previously proposed OTTO-cycle PBR design. This optimum design uses 17% U-235 enrichment with 4 g HM-loading per fuel pebble

  1. High-flux first-wall design for a small reversed-field pinch reactor

    Science.gov (United States)

    Cort, G. E.; Graham, A. L.; Christensen, K. E.

    To achieve the goal of a commercially economical fusion power reactor, small physical size and high power density should be combined with simplicity (minimized use of high technology systems). The Reversed-Field Pinch (RFP) is a magnetic confinement device that promises to meet these requirements with power densities comparable to those in existing fission power plants. To establish feasibility of such an RFP reactor, a practical design for a first wall capable of withstanding high levels of cyclic neutron wall loadings is needed. Associated with the neutron flux in the proposed RFP reactor is a time averaged heat flux of 4.5 MW/sq m with a conservatively estimated transient peak approximately twice the average value. The design for a modular first wall made from a high-strength copper alloy that will meet these requirements of cyclic thermal loading is presented. The heat removal from the wall is by subcooled water flowing in straight tubes at high linear velocities.

  2. Biodiesel Reactor Design with Glycerol Separation to Increase Biodiesel Production Yield

    Directory of Open Access Journals (Sweden)

    Budy Rahmat

    2013-09-01

    Full Text Available The study consisted of reactor design used for transesterification process, effect of glycerol separation ontransesterification reaction, determination of biodiesel quality, and mass balance analysis. The reactor was designed byintegrating circulated pump/stirrer, static mixer, and sprayer that intensify the reaction in the outer tank reactor. The objective was to reduce the use of methanol in excess and to shorten the processing time. The results showed that thereactor that applied the glycerol separation was able to compensate for the decreased use of the reactant methanol from 6:1 to 5:1 molar ratio, and changed the mass balance in the product, including: (i the increase of biodiesel productionfrom 42.37% to 49.34%, and (ii the reduction of methanol in excess from 42.37% to 32.89%. The results suggested that the efficiency of biodiesel production could be increased with the glycerol separation engineering.

  3. Reactor Pressure Vessel Temperature Analysis for Prismatic and Pebble-Bed VHTR Designs

    Energy Technology Data Exchange (ETDEWEB)

    H. D. Gougar; C. B. Davis

    2006-04-01

    Analyses were performed to determine maximum temperatures in the reactor pressure vessel for two potential Very-High Temperature Reactor (VHTR) designs during normal operation and during a depressurized conduction cooldown accident. The purpose of the analyses was to aid in the determination of appropriate reactor vessel materials for the VHTR. The designs evaluated utilized both prismatic and pebble-bed cores that generated 600 MW of thermal power. Calculations were performed for fluid outlet temperatures of 900 and 950 °C, corresponding to the expected range for the VHTR. The analyses were performed using the RELAP5-3D and PEBBED-THERMIX computer codes. Results of the calculations were compared with preliminary temperature limits derived from the ASME pressure vessel code.

  4. Photolytic treatment of atrazine-contaminated water: products, kinetics, and reactor design.

    Science.gov (United States)

    Ye, Xuejun; Chen, Daniel; Li, Kuyen; Wang, Bin; Hopper, Jack

    2007-08-01

    This study investigates the products, kinetics, and reactor design of atrazine photolysis under 254-nm ultraviolet-C (UVC) irradiation. With an initial atrazine concentration of 60 microg/L (60 ppbm), only two products remain in detectable levels. Up to 77% of decomposed atrazine becomes hydroxyatrazine, the major product. Both atrazine and hydroxyatrazine photodecompose following the first-order rate equation, but the hydroxyatrazine photodecomposition rate is significantly slower than that of atrazine. For atrazine photodecomposition, the rate constant is proportional to the square of UVC output, but inversely proportional to the reactor volume. For a photochemical reactor design, a series of equations are proposed to calculate the needed UVC output power, water treatment capacity, and atrazine outlet concentration.

  5. Exploratory Design of a Reactor/Fuel Cycle Using Spent Nuclear Fuel Without Conventional Reprocessing - 13579

    Energy Technology Data Exchange (ETDEWEB)

    Bertch, Timothy C.; Schleicher, Robert W.; Rawls, John D. [General Atomics 3550 General Atomics Court San Diego, CA 92130 (United States)

    2013-07-01

    General Atomics has started design of a waste to energy nuclear reactor (EM2) that can use light water reactor (LWR) spent nuclear fuel (SNF). This effort addresses two problems: using an advanced small reactor with long core life to reduce nuclear energy overnight cost and providing a disposal path for LWR SNF. LWR SNF is re-fabricated into new EM2 fuel using a dry voloxidation process modeled on AIROX/ OREOX processes which remove some of the fission products but no heavy metals. By not removing all of the fission products the fuel remains self-protecting. By not separating heavy metals, the process remains proliferation resistant. Implementation of Energy Multiplier Module (EM2) fuel cycle will provide low cost nuclear energy while providing a long term LWR SNF disposition path which is important for LWR waste confidence. With LWR waste confidence recent impacts on reactor licensing, an alternate disposition path is highly relevant. Centered on a reactor operating at 250 MWe, the compact electricity generating system design maximizes site flexibility with truck transport of all system components and available dry cooling features that removes the need to be located near a body of water. A high temperature system using helium coolant, electricity is efficiently produced using an asynchronous high-speed gas turbine while the LWR SNF is converted to fission products. Reactor design features such as vented fuel and silicon carbide cladding support reactor operation for decades between refueling, with improved fuel utilization. Beyond the reactor, the fuel cycle is designed so that subsequent generations of EM2 reactor fuel will use the previous EM2 discharge, providing its own waste confidence plus eliminating the need for enrichment after the first generation. Additional LWR SNF is added at each re-fabrication to replace the removed fission products. The fuel cycle uses a dry voloxidation process for both the initial LWR SNF re-fabrication and later for EM2

  6. Design and Analysis of the Power Control System of the Fast Zero Energy Reactor FR-0

    Energy Technology Data Exchange (ETDEWEB)

    Schuh, N.J.H.

    1966-12-15

    This report describes the power control by means of the fine-control rod and the design of the control system of the fast zero energy reactor FR-0 located in Studsvik, Sweden. System requirements and some operational conditions were used as design criteria. Manual and automatic control is possible. Variable electronic end-stops for the control rod have been designed, because of the special construction of the reactor and control rod. Noise in the control system caused by the reactor, detector and electronics caused disturbances of the control system at the lower power levels. The noise power-spectrum was measured. Statistical design methods, using the measured noise power spectrum, were used to design filters, which will reduce the influence of the noise at the lower power levels. Root Loci sketches and Bode diagrams were used for stability analyses. The system was simulated on an analogue computer, taking into account even nonlinearities of the control system and noise. Typical cases of reactor operation were simulated and stability analysis performed.

  7. Design of a redundant meteorological station for a BWR reactor; Diseno de una estacion meteorologica redundante para un reactor BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, R.; Celis del Angel, L.; Bucio, F.; Rivero, T.; Palacios, J. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: ramses@nuclear.inin.mx

    2008-07-01

    In this work the design of a meteorological station for a reactor type BWR is proposed. Two independent channels of data acquisition that allow him to have a bigger readiness is exposed. It is incorporate sensors without mobile parts to measure speed, wind direction and pluvial precipitation. It also counts, with sensors of global solar radiation, net radiation, barometric pressure, relative humidity and ambient temperature; with them they are possible to be calculated, moreover, other variables as temperature differential, dew point and atmospheric stability. The sensors are placed on a tower to different heights and send their information (each second) to a local registration system, the one which in turn, it remits the data to the monitoring office so that a computer is linked with the system, display and management the information in real time and automatic way. The redundant structure allows that in the event of maintenance the data acquisition is not interrupted, even if the information is transferred to another place. In all the station sections it is used protocols of standard communication to allow that a great quantity of devices can be connected without major problem. The above-mentioned would allow to the operators in the control room to have reliable information during the whole time of the reactor operation. (Author)

  8. Fusion Reactor and Fusion Reactor Materials:Concept Design of the ITER Test Blanket Modules

    Institute of Scientific and Technical Information of China (English)

    HUANGJinhua; LIZaixing; ZHUYukun; HUGang

    2003-01-01

    Performances required: prospect to be adopted in DEMO. Shielding for V.V. and TFC in ITER. Design principles: the peak temperature and stress should not exceed technical limits. The structure of test blanket modules (TBM) should be simple for easy fabrication, and TBM should be robust for reliability.

  9. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  10. CURRENT STATUS OF INTEGRITY ASSESSMENT BY SIPPING SYSTEM OF SPENT FUEL BUNDLES IRRADIATED IN CANDU REACTOR

    Directory of Open Access Journals (Sweden)

    JONG-YOUL PARK

    2014-12-01

    Full Text Available In terms of safety and the efficient management of spent fuel storage, detecting failed fuel is one of the most important tasks in a CANada Deuterium Uranium (CANDU reactor operation. It has been successfully demonstrated that in a CANDU reactor, on-power failed fuel detection and location systems, along with alarm area gamma monitors, can detect and locate defective and suspect fuel bundles before discharging them from the reactor to the spent fuel storage bay. In the reception bay, however, only visual inspection has been used to identify suspect bundles. Gaseous fission product and delayed neutron monitoring systems cannot precisely distinguish failed fuel elements from each fuel bundle. This study reports the use of a sipping system in a CANDU reactor for the integrity assessment of spent fuel bundles. The integrity assessment of spent fuel bundles using this sipping system has shown promise as a nondestructive test for detecting a defective fuel bundle in a CANDU reactor.

  11. Nuclear Systems Enhanced Performance Program, Maintenance Cycle Extension in Advanced Light Water Reactor Design

    Energy Technology Data Exchange (ETDEWEB)

    Professor Neill Todreas

    2001-10-01

    A renewed interest in new nuclear power generation in the US has spurred interest in developing advanced reactors with features which will address the public's concerns regarding nuclear generation. However, it is economic performance which will dictate whether any new orders for these plants will materialize. Economic performance is, to a great extent, improved by maximizing the time that the plant is on-line generating electricity relative to the time spent off-line conducting maintenance and refueling. Indeed, the strategy for the advanced light water reactor plant IRIS (International Reactor, Innovative and Secure) is to utilize an eight year operating cycle. This report describes a formalized strategy to address, during the design phase, the maintenance-related barriers to an extended operating cycle. The top-level objective of this investigation was to develop a methodology for injecting component and system maintainability issues into the reactor plant design process to overcome these barriers. A primary goal was to demonstrate the applicability and utility of the methodology in the context of the IRIS design. The first step in meeting the top-level objective was to determine the types of operating cycle length barriers that the IRIS design team is likely to face. Evaluation of previously identified regulatory and investment protection surveillance program barriers preventing a candidate operating PWR from achieving an extended (48 month) cycle was conducted in the context of the IRIS design. From this analysis, 54 known IRIS operating cycle length barriers were identified. The resolution methodology was applied to each of these barriers to generate design solution alternatives for consideration in the IRIS design. The methodology developed has been demonstrated to narrow the design space to feasible design solutions which enable a desired operating cycle length, yet is general enough to have broad applicability. Feedback from the IRIS design team

  12. Boundary Layer of Photon Absorption Applied to Heterogeneous Photocatalytic Solar Flat Plate Reactor Design

    Directory of Open Access Journals (Sweden)

    Héctor L. Otálvaro-Marín

    2014-01-01

    Full Text Available This study provides information to design heterogeneous photocatalytic solar reactors with flat plate geometry used in treatment of effluents and conversion of biomass to hydrogen. The concept of boundary layer of photon absorption taking into account the efficient absorption of radiant energy was introduced; this concept can be understood as the reactor thickness measured from the irradiated surface where 99% of total energy is absorbed. Its thickness and the volumetric rate of photons absorption (VRPA were used as design parameters to determine (i reactor thickness, (ii maximum absorbed radiant energy, and (iii the optimal catalyst concentration. Six different commercial brands of titanium dioxide were studied: Evonik-Degussa P-25, Aldrich, Merck, Hombikat, Fluka, and Fisher. The local volumetric rate of photon absorption (LVRPA inside the reactor was described using six-flux absorption-scattering model (SFM applied to solar radiation. The radiation field and the boundary layer thickness of photon absorption were simulated with absorption and dispersion effects of catalysts in water at different catalyst loadings. The relationship between catalyst loading and reactor thickness that maximizes the absorption of radiant energy was obtained for each catalyst by apparent optical thickness. The optimum concentration of photocatalyst Degussa P-25 was 0.2 g/l in 0.86 cm of thickness, and for photocatalyst Aldrich it was 0.3 g/l in 0.80 cm of thickness.

  13. A Study on Conceptual Design of Fischer-Tropsch Reactors in GTL Applications

    Directory of Open Access Journals (Sweden)

    Shin Jae Sun

    2016-01-01

    Full Text Available GTL (Gas-to-liquid process is becoming an attractive technology which can produce liquid petroleum products using natural gas. As a part of preliminary design of GTL-FPSO application, process simulation analysis for conceptual design and optimization of reformers and F-T reactors are performed in GTL-FPSO applications by implementing the user made subroutine programs of kinetic equations into PRO/II PROVISION simulator. As for the F-T reactors, Plug Flow Reactor (PFR model is used with detailed kinetics equations over two different Fe based catalysts (Fe-Cu-K and K/Fe-Cu-Al. Dry reformer is also studied with Plug Flow Reactor (PFR model. In this study, simulation results are compared with available experimental data and found well agreed with experimental data for both reformer and FT reactor. The Peng-Robinson equation of state is also used to calculate the vapor phase non-idealities and vapor-liquid equilibrium. The optimum operating conditions and process simulation analysis are also presented.

  14. Design improvement and performance evaluation of solar photocatalytic reactor for industrial effluent treatment.

    Science.gov (United States)

    Nair, Ranjith G; Bharadwaj, P J; Samdarshi, S K

    2016-12-01

    This work reports the details of the design components and materials used in a linear compound parabolic trough reactor constructed with an aim to use the photocatalyst for solar photocatalytic applications. A compound parabolic trough reactor has been designed and engineered to exploit both UV and visible part of the solar irradiation. The developed compound parabolic trough reactor could receive almost 88% of UV radiation along with a major part of visible radiation. The performance of the reactor has been evaluated in terms of degradation of a probe pollutant using the parameters such as rate constant, residence time and photonic efficiency. An attempt has been made to assess the performance in different ranges of solar spectrum. Finally the developed reactor has been employed for the photocatalytic treatment of a paper mill effluent using Degussa P25 as the photocatalyst. The paper mill effluent collected from Nagaon paper mill, Assam, India has been treated under both batch mode and continuous mode using Degussa P25 photocatalyst under artificial and natural solar radiation, respectively. The photocatalytic degradation kinetics of the paper mill effluent has been determined using the reduction in total organic carbon (TOC) values of the effluent.

  15. Rational Basis for Designing Horizontal-Flow Anaerobic Immobilized Sludge (HAIS Reactor for Wastewater Treatment

    Directory of Open Access Journals (Sweden)

    M. Zaiat

    1997-03-01

    Full Text Available The conception and development on a rational basis of a new configuration of anaerobic fixed-bed bioreactor for wastewater treatment, the horizontal-flow anaerobic immobilized sludge (HAIS reactor, is presented. Such a reactor containing immobilized sludge in polyurethane foam matrices was first assayed for treating paper industry wastewater. A very short start-up period was observed and the reactor achieved stable operation by the eighth day. Afterwards, fundamental aspects of the process were investigated in order to obtain a rational basis for HAIS reactor design. A sequence of experiments was carried out for evaluating the cell wash-out from polyurethane foam matrices, the liquid-phase mass transfer coefficient and the intrinsic kinetic parameters, besides the hydrodynamic flow pattern of the reactor. The knowledge of such fundamental phenomena is useful for improving the reactor’s design and operation. Besides, these fundamental studies are essential to provide parameters for simulation and optimization of processes that make use of immobilized biomass

  16. Development of MMIS design technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, In Soo; Park, H. Y.; Lee, C. K. and others

    1999-03-01

    The objective of this study is to establish the concept design of the SMART MMIS through the setup of a structure and design concept of MMIS, the development of design methodology, and essential technologies for MMIS considering computer based digital technologies and human factors. Main activities of this study divided into three categories such as the development of conceptual design and requirement of the SMART I and C, the development of conceptual design and requirement of the SMART MMI, and essential technology for the MMIS. The results of main activities that have been performed during the study are as follows; 1) The review of licensing requirements and operating experience of existing plants. 2) The establishment of a system design concept about development strategy, basic functions and requirements, and applicable technology through the adoption of digital technology and human factorsengineering. 3) The accomplishment of function, structure, and design requirement of each MMIS through the performance of conceptual design on system structure and actuation logic using function analysis. It is necessary to develop and apply new technologies for the MMIS design compatible with the design basis of SMART. Considered essential technologies for the MMIS in this study are 1) the development of a signal validation algorithm and sensor diagnosis technique 2) the development of a soft controller for operator's action 3) the establishment of a data communication structure for MMIS architecture 4) the development of an information optimization basis and automatic start-up algorithm 5) the development of sensor reduction requirements 6) the development of a qualification method of digital equipment for applied digital technology and commercial grade items. (author)

  17. Design Study of Modular Nuclear Power Plant with Small Long Life Gas Cooled Fast Reactors Utilizing MOX Fuel

    Science.gov (United States)

    Ilham, Muhammad; Su’ud, Zaki

    2017-01-01

    Growing energy needed due to increasing of the world’s population encourages development of technology and science of nuclear power plant in its safety and security. In this research, it will be explained about design study of modular fast reactor with helium gas cooling (GCFR) small long life reactor, which can be operated over 20 years. It had been conducted about neutronic design GCFR with Mixed Oxide (UO2-PuO2) fuel in range of 100-200 MWth NPPs of power and 50-60% of fuel fraction variation with cylindrical pin cell and cylindrical balance of reactor core geometry. Calculation method used SRAC-CITATION code. The obtained results are the effective multiplication factor and density value of core reactor power (with geometry optimalization) to obtain optimum design core reactor power, whereas the obtained of optimum core reactor power is 200 MWth with 55% of fuel fraction and 9-13% of percentages.

  18. Plasma engineering design of a Compact Reversed-Field Pinch Reactor (CRFPR)

    Science.gov (United States)

    Bathke, C. G.; Embrechts, M. J.; Hagenson, R. L.; Krakowski, R. A.; Miller, R. L.

    1983-11-01

    The rationale for and the characteristics of the high-power-density Compact Reversed-Field Pinch Reactor (CRFPR) are discussed. Particular emphasis is given to key plasma engineering aspects of the conceptual design, including plasma operations, current drive, and impurity/ash control by means of pumped limiters or magnetic divertors. A brief description of the Fusion-Power-Core integration is given.

  19. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the si

  20. Safe design and operation of tank reactors for multiple-reaction networks: uniqueness and multiplicity

    NARCIS (Netherlands)

    Westerterp, K.R.; Westerink, E.J.

    1990-01-01

    A method is developed to design a tank reactor in which a network of reactions is carried out. The network is a combination of parallel and consecutive reactions. The method ensures unique operation. Dimensionless groups are used which are either representative of properties of the reaction system

  1. Overview of Progress on the EU DEMO Reactor Magnet System Design

    NARCIS (Netherlands)

    Zani, L.; Bayer, C.; biancolini, M.E.; Bonifetto, R.; Nijhuis, Arend; Yagotintsev, K.

    2016-01-01

    The DEMO reactor is expected to be the first application of fusion for electricity generation in the near future. To this aim, conceptual design activities are progressing in Europe (EU) under the lead of the EUROfusion Consortium in order to drive on the development of the major tokamak systems. In

  2. Design characteristics for pressurized water small modular nuclear power reactors with focus on safety

    Energy Technology Data Exchange (ETDEWEB)

    Kani, Iraj Mahmoudzadeh [Tehran Univ. (Iran, Islamic Republic of). Civil Faculty; Zandieh, Mehdi [Tehran Univ. (Iran, Islamic Republic of). Civil Faculty; International Univ. of Imam Khomeini (Iran, Islamic Republic of). Architecture Faculty; Abadi, Saeed Kheirollahi Hossein [International Univ. of Imam Khomeini (Iran, Islamic Republic of). Architecture Faculty

    2016-05-15

    Small Modular Reactors (SMRs) are a technology, attracting attention. Light water SMR possess an upgraded design case and emphasize the significance of integral models. Beside of these advantages, SMRs has faced numerous challenges, e.g. licensing, cost/investment, safety and security observation, social and environmental issues in building new plants.

  3. Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-01-01

    This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified.

  4. Design of Seismic Test Rig for Control Rod Drive Mechanism of Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Jongoh; Kim, Gyeongho; Yoo, Yeonsik; Cho, Yeonggarp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The reactor assembly is submerged in a reactor pool filled with water and its reactivity is controlled by locations of four control absorber rods(CARs) inside the reactor assembly. Each CAR is driven by a stepping motor installed at the top of the reactor pool and they are connected to each other by a tie rod and an electromagnet. The CARs scram the reactor by de-energizing the electromagnet in the event of a safe shutdown earthquake(SSE). Therefore, the safety function of the control rod drive mechanism(CRDM) which consists of a drive assembly, tie rod and CARs is to drop the CAR into the core within an appropriate time in case of the SSE. As well known, the operability for complex equipment such as the CRDM during an earthquake is very hard to be demonstrated by analysis and should be verified through tests. One of them simulates the reactor assembly and the guide tube of the CAR, and the other one does the pool wall where the drive assembly is installed. In this paper, design of the latter test rig and how the test is performed are presented. Initial design of the seismic test rig and excitation table had its first natural frequency at 16.3Hz and could not represent the environment where the CRDM was installed. Therefore, experimental modal analyses were performed and an FE model for the test rig and table was obtained and tuned based on the experimental results. Using the FE model, the design of the test rig and table was modified in order to have higher natural frequency than the cutoff frequency. The goal was achieved by changing its center of gravity and the stiffness of its sliding bearings.

  5. Current Status of Development of High Nickel Low Alloy Steels for Commercial Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Min Chul; Lee, B. S.; Park, S. G.; Lee, K. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-12-15

    SA508 Gr.3 Mn-Mo-Ni low alloy steels have been used for nuclear reactor pressure vessel steels up to now. Currently, the design goal of nuclear power plant is focusing at larger capacity and longer lifetime. Requirements of much bigger pressure vessels may cause critical problems in the manufacturing stage as well as for the welding stage. Application of higher strength steel may be required to overcome the technical problems. It is known that a higher strength and fracture toughness of low alloy steels such as SA508 Gr.4N low alloy steel could be achieved by increasing the Ni and Cr contents. Therefore, SA508 Gr.4N low alloy steel is very attractive as eligible RPV steel for the next generation PWR systems. In this report, we propose the possibility of SA508 Gr.4N low alloy steel for an application of next generation commercial RPV, based on the literature research result about development history of the RPV steels and SA508 specification. In addition, we have surveyed the research result of HSLA(High Strength Low Alloy steel), which has similar chemical compositions with SA508 Gr.4N, to understand the problems and the way of improvement of SA508 Gr.4N low alloy steel. And also, we have investigated eastern RPV steel(WWER-1000), which has higher Ni contents compared to western RPV steel.

  6. Mechanical design aspects of a tandem mirror fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Neef, W.S. Jr.

    1977-04-25

    Two ''plugs'' of dense plasma at either end of a central solenoid cell form the basis of a new mirror fusion power plant concept. A central cell blanket design is presented. Modules on crawler tracks serviced by remote welding and handling machines of very simple design are important features resulting from linear axisymmetric geometry. Three blanket designs are considered and the best one presented in some detail. It has lithium as the breeder material, helium cooled. ''Plug'' magnet field strengths must be high. A novel magnet is presented to satisfy the physics of the end plugs. Beam sources at 1,200 KV present special problems. Methods of voltage standoff, arc damage control, and neutralization are discussed. New secondary containment ideas are presented to allow removable roof sections of balanced design.

  7. A Metropolis algorithm combined with Nelder-Mead Simplex applied to nuclear reactor core design

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner F. [Depto. de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro, R. Alberto Rangel, s/n, P.O. Box 972285, Nova Friburgo, RJ 28601-970 (Brazil)], E-mail: wfsacco@iprj.uerj.br; Filho, Hermes Alves; Henderson, Nelio [Depto. de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro, R. Alberto Rangel, s/n, P.O. Box 972285, Nova Friburgo, RJ 28601-970 (Brazil); Oliveira, Cassiano R.E. de [Nuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States)

    2008-05-15

    A hybridization of the recently introduced Particle Collision Algorithm (PCA) and the Nelder-Mead Simplex algorithm is introduced and applied to a core design optimization problem which was previously attacked by other metaheuristics. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a three-enrichment-zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. The new metaheuristic performs better than the genetic algorithm, particle swarm optimization, and the Metropolis algorithms PCA and the Great Deluge Algorithm, thus demonstrating its potential for other applications.

  8. Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Rene Gerardo [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hutchinson, Jesson D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mcclure, Patrick Ray [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, William L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-08-20

    The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a deep space exploration mission. The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a “nuclear generated” heat source. This series of experiments is a larger scale follow up to the DUFF series of experiments1,2 that were performed using the Flat-Top assembly.

  9. Ferritic steels for sodium-cooled fast reactors: Design principles and challenges

    Science.gov (United States)

    Raj, Baldev; Vijayalakshmi, M.

    2010-09-01

    An overview of the current status of development of ferritic steels for emerging fast reactor technologies is presented in this paper. The creep-resistant 9-12Cr ferritic/martensitic steels are classically known for steam generator applications. The excellent void swelling resistance of ferritic steels enabled the identification of their potential for core component applications of fast reactors. Since then, an extensive knowledge base has been generated by identifying the empirical correlations between chemistry of the steels, heat treatment, structure, and properties, in addition to their in-reactor behavior. A few concerns have also been identified which pertain to high-temperature irradiation creep, embrittlement, Type IV cracking in creep-loaded weldments, and hard zone formation in dissimilar joints. The origin of these problems and the methodologies to overcome the limitations are highlighted. Finally, the suitability of the ferritic steels is re-evaluated in the emerging scenario of the fast reactor technology, with a target of achieving better breeding ratio and improved thermal efficiency.

  10. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

    Science.gov (United States)

    Marcille, T. F.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.; Poston, D. I.; Kapernick, R. J.

    2006-01-01

    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  11. Human factors regulation in response to advanced reactor design trends

    Energy Technology Data Exchange (ETDEWEB)

    Innes, L.; Harrison, F. (Atomic Energy Control Board, Ottawa, Ontario (Canada)); Rhodes, W. (Rhodes Associates Inc., Willowdale, Ontario (Canada))

    1993-01-01

    This paper presents the direction being taken in the Canadian regulatory program to guide the incorporation of human factors principles, knowledge, information, and methods in activities such as those that address the design and implementation of the advanced control room (ACR) in nuclear power plants. Within the current activity in the international nuclear industry aimed at the development of ACR concepts, the steps that are being taken to develop design review guidelines for ACRs, such as the work sponsored by the U.S. Nuclear Regulatory Commission are filling an identified important need. The salient feature of most ACR projects is the increased use of digital technology for the analysis, synthesis, management, and display of information. The greatly increased functionality of the devices that can be introduced during such projects allows changes to be made at the design and implementation stages that are not necessarily based on prior function allocation or task analyses.

  12. Design status of heavy ion injector program

    Energy Technology Data Exchange (ETDEWEB)

    Ballard, E.O.; Meyer, E.A.; Riepe, K.B.; Rutkowski, H.L.; Shurter, R.P.; Van Haaften, F.W.

    1985-10-01

    Design and development of a sixteen beam, heavy ion injector is in progress at Los Alamos National Laboratory (LANL) to demonstrate the injector technology for the High Temperature Experiment (HTE) proposed by Lawrence Berkeley Laboratory. The injector design provides for individual ion sources mounted to a support plate defining the sixteen beam array. The beamlets are electrostatically accelerated through a series of electrodes inside an evacuated (10 X torr) high voltage (HV) accelerating column. The column consists of two 28-inch diameter insulator modules made of 85 percent Al2O3 ceramic rings brazed to niobium feedthrough rings to which the electrodes are mechanically attached. Field shaping is used to minimize electron avalanche induced flashover along the inside surface of the ceramic rings. The column is self-supporting and is cantilevered from one end of the containment vessel. A brazed assembly was chosen to provide the required bond strength and high vacuum capability. The HV pulsed power supply is a 2MV Marx generator cantilevered from the opposite end of the containment vessel. The stainless steel pressure vessel (PV) contains a 65 psig mixture of SF6(30%) and nitrogen (70%) to provide the electrical insulation.

  13. Conceptual Design of I and C Architecture for a New Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Yong Suk; Park, Jae Kwan; Kim, Taek Kyu; Bae, Sang Hoon; Baang, Dane; Kim, Young Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    NRR) by 2016 was launched by KAERI in 2012. The purposes of the project are to meet domestic industrial needs of a research reactor and to secure an internationally competitive NRR. For instrumentation and control (I and C) systems of the NRR, it shall also be designed to secure the competition of the NRR. The I and C should account for the competition in terms of three aspects: safety, performance and cost. A I and C conceptual design activity should be carried out at the early stage of the project to figure out the I and C on the high level. Constructing I and C architecture is to accomplish the high level design. During the I and C architecture construction, the three aspects should be taken into account. This paper conceptually constructs the I and C architecture for the NRR by comparing to the cases of a Jordan training and research reactor (JRTR) project and a RA-10 multipurpose research reactor. The JRTR is an on-going project launched by KAERI and DAEWOO consortium in 2011. The I and C of the JRTR was digitalized based on the I and C functions of the HANARO research reactor, as shown in Fig. 1. The RA-10 was launched by INVAP of Argentina in 2011. The digital I and C developed by INVAP was built in an OPAL of Australia, as shown in Fig. 2

  14. Conceptual System Design of a Supercritical CO2 cooled Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong Gu; Oh, Bongseong; Baik, Seung Joon; Yu, Hwanyeal; Kim, Yonghee; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The S-CO2 Brayton cycle has many advantages for SMR's power conversion system. The S-CO2 cycle can achieve small component size and simple cycle layout as shown in Fig. 1. Therefore, a concept of one module containing the S-CO2 cooled fast reactor core and power conversion system is realizable. Thanks to the compact heat exchanger technology such as Printed Circuit Heat Exchanger (PCHE), the supercritical fluid with mediocre heat transfer performance can be utilized to a thermal cycle. This concept of fully modularized reactor is named as KAIST Micro Modular Reactor (MMR). It can achieve large economic by production in series, and transported in the land way or sea way. Based on the design results and dimensions of the reactor core and cycle components, the authors propose a conceptual layout of KAIST MMR. Based on this concept of reactor core, power conversion system, and decay heat removal system, the seasonal operation and transient analysis will be performed in the further works.

  15. Preliminary safety calculations to improve the design of Molten Salt Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brovchenko, M.; Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Capellan, N.; Ghetta, V.; Laureau, A. [LPSC, CNRS/IN2P3, Grenoble INP, 53,rue des Martyrs, 38026 Grenoble Cedex (France)

    2012-07-01

    Molten salt reactors are liquid fuel reactors so that they are flexible in operation but very different in the safety approach from solid fuel reactors. This study bears on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the R and D studies. This paper presents the first step of the safety approach: the systematic description of the MSFR, limited here to the main systems surrounding the core. This systematic description is the basis on which we will be able to devise accidental scenarios. Thanks to the negative reactivity feedback coefficient, most accidental scenarios lead to reactor shut down. Because of the decay heat generated in the fuel salt, it must be cooled. After the description of the tools developed to calculate the residual heat, the different contributions are discussed in this study. The decay heat of fission products in the MSFR is evaluated to be low (3% of nominal power), mainly due to the reprocessing that transfers the fission products to the gas reprocessing unit. As a result, the contribution of the actinides is significant (0.5% of nominal power). The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable (> 1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system. (authors)

  16. Design of pyrolysis reactor for production of bio-oil and bio-char simultaneously

    Science.gov (United States)

    Aladin, Andi; Alwi, Ratna Surya; Syarif, Takdir

    2017-05-01

    The residues from the wood industry are the main contributors to biomass waste in Indonesia. The conventional pyrolysis process, which needs a large energy as well as to produce various toxic chemical to the environment. Therefore, a pyrolysis unit on the laboratory scale was designed that can be a good alternative to achieve zero-waste and low energy cost. In this paper attempts to discuss design and system of pyrolysis reactor to produce bio-oil and bio-char simultaneously.

  17. Fusion-power-core design of a Compact Reversed-Field Pinch Reactor (CRFPR)

    Science.gov (United States)

    Copenhaver, C.; Schnurr, N. M.; Krakowski, R. A.; Hagenson, R. L.; Mynard, R. C.; Cappiello, C.; Lujan, R. E.; Davidson, J. W.; Chaffee, A. D.; Battat, M. E.

    A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuraton, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

  18. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  19. Conceptual design of the Clinch River Breeder Reactor spent-fuel shipping cask

    Energy Technology Data Exchange (ETDEWEB)

    Pope, R B; Diggs, J M [eds.

    1982-04-01

    Details of a baseline conceptual design of a spent fuel shipping cask for the Clinch River Breeder Reactor (CRBR) are presented including an assessment of shielding, structural, thermal, fabrication and cask/plant interfacing problems. A basis for continued cask development and for new technological development is established. Alternates to the baseline design are briefly presented. Estimates of development schedules, cask utilization and cost schedules, and of personnel dose commitments during CRBR in-plant handling of the cask are also presented.

  20. Tokamak experimental power reactor conceptual design. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1976-08-01

    Volume II contains the following appendices: (1) summary of EPR design parameters, (2) impurity control, (3) plasma computational models, (4) structural support system, (5) materials considerations for the primary energy conversion system, (6) magnetics, (7) neutronics penetration analysis, (8) first wall stress analysis, (9) enrichment of isotopes of hydrogen by cryogenic distillation, and (10) noncircular plasma considerations. (MOW)

  1. Design of a PID Controller for a PCR Micro Reactor

    Science.gov (United States)

    Dinca, M. P.; Gheorghe, M.; Galvin, P.

    2009-01-01

    Proportional-integral-derivative (PID) controllers are widely used in process control, and consequently they are described in most of the textbooks on automatic control. However, rather than presenting the overall design process, the examples given in such textbooks are intended to illuminate specific focused aspects of selection, tuning and…

  2. Options Study Documenting the Fast Reactor Fuels Innovative Design Activity

    Energy Technology Data Exchange (ETDEWEB)

    Jon Carmack; Kemal Pasamehmetoglu

    2010-07-01

    This document provides presentation and general analysis of innovative design concepts submitted to the FCRD Advanced Fuels Campaign by nine national laboratory teams as part of the Innovative Transmutation Fuels Concepts Call for Proposals issued on October 15, 2009 (Appendix A). Twenty one whitepapers were received and evaluated by an independent technical review committee.

  3. Design of Energy Storage Reactors for Dc-To-Dc Converters. Ph.D. Thesis

    Science.gov (United States)

    Chen, D. Y.

    1975-01-01

    Two methodical approaches to the design of energy-storage reactors for a group of widely used dc-to-dc converters are presented. One of these approaches is based on a steady-state time-domain analysis of piecewise-linearized circuit models of the converters, while the other approach is based on an analysis of the same circuit models, but from an energy point of view. The design procedure developed from the first approach includes a search through a stored data file of magnetic core characteristics and results in a list of usable reactor designs which meet a particular converter's requirements. Because of the complexity of this procedure, a digital computer usually is used to implement the design algorithm. The second approach, based on a study of the storage and transfer of energy in the magnetic reactors, leads to a straightforward design procedure which can be implemented with hand calculations. An equation to determine the lower-bound volume of workable cores for given converter design specifications is derived. Using this computer lower-bound volume, a comparative evaluation of various converter configurations is presented.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-01

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

  5. Disposal of irradiated fuel elements from German research reactors. Status and outlook

    Energy Technology Data Exchange (ETDEWEB)

    Thamm, G. [Central Research Reactor and Nuclear Operations Division, Research Centre Juelich, Forschungszentrum Juelich GmbH, Juelich (Germany)

    1999-07-01

    There will be a quantity of highly radioactive spent nuclear fuel (snf) from German research reactors amounting to about 9.1 t by the end of the next decade, which has to be disposed of. About 4.1 t of this quantity are intended to be returned to the USA. The remaining approximately 5 t can be loaded into approximately 30 CASTOR-2 casks and will be stored in a central German dry interim store for about 30 to 50 years (first step of the domestic disposal concept). Of course, snf arising from the operation of research reactors beyond 2010 has to be disposed of in the same way (3 MTR-2 casks every two years for BER-II and FRM-II). It is expected that snf from the zero-power facilities probably will be recycled for reusing the uranium. Due to the amendment of the German Atomic Energy Act intended by the new Federal German Government, the interim dry storage of snf from power reactors in central storage facilities like Ahaus or Gorleben will be stopped and the power reactors have to store snf at their own sites. Although the amendment only concerns nuclear power reactors, it could not be excluded that snf from research reactors, too, cannot be stored at Ahaus or Gorleben at present. (author)

  6. A dynamic model of a passively cooled small modular reactor for controller design purposes

    Energy Technology Data Exchange (ETDEWEB)

    Arda, Samet E., E-mail: s.e.arda@asu.edu; Holbert, Keith E., E-mail: holbert@asu.edu

    2015-08-15

    Highlights: • A mathematical dynamic model is developed for a passively cooled small modular reactor. • Reactor response associated single-phase natural circulation is analyzed. • A moving boundary model for a helical-coil steam generator is analyzed. • Dynamic responses of the overall model to representative perturbations are evaluated. • This compact model can be utilized for control system design. - Abstract: An analytical dynamic model for a passively cooled small modular reactor (SMR) is developed using a state-variable lumped parameter approach. Reactor power is represented by the generation time formulation of the point kinetics equations with a single combined neutron precursor group. The heat transfer process in the core is described via an overall heat transfer coefficient by defining two coolant lumps paired to a single fuel lump. In addition, a thermal–hydraulics model for single-phase natural circulation is incorporated. For the helical-coil steam generator, a moving-boundary model including subcooled, two-phase, and superheated regions is utilized. Finally, the hot leg riser and downcomer regions are expressed by first-order lags. The performance of the overall system described by ordinary differential equations (ODEs) is evaluated by the Simulink dynamic environment and directly using a MATLAB ODE solver recommended for stiff systems. Simulation results based on NuScale SMR design data show that the initial steady-state values for 100% power are within range of the design data and the model can predict the system dynamics due to typical perturbations, e.g., control rod movement and change in feedwater mass flow rate and temperature. The model developed in this work can be utilized as a foundation for designing and testing a suitable control algorithm for reactor thermal power.

  7. Reactor operation

    CERN Document Server

    Shaw, J

    2013-01-01

    Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

  8. Reactor Physics

    Energy Technology Data Exchange (ETDEWEB)

    Ait Abderrahim, A

    2001-04-01

    The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

  9. The Pressure Relief System Design for Industrial Reactors

    Directory of Open Access Journals (Sweden)

    Iztok Hace

    2013-01-01

    Full Text Available A quick and simple approach for reactor—emergency relief system design—for runaway chemical reactions is presented. A cookbook for system sizing with all main characteristic dimensions and parameters is shown on one realistic example from process industry. System design was done based on existing theories, standards, and correlations obtained from the literature, which were implemented for presented case. A simple and effective method for emergency relief system is shown, which may serve as an example for similar systems design. Obtained results may contribute to better understanding of blow down system frequently used in industrial plants, for increasing safety, decreasing explosion damage, and alleviating the ecological problems together with environmental pollution in case of industrial accidents.

  10. Optimum sensor placement in microchannel reactors: design tool applications

    Science.gov (United States)

    Sen, Mehmet; Fiering, Jason; Kowalski, Gregory J.; Larson, Dale

    2011-02-01

    A computational analysis of a microchannel reacting flow that includes diffusion and heat transfer processes to determine design rules for sensor placement is described. The objective is to optimize the positioning of nanohole array sensors which measure concentration and temperature and to analyze the characteristics of the local quantities sensed by nanohole arrays. Because the position and minimum spacing of the sensors are limited by material and fabrication constraints, the computational analysis is used to verify the effectiveness and limitations of this approach. Thermal boundary analysis is performed to analyze the relation between the sensed layer (micro-sensing region) over the nanohole array sensors and the boundary layer development. The relationship between the sensor position and the nodes of the numerical solution that limit this design process are discussed.

  11. Design of a fusion reactor for eutectic alloys Pb- Li; Diseno de un reactor de fusion para aleaciones autecticas Pb-Li

    Energy Technology Data Exchange (ETDEWEB)

    Quinones, J.; Barrena Perez, M. I.; Gomez de Salazar, J. M.; Serrano, L.; Duran, S.; Conde, E.; Barrado, A. I.; Fernandez, M.; Sedano, L.; Soria Munoz, A.

    2010-07-01

    Given the interest that have the Pb-Li eutectic alloys in the field of production of tritium, designed to optimize energy through nuclear fusion processes, in this paper, we present the design, construction and commissioning of a fusion reactor of Pb-Li alloys eutectic and the optimal process conditions, for these alloys.

  12. The effects of spectral shift absorbers on the design and safety of fast spectrum space reactors

    Science.gov (United States)

    King, Jeffrey Charles

    Spectral Shift Absorbers (SSAs) are incorporated into space reactors to maintain them sufficiently subcritical when submerged in seawater or wet sand and subsequently flooded, following a launch abort accident. The effect of four SSAs (samarium-149, europium-151, gadolinium-155, and gadolinium-157) on the submersion criticality, operation, and temperature reactivity feedback of the thermal spectrum reactors developed in the Systems for Nuclear Auxilary Power (SNAP) program is extensively documented. Recent work on SSAs in fast spectrum space reactors, preferred for compactness and higher powers, has focused on rhenium as the primary SSA. In addition to identifying additional SSAs, the present work investigates the effects of SSAs on the overall size and mass, temperature reactivity feedback, and operational lifetime of fast spectrum space reactors. The fast spectrum S4 reactor has a sectored Mo-14%Re solid-core, loadedwith UN fuel, cooled by He-30%Xe, and designed to avoid single point failures at a steady thermal power of 550 kWth. The addition of SSAs to the reactor core increases the fuel enrichment and decreases the size and mass of the reactor and the radiation shadow shield. SSA additions of boron-10, europium-151, gadolinium-155 and iridium result in the smallest and lightest S4 reactors. The effects of SSA additions on the operational lifetime and the temperature and burnup reactivity coefficients of the S^4 reactor are studied. An increasein fuel enrichment with SSAs markedly increases the operational lifetime by decreasing the burnup reactivity coefficient with only a slight decrease in the temperature reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt%), the temperature and burnup reactivity coefficients are the highest (-0.2709 ¢/K and -1.3470 /atom%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to -0.2649 ¢/K and -1.0230 /atom%, and

  13. Application of Decomposition Methodology to Solve Integrated Process Design and Controller Design Problems for Reactor-Separator-Recycle System

    DEFF Research Database (Denmark)

    Abd.Hamid, Mohd-Kamaruddin; Sin, Gürkan; Gani, Rafiqul

    2010-01-01

    This paper presents the integrated process design and controller design (IPDC) for a reactor-separator-recycle (RSR) system and evaluates a decomposition methodology to solve the IPDC problem. Accordingly, the IPDC problem is solved by decomposing it into four hierarchical stages: (i) pre...... the design of a RSR system involving consecutive reactions, A B -> C and shown to provide effective solutions that satisfy design, control and cost criteria. The advantage of the proposed methodology is that it is systematic, makes use of thermodynamic-process knowledge and provides valuable insights...

  14. Large scale breeder reactor plant prototype mechanical pump conceptual design study

    Energy Technology Data Exchange (ETDEWEB)

    1976-07-01

    This final report is a complete conceptual design study of a mechanical pump for a large scale breeder reactor plant. The pumps are located in the cold leg side of the loops. This makes the net positive suction head available - NPSHA - low, and is, in fact, a major influencing factor in the design. Where possible, experience gained from the Clinch River Project and the FFTF is used in this study. Experience gained in the design, manufacturer, and testing of pumps in general and sodium pumps in particular is reflected in this report. The report includes estimated cost and time schedule for design, manufacture, and testing. It also includes a recommendation for development needs.

  15. New reactor concepts. An analysis of the actual research status; Neue Reaktorkonzepte. Eine Analyse des aktuellen Forschungsstands

    Energy Technology Data Exchange (ETDEWEB)

    Pistner, Christoph; Englert, Matthias

    2017-04-15

    The report on new reactor concepts covers the following issues: characterization and survey of new reactor concepts; evaluation criteria: safety, resources for fuel supply, waste problems, economy and proliferation; comprehensive relevant aspects: thorium as alternative resource, partitioning and transmutation; actual developments and preliminary experiences for fast breeding reactor (FBR), high-temperature reactor (HTR), molten salt reactor (MSR), small modular reactor (SMR).

  16. Dismantling design for a reference research reactor of the WWR type

    Energy Technology Data Exchange (ETDEWEB)

    Lobach, Yu.N., E-mail: lobach@kinr.kiev.ua [Institute for Nuclear Research, Pr. Nauki, 47, Kiev 03680 (Ukraine); Cross, M.T., E-mail: Martin.Cross@nuvia.co.uk [Nuvia Ltd., Robinson House, Crow Park Way, Westlakes Science and Technology Park, Moor Row, Cumbria CA24 3HY (United Kingdom)

    2014-01-15

    Highlights: • Design features of WWRs relevant to decommissioning have been analysed. • The technical basis for the preparation and implementation of dismantling has been established for a reference WWR. • The applicability of existing proven dismantling technologies has been established. -- Abstract: A decommissioning study has been carried out for a reference research reactor of the WWR type. Many such reactors were constructed more than 50 years ago and most of them are still in operation. Decommissioning has now become an important consideration. This paper summarizes the main decommissioning steps and, on the basis of the reactor design features, technical aspects of the dismantling and removal of the contaminated/activated components have been analysed. The advisability of the removal of large components, such as the reactor vessel and the heat-exchangers, as one piece items has also been demonstrated. Additionally, a work schedule and an estimation of the collective dose for the preparation and implementation of dismantling have been established. The applicability of existing proven dismantling technologies has been identified together with some additional features for the dismantling.

  17. Design and operation of a rotating drum radio frequency plasma reactor for the modification of free nanoparticles.

    Science.gov (United States)

    Shearer, Jeffrey C; Fisher, Ellen R

    2013-06-01

    A rotating drum rf plasma reactor was designed to functionalize the surface of nanoparticles and other unusually shaped substrates through plasma polymerization and surface modification. This proof-of-concept reactor design utilizes plasma polymerized allyl alcohol to add OH functionality to Fe2O3 nanoparticles. The reactor design is adaptable to current plasma hardware, eliminating the need for an independent reactor setup. Plasma polymerization performed on Si wafers, Fe2O3 nanoparticles supported on Si wafers, and freely rotating Fe2O3 nanoparticles demonstrated the utility of the reactor for a multitude of processes. X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy were used to characterize the surface of the substrates prior to and after plasma deposition, and scanning electron microscopy was used to verify that no extensive change in the size or shape of the nanoparticles occurred because of the rotating motion of the reactor. The reactor design was also extended to a non-depositing NH3 plasma modification system to demonstrate the reactor design is effective for multiple plasma processes.

  18. Evaluation for the status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Sook; Lee, Byung Doo

    2007-11-15

    Safeguards implementation of nuclear material was carried out at facility level in an effect to support the peaceful nuclear activities in KAERI. Safeguards implementation is to fulfill the obligations associated with international agreements such as IAEA comprehensive safeguards agreement and additional protocol. IAEA inspection is the most important and basic factor of the safeguards implementation for the purpose of verifying whether all source or special fissionable material is diverted to nuclear weapons or other nuclear explosive devices. The status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor during 2001-2006 is evaluated in this report.

  19. CO2 conversion by plasma technology: insights from modeling the plasma chemistry and plasma reactor design

    Science.gov (United States)

    Bogaerts, A.; Berthelot, A.; Heijkers, S.; Kolev, St.; Snoeckx, R.; Sun, S.; Trenchev, G.; Van Laer, K.; Wang, W.

    2017-06-01

    In recent years there has been growing interest in the use of plasma technology for CO2 conversion. To improve this application, a good insight into the underlying mechanisms is of great importance. This can be obtained from modeling the detailed plasma chemistry in order to understand the chemical reaction pathways leading to CO2 conversion (either in pure form or mixed with another gas). Moreover, in practice, several plasma reactor types are being investigated for CO2 conversion, so in addition it is essential to be able to model these reactor geometries so that their design can be improved, and the most energy efficient CO2 conversion can be achieved. Modeling the detailed plasma chemistry of CO2 conversion in complex reactors is, however, very time-consuming. This problem can be overcome by using a combination of two different types of model: 0D chemical reaction kinetics models are very suitable for describing the detailed plasma chemistry, while the characteristic features of different reactor geometries can be studied by 2D or 3D fluid models. In the first instance the latter can be developed in argon or helium with a simple chemistry to limit the calculation time; however, the ultimate aim is to implement the more complex CO2 chemistry in these models. In the present paper, examples will be given of both the 0D plasma chemistry models and the 2D and 3D fluid models for the most common plasma reactors used for CO2 conversion in order to emphasize the complementarity of both approaches. Furthermore, based on the modeling insights, the paper discusses the possibilities and limitations of plasma-based CO2 conversion in different types of plasma reactors, as well as what is needed to make further progress in this field.

  20. The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

    Science.gov (United States)

    Scarlat, Raluca O.; Peterson, Per F.

    2014-01-01

    The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

  1. Status of Fuel Development and Manufacturing for Space Nuclear Reactors at BWX Technologies

    Science.gov (United States)

    Carmack, W. J.; Husser, D. L.; Mohr, T. C.; Richardson, W. C.

    2004-02-01

    New advanced nuclear space propulsion systems will soon seek a high temperature, stable fuel form. BWX Technologies Inc (BWXT) has a long history of fuel manufacturing. UO2, UCO, and UCx have been fabricated at BWXT for various US and international programs. Recent efforts at BWXT have focused on establishing the manufacturing techniques and analysis capabilities needed to provide a high quality, high power, compact nuclear reactor for use in space nuclear powered missions. To support the production of a space nuclear reactor, uranium nitride has recently been manufactured by BWXT. In addition, analytical chemistry and analysis techniques have been developed to provide verification and qualification of the uranium nitride production process. The fabrication of a space nuclear reactor will require the ability to place an unclad fuel form into a clad structure for assembly into a reactor core configuration. To this end, BWX Technologies has reestablished its capability for machining, GTA welding, and EB welding of refractory metals. Specifically, BWX Technologies has demonstrated GTA welding of niobium flat plate and EB welding of niobium and Nb-1Zr tubing. In performing these demonstration activities, BWX Technologies has established the necessary infrastructure to manufacture UO2, UCx, or UNx fuel, components, and complete reactor assemblies in support of space nuclear programs.

  2. Status and future of the WWR-M research reactor in Kiev

    Energy Technology Data Exchange (ETDEWEB)

    Bazavov, D.A.; Gavrilyuk, V.I.; Kirischuk, V.I.; Kochetkov, V.V.; Lysenko, M.V.; Makarovskiy, V.N.; Scherbachenko, A.M.; Shevel, V.N.; Slisenko, V.I. [Institute for Nuclear Research, Kiev (Ukraine)

    2001-07-01

    Kiev WWR-M Research Reactor, operated at maximum power of 10 MW, was put into operation in 1960 and during its 40-years history has been used to perform numerous studies in different areas of science and technology. Due to a number of technical problems the Research Reactor, the only one in Ukraine, was shut down in 1993 and then put into operation in 1999 again. Now there is an intention to reconstruct Kiev Research Reactor. The upgraded Research Reactor would allow solving such problems as the safe operation of Ukrainian NPPs, radioisotope production and, naturally, fundamental and applied research. The main problem for the successful operation of Kiev Research Reactor is the management and storage of spent fuel at the site, since after core unloading the spent fuel storage appears to be practically completed. So it is absolutely necessary to ship the most part of the spent fuel for reprocessing and as soon as possible. Besides, there is a need to build up the new spent fuel storage, because the tank of available storage requires careful inspection for corrosion. (author)

  3. Experimental Design for Evaluating Selected Nondestructive Measurement Technologies - Advanced Reactor Technology Milestone: M3AT-16PN2301043

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Hirt, Evelyn H. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Pitman, Stan G. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Dib, Gerges [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Roy, Surajit [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Good, Morris S. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Walker, Cody M. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-07-16

    The harsh environments in advanced reactors (AdvRx) increase the possibility of degradation of safety-critical passive components, and therefore pose a particular challenge for deployment and extended operation of these concepts. Nondestructive evaluation technologies are an essential element for obtaining information on passive component condition in AdvRx, with the development of sensor technologies for nondestructively inspecting AdvRx passive components identified as a key need. Given the challenges posed by AdvRx environments and the potential needs for reducing the burden posed by periodic in-service inspection of hard-to-access and hard-to-replace components, a viable solution may be provided by online condition monitoring of components. This report identifies the key challenges that will need to be overcome for sensor development in this context, and documents an experimental plan for sensor development, test, and evaluation. The focus of initial research and development is on sodium fast reactors, with the eventual goal of the research being developing the necessary sensor technology, quantifying sensor survivability and long-term measurement reliability for nondestructively inspecting critical components. Materials for sensor development that are likely to withstand the harsh environments are described, along with a status on the fabrication of reference specimens, and the planned approach for design and evaluation of the sensor and measurement technology.

  4. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a {sup 23}Na(n,g){sup 24}Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B{sub 4}C shielding inside the subassembly.

  5. The SPES3 Experimental Facility Design for the IRIS Reactor Simulation

    Directory of Open Access Journals (Sweden)

    Mario Carelli

    2009-01-01

    Full Text Available IRIS is an advanced integral pressurized water reactor, developed by an international consortium led by Westinghouse. The licensing process requires the execution of integral and separate effect tests on a properly scaled reactor simulator for reactor concept, safety system verification, and code assessment. Within the framework of an Italian R&D program on Nuclear Fission, managed by ENEA and supported by the Ministry of Economic Development, the SPES3 facility is under design and will be built and operated at SIET laboratories. SPES3 simulates the primary, secondary, and containment systems of IRIS with 1 : 100 volume scale, full elevation, and prototypical thermal-hydraulic conditions. The simulation of the facility with the RELAP5 code and the execution of the tests will provide a reliable tool for data extrapolation and safety analyses of the final IRIS design. This paper summarises the main design steps of the SPES3 integral test facility, underlying choices and phases that lead to the final design.

  6. Advanced tokamak reactors based on the spherical torus (ATR/ST). Preliminary design considerations

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.L.; Krakowski, R.A.; Bathke, C.G.; Copenhaver, C.; Schnurr, N.M.; Engelhardt, A.G.; Seed, T.J.; Zubrin, R.M.

    1986-06-01

    Preliminary design results relating to an advanced magnetic fusion reactor concept based on the high-beta, low-aspect-ratio, spherical-torus tokamak are summarized. The concept includes resistive (demountable) toroidal-field coils, magnetic-divertor impurity control, oscillating-field current drive, and a flowing liquid-metal breeding blanket. Results of parametric tradeoff studies, plasma engineering modeling, fusion-power-core mechanical design, neutronics analyses, and blanket thermalhydraulics studies are described. The approach, models, and interim results described here provide a basis for a more detailed design. Key issues quantified for the spherical-torus reactor center on the need for an efficient drive for this high-current (approx.40 MA) device as well as the economic desirability to increase the net electrical power from the nominal 500-MWe(net) value adopted for the baseline system. Although a direct extension of present tokamak scaling, the stablity and transport of this high-beta (approx.0.3) plasma is a key unknown that is resoluble only by experiment. The spherical torus generally provides a route to improved tokamak reactors as measured by considerably simplified coil technology in a configuration that allows a realistic magnetic divertor design, both leading to increased mass power density and reduced cost.

  7. Seismic Analysis Issues in Design Certification Applications for New Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Miranda, M.; Morante, R.; Xu, J.

    2011-07-17

    The licensing framework established by the U.S. Nuclear Regulatory Commission under Title 10 of the Code of Federal Regulations (10 CFR) Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” provides requirements for standard design certifications (DCs) and combined license (COL) applications. The intent of this process is the early reso- lution of safety issues at the DC application stage. Subsequent COL applications may incorporate a DC by reference. Thus, the COL review will not reconsider safety issues resolved during the DC process. However, a COL application that incorporates a DC by reference must demonstrate that relevant site-specific de- sign parameters are within the bounds postulated by the DC, and any departures from the DC need to be justified. This paper provides an overview of several seismic analysis issues encountered during a review of recent DC applications under the 10 CFR Part 52 process, in which the authors have participated as part of the safety review effort.

  8. Pollution control. sulphur recovery from industrial effluents. design of reactor

    Energy Technology Data Exchange (ETDEWEB)

    Venkatram, T.; Pachaiyappan, V.

    1978-05-01

    In a process for the recovery of sulfur from a refinery gas containing 10.8-17.8% hydrogen sulfide, 13.4-27.6% hydrogen, 19.7-23.6% methane, and 34.2-40.7% C/sub 2/-C/sub 5/ hydrocarbons, the hydrogen sulfide is selectively absorbed in monoethanol amine (MEA) solution. The hydrogen sulfide gas is stripped off from the MEA solution and sent to a furnace chamber with air at an air/hydrogen sulfide ratio of 2.5 by vol to convert 33% of the hydrogen sulfide to sulfur dioxide and the hydrocarbons to carbon dioxide and water. The product gases are cooled from 1200/sup 0/C to a final temperature of 232/sup 0/C, and any sulfur formed in the furnace is condensed and drained to a sulfur pit. The outlet gases from the final pass are routed to the Claus converter in which hydrogen sulfide and sulfur dioxide react over a bauxite catalyst to form elemental sulfur. Noncondensable gases are sent to a second converter for completion of the reaction. Any remaining noncondensable gases are sent to a coalescer to recover entrained sulfur and then to an incinerator. Design of the Claus converter and of the shell-and-tube heat exchanger used for cooling the gases from the furnace and for steam generation is illustrated for various operating conditions. Diagrams.

  9. Steady-state and loss-of-pumping accident analyses of the Savannah River new production reactor representative design

    Energy Technology Data Exchange (ETDEWEB)

    Pryor, R.J.; Maloney, K.J.

    1990-10-01

    This document contains the steady-state and loss-of-pumping accident analysis of the representative design for the Savannah River heavy water new production reactor. A description of the reactor system and computer input model, the results of the steady-state analysis, and the results of four loss-of-pumping accident calculations are presented. 5 refs., 37 figs., 4 tabs.

  10. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  11. Design Features of a Core Protection System for an Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Bon Seung; In, Wang Kee; Kim, Keung Koo; Lee, Chung Chan; Zee, Sung Quun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    A system-integrated modular advanced research reactor is under development in the KAERI. Therefore, it is required to design an advanced core protection system for an integral reactor and an online digital core protection system, SCOPS is being developed as a part of plant protection system. SCOPS calculates the minimum CHFR(Critical Heat Flux Ratio) and maximum LPD(Local Power Density) based on the several online measured system parameters, such as the excore detector signal, CEA positions, MCP pump speed, pressure and temperature. Calculated values are compared with predetermined limiting values and the trip signal is generated if necessary. This paper describes the basic design features of SCOPS and several output parameters for a simple test case are presented.

  12. Design of the cold neutron triple-axis spectrometer at the China Advanced Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, P.; Zhang, Hongxia; Bao, W. [Department of Physics, Beijing Key Laboratory of Opto-electronic Functional Materials & Micro-nano Devices, Renmin University of China, Beijing 100872 (China); Schneidewind, A. [Jülich Center for Neutron Science (JCNS), Forschungszentrum Jülich GmbH, Outstation at Heinz MaierCLeibnitz Zentrum (MLZ), D-85747 Garching (Germany); Link, P. [Heinz Maier-Leibnitz Zentrum, Technische Universität München, D-85748 Garching (Germany); Grünwald, A.T.D. [II. Physikalisches Institut, Universität zu Köln, D-50937 Köln (Germany); Georgii, R. [Heinz Maier-Leibnitz Zentrum, Technische Universität München, D-85748 Garching (Germany); Hao, L.J.; Liu, Y.T. [China Institute of Atomic Energy, PO Box-275-30, Beijing 102413 (China)

    2016-06-11

    The design of the first cold neutron triple-axis spectrometer at the China Advanced Research Reactor is presented. Based on the Monte Carlo simulations using neutron ray-tracing program McStas, the parameters of major neutron optics in this instrument are optimized. The neutron flux at sample position is estimated to be 5.6 ×10{sup 7} n/cm{sup 2}/s at neutron incident energy E{sub i}=5 meV when the reactor operates normally at the designed 60 MW power. The performances of several neutron supermirror polarizing devices are compared and their critical parameters are optimized for this spectrometer. The polarization analysis will be realized with a flexible switch from the unpolarized experimental mode.

  13. Design of the cold neutron triple-axis spectrometer at the China Advanced Research Reactor

    Science.gov (United States)

    Cheng, P.; Zhang, Hongxia; Bao, W.; Schneidewind, A.; Link, P.; Grünwald, A. T. D.; Georgii, R.; Hao, L. J.; Liu, Y. T.

    2016-06-01

    The design of the first cold neutron triple-axis spectrometer at the China Advanced Research Reactor is presented. Based on the Monte Carlo simulations using neutron ray-tracing program McStas, the parameters of major neutron optics in this instrument are optimized. The neutron flux at sample position is estimated to be 5.6 ×107 n/cm2/s at neutron incident energy Ei=5 meV when the reactor operates normally at the designed 60 MW power. The performances of several neutron supermirror polarizing devices are compared and their critical parameters are optimized for this spectrometer. The polarization analysis will be realized with a flexible switch from the unpolarized experimental mode.

  14. The ATLAS Data Acquisition and Trigger concept, design and status

    CERN Document Server

    Kordas, K; Alexandrov, I; Amorim, A; Aracena, I; Armstrong, S; Badescu, E; Baines, J T M; Barros, N; Beck, H P; Bee, C; Bellomo, M; Biglietti, M; Blair, R; Bogaerts, J A C; Bold, T; Bosman, M; Burckhart-Chromek, D; Caprini, M; Caramarcu, C; Carlino, G; Caron, B; Casado, M P; Cataldi, G; Ciobotaru, M; Comune, G; Conde-Muíño, P; Conventi, F; Corso-Radu, A; Cranfield, R; Cranmer, K; Crone, G; Damazio, D; Dawson, J; De Santo, A; Del Prete, T; Della Pietra, M; Di Mattia, A; Diaz-Gomaz, M; Dobinson, Robert W; Dobson, M; Dos Anjos, A; Dotti, A; Drake, G; Ellis, Nick; Emeliyanov, D; Ermoline, Y; Ertorer, E; Falciano, S; Ferrari, R; Ferrer, M L; Francis, D; Gadomski, S; Gameiro, S; Garitaonandia, H; Gaudio, G; Gaumer, O; George, S; Gesualdi-Mello, A; Goncalo, R; Gorini, B; Gorini, E; Green, B; Haas, S; Haberichter, W N; Hadavand, H; Haeberli, C; Haller, J; Hansen, J; Hauser, R; Hillier, S J; Höcker, A; Hughes-Jones, R E; Joos, M; Kabana, S; Kazarov, A; Khomich, A; Kieft, G; Kilvington, G; Kirk, J; Klous, S; Kohno, T; Kolos, S; Konstantinidis, N P; Kootz, A; Korcyl, K; Kotov, V; Kugel, A; Landon, M; Lankford, A; Leahu, L; Leahu, M; Lehmann-Miotto, G; Le Vine, M J; Liu, W; Lowe, A; Luminari, L; Maeno, T; Männer, R; Mapelli, L; Martin, B; Marzano, F; Masik, J; McLaren, R; McMahon, T; Meessen, C; Meirosu, C; Mineev, M; Misiejuk, A; Moore, R; Morettini, P; Mornacchi, G; Müller, M; Murillo-García, R; Nagasaka, Y; Negri, A; Nisati, A; Osuna, C; Padilla, C; Panikashvili, N; Parodi, F; Pasqualucci, E; Pauly, T; Perera, V; Pérez-Réale, V; Petersen, J; Pinfold, J L; Pope, B; Portes de Albuquerqu, M; Potter, C; Pretzl, K; Prigent, D; Primavera, M; Rheaum, P; Robertson, S; Roda, C; Ryabov, Yu; Salvatore, D; Santamarina-Rios, C; Scannicchio, D A; Schiavi, C; Schlereth, J L; Scholtes, I; Seixas, M; Sidoti, A; Sivoklokov, S Yu; Sloper, J; Sole-Segura, E; Soloviev, I; Soluk, R A; Spagnolo, S; Spiwoks, R; Stamen, R; Stancu, S; Stefanidis, E; Strong, J; Sushkov, S; Sutton, M; Szymocha, T; Tapprogge, S; Tarem, S; Tarem, Z; Teixeira-Dias, P; Thomas, E; Torres, R; Touchard, F; Tremblet, L; Unel, N G; Usai, G; Vachon, B; Van Wasen, J; Vandelli, W; Vaz-Gil-Lopes, L; Ventura, A; Vercesi, V; Vermeulen, J; von der Schmitt, H; Warburton, A; Watson, A; Wengler, T; Werner, P; Wheeler, S; Wickens, F; Wiedenmann, W; Wielers, M; Wiesmann, M; Woerling, E E; Wu, X; Yasu, Y; Yu, M; Zema, F; Zobernig, H; 10th Topical Seminar on Innovative Particle and Radiation Detectors

    2007-01-01

    This article presents the base-line design and implementation of the ATLAS Trigger and Data Acquisition system, in particular the Data Flow and High Level Trigger components. The status of the installation and commissioning of the system is also presented.

  15. Conceptual design of superconducting magnet systems for the Argonne Tokamak Experimental Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wang, S.T.; Turner, L.R.; Mills, F.E.; DeMichele, D.W.; Smelser, P.; Kim, S.H.

    1976-01-01

    As an integral effort in the Argonne Tokamak Experimental Power Reactor Conceptual Design, the conceptual design of a 10-tesla, pure-tension superconducting toroidal-field (TF) coil system has been developed in sufficient detail to define a realistic design for the TF coil system that could be built based upon the current state of technology with minimum technological extrapolations. A conceptual design study on the superconducting ohmic-heating (OH) coils and the superconducting equilibrium-field (EF) coils were also completed. These conceptual designs are developed in sufficient detail with clear information on high current ac conductor design, cooling, venting provision, coil structural support and zero loss poloidal coil cryostat design. Also investigated is the EF penetration into the blanket and shield.

  16. A SMALL MODULAR REACTOR DESIGN FOR MULTIPLE ENERGY APPLICATIONS: HTR50S

    Directory of Open Access Journals (Sweden)

    X. YAN

    2013-06-01

    Full Text Available HTR50S is a small modular reactor system based on HTGR. It is designed for a triad of applications to be implemented in successive stages. In the first stage, a base plant for heat and power is constructed of the fuel proven in JAEA's 950°C, 30MWt test reactor HTTR and a conventional steam turbine to minimize development risk. While the outlet temperature is lowered to 750°C for the steam turbine, thermal power is raised to 50MWt by enabling 40% greater power density in 20% taller core than the HTTR. However the fuel temperature limit and reactor pressure vessel diameter are kept. In second stage, a new fuel that is currently under development at JAEA will allow the core outlet temperature to be raised to 900°C for the purpose of demonstrating more efficient gas turbine power generation and high temperature heat supply. The third stage adds a demonstration of nuclear-heated hydrogen production by a thermochemical process. A licensing approach to coupling high temperature industrial process to nuclear reactor will be developed. The low initial risk and the high longer-term potential for performance expansion attract development of the HTR50S as a multipurpose industrial or distributed energy source.

  17. Design and implementation of a novel portable atomic layer deposition/chemical vapor deposition hybrid reactor.

    Science.gov (United States)

    Selvaraj, Sathees Kannan; Jursich, Gregory; Takoudis, Christos G

    2013-09-01

    We report the development of a novel portable atomic layer deposition chemical vapor deposition (ALD/CVD) hybrid reactor setup. Unique feature of this reactor is the use of ALD/CVD mode in a single portable deposition system to fabricate multi-layer thin films over a broad range from "bulk-like" multi-micrometer to nanometer atomic dimensions. The precursor delivery system and control-architecture are designed so that continuous reactant flows for CVD and cyclic pulsating flows for ALD mode are facilitated. A custom-written LabVIEW program controls the valve sequencing to allow synthesis of different kinds of film structures under either ALD or CVD mode or both. The entire reactor setup weighs less than 40 lb and has a relatively small footprint of 8 × 9 in., making it compact and easy for transportation. The reactor is tested in the ALD mode with titanium oxide (TiO2) ALD using tetrakis(diethylamino)titanium and water vapor. The resulting growth rate of 0.04 nm/cycle and purity of the films are in good agreement with literature values. The ALD/CVD hybrid mode is demonstrated with ALD of TiO2 and CVD of tin oxide (SnOx). Transmission electron microscopy images of the resulting films confirm the formation of successive distinct TiO2-ALD and SnO(x)-CVD layers.

  18. Design Study of Small Pb-Bi Cooled Modified Candle Reactors

    Science.gov (United States)

    Su'ud, Zaki; Sekimoto, H.

    2010-06-01

    In this study application of modified CANDLE burnup scheme based long life Pb-Bi Cooled Fast Reactors for small long life reactors with natural Uranium as Fuel Cycle Input has been performed. The reactor cores are subdivided into several parts with the same volume in the axial directions. The natural uranium is initially put in region 1, after one cycle of 10 years of burn-up it is shifted to region 2, and 10 years after that it is shifted to region 3. This concept is applied to all regions, i.e. shifted the core of I'th region into I+1 region after the end of 10 years burn-up cycle. The first region 1 is filled by fresh natural uranium fuel. Compared to the previous works, in a smaller reactor core the criticality need to be considered more carefully especially at the beginning of life. As an optimized design, a core of 85 cm radius and 150 cm height with 300 MWt power are selected. This core can be operated 10 years without refueling or fuel shuffling. The average discharge burn-up is 350 GWd/ton HM.

  19. Design and implementation of a novel portable atomic layer deposition/chemical vapor deposition hybrid reactor

    Science.gov (United States)

    Selvaraj, Sathees Kannan; Jursich, Gregory; Takoudis, Christos G.

    2013-09-01

    We report the development of a novel portable atomic layer deposition chemical vapor deposition (ALD/CVD) hybrid reactor setup. Unique feature of this reactor is the use of ALD/CVD mode in a single portable deposition system to fabricate multi-layer thin films over a broad range from "bulk-like" multi-micrometer to nanometer atomic dimensions. The precursor delivery system and control-architecture are designed so that continuous reactant flows for CVD and cyclic pulsating flows for ALD mode are facilitated. A custom-written LabVIEW program controls the valve sequencing to allow synthesis of different kinds of film structures under either ALD or CVD mode or both. The entire reactor setup weighs less than 40 lb and has a relatively small footprint of 8 × 9 in., making it compact and easy for transportation. The reactor is tested in the ALD mode with titanium oxide (TiO2) ALD using tetrakis(diethylamino)titanium and water vapor. The resulting growth rate of 0.04 nm/cycle and purity of the films are in good agreement with literature values. The ALD/CVD hybrid mode is demonstrated with ALD of TiO2 and CVD of tin oxide (SnOx). Transmission electron microscopy images of the resulting films confirm the formation of successive distinct TiO2-ALD and SnOx-CVD layers.

  20. Optimal catalyst curves: Connecting density functional theory calculations with industrial reactor design and catalyst selection

    DEFF Research Database (Denmark)

    Jacobsen, C.J.H.; Dahl, Søren; Boisen, A.

    2002-01-01

    For ammonia synthesis catalysts a volcano-type relationship has been found experimentally. We demonstrate that by combining density functional theory calculations with a microkinetic model the position of the maximum of the volcano curve is sensitive to the reaction conditions. The catalytic ammo......-principle quantum mechanical calculations of gas-surface interactions, reactor design, and catalyst selection has been established for the first time....

  1. Preliminary design of a Binary Breeder Reactor; Diseno preliminar de un reactor esferico de quema/cria

    Energy Technology Data Exchange (ETDEWEB)

    Garcia C, E. Y.; Francois, J. L.; Lopez S, R. C., E-mail: eliasgarcerv@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, 62550 Jiutepec, Morelos (Mexico)

    2014-10-15

    A binary breeder reactor (BBR) is a reactor that by means of the transmutation and fission process can operates through the depleted uranium burning with a small quantity of fissile material. The advantages of a BBR with relation to other nuclear reactor types are numerous, taking into account their capacity to operate for a long time without requiring fuel reload or re-arrangement. In this work four different simulations are shown carried out with the MCNPX code with libraries Jeff-3.1 to 1200 K. The objective of this study is to compare two different models of BBR: a spherical reactor and a cylindrical one, using two fuel cycles for each one of them (U-Pu and Th-U) and different reflectors for the two different geometries. For all the models a super-criticality state was obtained at least 10.9 years without carrying out some fuel re-arrangement or reload. The plutonium-239 production was achieved in the models where natural uranium was used in the breeding area, while the production of uranium-233 was observed in the cases where thorium was used in the fertile area. Finally, a behavior of stationary wave reactor was observed inside the models of spherical reactor when contemplating the power uniform increment in the breeding area, while inside the cylindrical models was observed the behavior of a traveling wave reactor when registering the displacement of the burnt wave along the cylindrical model. (Author)

  2. Status of DOE efforts to renew acceptance of foreign research reactor spent nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Head, C.R.

    1997-08-01

    This presentation summarizes the efforts being made by the Department of Energy to renew acceptance of spent nuclear fuel shipments from foreign research reactors. The author reviews the actions undertaken in this process in a fairly chronological manner, through the present time, as well as the development of an environmental impact statement to support the proposed actions.

  3. 78 FR 46621 - Status of the Office of New Reactors' Implementation of Electronic Distribution of Advanced...

    Science.gov (United States)

    2013-08-01

    ... Pike, Rockville, Maryland 20852. FOR FURTHER INFORMATION CONTACT: Cameron S. Goodwin, Office of New...: Cameron.Goodwin@nrc.gov . Further information This electronic distribution process was first utilized by... S. Goodwin, Project Manager, Small Modular Reactor Licensing Branch 2, Division of Advanced...

  4. Nordic Forum for Generation IV Reactors,Status and activities in 2012

    DEFF Research Database (Denmark)

    Van Nieuwenhove, Rudi; Lauritzen, Bent; Nonbøl, Erik

    The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risø, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many as-pects in Gen-IV reactor research and gave a good overview of the activi-ties within this fi...

  5. Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior

    Directory of Open Access Journals (Sweden)

    Ordóñez Ródenas José

    2016-01-01

    Full Text Available One of the goals of Generation IV reactors is to increase safety from those of previous generations. Different research platforms have been identified the need to improve the reliability of the simulation tools to ensure the capability of the plant to accommodate the design basis transients established in preliminary safety studies. The paper describes the modelling of primary pumps in advanced sodium cooled reactors using the TRACE code. Following the implementation of the models, the results obtained in the analysis of different design basis transients are compared with the simplifying approximations used in reference models. The paper shows the process to obtain a consistent pump model of the ESFR (European Sodium Fast Reactor design and the analysis of loss of flow transients triggered by pumps coast–down analyzing the thermal hydraulic neutronic coupled system response. A sensitivity analysis of the system pressure drops effect and the other relevant parameters that influence the natural convection after the pumps coast–down is also included.

  6. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  7. The Design and Control of Distillation Column with Side Reactors for Chlorobenzene Production

    Institute of Scientific and Technical Information of China (English)

    薄翠梅; 汤吉海; 柏杨进; 乔旭; 丁良辉; 张湜

    2012-01-01

    The distillation column with side reactors (SRC) can overcome the temperature/pressure mismatch in the traditional reactive distillation, the column operates at temperature/pressure favorable for vapor-liquid separation, while the reactors operate at temperatures/pressures favorable for reaction kinetics. According to the smooth operation and automatic control problem of the distillation column with side reactors (SRC), the design, simulation calculation and dynamic control of the SCR process for chlorobenzene production are discussed in the paper. Firstly, the mechanism models, the integrated structure optimal design and process simulation systems are established, respectively. And then multivariable control schemes are designed, the controllability of SRC process based on the optimal steady-state integrated structure is explored. The dynamic response performances of closed-loop system against several disturbances are discussed to verify the effectiveness of control schemes for the SRC process. The simulating results show that the control structure using conventional control strategies can effectively overcome feeding disturbances in a specific range.

  8. Combining COMSOL modeling with acoustic pressure maps to design sono-reactors.

    Science.gov (United States)

    Wei, Zongsu; Weavers, Linda K

    2016-07-01

    Scaled-up and economically viable sonochemical systems are critical for increased use of ultrasound in environmental and chemical processing applications. In this study, computational simulations and acoustic pressure maps were used to design a larger-scale sono-reactor containing a multi-stepped ultrasonic horn. Simulations in COMSOL Multiphysics showed ultrasonic waves emitted from the horn neck and tip, generating multiple regions of high acoustic pressure. The volume of these regions surrounding the horn neck were larger compared with those below the horn tip. The simulated acoustic field was verified by acoustic pressure contour maps generated from hydrophone measurements in a plexiglass box filled with water. These acoustic pressure contour maps revealed an asymmetric and discrete distribution of acoustic pressure due to acoustic cavitation, wave interaction, and water movement by ultrasonic irradiation. The acoustic pressure contour maps were consistent with simulation results in terms of the effective scale of cavitation zones (∼ 10 cm and acoustic field and identified cavitation location, a cylindrically-shaped sono-reactor with a conical bottom was designed to evaluate the treatment capacity (∼ 5 L) for the multi-stepped horn using COMSOL simulations. In this study, verification of simulation results with experiments demonstrates that coupling of COMSOL simulations with hydrophone measurements is a simple, effective and reliable scientific method to evaluate reactor designs of ultrasonic systems.

  9. Status and perspective of development of cold moderators at the IBR-2 reactor

    Science.gov (United States)

    Kulikov, S.; Shabalin, E.

    2012-03-01

    The modernized IBR-2M reactor will start its operation with three water grooved moderators in 2011. Afterwards, they will be exchanged by a new complex of moderators. The complex consists of three so-called kombi-moderators, each of them containing a pre-moderator, a cold moderator, grooved ambient water moderators and post-moderators. They are mounted onto three moveable trolleys that serve to deliver and install moderators near the reactor core. The project is divided in three stages. In 2012 the first stage of development of complex of moderators will be finished. The water grooved moderator will be replaced with the new kombi-moderator for beams #7, 8, 10, 11. Main parameters of moderators for this direction will be studied then. The next stages will be done for beams #2-3 and for beams #1, 9, 4-6, consequently. Cold moderator chambers at the modernized IBR-2 reactor are filled with thousands of beads (~3.5 - 4 mm in diameter) of moderating material. The cold helium gas flow delivers beads from the charging device to the moderator during the fulfillment process and cools down them during the reactor cycle. The mixture of aromatic hydrocarbons (mesithylen and m-xylen) has been chosen as moderating material. The explanation of the choice of material for novel cold neutron moderators, configuration of moderator complex for the modernized IBR-2 reactor and the main results of optimization of moderator complex for the third stage of moderator development are discussed in the article.

  10. Evaluation of a Business Case for Safeguards by Design in Nuclear Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Thomas W.; Seward, Amy M.; Lewis, Valerie A.; Gitau, Ernest TN; Zentner, Michael D.

    2012-12-01

    Safeguards by Design (SbD) is a well-known paradigm for consideration and incorporation of safeguards approaches and associated design features early in the nuclear facility development process. This paradigm has been developed as part of the Next Generation Safeguards Initiative (NGSI), and has been accepted as beneficial in many discussions and papers on NGSI or specific technologies under development within NGSI. The Office of Nuclear Safeguards and Security funded the Pacific Northwest National Laboratory to examine the business case justification of SbD for nuclear power reactors. Ultimately, the implementation of SbD will rely on the designers of nuclear facilities. Therefore, it is important to assess the incentives which will lead designers to adopt SbD as a standard practice for nuclear facility design. This report details the extent to which designers will have compelling economic incentives to adopt SbD.

  11. Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters

    Energy Technology Data Exchange (ETDEWEB)

    Darmann, Frank [Zenergy Power, Inc., Burlingame, CA (United States); Lombaerde, Robert [Zenergy Power, Inc., Burlingame, CA (United States); Moriconi, Franco [Zenergy Power, Inc., Burlingame, CA (United States); Nelson, Albert [Zenergy Power, Inc., Burlingame, CA (United States)

    2012-03-01

    Zenergy Power has successfully designed, built, tested, and installed in the US electrical grid a saturable reactor Fault Current Limiter. Beginning in 2007, first as SC Power Systems and from 2008 as Zenergy Power, Inc., ZP used DOE matching grant and ARRA funds to help refine the design of the saturated reactor fault current limiter. ZP ultimately perfected the design of the saturated reactor FCL to the point that ZP could reliably design a suitable FCL for most utility applications. Beginning with a very basic FCL design using 1G HTS for a coil housed in a LN2 cryostat for the DC bias magnet, the technology progressed to a commercial system that was offered for sale internationally. Substantial progress was made in two areas. First, the cryogenics cooling system progressed from a sub-cooled liquid nitrogen container housing the HTS coils to cryostats utilizing dry conduction cooling and reaching temperatures down to less than 20 degrees K. Large, round cryostats with warm bore diameters of 1.7 meters enabled the design of large tanks to hold the AC components. Second, the design of the AC part of the FCL was refined from a six legged spider design to a more compact and lighter design with better fault current limiting capability. Further refinement of the flux path and core shape led to an efficient saturated reactor design requiring less Ampere-turns to saturate the core. In conclusion, the development of the saturable reactor FCL led to a more efficient design not requiring HTS magnets and their associated peripheral equipment, which yielded a more economical product in line with the electric utility industry expectations. The original goal for the DOE funding of the ZP project Design, Test and Demonstration of Saturable Reactor High-Temperature Superconductor Fault Current Limiters was to stimulate the HTS wire industry with, first 1G, then 2G, HTS wire applications. Over the approximately 5 years of ZP's product development program, the amount of HTS

  12. A Fusion Reactor Design with a Liquid First Wall and Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Nygren, R E; Rognlien, T D; Rensink, M E; Smolentsev, S S; Youssef, M E; Sawan, M Z; Merrill, B J; Eberle, C; Fogarty, P J; Nelson, B E; Sze, D K; Majeski, R

    2003-11-13

    Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840MW of fusion power of which 767MW is in the form of energetic particles (alpha power) and 3073MW is in the form of neutrons. The alpha plus auxiliary power total 909MW of which 430MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.

  13. Adapting computational optimization concepts from aeronautics to nuclear fusion reactor design

    Directory of Open Access Journals (Sweden)

    Baelmans M.

    2012-10-01

    Full Text Available Even on the most powerful supercomputers available today, computational nuclear fusion reactor divertor design is extremely CPU demanding, not least due to the large number of design variables and the hybrid micro-macro character of the flows. Therefore, automated design methods based on optimization can greatly assist current reactor design studies. Over the past decades, “adjoint methods” for shape optimization have proven their virtue in the field of aerodynamics. Applications include drag reduction for wing and wing-body configurations. Here we demonstrate that also for divertor design, these optimization methods have a large potential. Specifically, we apply the continuous adjoint method to the optimization of the divertor geometry in a 2D poloidal cross section of an axisymmetric tokamak device (as, e.g., JET and ITER, using a simplified model for the plasma edge. The design objective is to spread the target material heat load as much as possible by controlling the shape of the divertor, while maintaining the full helium ash removal capabilities of the vacuum pumping system.

  14. Schedule and status of irradiation experiments

    Energy Technology Data Exchange (ETDEWEB)

    Rowcliffe, A.F.; Grossbeck, M.L. [Oak Ridge National Lab., TN (United States)

    1997-04-01

    To provide an updated summary of the status of irradiation experiments for the neutron-interactive materials program. The current status of reactor irradiation experiments is presented in tables summarizing the experimental objectives, conditions, and schedule. Currently, the program has two irradiation experiments in reactor; and 8 experiments in the planning or design stages. Postirradiation examination and testing is in progress on 18 experiments.

  15. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2016-05-15

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  16. Preliminary conceptual design of inspection and maintenance for KALIMER reactor system

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Kim, Seok Hun; Yoo, Bong

    2000-08-01

    In-service inspection and maintenance are very important for improving the safety and availability of nuclear power plants. The conceptual requirements of in-service inspection and maintenance should be reflected in the earlier design process for the verification of the plant operability and reliability. In this report the fundamental approaches of the inspection and maintenance for KALIMER are established to ensure the structural integrity and operability for KALIMER. The general strategy and methodology of maintenance and inspection for the reactor system and components are proposed and described for satisfying the intents of the section XI, division 3, of ASME code and considering the design characteristics of KALIMER.

  17. Relevant thermal-hydraulic aspects in the design of the RRR (Replacement Research Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Doval, Alicia S.; Mazufri, Claudio M. [INVAP SE, Bariloche (Argentina)

    2002-07-01

    A description of the main thermal-hydraulic features and challenges of the Replacement Research Reactor, for the Australian Nuclear Science and Technology Organization (ANSTO), is presented. Different hydraulic and thermal-hydraulic aspects are considered, core cooling during full power operation and the way it affects the design, design criteria, engineered safety features and computational tools, amongst others. A special section is devoted to the thermal-hydraulic aspects inside the reflector tank, as well as the cooling of irradiation facilities, particularly, the Molybdenum production facility. (author)

  18. The conceptual design of a robust, compact, modular tokamak reactor based on high-field superconductors

    Science.gov (United States)

    Whyte, D. G.; Bonoli, P.; Barnard, H.; Haakonsen, C.; Hartwig, Z.; Kasten, C.; Palmer, T.; Sung, C.; Sutherland, D.; Bromberg, L.; Mangiarotti, F.; Goh, J.; Sorbom, B.; Sierchio, J.; Ball, J.; Greenwald, M.; Olynyk, G.; Minervini, J.

    2012-10-01

    Two of the greatest challenges to tokamak reactors are 1) large single-unit cost of each reactor's construction and 2) their susceptibility to disruptions from operation at or above operational limits. We present an attractive tokamak reactor design that substantially lessens these issues by exploiting recent advancements in superconductor (SC) tapes allowing peak field on SC coil > 20 Tesla. A R˜3.3 m, B˜9.2 T, ˜ 500 MW fusion power tokamak provides high fusion gain while avoiding all disruptive operating boundaries (no-wall beta, kink, and density limits). Robust steady-state core scenarios are obtained by exploiting the synergy of high field, compact size and ideal efficiency current drive using high-field side launch of Lower Hybrid waves. The design features a completely modular replacement of internal solid components enabled by the demountability of the coils/tapes and the use of an immersion liquid blanket. This modularity opens up the possibility of using the device as a nuclear component test facility.

  19. Design of a self-tuning regulator for temperature control of a polymerization reactor.

    Science.gov (United States)

    Vasanthi, D; Pranavamoorthy, B; Pappa, N

    2012-01-01

    The temperature control of a polymerization reactor described by Chylla and Haase, a control engineering benchmark problem, is used to illustrate the potential of adaptive control design by employing a self-tuning regulator concept. In the benchmark scenario, the operation of the reactor must be guaranteed under various disturbing influences, e.g., changing ambient temperatures or impurity of the monomer. The conventional cascade control provides a robust operation, but often lacks in control performance concerning the required strict temperature tolerances. The self-tuning control concept presented in this contribution solves the problem. This design calculates a trajectory for the cooling jacket temperature in order to follow a predefined trajectory of the reactor temperature. The reaction heat and the heat transfer coefficient in the energy balance are estimated online by using an unscented Kalman filter (UKF). Two simple physically motivated relations are employed, which allow the non-delayed estimation of both quantities. Simulation results under model uncertainties show the effectiveness of the self-tuning control concept.

  20. Challenges in polyoxometalate-mediated aerobic oxidation catalysis: catalyst development meets reactor design.

    Science.gov (United States)

    Lechner, Manuel; Güttel, Robert; Streb, Carsten

    2016-11-14

    Selective catalytic oxidation is one of the most widely used chemical processes. Ideally, highly active and selective catalysts are used in combination with molecular oxygen as oxidant, leading to clean, environmentally friendly process conditions. For homogeneous oxidation catalysis, molecular metal oxide anions, so-called polyoxometalates (POMs) are ideal prototypes which combine high reactivity and stability with chemical tunability on the molecular level. Typically, POM-mediated aerobic oxidations are biphasic, using gaseous O2 and liquid reaction mixtures. Therefore, the overall efficiency of the reaction is not only dependent on the chemical components, but requires chemical engineering insight to design reactors with optimized productivity. This Perspective shows that POM-mediated aerobic liquid-phase oxidations are ideal reactions to be carried out in microstructured flow reactors as they enable facile mass and energy transfer, provide large gas-liquid interfaces and can be easily upscaled. Recent advances in POM-mediated aerobic catalytic oxidations are therefore summarized with a focus on technological importance and mechanistic insight. The principles of reactor design are discussed from a chemical engineering point of view with a focus on homogeneous oxidation catalysis using O2 in microfluidic systems. Further, current limitations to catalytic activity are identified and future directions based on combined chemistry and chemical engineering approaches are discussed to show that this approach could lead to sustainable production methods in industrial chemistry based on alternative energy sources and chemical feedstocks.

  1. MULTI-LOOP CONTROL DESIGN IN MULTIVARIABLE (2X2 CONTINUOUS STIRRED TANK REACTOR

    Directory of Open Access Journals (Sweden)

    Abdul Wahid

    2015-06-01

    Full Text Available With this study, the design and tuning of multi-loop for multivariable (2x2 CSTR will be made in order to achieve optimum CSTR control performance. This study used Bequette model reactor and MATLAB software and is expected to be able to cope with disturbances in the reactor so that the reactor system is able to stabilize quickly despite the distractions. In this study, the design will be made using multi-loop approach, along with PI controller as the next step. Then, BLT and auto-tune tuning method will be used in PI controller and given disturbances to both of tuning method. The controller performances are then compared. Results of the study are then analyzed for discussions and conclusions. Results from this study have shown that in terms of disturbance rejection, BLT is better than auto-tune based on comparison between both of controller performances. For IAE for the case of temperature, BLT is 30% better than auto-tune, but it is almost the same for the case of concentration. For settling time for the case of concentration, BLT is 30% better than auto-tune, and for the case of temperature, BLT is 18% better than auto-tune. For rise time for the case of concentration and temperature, BLT is 30% better than auto-tune.

  2. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  3. 2400MWt GAS-COOLED FAST REACTOR DHR STUDIES STATUS UPDATE.

    Energy Technology Data Exchange (ETDEWEB)

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01

    A topical report on demonstrating the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor was published in March 2006. The analysis was performed with the system code RELAP5-3D (version 2.4.1.1a) and the model included the full complement of the power conversion unit (PCU): heat exchange components (recuperator, precooler, intercooler) and rotating machines (turbine, compressor). A re-analysis of the success case in Ref is presented in this report. The case was redone to correct unexpected changes in core heat structure temperatures when the PCU model was first integrated with the reactor model as documented in Ref [1]. Additional information on the modeling of the power conversion unit and the layout of the heat exchange components is provided in Appendix A.

  4. Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea

    OpenAIRE

    Hyun-Gil Kim; Jae-Ho Yang; Weon-Ju Kim; Yang-Hyun Koo

    2016-01-01

    For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF) became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell UO2 and high-density compo...

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

    Energy Technology Data Exchange (ETDEWEB)

    Adams, S.R.

    1985-10-01

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

  6. Small-break loss-of-coolant accidents in the updated PIUS 600 advanced reactor design

    Energy Technology Data Exchange (ETDEWEB)

    Boyack, B.E.; Steiner, J.L.; Harmony, S.C. [Los Alamos National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness of the PIUS concept to severe off-normal conditions having a very low probability of occurrence.

  7. Cost-based optimization of a nuclear reactor core design: a preliminary model

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner F.; Alves Filho, Hermes [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Inst. Politecnico. Dept. de Modelagem Computacional]. E-mails: wfsacco@iprj.uerj.br; halves@iprj.uerj.br; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil). Div. de Reatores]. E-mail: cmnap@ien.gov.br

    2007-07-01

    A new formulation of a nuclear core design optimization problem is introduced in this article. Originally, the optimization problem consisted in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the radial power peaking factor in a three-enrichment zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. Here, we address the same problem using the minimization of the fuel and cladding materials costs as the objective function, and the radial power peaking factor as an operational constraint. This cost-based optimization problem is attacked by two metaheuristics, the standard genetic algorithm (SGA), and a recently introduced Metropolis algorithm called the Particle Collision Algorithm (PCA). The two algorithms are submitted to the same computational effort and their results are compared. As the formulation presented is preliminary, more elaborate models are also discussed (author)

  8. Research reactor power controller design using an output feedback nonlinear receding horizon control method

    Energy Technology Data Exchange (ETDEWEB)

    Etchepareborda, Andres [Department of Nuclear Engineering, Argentine National Atomic Energy Commission, Centro Atomico Bariloche, Av. E. Bustillo 9500, Bariloche 8400 (Argentina)]. E-mail: etche@cab.cnea.gov.ar; Lolich, Jose [INVAP S.E., Moreno 1089, Bariloche 8400 (Argentina)

    2007-02-15

    A constrained, output feedback nonlinear receding horizon control (NRHC) method is applied to design a research reactor power controller. The method uses a nonlinear plant model subject to state, control and terminal set constraints; a nonlinear cost function; and a high gain observer. The controller regulates reactor power from 1% to 100% of full power; considers known disturbances, such as reactivity insertions and changes in core inlet flow and temperature; and includes upper limits constraints on neutron flux, neutron flux rate, core outlet temperature and core inlet-outlet temperature difference. Simulation results show an excellent performance for power regulation and known disturbances rejection: all process variables are kept within the admissible limits avoiding the actuation of the safety systems.

  9. The design and simulation of TCR(thyristor control reactor) reactive power compensation system based on Arene

    Institute of Scientific and Technical Information of China (English)

    WANG Shu-fang; ZHANG Li; JIANG Jian-guo; WANG Ru-lin

    2004-01-01

    Inevitably, the question of reactive power compensation was aroused by applied of power electronics. Based on the study of the instantaneous reactive power theory, the designs of TCR(thyristor control reactor) thyristor control reactor reactive power compensation system and TCR single closed loop strategy was proposed. In addition, as digital simulation software, Arene was applied to simulate the Jining coal mine No.2 system. The simulation results validate that the design is effective to improve power factor and stabilization of the system.

  10. Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    2016-08-19

    FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for nuclear applications due to their high temperature steam oxidation resistance and hypothesized enhanced radiation tolerance. A need currently exists to determine the radiation tolerance of these newly developed alloys. To address this need, a preliminary study was conducted using the High Flux Isotope Reactor (HFIR) to irradiate an early generation FeCrAl ODS alloy, 125YF. Preliminary post-irradiation examination (PIE) on these irradiated specimens have shown good radiation tolerance at elevated temperatures (≥330°C) but possible radiation-induced hardening and embrittlement at irradiations of 200°C to a damage level of 1.9 displacement per atom (dpa). Building on this experience, a new series of irradiations are currently being conceptualized. This irradiation series called the FCAD irradiation program will irradiate the latest generation FeCrAl ODS and FeCr ODS alloys to significantly higher doses. These experiments will provide the necessary information to determine the mechanical performance of irradiated FeCrAl ODS alloys at light water reactor and fast reactor conditions.

  11. Status of FeCrAl ODS Irradiations in the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD); Howard, Richard H. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States). Fuel Cycle Research and Development (FCRD)

    2016-08-19

    FeCrAl oxide-dispersion strengthened (ODS) alloys are an attractive sub-set alloy class of the more global FeCrAl material class for nuclear applications due to their high-temperature steam oxidation resistance and hypothesized enhanced radiation tolerance. A need currently exists to determine the radiation tolerance of these newly developed alloys. To address this need, a preliminary study was conducted using the High Flux Isotope Reactor (HFIR) to irradiate an early generation FeCrAl ODS alloy, 125YF. Preliminary post-irradiation examination (PIE) on these irradiated specimens have shown good radiation tolerance at elevated temperatures (≥330°C) but possible radiation-induced hardening and embrittlement at irradiations of 200°C to a damage level of 1.9 displacement per atom (dpa). Building on this experience, a new series of irradiations are currently being conceptualized. This irradiation series called the FCAD irradiation program will irradiate the latest generation FeCrAl ODS and FeCr ODS alloys to significantly higher doses. These experiments will provide the necessary information to determine the mechanical performance of irradiated FeCrAl ODS alloys at light water reactor and fast reactor conditions.

  12. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Science.gov (United States)

    Rabir, Mohamad Hairie; Hah, Chang Joo; Ju, Cho Sung

    2015-04-01

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  13. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  14. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    Science.gov (United States)

    Anuar, Nuraslinda; Muhamad Pauzi, Anas

    2016-01-01

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the βmin is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the βmin, resulting in a list of candidate designs that possess the β value that is larger than the βmin. The proposed methodology can also be applied to purposes other than technological foresight.

  15. A mechanism for proven technology foresight for emerging fast reactor designs and concepts

    Energy Technology Data Exchange (ETDEWEB)

    Anuar, Nuraslinda, E-mail: nuraslinda@uniten.edu.my; Muhamad Pauzi, Anas, E-mail: anas@uniten.edu.my [College of Engineering, Universiti Tenaga Nasional, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    The assessment of emerging nuclear fast reactor designs and concepts viability requires a combination of foresight methods. A mechanism that allows for the comparison and quantification of the possibility of being a proven technology in the future, β for the existing fast reactor designs and concepts is proposed as one of the quantitative foresight method. The methodology starts with the identification at the national or regional level, of the factors that would affect β. The factors are then categorized into several groups; economic, social and technology elements. Each of the elements is proposed to be mathematically modelled before all of the elemental models can be combined. Once the overall β model is obtained, the β{sub min} is determined to benchmark the acceptance as a candidate design or concept. The β values for all the available designs and concepts are then determined and compared with the β{sub min}, resulting in a list of candidate designs that possess the β value that is larger than the β{sub min}. The proposed methodology can also be applied to purposes other than technological foresight.

  16. Safeguards and Security by Design (SSBD) for Small Modular Reactors (SMRs) through a Common Global Approach

    Energy Technology Data Exchange (ETDEWEB)

    Badwan, Faris M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Demuth, Scott Francis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Miller, Michael Conrad [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pshakin, Gennady [Obninsk Institute of Physics and Power Engineering (Russian Federation)

    2015-02-23

    Small Modular Reactors (SMR) with power levels significantly less than the currently standard 1000 to 1600-MWe reactors have been proposed as a potential game changer for future nuclear power. SMRs may offer a simpler, more standardized, and safer modular design by using factory built and easily transportable components. Additionally, SMRs may be more easily built and operated in isolated locations, and may require smaller initial capital investment and shorter construction times. Because many SMRs designs are still conceptual and consequently not yet fixed, designers have a unique opportunity to incorporate updated design basis threats, emergency preparedness requirements, and then fully integrate safety, physical security, and safeguards/material control and accounting (MC&A) designs. Integrating safety, physical security, and safeguards is often referred to as integrating the 3Ss, and early consideration of safeguards and security in the design is often referred to as safeguards and security by design (SSBD). This paper describes U.S./Russian collaborative efforts toward developing an internationally accepted common approach for implementing SSBD/3Ss for SMRs based upon domestic requirements, and international guidance and requirements. These collaborative efforts originated with the Nuclear Energy and Nuclear Security working group established under the U.S.-Russia Bilateral Presidential Commission during the 2009 Presidential Summit. Initial efforts have focused on review of U.S. and Russian domestic requirements for Security and MC&A, IAEA guidance for security and MC&A, and IAEA requirements for international safeguards. Additionally, example SMR design features that can enhance proliferation resistance and physical security have been collected from past work and reported here. The development of a U.S./Russian common approach for SSBD/3Ss should aid the designer of SMRs located anywhere in the world. More specifically, the application of this approach may

  17. The near boiling reactor: Conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    Science.gov (United States)

    Cole, Christopher J. P.

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the Victoria Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96°C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has highlighted additional areas

  18. The near boiling reactor : conceptual design of a small inherently safe nuclear reactor to extend the operational capability of the Victoria Class submarine

    Energy Technology Data Exchange (ETDEWEB)

    Cole, C.J.P

    2005-07-01

    Nuclear power has several unique advantages over other air independent energy sources for nuclear combat submarines. An inherently safe, small nuclear reactor, capable of supply the hotel load of the 'Victoria' Class submarines, has been conceptually developed. The reactor is designed to complement the existing diesel electric power generation plant presently onboard the submarine. The reactor, rated at greater than 1 MW thermal, will supply electricity to the submarine's batteries through an organic Rankine cycle energy conversion plant at 200 kW. This load will increase the operational envelope of the submarine by providing up to 28 continuous days submerged, allowing for an enhanced indiscretion ratio (ratio of time spent on the surface versus time submerged) and a limited under ice capability. The power plant can be fitted into the existing submarine by inserting a 6 m hull plug. With its simplistic design and inherent safety features, the reactor plant will require a minimal addition to the crew. The reactor employs TRISO fuel particles for increased safety. The light water coolant remains at atmospheric pressure, exiting the core at 96{sup o}C. Burn-up control and limiting excess reactivity is achieved through movable reflector plates. Shut down and regulatory control is achieved through the thirteen hafnium control rods. Inherent safety is achieved through the negative prompt and delayed temperature coefficients, as well as the negative void coefficient. During a transient, the boiling of the moderator results in a sudden drop in reactivity, essentially shutting down the reactor. It is this characteristic after which the reactor has been named. The design of the reactor was achieved through modelling using computer codes such as MCNP5, WIMS-AECL, FEMLAB, and MicroShield5, in addition to specially written software for kinetics, heat transfer and fission product poisoning calculations. The work has covered a broad area of research and has

  19. Chemical looping reactor system design double loop circulating fluidized bed (DLCFB)

    Energy Technology Data Exchange (ETDEWEB)

    Bischi, Aldo

    2012-05-15

    Chemical looping combustion (CLC) is continuously gaining more importance among the carbon capture and storage (CCS) technologies. It is an unmixed combustion process which takes place in two steps. An effective way to realize CLC is to use two interconnected fluidized beds and a metallic powder circulating among them, acting as oxygen carrier. The metallic powder oxidizes at high temperature in one of the two reactors, the air reactor (AR). It reacts in a highly exothermic reaction with the oxygen of the injected fluidising air. Afterwards the particles are sent to the other reactor where the fuel is injected, the fuel reactor (FR). There, they transport heat and oxygen necessary for the reaction with the injected fuel to take place. At high temperatures, the particle's oxygen reacts with the fuel producing Co2 and steam, and the particles are ready to start the loop again. The overall reaction, the sum of the enthalpy changes of the oxygen carrier oxidation and reduction reactions, is the same as for the conventional combustion. Two are the key features, which make CLC promising both for costs and capture efficiency. First, the high inherent irreversibility of the conventional combustion is avoided because the energy is utilized stepwise. Second, the Co2 is intrinsically separated within the process; so there is in principle no need either of extra carbon capture devices or of expensive air separation units to produce oxygen for oxy-combustion. A lot of effort is taking place worldwide on the development of new chemical looping oxygen carrier particles, reactor systems and processes. The current work is focused on the reactor system: a new design is presented, for the construction of an atmospheric 150kWth prototype working with gaseous fuel and possibly with inexpensive oxygen carriers derived from industrial by-products or natural minerals. It consists of two circulating fluidized beds capable to operate in fast fluidization regime; this will increase the

  20. Conceptual Design study of Small Long-life Gas Cooled Fast Reactor With Modified CANDLE Burn-up Scheme

    Science.gov (United States)

    Nur Asiah, A.; Su'ud, Zaki; Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this paper, conceptual design study of Small Long-life Gas Cooled Fast Reactors with Natural Uranium as Fuel Cycle Input has been performed. In this study Gas Cooled Fast Reactor is slightly modified by employing modified CANDLE burn-up scheme so that it can use Natural Uranium as fuel cycle input. Due to their hard spectrum, GCFR in this study showed very good performance in converting U-238 to plutonium in order to maintain the operation condition requirement of long-life reactors. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. With such condition we got an optimal design of 325 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input. The average discharge burn-up is about 290 GWd/ton HM.

  1. Preliminary Conceptual Design and Development of Core Technology of Very High Temperature Gas-Cooled Reactor Hydrogen Production

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Jong Hwa; Kang, H. S.; Gil, C. S. and others

    2006-05-15

    For the nuclear hydrogen production system, the VHTR technology and the IS cycle technology are being developed. A comparative evaluation on the block type reactor and the pebble type reactor is performed to decide a proper nuclear hydrogen production reactor. 100MWt prismatic type reactor is tentatively decided and its safety characteristics are roughly investigated. Computation codes of nuclear design, thermo-fluid design, safety-performance analysis are developed and verified. Also, the development of a risk informed design technology is started. Experiments for metallic materials and graphites are carried out for the selection of materials of VHTR components. Diverse materials for process heat exchanger are studied in various corrosive environments. Pyrolytic carbon and SiC coating technology is developed and fuel manufacturing technology is basically established. Computer program is developed to evaluate the performance of coated particle fuels.

  2. Status of Wrought FeCrAl-UO2 Capsules Irradiated in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harp, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Core, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Linton, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    Candidate cladding materials for accident tolerant fuel applications require extensive testing and validation prior to commercial deployment within the nuclear power industry. One class of cladding materials, FeCrAl alloys, is currently undergoing such effort. Within these activities is a series of irradiation programs within the Advanced Test Reactor. These programs are developed to aid in commercial maturation and understand the fundamental mechanisms controlling the cladding performance during normal operation of a typical light water reactor. Three different irradiation programs are on-going; one designed as a simple proof-of-principle concept, the other to evaluate the susceptibility of FeCrAl to fuel-cladding chemical interaction, and the last to fully simulate the conditions of a pressurized water reactor experimentally. To date, nondestructive post-irradiation examination has been completed on the rodlet deemed FCA-L3 from the simple proof-of-concept irradiation program. Initial results show possible breach of the rodlet under irradiation but further studies are needed to conclusively determine whether breach has occurred and the underlying reasons for such a possible failure. Further work includes characterizing additional rodlets following irradiation.

  3. Design and Analysis of Thorium-fueled Reduced Moderation Boiling Water Reactors

    Science.gov (United States)

    Gorman, Phillip Michael

    The Resource-renewable Boiling Water Reactors (RBWRs) are a set of light water reactors (LWRs) proposed by Hitachi which use a triangular lattice and high void fraction to incinerate fuel with an epithermal spectrum, which is highly atypical of LWRs. The RBWRs operate on a closed fuel cycle, which is impossible with a typical thermal spectrum reactor, in order to accomplish missions normally reserved for sodium fast reactors (SFRs)--either fuel self-sufficiency or waste incineration. The RBWRs also axially segregate the fuel into alternating fissile "seed" regions and fertile "blanket" regions in order to enhance breeding and leakage probability upon coolant voiding. This dissertation focuses on thorium design variants of the RBWR: the self-sufficient RBWR-SS and the RBWR-TR, which consumes reprocessed transuranic (TRU) waste from PWR used nuclear fuel. These designs were based off of the Hitachi-designed RBWR-AC and the RBWR-TB2, respectively, which use depleted uranium (DU) as the primary fertile fuel. The DU-fueled RBWRs use a pair of axially segregated seed sections in order to achieve a negative void coefficient; however, several concerns were raised with this multi-seed approach, including difficulty with controlling the reactor and unacceptably high axial power peaking. Since thorium-uranium fuel tends to have much more negative void feedback than uranium-plutonium fuels, the thorium RBWRs were designed to use a single elongated seed to avoid these issues. A series of parametric studies were performed in order to find the design space for the thorium RBWRs, and optimize the designs while meeting the required safety constraints. The RBWR-SS was optimized to maximize the discharge burnup, while the RBWR-TR was optimized to maximize the TRU transmutation rate. These parametric studies were performed on an assembly level model using the MocDown simulator, which calculates an equilibrium fuel composition with a specified reprocessing scheme. A full core model was

  4. Design issues on using FPGA-based I and C systems in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Farias, Marcos S.; Carvalho, Paulo Victor R. de; Santos, Isaac Jose A.L. dos; Lacerda, Fabio de, E-mail: msantana@ien.gov.br, E-mail: paulov@ien.gov.br, E-mail: luquetti@ien.gov.br, E-mail: acerda@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Div. de Engenharia Nuclear

    2015-07-01

    The FPGA (field programmable gate array) is widely used in various fields of industry. FPGAs can be used to perform functions that are safety critical and require high reliability, like in automobiles, aircraft control and assistance and mission-critical applications in the aerospace industry. With these merits, FPGAs are receiving increased attention worldwide for application in nuclear plant instrumentation and control (I and C) systems, mainly for Reactor Protection System (RPS). Reasons for this include the fact that conventional analog electronics technologies are become obsolete. I and C systems of new Reactors have been designed to adopt the digital equipment such as PLC (Programmable Logic Controller) and DCS (Distributed Control System). But microprocessors-based systems may not be simply qualified because of its complex characteristics. For example, microprocessor cores execute one instruction at a time, and an operating system is needed to manage the execution of programs. In turn, FPGAs can run without an operating system and the design architecture is inherently parallel. In this paper we aim to assess these and other advantages, and the limitations, on FPGA-based solutions, considering the design guidelines and regulations on the use of FPGAs in Nuclear Plant I and C Systems. We will also examine some circuit design techniques in FPGA to help mitigate failures and provide redundancy. The objective is to show how FPGA-based systems can provide cost-effective options for I and C systems in modernization projects and to the RMB (Brazilian Multipurpose Reactor), ensuring safe and reliable operation, meeting licensing requirements, such as separation, redundancy and diversity. (author)

  5. A Small Modular Reactor Core Design using FCM Fuel and BISO BP particles

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Yeon; Hwang, Dae Hee; Yoo, Ho Seong; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    The objective of this work is to design a PWR small modular reactor which employs the advanced fuel technology of FCM particle fuels including BISO burnable poisons and advanced cladding of SiC in order to improve the fuel economy and safety by increasing fuel burnup and temperature, and by reducing hydrogen generation under accidents. Recently, many countries including USA have launched projects to develop the accident tolerant fuels (ATF) which can cope with the accidents such as LOCA (Loss of Coolant Accident). In general, the ATF fuels are required to meet the PWR operational, safety, and fuel cycle constraints which include enhanced burnup, lower or no generation of hydrogen, lower operating temperatures, and enhanced retention of fission products. Another stream of research and development in nuclear society is to develop advanced small modular reactors in order to improve inherent passive safety and to reduce the risk of large capital investment. In this work, a small PWR modular reactor core was neutronically designed and analyzed. The SMR core employs new 13x13 fuel assemblies which are loaded with thick FCM fuel rods in which TRISO fuel particles AO and also the first cycle has the AOs which are within the typical design limit. Also, this figure shows that the evolutions of AO for the cycles 6 and 7 are nearly the same. we considered the SiC cladding for reduction of hydrogen generation under accidents. From the results of core design and analysis, it is shown that the core has long cycle length of 732 -1191 EFPDs, high discharge burnup of 101-105 MWD/kg, low power peaking factors, low axial offsets, negative MTCs, and large shutdown margins except for BOC of the first cycle. So, it can be concluded that the new SMR core is neutronically feasible.

  6. Hydrocarbon reforming catalysts and new reactor designs for compact hydrogen generators

    Energy Technology Data Exchange (ETDEWEB)

    Schaefer, A.; Schwab, E.; Urtel, H. [BASF SE, Ludwigshafen (Germany); Farrauto, R. [BASF Catalysts LLC, Iselin, NJ (United States)

    2010-12-30

    A hydrogen based future energy scenario will use fuel cells for the conversion of chemically stored energy into electricity. Depending upon the type of fuel cell, different specifications will apply for the feedstock which is converted in the cell, ranging from very clean hydrogen for PEM-FC's to desulfurized methane for SOFC and MCFC technology. For the foreseeable future, hydrogen will be supplied by conventional reforming, however operated in compact and dynamic reformer designs. This requires that known catalyst formulations are offered in specific geometries, giving flexibility for novel reactor design options. These specific geometries can be special tablet shapes as well as monolith structures. Finally, also nonhydrocarbon feedstock might be used in special applications, e.g. bio-based methanol and ethanol. BASF offers catalysts for the full process chain starting from feedstock desulfurization via reforming, high temperature shift, low temperature shift to CO fine polishing either via selective oxidation or selective methanation. Depending upon the customer's design, most stages can be served either with precious metal based monolith solutions or base metal tablet solutions. For the former, we have taken the automobile catalyst monolith support and extended its application to the fuel cell hydrogen generation. Washcoats of precious metal supported catalysts can for example be deposited on ceramic monoliths and/or metal heat exchangers for efficient generation of hydrogen. Major advantages are high through puts due to more efficient heat transfer for catalysts on metal heat exchangers, lower pressure drop with greater catalyst mechanical and thermal stability compared to particulate catalysts. Base metal tablet catalysts on the other hand can have intrinsic cost advantages, larger fractions of the reactor can be filled with active mass, and if produced in unconventional shape, again novel reactor designs are made possible. Finally, if it comes to

  7. Design Parameters for a Natural Uranium UO{sub 3} or U{sub 3}O{sub 8} Fueled Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hopper, C.M.

    2002-11-15

    A recent Oak Ridge National Laboratory report provided preliminary analyses to propose alternative design parameters for a nuclear reactor that could be fueled with natural UO{sub 3} or U{sub 3}O{sub 8} and moderated with either heavy water or reactor-grade graphite. This report provides more specific reactor design and operating parameters for a heavy water-moderated reactor only. The basic assumptions and analytical approach are discussed together with the results of the analysis.

  8. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  9. Use of freeze-casting in advanced burner reactor fuel design

    Energy Technology Data Exchange (ETDEWEB)

    Lang, A. L.; Yablinsky, C. A.; Allen, T. R. [Dept. of Engineering Physics, Univ. of Wisconsin Madison, 1500 Engineering Drive, Madison, WI 53711 (United States); Burger, J.; Hunger, P. M.; Wegst, U. G. K. [Thayer School of Engineering, Dartmouth College, 8000 Cummings Hall, Hanover, NH 03755 (United States)

    2012-07-01

    This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models

  10. Integrating Safety, Operations, Security, and Safeguards (ISOSS) into the design of small modular reactors : a handbook.

    Energy Technology Data Exchange (ETDEWEB)

    Middleton, Bobby D.; Mendez, Carmen Margarita

    2013-10-01

    The existing regulatory environment for nuclear reactors impacts both the facility design and the cost of operations once the facility is built. Delaying the consideration of regulatory requirements until late in the facility design - or worse, until after construction has begun - can result in costly retrofitting as well as increased operational costs to fulfill safety, security, safeguards, and emergency readiness requirements. Considering the scale and scope, as well as the latest design trends in the next generation of nuclear facilities, there is an opportunity to evaluate the regulatory requirements and optimize the design process for Small Modular Reactors (SMRs), as compared to current Light Water Reactors (LWRs). To this end, Sandia has embarked on an initiative to evaluate the interactions of regulations and operations as an approach to optimizing the design of SMR facilities, supporting operational efficiencies, as well as regulatory requirements. The early stages of this initiative consider two focus areas. The first focus area, reported by LaChance, et al. (2007), identifies the regulatory requirements established for the current fleet of LWR facilities regarding Safety, Security, Operations, Safeguards, and Emergency Planning, and evaluates the technical bases for these requirements. The second focus area, developed in this report, documents the foundations for an innovative approach that supports a design framework for SMR facilities that incorporates the regulatory environment, as well as the continued operation of the facility, into the early design stages, eliminating the need for costly retrofitting and additional operating personnel to fulfill regulatory requirements. The work considers a technique known as Integrated Safety, Operations, Security and Safeguards (ISOSS) (Darby, et al., 2007). In coordination with the best practices of industrial operations, the goal of this effort is to develop a design framework that outlines how ISOSS

  11. Design considerations for post accident monitoring system of a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Gwi Sook; Park, Je Yun; Kim, Young Ki [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The Post Accident Monitoring System (PAMS) provides primary information for operators to assess the plant conditions and perform their role in bringing the plant to a safe condition during an accident. The PAMS of NPP (Nuclear Power Plant) in KOREA provides the continuous display of the PAM category 1 parameters specified in R.G 1.97, Rev. 03. Recently the PAMS of NPP has been designed according to R.G 1.97, Rev. 04. There is no PAMS at the HANARO in KOREA, but recently RRs (Research Reactors) around the world are going to have PAMS for various multi purposes. We should determine the design considerations for PAMS in a Korean RR based on the design state analysis. Thus, this paper proposes strategies on the design considerations for the PAMS of a Korean RR.

  12. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    Science.gov (United States)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  13. Liquid immersion blanket design for use in a compact modular fusion reactor

    Science.gov (United States)

    Sorbom, Brandon; Ball, Justin; Barnard, Harold; Haakonsen, Christian; Hartwig, Zachary; Olynyk, Geoffrey; Sierchio, Jennifer; Whyte, Dennis

    2012-10-01

    Traditional tritium breeding blankets in fusion reactor designs include a large amount of structural material. This results in complex engineering requirements, complicated sector maintenance, and marginal tritium breeding ratios (TBR). We present a conceptual design of a fully liquid blanket. To maximize tritium breeding volume, the vacuum vessel is completely immersed in a continuously recycled FLiBe blanket, with the exception of small support posts. FLiBe has a wide liquid temperature window (459 C to 1430 C), low electrical conductivity to minimize MHD effects, similar thermal/fluid characteristics to water, and is chemically inert. While tritium breeding with FLiBe in traditional blankets is poor, we use MCNP neutronics analysis to show that the immersion blanket design coupled with a beryllium neutron multiplier results in TBR > 1. FLiBe is shown to be a sufficient radiation shield for the toroidal field magnets and can be used as a coolant for the vacuum vessel and divertor, allowing for a simplified single-phase, low-pressure, single-fluid cooling scheme. When coupled with a high-field compact reactor design, the immersion blanket eliminates the need for complex sector maintenance, allows the vacuum vessel to be a replaceable component, and reduces financial cost.

  14. Design and Optimization for the Windowless Target of the China Nuclear Waste Transmutation Reactor

    Directory of Open Access Journals (Sweden)

    Desheng Cheng

    2016-04-01

    Full Text Available A windowless spallation target can provide a neutron source and maintain neutron chain reaction for a subcritical reactor, and is a key component of China's nuclear waste transmutation of coupling accelerator and subcritical reactor. The main issue of the windowless target design is to form a stable and controllable free surface that can ensure that energy spectrum distribution is acquired for the neutron physical design when the high energy proton beam beats the lead–bismuth eutectic in the spallation target area. In this study, morphology and flow characteristics of the free surface of the windowless target were analyzed through the volume of fluid model using computational fluid dynamics simulation, and the results show that the outlet cross section size of the target is the key to form a stable and controllable free surface, as well as the outlet with an arc transition. The optimization parameter of the target design, in which the radius of outlet cross section is 60 ± 1 mm, is verified to form a stable and controllable free surface and to reduce the formation of air bubbles. This work can function as a reference for carrying out engineering design of windowless target and for verification experiments.

  15. Design and optimization for the windowless target of the China Nuclear Waste Transmutation Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, De Sheng; Wang, Weihua; Yang, Shi Jun; Deng, Haifei; Wang, Rong Fei; Wang, Bin Jun [Institute Applied Physics of AOA, Hefei (China)

    2016-04-15

    A windowless spallation target can provide a neutron source and maintain neutron chain reaction for a subcritical reactor, and is a key component of China's nuclear waste transmutation of coupling accelerator and subcritical reactor. The main issue of the windowless target design is to form a stable and controllable free surface that can ensure that energy spectrum distribution is acquired for the neutron physical design when the high energy proton beam beats the lead-bismuth eutectic in the spallation target area. In this study, morphology and flow characteristics of the free surface of the windowless target were analyzed through the volume of fluid model using computational fluid dynamics simulation, and the results show that the outlet cross section size of the target is the key to form a stable and controllable free surface, as well as the outlet with an arc transition. The optimization parameter of the target design, in which the radius of outlet cross section is 60 ± 1 mm, is verified to form a stable and controllable free surface and to reduce the formation of air bubbles. This work can function as a reference for carrying out engineering design of windowless target and for verification experiments.

  16. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH{sub 2} and B{sub 4}C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor.

  17. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    Energy Technology Data Exchange (ETDEWEB)

    Korobkina, E., E-mail: ekorobk@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Medlin, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Wehring, B.; Hawari, A.I. [Department of Nuclear Engineering, North Carolina State University, 2500 Stinson Drive, Box 7909, Raleigh, NC 27695 (United States); Huffman, P.R.; Young, A.R. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States); Beaumont, B. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Palmquist, G. [Department of Physics, North Carolina State University, 2401 Stinson Drive, Box 8202, Raleigh, NC 27695 (United States); Triangle Universities Nuclear Laboratory, 116 Science Drive, Box 90308, Durham, NC 27708 (United States)

    2014-12-11

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ∼40K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  18. Ultracold neutron source at the PULSTAR reactor: Engineering design and cryogenic testing

    Science.gov (United States)

    Korobkina, E.; Medlin, G.; Wehring, B.; Hawari, A. I.; Huffman, P. R.; Young, A. R.; Beaumont, B.; Palmquist, G.

    2014-12-01

    Construction is completed and commissioning is in progress for an ultracold neutron (UCN) source at the PULSTAR reactor on the campus of North Carolina State University. The source utilizes two stages of neutron moderation, one in heavy water at room temperature and the other in solid methane at ~ 40 K, followed by a converter stage, solid deuterium at 5 K, that allows a single down scattering of cold neutrons to provide UCN. The UCN source rolls into the thermal column enclosure of the PULSTAR reactor, where neutrons will be delivered from a bare face of the reactor core by streaming through a graphite-lined assembly. The source infrastructure, i.e., graphite-lined assembly, heavy-water system, gas handling system, and helium liquefier cooling system, has been tested and all systems operate as predicted. The research program being considered for the PULSTAR UCN source includes the physics of UCN production, fundamental particle physics, and material surface studies of nanolayers containing hydrogen. In the present paper we report details of the engineering and cryogenic design of the facility as well as results of critical commissioning tests without neutrons.

  19. Design of a new wet storage rack for spent fuels from IEA-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rodrigues, Antonio C.I.; Madi Filho, Tufic; Siqueira, Paulo T.D.; Ricci Filho, Walter, E-mail: acirodri@ipen.br, E-mail: tmfilho@ipen.br, E-mail: ptsiquei@ipen.br, E-mail: wricci@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The IEA-R1 research reactor operates in a regimen of 64h weekly, at the power of 4.5 MW. In these conditions, the racks of the spent fuel elements have less than half of its initial capacity. Thus, maintaining these operating conditions, the storage will have capacity for about six years. Since the estimated useful life of the IEA-R1 is about another 20 years, it will be necessary to increase the storage capacity of spent fuel. Dr. Henrik Grahn, expert of the International Atomic Energy Agency on wet storage, visiting the IEA-R1 Reactor (September/2012) made some recommendations: among them, the design and installation of racks made with borated stainless steel and internally coated with an aluminum film, so that corrosion of the fuel elements would not occur. After an extensive literature review of material options given for this type of application we got to Boral® manufactured by 3M due to numerous advantages. This paper presents studies on the analysis of criticality using the computer code MCNP 5, demonstrating the possibility of doubling the storage capacity of current racks to attend the demand of the IEA-R1 reactor while attending the safety requirements the International Atomic Energy Agency. (author)

  20. State Observer Design for Monitoring the Degree of Polymerization in a Series of Melt Polycondensation Reactors

    Directory of Open Access Journals (Sweden)

    Chen Ling

    2016-01-01

    Full Text Available A nonlinear reduced-order state observer is applied to estimate the degree of polymerization in a series of polycondensation reactors. The finishing stage of polyethylene terephthalate synthesis is considered in this work. This process has a special structure of lower block triangular form, which is properly utilized to facilitate the calculation of the state-dependent gain in the observer design. There are two possible on-line measurements in each reactor. One is continuous, and the other is slow-sampled with dead time. For the slow-sampled titration measurement, inter-sample behavior is estimated from an inter-sample output predictor, which is essential in providing continuous corrections on the observer. Dead time compensation is carried out in the same spirit as the Smith predictor to reduce the effect of delay in the measurement outputs. By integrating the continuous-time reduced-order observer, the inter-sample predictor and the dead time compensator together, the degree of polymerization is accurately estimated in all reactors. The observer performance is demonstrated by numerical simulations. In addition, a pre-filtering technique is used in the presence of sensor noise.

  1. Design of a tokamak fusion reactor first wall armor against neutral beam impingement

    Energy Technology Data Exchange (ETDEWEB)

    Myers, R.A.

    1977-12-01

    The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiation damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem.

  2. Thermodynamic modelling and solar reactor design for syngas production through SCWG of algae

    Science.gov (United States)

    Venkataraman, Mahesh B.; Rahbari, Alireza; Pye, John

    2017-06-01

    Conversion of algal biomass into value added products, such as liquid fuels, using solar-assisted supercritical water gasification (SCWG) offers a promising approach for clean fuel production. SCWG has significant advantages over conventional gasification in terms of flexibility of feedstock, faster intrinsic kinetics and lower char formation. A relatively unexplored avenue in SCWG is the use of non-renewable source of energy for driving the endothermic gasification. The use of concentrated solar thermal to provide the process heat is attractive, especially in the case of expensive feedstocks such as algae. This study attempts to identify the key parameters and constraints in designing a solar cavity receiver/reactor for on-sun SCWG of algal biomass. A tubular plug-flow reactor, operating at 24 MPa and 400-600 °C with a solar input of 20MWth is modelled. Solar energy is utilized to increase the temperature of the reaction medium (10 wt.% algae solution) from 400 to 605 °C and simultaneously drive the gasification. The model additionally incorporates material constraints based on the allowable stresses for a commercially available Ni-based alloy (Inconel 625), and exergy accounting for the cavity reactor. A parametric evaluation of the steady state performance and quantification of the losses through wall conduction, external radiation and convection, internal convection, frictional pressure drop, mixing and chemical irreversibility, is presented.

  3. Design considerations regarding slug ruptures in the intermediate power level reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pearl, W.L.; Pursel, C.A.

    1954-11-01

    The minimum shutdown time, to permit accessibility, for the Intermediate Power Reactor is estimated to be 38 hours. In case the reactor were shutdown following each rupture this long shutdown period would have serious disadvantages. The desirability of being able to make firm power commitments (independent of slug ruptures) has led to a study of the possibility of continuous operation following a rupture. There is evidence to indicate that, at the proposed water temperature, the rate of corrosion of uranium may be so high that at least a major portion of the rupture products may have entered the system before the reactor can be shutdown. A pushout of the affected column would then be a pushout of only those slugs which are still intact and the problem would still remain of removing the rupture products from the system. The first portion of this report is concerned with the rate of corrosion of a slug following rupture and the possible limitations to the principle of non-shutdown operation. These limitations include a flow stoppage by the ruptured can, undue increase in gamma activity, increased corrosion by the rupture products, and adherence of rupture products to the piping. The latter portion of the document is concerned with design considerations of the shielding and water plant so as to eliminate or minimize the effects of the introduction of rupture products into the cooling system. 7 refs., 2 figs.

  4. Nordic forum for generation IV reactors, status and activities in 2012

    Energy Technology Data Exchange (ETDEWEB)

    Van Nieuwenhove, R. [Institutt for Energiteknikk, OECD Halden Reactor Project, Kjeller (Norway); Lauritzen, B.; Nonboel, E. [Technical Univ. of Denmark. DTU Nutech, Roskilde (Denmark)

    2012-12-15

    The Nordic-Gen4 (continuation from NOMAGE4) seminar was this year hosted by DTU Nutech at Risoe, Denmark. The seminar was well attended (49 participants from 12 countries). The presentations covered many aspects in Gen-IV reactor research and gave a good overview of the activities within this field at the various institutes and universities. The present report contains book of abstracts. The individual Power Point presentations are indexed in INIS and may be found at http://nordic-gen4.org/seminars/nordic-gen4-riso-2012-2/ (LN)

  5. Licensed operating reactors: Status summary report, data as of December 31, 1995. Volume 20

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-06-01

    The US Nuclear Regulatory Commission`s monthly summary of licensed nuclear power reactor data is based primarily on the operating data report submitted by licensees for each unit. This report is divided into two sections: the first contains summary highlights and the second contains data on each individual unit in commercial operation. Section 1 availability factors, capacity factors, and forced outage rates are simple arithmetic averages. Section 2 items in the cumulative column are generally as reported by the licensees and notes to the use of weighted averages and starting dates other than commercial operation are provided.

  6. Design of an Air-Sparged Tubular Photocatalytic Reactor for the Degradation of Methylene Blue: Mass-Transfer Limitation Studies

    Directory of Open Access Journals (Sweden)

    Ramoso Patrick

    2016-01-01

    Full Text Available An alternative process for the removal of organic pollutants in aqueous systems is photocatalysis. The challenges hindering its industrial use are electron-hole recombination and mass-transfer limitations. In order to address these problems, the objective of this study is to introduce air by sparging, and design an air-sparged photocatalytic reactor using titanium dioxide immobilized on borosilicate glass. The performance of the reactor on the removal of the model pollutant, methylene blue (MB, was evaluated and compared against the reactor operated without sparging. The effect of mass-transfer limitations on reactor performance was also investigated by regression using a Langmuir-type model equation. The sparged photocatalytic reactor was able to degrade 57% MB in 2 hours, an improvement of 40% compared to no sparging, and is comparable to similar reactors in literature, but with the advantage of using less expensive materials of construction and simpler immobilization technique. Mass-transfer limitation studies showed a good fitting of the initial reaction rate r, with r = 0.1399Q / (0.6120 + Q for the sparged operation, and Q is the volumetric flowrate of water (L/min. The model also shows that the reactor operates near the reaction-limited regime, and that the extent of mass-transfer limitation effects was reduced by the present reactor.

  7. Evaluation for status and utilization of neutron activation analysis facility of HANARO research reactor (1996-2003)

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yong Sam; Moon, Jong Hwa; Chung, Young Ju; Kim, Sun Ha; Baek, Sung Ryel; Cho, Hyun Je; Park, Kwang Won; Kang, Sang Hun; Lim, Jong Myoung; Kim, Young Jin

    2004-05-01

    This report is written for the evaluation of the utilization and the status of Neutron Activation Analysis (NAA) facility of HANARO research reactor (1996-2003) which is accomplished by NAA user supporting system ; (i) the results by the number of sample is classified the research and development and analytical service, each item which is divided into the type of supporting (component analysis, nuclide analysis, neutron irradiation) and the field of utilization (environment, human health, industrial materials, standard materials, social and culture and others) are summarized with relevant contents in detail, (ii) The trends of the utilization is evaluated with a technology, a field, an organization of utilization, (iii) the results of training and education for students and users, the survey of user's demand, the investigation of self-satisfaction for NAA user supporting system.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  9. Present status of neutron beam facilities at the research reactor, HANARO, and its future prospect

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang-Hee; Kang, Young-Hwan; Kuk, Il-Hiun [Korea Atomic Energy Research Institute, Taejon (Korea)

    2001-03-01

    Korea has been operating its new research reactor, HANARO, since its first criticality in 1995. It is an open-tank-in-pool type reactor using LEU fuel with thermal neutron flux of 2 x 10{sup 14} nominally at the nose in the D{sub 2}O reflector having 7 horizontal beam ports and a provision of vertical hole for cold neutron source installation. KAERI has pursued an extensive instrument development program since 1992 by the support of the nuclear long-term development program of the government and there are now 4 working instruments. A high resolution powder diffractometer and a neutron radiography facility has been operational since late 1997 and 1996, respectively. A four-circle diffractometer has been fully working since mid 1999 and a small angle neutron spectrometer is just under commissioning phase. With the development of linear position sensitive detector with delay-line readout electronics, we have developed a residual stress instrument as an optional machine to the HRPD for last two years. Around early 1998 informal users program started with friendly users and it became a formal users support program by the ministry of science and technology. Short description for peer group formation and users activities is given. (author)

  10. Design considerations of the irradiation test vehicle for the advanced test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, H.; Gomes, I.C.; Smith, D.L. [Argonne National Lab., IL (United States)] [and others

    1997-08-01

    An irradiation test vehicle (ITV) for the Advanced Test Reactor (ATR) is being jointly developed by the Lockheed Martin Idaho Technologies Company (LMIT) and the U.S. Fusion Program. The vehicle is intended for neutron irradiation testing of candidate structural materials, including vanadium-based alloys, silicon carbide composites, and low activation steels. It could possibly be used for U.S./Japanese collaboration in the Jupiter Program. The first test train is scheduled to be completed by September 1998. In this report, we present the functional requirements for the vehicle and a preliminary design that satisfies these requirements.

  11. Research Activities on Development of Piping Design Methodology of High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Nam-Su [Seoul National Univ. of Science and Technology, Seoul(Korea, Republic of); Won, Min-Gu [Sungkyukwan Univ., Suwon (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering and Construction Co. Inc., Gimcheon (Korea, Republic of); Lee, Hyeog-Yeon; Kim, Yoo-Gon [Korea Atomic Energy Research Institute, Daejeon(Korea, Republic of)

    2016-10-15

    A SFR is operated at high temperature and low pressure compared with commercial pressurized water reactor (PWR), and such an operating condition leads to time-dependent damages such as creep rupture, excessive creep deformation, creep-fatigue interaction and creep crack growth. Thus, high temperature design and structural integrity assessment methodology should be developed considering such failure mechanisms. In terms of design of mechanical components of SFR, ASME B and PV Code, Sec. III, Div. 5 and RCC-MRx provide high temperature design and assessment procedures for nuclear structural components operated at high temperature, and a Leak-Before-Break (LBB) assessment procedure for high temperature piping is also provided in RCC-MRx, A16. Three web-based evaluation programs based on the current high temperature codes were developed for structural components of high temperature reactors. Moreover, for the detailed LBB analyses of high temperature piping, new engineering methods for predicting creep C*-integral and creep COD rate based either on GE/EPRI or on reference stress concepts were proposed. Finally, the numerical methods based on Garofalo's model and RCC-MRx have been developed, and they have been implemented into ABAQUS. The predictions based on both models were compared with the experimental results, and it has been revealed that the predictions from Garafalo's model gave somewhat successful results to describe the deformation behavior of Gr. 91 at elevated temperatures.

  12. Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

    Science.gov (United States)

    Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

    2007-01-01

    A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

  13. Application of computational fluid dynamics (CFD) codes as design tools for inertial confinement fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abanades, A; MartInez-Val, J M [Instituto de Fusion Nuclear, c/Jose Gutierrez Abascal 2, 28006 - Madrid (Spain); Sordo, F; Lafuente, A [Escuela Tecnica Superior de Ingenieros Industriales-UPM, c/Jose Gutierrez Abascal 2, 28006 - Madrid (Spain); Munoz, J [Fundacion para el Fomento de la Innovacion Industrial, c/Jose Gutierrez Abascal 2, 28006 - Madrid (Spain)], E-mail: abanades@etsii.upm.es

    2008-05-15

    The engineering design of the new innovative fusion reactors constitutes a clear challenge for the need to overcome several new technological edges in every engineering aspect. The great amount of thermal energy delivered into any inertial fusion chamber and the large temperatures and thermal gradients that are envisaged, joined to the even more demanding aspects related to neutron activation, Tritium breeding and the characteristics that are imposed to the coolant that could be used for that purpose, converged into material selection in which liquid metal seems to be one of the most interesting options. The safety assessment of such Fusion reactors should be clearly provided to fulfill the requirements asked by the Regulatory Bodies in a near-term future, when licensing will be a must. Therefore the availability of well proven and validated engineering design tools is a must. In this context, CFD is one of the tools that are potentially needed for thermal-hydraulic design of such complex machines. The state-of-the-art of CFD technologies will be shown, in particular in relation with liquid metals.

  14. Application of computational fluid dynamics (CFD) codes as design tools for inertial confinement fusion reactor

    Science.gov (United States)

    Abánades, A.; Sordo, F.; Lafuente, A.; Muñoz, J.; Martínez-Val, J. M.

    2008-05-01

    The engineering design of the new innovative fusion reactors constitutes a clear challenge for the need to overcome several new technological edges in every engineering aspect. The great amount of thermal energy delivered into any inertial fusion chamber and the large temperatures and thermal gradients that are envisaged, joined to the even more demanding aspects related to neutron activation, Tritium breeding and the characteristics that are imposed to the coolant that could be used for that purpose, converged into material selection in which liquid metal seems to be one of the most interesting options. The safety assessment of such Fusion reactors should be clearly provided to fulfill the requirements asked by the Regulatory Bodies in a near-term future, when licensing will be a must. Therefore the availability of well proven and validated engineering design tools is a must. In this context, CFD is one of the tools that are potentially needed for thermal-hydraulic design of such complex machines. The state-of-the-art of CFD technologies will be shown, in particular in relation with liquid metals.

  15. Optimizing the Design of Small Fast Spectrum Battery-Type Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Staffan Qvist

    2014-07-01

    Full Text Available This study is focused on defining and optimizing the design parameters of inherently safe “battery” type sodium-cooled metallic-fueled nuclear reactor cores that operate on a single stationary fuel loading at full power for 30 years. A total of 29 core designs were developed with varying power and flow conditions, including detailed thermal-hydraulic, structural-mechanical